REACTOR PHYSICS ACTIVITIES IN OECD COUNTRIES · (ITP/U Innsbruck) Gsterreichisches...

t NEA COMMITTEEON REACTOR PHYSICS !

REACTOR PHYSICS ACTIVITIES IN OECD COUNTRIES

October 1979-September 1980

OECD NUCLEAR ENERGY AGENCY 38 boulevard Suchet 75016 Paris

NEA COMMITTEE ON REACTOR P H Y S I C S

REACTOR P H Y S I C S A C T I V I T I E S I N

NEA MEMBER COUNTRIES

O c t o b e r 1979 - ~ e ~ t e m b e r 1980

OECD NUCLEAR ENERGY AGENCY 38 B o u l e v a r d Suchet, 75016 P A R I S

REACTOR PHYSICS ACTIVITIES IN NEA MEMBER COUNTRIES

This document is presented to the Reactor Physics. Idaho. from 22nd

a compilation of national activity reports Twenty-Third Meeting of the NEA Committee on held at Argonne National Laboratory.West. to 26th September 1980 .

Australia Austria ................................... 3 Belgium ................................... 11

Canada ................................... 22

Denmark ................................... 24 Finland ................................... 32 France ................................... 37 F.R. Germany ................................... 47 Italy ................................... 93 Japan ................................... 100

Netherlands ................................... 129 Norway ................................... 138 Spain ................................... 148 Sweden ................................... 160 Switzerland ................................... 166 United Kingdom ................................... 173 United States ................................... 186 JRC-Ispra ................................... 193

RERCMR PHYSICS ACTIVITIES I N AUSTRALIA

October 1979 - September 1980

D.B. MCCULLOCH

Australian Atomic Energy Commission Research Establishment Lucas Heights, New South Wales, Australia

1. REACTOR CODES DEVE,LOPMENT . Work has continued on POW-3D,the three-dimensional diffusion theory

'work-horse' module of the AUS scheme.

The e s sen t i a l mathematical rout ines to solve the large sparse system of l i nea r equations have been completed and extensively tested. Three methods capable of solving such systems a r e available:- Successive Line Overrelaxation (SLOR), Method of Incomplete Conjugate Gradients (ICCG) and Method of Implici t Nonstationary I t e r a t ion ( M I N I ) . Tests show t h a t ICCG is s l i g h t l y superior to M I N I i n the use of machine time f o r some problems, but addi t ional 1/0 a c t i v i t y is required, par t icu lar ly fo r the three-dimensional problem. Convergence of group equations is ass i s ted through the use of M I N I .

The code is serving a s a t e s t vehicle to study the e f f e c t of var ia t iona l methods a s a secondary means of accelerat ing convergence of large l i nea r systems. The dis junct ive par t i t ioning and dis junct ive weighting (DPDW) coarse mesh rebalancing lnethod used e a r l i e r in the two dimensional code POW

.was successfully extended t o three-dimensions, and has proved compatible w i t h a l l th ree i t e r a t i v e schemes.

~ w o more sophisticated systems a r e being tes ted. One involves a multiplicative pyramid correction fonn combined w i t h d is junct ive weighting (MPDw), and the other an addi t ive pyramid with dis junct ive weighting (APDW). MPDW involves considerable overhead but provides much superior performance t o APDW on ce r t a in t e s t problems. Compared with DPDW, however, the addi t ional overheads appear to date t o outweigh any improvement i n convergence, but this s i tua t ion may well be reversed when coarse mesh rebalancing i s applied f i l l y t o

b the eigen value problem

Although both new schemes seem t o be compatible with the three i t e r a t i v e methods on a number of t e s t problems, DPDW leads t o a smaller system of . equations t h a t is e f f i c i e n t l y solved with M I N I , while MPDW and APDW f a i l i n general to preserve the mathematical propert ies of t h e or ig ina l system, leaving d i r e c t methods a s the most e f f i c i e n t means of solution of the reduced System.

2. BURNUP METHODS

The burnup methods used within the ALjS scheme a r e current ly being upgraded. To date , burnup calculat ions have re l ied on CHAR wh.ich is a multi-region burnup module using an ana ly t ic method t o solve the nuclide depletion equations. CHAR has been applied i n the past mainly t o l a t t i c e burnup calculations. Its appl icabi l i ty t o global calculat ions was l imited to few-region calculat ions by the necessi ty i n most reactor types t o perform subsidiary l a t t i c e calculat ions a t each time s t ep for each region.

Although t h e methods development i s being done i n conjunction with an inves t iga t ion of burnup modelling i n l a r g e f a s t r e a c t o r s , requirements f o r thermal r e a c t o r burnup a r e a l s o being considered. The inves t iga t ion inc ludes t h e e f f e c t s of spec t ra used i n group condensation, mesh i n t e r v a l s , d i f fus ion theory versus SN method, he terogeni ty , number of regions with cons tan t burnup, time s t e p , v a r i a t i o n of i so tope c r o s s s e c t i o n s with i r r a d i a t i o n , and f i s s i o n product representa t ion .

The changes made t o t h e AUS scheme include:

(a) provis ion o f e d i t i n g f a c i l i t i e s i n CHAR,

(b ) allowing i so tope c r o s s sec t ions t o be i r r a d i a t i o n dependent which extends the use of CHAR i n g lobal c a l c u l a t i o n s ,

( c ) provision f o r energy condensation of i so topes over spec t ra from a g lobal ca lcu la t ion , and

(d ) add i t ion of a module which group-condenses t h e main c r o s s sec t ion l i b r a r y .

A new f i s s i o n product l i b r a r y is being generated from ENDFB using t h e OWL code XLACS. Some changes t o the XLACS resonance t rea tment were made t o reduce the required computer time.

A simple burnup module,BUWMAC,which adopts t h e usual assumption t h a t macroscopic l a t t i c e d a t a may be tabula ted aga ins t i r r a d i a t i o n has a l s o been w r i t t e n f o r AUS.

3. GROUP CROSS SECTION LIBRARY

Because a v a i l a b i l i t y of ENDFBV da ta is r e s t r i c t e d and t h e r e i s a l a r g e d i s c r e ancy i n 2 3 8 ~ resonance captures using ENDFB E, a modified ENDFB f i l e f o r 23iiU has been formed using t h e resonance da ta e z u a t e d by de Saussure e t a l t f o r ENDFBV. This modified da ta f i l e was used t o prepare new AUS c r o s s s e c t i o n s f z r 2 3 8 ~ which were t e s t e d i n AUE; c a l c u l a t i o n s o f t h e TRX-1 l a t t i c e experiment. Compared with E N D F B ~ data,keff increased by 0.2% and p2* decreased by 1%, s t i l l leaving a?% e r r o r i n p 2 @ compared with experiment. The matter was n o t pursued f u r t h e r because t h e change was r e l a t i v e l y small .

i de Saussure G . , Olsen D.K. , Perez R.B. and D i f i l i p p F.C. - ORNL/TM-6152

NEACRP-L-244 AUSTRIA

REACTOR PHYSICS ACTIVITIES IN AUSTRIA

September 1979 - September 1980

compiled by

B. Putz -

List of contributing organizations:

Atominstitut der Gsterreichischen Universitaten, Wien (AI)

Institut fiir Theoretische Physik der Technischen Universitlt Graz (ITE/TU Graz)

Institut fiir Theoretische Physik der Universitlt Innsbruck (ITP/U Innsbruck)

Gsterreichisches Forschungszentrum Seibersdorf Ges.m.b.H. (former Gsterreichische Studiengesellschaft fiir Atomenergie Ges.m.b.H.) (FZS)

1. REACTOR THEORY

1.1 Reactor Analysis

The investigations carried out at FZS on the consequences

a of the reduction of the fuel enrichment in the research

reactor ASTRA have been completed. The lower enrichment

necessitates a higher total uranium inventory per fuel

element which may entail metallurgical problems.

. Criticality calculations have been performed at the same

institute for the storage of BWR fuel elements in a high

density fuel rack made of boronated steel. The objective

of the studies had been the dependence of the results on

different calculational methods and on important design

parameters such as the concentration of boron, the thick-

ness of the boronated steel plates and the width of the

watergap /I/. ,

The neu t ron f l u x d i s t r i b u t i o n i n mu l t i sphe re c o n f i g u r a t i o n s

h a s been s t u d i e d a t ITP/TU Graz. The f l u x modulat ion by

t h e i n f l u e n c e o f ne ighbour ing sphe re s has been t aken i n t o

account by de t e rmin ing t h e f l u x o f a p o i n t s o u r c e o f neu-

t r o n s i n a n i n f i n i t e medium, which c o n t a i n s a s p h e r i c a l

p e . r t u r b a t i o n zone e c c e n t r i c t o t h e p o i n t s o u r c e . An i t e r a -

t i o n method a l l ows c o n t i n u a l l y improving approximat ions .

/ 2 / .

An a t t e m p t w a s made a t t h e above : i n s t i t u t e t o f i n d o u t

whether t h e i d e a l i z a t i o n o f a s p h e r i c a l u n i t ce l l i s

j u s t i f i e d o r n o t when c a l c u l a t i n g t h e : . n e u t r d n _ . s p e c t r ~ m .

o f pebble-bed co re s . To t h i s end t h e f i n e s t r u c t u r e

o f t h e f l u x d i s t r i b u t i o n i n a s p h e r i c a l f u e l e lement

and i n i t s ambient medium h a s b e e n determined u s i n g

i n t e g r a l t r a n s p o r t t h e o r y f3 / .

The c r i t i c a l masses f o r f u e l e lements o f r e s e a r c h r e a c t o r s

haye been c a l c u l a t e d a t ITPfTU Griiz f o r medium and h i g h

enr ichment and compared w i t h lowly e n r i c h e d uranium

dioxide-water-systems i n o r d e r t o i n v e s t i g a t e t h e i n f l u -

ence o f t h e h e t e r o g e n e i t y and o f t h e c l a d d i n g on t h e

c r i t i c a l i t y 1 4 J . a

S t u d i e s underway a t pZS on t h e method of Doppler we igh t ing ,

which e n a b l e s space dependent t empera ture e f f e c t s be ing . t aken i n t o account i n t h e p o i n t k i n e t i c s e q u a t i o n s , have

cont inued .

Resonance Absorption

The less significant and usually neglected temperature

dependence of resonances in the thermal region has

been investigated at FZS comparing the results obtained

with the y-J-formalism to those of a more exact

formula /5 / .

Neutron Thermalization

The development of a thermalization method combining

Selengut's method of overlapping neutron spectra with

the multicollision probability method has been completed

at FZS. The new method is suitable to a wider field of

application, especially to the homogenization of reactor

cells. Subdividing the cell into N regions leads to a

system of 2 N~ unknowns for the neutron currents. By

a recurrence formalism this system can be reduced to a

system with only N unknowns resulting in a considerable

saving of computer time / 6 / .

Synergetic Fusion-Fission Systems

Study of synergetic fusion-fission systems has continued

to be an important activity at ITP/TU Graz and at ITP/U

Innsbruck.

One of the topics dealt with at TU Graz is the impact of the integration of conventional fission reactors with

non-fission neutron sources (such as spallation neutrons

and fusion neutron sources) on the overal thermal-to-

electric conversion efficiency of the system /7/. Some

effort were directed toward the identification of a set

of efficiency merit parameters which more accurately

assesses conventional and synergetic nuclear energy systems

/ a / -

I n v e s t i g a t i o n s c a r r i e d o u t a t TU Graz on t h e mathe- '

m a t i c a l - p h y s i c a l s i m u l a r i t i e s and d i f f e r e n c e s between

f u s i o n and f i s s i o n m u l t i p l i c a t i o n p r o c e s s e s showed

t h a t advanced f u s i o n c y c l e s can s u s t a i n e x c u r s i o n

t e n d e n c i e s e s s e n t i a l l y analogous t o c o n v e n t i o n a l

f i s s i o n c y c l e s /9 / .

The energy break-even c o n d i t i o n s of a f u s i o n - f i s s i o n

r e a c t o r system, i n which t h e f u s i o n d e v i c e i s f u e l e d

w i t h deu te r ium o n l y and d r i v e n by n e u t r a l beam i n j e c t i o n ,

were s t u d i e d a t U Innsbruck . The i n t e r r e l a t i o n s h i p be t -

ween t h e f u s i o n neu t ron p roduc t ion r a t e , t h e plasma f u s i o n

g a i n and t h e p roduc t n e T E , and p a r t i c u l a r i l y t h e . e f f e c t

o f t h e i n j e c t e d h i g h e n e r g e t i c d u e t e r o n s on t h e s e para-.

m e t e r s were examined. The r e s u l t s i n d i c a t e t h a t even

t h e D-D f u s i o n p r o c e s s may be viewed a s 3 neut ron sou rce

s u f f i c i e n t t o d r i v e a s u b c r i t i c a l E i s s i o n / c o n v e r s i o n

assembly / lo( .

A t ITP/U Innsbruck a n a l y t i c a l app rox ima t ions have been

developed f ~ r t h e c h a r a c t e r i s t i c p a r a m e t e r s o f a hybr id

b r e e d e r , which e n a b l e s a f u l l y a n a l y t i c a l d e s c r i p t i o n

o f t h e b l a n k e t performance va ry ing w i t h f u e l r e s i d e n c e

t ime . A t t e n t i o n h a s been p l a c e d On the i n f l u e n c e of f i s s i l e f u e l enr ichment and on t h e b u i l d up o f f i s s i o n

p rqduc t s [11[.

EXPERIMENTAL REACTOR PHYSICS

Water I n g r e s s i n t o Graph i t e Assemblies

Using t h e r e a c t o r code GAMTEREX t h e l a y o u t of exper iments

h a s been determined a t ITP/TUGraz aimed a t s t u d y i n g t h e

w a t e r i n g r e s s i n t o pebble beds o f AVR f u e l e lements . These

experiments are planned to be performed at the siemens-

Argonout-Reactor (SAR) in Graz and first measurements

of reaction rates in "dry" pebble beds are provided as

preliminary tests for the experiments at the "wet"

core /12/.

At the same institute some effort has been devoted to

investigitions on how insertion of water in the internal

graphite reflector of the SAR influences the reactivity

/ 13 / .

0 2.2 Neutron Flux Control

A code enabling the adjustment of a constant neutron

flux in a research reactor has been written at ITP/TU

Graz for the microprocessor MC 6800 /14/.

2.3 Neutron Spectrum

The fast neutron emission spectrum of 252~f has been

investigated at A1 by means of proton recoil spectro-

meters. With a large counter tube of 900 mrn length the

neutron distribution between 0.9 MeV and 10 MeV could

be determined. Monte Carlo calculated response functions

were applied to infold the measured proton recoil

distributions. The energy interval between 1 MeV and

3 MeV had been examined with a smaller tube (466 man)

in a search for neutron fine-structure groups. No such

groups could be established. . 2.4 Cross Section Measurement

As an application of photoneutron sources absolute

measurements of the absorption cross section of tan-

talum were performed at A1 using a transmission method

and a long counter as neutron detector. The neutron

e n e r g i e s were 20.9 keV, 121.8 keV, 215.3 keV and

837.8 keV. The r e s u l t s may, i n p r i n c i p l e ; be u s e d ' a s

r e f e r e n c e v a l u e s f o r r e l a t i v e measurements.

2 . 5 Delayed Neutron Measurement

A method u s i n g a s o l i d s t a t e n u c l e a r t r a c k d e t e c t o r

h a s been developed a t A 1 t o t e t e c t d e l a y e d n e u t r o n s

from f i s s i o n p r o d u c t s con ta ined i n t h e p r imary c o o l a n t

o f a n u c l e a r r e a c t o r . I n an in -core l o o p o f t h e TRIGA

r e a c t o r Vienna a sma l l sample o f 93% e n r i c h e d uranium

was i r r a d i a t e d and t h e f i s s i o n p r o d u c t s w e r e t r a n s -

p o r t e d by a purg ing system t o t h e t r a c k d e t e c t o r . A

c o r r e l a t i o n could be o b t a i n e d between t h e number o f

t r a c k s and t h e r e a c t o r power and t h r e e g r o u p s o f delayed

n e u t r o n s w e r e i d e n t i f i e d /15 / .

3. GENERAL

A d e t a i l e d su rvey h a s been worked o u t on t h e e x p e r i -

mental and t h e o r e t i c a l s t u d i e s t h a t have been performed

a t ITP/TU Graz between 1973 and 1979 conce rn ing i n v e s t i -

g a t i o n s on t h e n u c l e a r p h y s i c a l behav iou r o f wa te r

moderated pebble beds and t h e i r E e a s i b i l i t y f o r power

p l a n t s / I 6/.

Based on t h e e q u i v a l e n t f u e l concep t and t h e f u e l s t o c k p i l e

concept t h e f i s s i l e f u e l t r a j e c t o r y c o n c e p t were developed

and a p p l i e d t o b u r n e r , c o n v e r t e r and b r e e d e r r e a c t o r s / l 7 / . .

REFERENCES :

F. WOLOCH, G. SDOUZ, M. SUDA, Neutronenphysikalische Aspekte der NaRlagerung von SWR-BE-Bundeln. ATKE - 35, 166 (1980).

F. SCHuRRER, A Diffusion-Theoretical Method to Cal- culate the Neutron Flux Distribution in Multisphere Configurations. ATKE - 35, 179 (1980).

F. SCHORRER, Successive Approximation of the Neutron Flux Distribution in Spherical Configurations. Acta Physics Austriaca (in the press).

H. MULLER, H. RABITSCH, F. SCHtiRRER, The Criticality of Water Reflected Homogeneous Arrays and of Heterogeneous Reactor Fuel Elements. Acta Physica Austriaca (in the press).

G. KAMELANDER, Reactor Physical Effects of Thermal Resonances. ATKE (in the press).

G. KAMELANDER, F. PUTZ, Application of the Multigroup Collision Probability Method to Selengut's Theory.of Overlapping Neutron Spectra. Nuc1.S~. Eng. - 74, 13 (1980).

M. HENDLER, A.A. HARMS, The Efficiency Decrement of Self-sufficient Nuclear Energy Systems. Trans. Am. Nucl. Soc. - 33, 785 (1979).

M. HEINDLER, A.A. HARMS, Efficiency Merit Assessment of emerging Synergetic Nuclear Energy Systems. ATKE - 36, 7 (1980).

A.A. HARMS, M. HEINDLER, The Existence and Characteri- zation of Self-sustaining Multiplicative Fuston and Fission Reaction Chains. Acta Physica Austriaca - 52 CDec. 1980) (in the press) . K.F. SCH~PF, Beam Driven D-Fusion Plasma within a Fusion-Fission Hybrid System. ATKE - 36, 26 (1980).

. - 1 K. SCHBPF, G. STRASSER, Analytical Description of the Fuel Dynamics in a Hybrid Fusion Breeder. ATKE (in the press).

/12/ F. SCHtiRRER, First Series of Measurements to the Project "Water Ingress into Pebble Beds of AVR Fuel Elements". Interal Report ITPR-79009, TU Graz 1979.

/13/ Hj. MuLLER, W. NINAUS, K.OSWALD, Changes in Reactivity by Insertion of Water in the Internal Graphite Reflector of the Argonaut. Acta Physica Austriaca (in the press).

/14/ W. NINAUS, G . KAHR, Neut ron F l u x C o n t r o l o f a ~ e s e a r c h R e a c t o r by a Microcomputer System. A c t a P h y s i c a A u s t r i a c a ( i n t h e p r e s s ) .

/15/ H. BBCK, D e t e c t i o n o f Delayed N e u t r o n s i n a N u c l e a r R e a c t o r Using t h e S o l i d S t a t e Track E t c h T e c h n i q u e . P a p e r p r e s e n t e d a t t h e 1 0 t h Int . .Conf. on S o l i d S t a t e N u c l e a r Track D e t e c t o r s , 2nd-7th J u l y 1 9 7 9 , Lyon, F r a n c e .

/ 1 6 / E. LEDINEGG, M. HEINDLER, Hj. MULLER, W. NINAUS, H . RBBITSCH, F . SCHuRRER, N u c l e a r P h y s i c a l Behav iour o f Water Moderated P e b b l e Beds. I n t e r n a l r e p o r t ITPR- 79010, T U Graz , 1979.

/17/ A.A.HARMS, M . HEINDLER, L i f e t i m e F u e l T r a j e c t o r i e s f o r F i s s i o n R e a c t o r s . T r a n s . An,. Nucl . SO;. - 3 3 , 125 ( 1 9 7 9 ) .

WACRP-L-24 4 BELGIUM

REACTOR PYYSICS ACTIVITIES IN BELGIUM

Progress report to the NEA Committee on Reactor Physics

Compiled by J. DEBRUE, SCK-CEN, Mol

Septeaber 1980

EiERMAL REACTORS

1 ,. Fuel Cycle'

a) Low power experiments in the VENUS criticality facility ....................................................... The variation of the reactivity of Pu02-U02 fuel ccnfigurations over

long periods of time (w 10 ears) was further investigated. This

variation is due to the 241~m build-u? resulting from the natural decay

of 241p~ (half-life : 14.4 years), The theoretical analysis has been

performed with the DLC 43B/CSRL cross-section library (218 groups, P3)

based on ENDF/B IV. The code packages AMPX-I1 A and MARS have been

used to produce 68 group cross-section sets which allowed to calculate '

weighting spectra in the different regions of the loadings by means

of ANISN; collapsing in 7 groups was finally made to perform 2 dinen- sional XY calculations with DOT 3,5. Comparing the calculated keff

with the experimental values, as obtained from 1969 to 1979 for near

critical configurations (adjustment is made by adding peripheral rods), 241 . indicates that the neutron capture in Am could be overestimated in

the calculation by about 10 to 20 %. This assuves that the effect of 241

the Pu decay is exactly calculated i,e, that the 241 Pu cross-sections

, in the thermal and resonance energy range are correct in ENDF/B IV.

However, according to the recent 241

Pu capture cross-section measurement

by Weston and Todd [I], the resonance capture for this isotope is

significantly underestimated in ENDF/B IV. Taking into account the

Weston and Todd date, the calculations agree with thc criticzlity

measurements in VENUS within the experimental error margin.

b) Power reactor calculations .......................... - The application of the calculation methods currently in use at Electrobel for three-dimensional simulation of P'fl reactors was

presented at the NEACRP sponsored Specralists' Meeting in November

1979 [ 2 ] . In order to avoid true 3 D calculations, the core is

mdelled as a perturbation of the "base" reference situation calcu-

lated by the MERCATOR-XY nodal simulator [j]. MERCATOR-Z operates

on the macroscopic cross-sections, condensed in each Z plane, to

solve the diffusion equation in one dimension, in terms of XY ave-

rages of the flux at each Z level. Most of the experience with

the LWR-WIHS/blERCATOR XY/MERCATOR Z chain, has been gained on fol-

lowing the TIHANGE I reactor. Control bank insertion, boron con-

centration and axial profiles as calculated for different steady

and transient conditions have been satisfactorily compared with

experimental observation.

- The TRILUX code, initially developed by GUNF, has been improved and extended by BELGONUCLEAIRE. TRILUX calculates 3 D nodal power density

distributions using a modified one-group nodal coupling calculation 2

in which each fuel node is characterized by km and M . The LVR-WIMS'

code is used as assembly constant generator. Two options have been

added to TRILUX : XENOLUX for the evaluation of xenon transients and a

MICROLUX for the determination of the pin power distribution over

selected nodes or over an "average" plane of the whole core. These

calculation tools have been compared with more sophisticated codes. . Operation data from SENA,and TIHANGE I were also recalculated for

checking the validity of the TRILUX - MICROLUX system. The interest

of using the MICROLUX option has been demonstrated in the calculation

of power and burn-up maps when important gradients of the thermal flux

exist at the interface of fuel assemblies, e.g. in Pu recycle confi-

gurations. Taking into account these gradients increases the power

release in a Pu assembly by about 6 7; [ L I ] .

- The parane , t r i c s tudy of s teady st a t e cond i t i ons and acc iden t ' ana -

l y s i s f o r a P m of 900 MWe loaded with 0 , 30 and 70 % plutonium

assembl ies has been completed a t BELGONUCLEAIRE wi th in t h e frame

of a c o n t r a c t wi th t h e CEC.

c ) Out-of-pile f u e l cyc l e ...................... - The m u l t i p l i c a t i o n f a c t o r of f r e s h f u e l s t o r a g e racks was c a l c u l a t e d ,

assuming t h a t t h e system is sub jec t ed t o water of a d e n s i t y vary ing

from 0 t o 100 %. S a f e t y c r i t e r i a (ANSI 18.2) impose t h a t ke f f must

be lower than 0.98 f o r t h e opt imal d e n s i t y ; t h i s optimal d e n s i t y is

of t he ' o rde r of 1 0 % t o 20 % f o r u sua l s t o r a g e racks of P'dR assem-

b l i e s . A t nominal d e n s i t y , kef f may not reach 0.95.

A p r a c t i c a l arrangement is a squa re a r r a y of assembl ies , each of them

being loaded i n a s t a i n l e s s s t e e l can of 5 mm th ickness . Such a

system, with an assembly p i t c h of 35 cm, is reasonably compact f o r

f r e s h f u e l s t o r a g e (10 x 1 0 assembl ies f o r example), provides t h e

mechanical p r o t e c t i o n normally necessary b u t does not r e q u i r e s p e c i a l

absorbing m a t e r i a l l i k e boron s t e e l . Moreover, t h e presence of t h e

s t e e l cas ing reduces s i g n i f i c a n t l y t he keff peaking a t low dens i ty .

The c a l c u l a t i o n s were performed, f o r a 2 3 5 ~ enrichment of 3.5 %, w i t h

neutron t r a n s p o r t codes :

- DTF-IV wi th 40 neutron energy groups : 1 D c y l i n d r i c a l approxima-

t i o n f o r a s i n g l e assembly i n an i n f i n i t e a r r a y

- DOT 3.5 wi th 6 neut ron energy groups f o r 2 D c a l c u l a t i o n s .

The v a l i d a t i o n of c a l c u l a t i o n methods remains a problem a s long a s

s y s t e n a t i c a l in tercomparisons a r e not a v a i l a b l e .

- I n r e l a t i o n with t h e s a f e t y a s p e c t s of handl ing and s t o r a g e of mixed

oxide f u e l assembl ies f o r PWR's, neu t ron and gamma dose r a t e s i n t h e

v i c i n i t y of Pu02-U02 f u e l assembl ies were c a l c u l a t e d by BELGONUCLEAIRE

f o r comparison with measured va lues on d i f f e r e n t types of assembl ies

manufactured i n t h e p a s t f o r B R 3 , SENA and DODEYAARD. The c a l c u l a t i o n

methods were app l i ed t o p lu ton i -~m assembl ies designed f o r a TIAANGE

type r e a c t o r . C r i t i c a l i t y c n l c u l n t i o n s were a l s o perforned i n t h e ' g!pJ-ld 16

light of the criteria ANSI 18.2, for fresh fuel assenblies in normal1.y

dry storage rlcks and for spent fuel assemblies in high density sto-

rage racks with boron containing cans around the assemblies. Mixed

oxide fuel as well as uranium oxide fuel were considered in this

evaluation [5].

2. Pressure vessel studies

The objective of these studies is to improve the neutronic aspects of LVR

pressure vessel surveillance methods and to validzte the neutron embrittle-

ment characteristics for the steel type used in the new belgian power plants.

The interlaboratory cooperation with US laboratories has been pursued in

the framework of the L1tfR Pressure Vessel Irradiation Surveillance Dosimetry

progranme supported by the NRC [ 6 ] .

Host of the efforts have been devoted to the analysis of the experimental

results obtained at the O R N L Pool Critical Assembly (PCA) and to the

characterization of the Pool Side Facility (PSF) at the O R R where the

irradiation of steel specimens has been started.

- Two configurations of the PCA Pressure Vessel mock-up were studied, corresponding to two positions of the thermal shield and pressure vessel ,

simulators with respect to the PCA reactor core. A detailed map of the

fission density in the core itself was first performed to provide an

exact picture of the fast neutron source distribution on an absolute

basis.

Threshold reaction rates were measured with fission chambers ( 2 B U 1 237Np)

and activation detectors ('031?h, 'I5.In, 58Ni, 27~1) at different neutron 6

penetration depths in water and in the simulators. ~i(n,a) neutron

spectrometry measurenents were finally made at three locations within

,the pressure vessel steel.

The accuracy of the measurements is better than - + 5 % for the spectral

indices and better than - + 7 % for the absolute equivalent fission fluxes

per unit PCA core neutron strength. A transport theory analysis of both

configurations was carried out : one- and two-dimensional multigroup

S8 - P3 cnlculations were performed including coupled neutron-gamma

calculations to correct neutron dosimeters for gamma ray induced res-

ponses.

The calculations reproduce the integral measurement results to within

an uncertainty of 2 25 :A. However it remains to ascertain the signi-

ficance of errors associated to the treatment of vertical neutron

leakage effects in the 2 D calculations, resulting from the more

limited height of the PCA core as compared with power reactors. This

calla in particular for pursuing the leakage sensitivity study under-

taken at SCK/CEM and included in the report presented at the PCA 6

"Blind Test" meeting [7]. The consistency of the Li(n,a) neutron

spectra -[a] with the spectral indices also requires further investi-

gation although a general agreement of + 10 % has been reached when considering only the reaction rates mostly sensitive in the neutron

energy range covered by the spectrometry technique.

The Blind Test exercise was organized in the frame of the validation

effort of transport theory computations needed to extrapolate, into

the pressure vessel of a power reactor, the dosimetry results obtained

at a surveillance position. The initial comparison of the solutions

proposed by the participants took place in May, 1980, at the NBS [g].

- The dosimetry radiometric measurements performed at PCA have Seen repeated in the actual PSF configuration at low power (10 - 20 \Id)

and at high power (30 MW); the equivalent fission fluxes for 03~h,

lq51n, 58~i and 27~1, reported to a unit core power defined by fission

chanber neasurements at PCA and by thermal balance at ORR, have been

found identical within - + 3 7L or better.

As part of this PSF "start up" characterization programme, various

damage exposure parameters, such as jb > 1 MeV, fl > 0.1 MeV and dpa

were derived from the radiometric results coupled with transport

theory calculations. It is concluded that in-vessel projections of

surveillance capsule enbrittlement data based on ,6 > 1 MeV as the damage correlation parameter may be non-conservative by up to 15 %

if dpa proves adequate and by up to 40 if @ > 0.1 MeV proves adequate,

a s s u g g e s t e d by some e x p e r i m e n t s . The s t e e l spec imen i r r a d i i t i o n s i n

PSF, a t t h e s u r v e i l l a n c e p o s i t i o n and i n t h e p r e s s u r e v e s s e l s i m u l a t o r ,

w i l l h e l p t o c l a r i f y t h i s i m p o r t a n t problem.

- Dosimetry measurements a r e under p r o g r e s s i n BR3 f o r comparison w i t h

c a l c u l a t i o n s i n t h e r e a l c o n d i t i o n s of a p p l i c a t i o n of t h e t h e o r e t i c a l

methods. A f i r s t s e r i e s o f r e s u l t s were o b t a i n e d i n B R 3 a t v a r i o u s

l o c a t i o n s be tween t h e p e r i p h e r y o f t h e c o r e and t h e o u t e r s i d e o f

t h e p r e s s u r e v e s s e l . On t h e o t h e r h a n d , s e v e r a l s u r v e i l l a n c e c a p s u l e s

were un loaded from t h e TIHANGE and DOEL r e a c t o r s . The r a d i a l and

a z i m u t h a l g r a d i e n t s o f t h e f a s t n e u t r o n f l u x i n t h e c a p s u l e s have been

de te rmined a c c u r a t e l y by measur ing t h e 54Mn a c t i v i t y i n t h e remnants

a v a i l a b l e a f t e r t h e mechan ica l t e s t s . A s a l a r g e number o f niobium

d o s i m e t e r s had been l o a d e d i n one of t h e s e c a p s u l e s , i t h a s been pos-

s i b l e t o d e f i n e a n e f f e c t i v e a c t i v a t i o n c r o s s - s e c t i o n f o r t h e r e a c t i o n

9 3 ~ b ( n , n ' ) 9 3 m ~ b on t h e b a s i s of t h e measured f l u e n c e by means of t h e

more c o n v e n t i o n a l d o s i m e t e r s . Fo r l o n g i r r a d i a t i o n t i m e s , n iobium,

w i t h i t s h a l f - l i f e of 16.4 y e a r s , w i l l b e t h e most v a l u a b l e f l u e n c e

mon i to r .

3. R e a c t o r o p e r a t i o n

The BR2 r e a c t o r h a s been r e s t a r t e d a t t h e b e g i n n i n g o f J u l y , 1980 , a f t e r

r ep lacemen t o f t h e b e r y l l i u m m a t r i x . The r e a c t o r was shutdown s i n c e 1 8

months. The o r i g i n a l m a t r i x had been s u b m i t t e d t o a n e u t r o n f l u e n c e of

a b o u t 8 x 1 0 ~ ~ n / c m ~ (> 1 MeV) i n t h e h i g h e s t r a t e d p i e c e s . S y s t e m a t i c a l

measurements were c a r r i e d o u t b e f o r e d i s m a n t l i n g i n o r d e r t o c o r r e l a t e the

d i m e n s i o n a l c h a n e e s a t d i f f e r e n t p o s i t i o n s w i t h t h e l o c a l f l u e n c e v a l u e s ,

The d i a m e t e r of t h e c h a n n e l s , which i s n o m i n a l l y 8 4 m m , i n c r e a s e d w i t h t h e

22 2 f l u e n c e t o r e a c h 1.1 m m a t 8 x 1 0 n/cm . By r e a s o n of t h e t r e n d o f t h e

c u r v e , i t seems t h a t t h e c r i t i c a l t e m p e r a t u r e f o r b e r y l l i u m equa led t h e 2 2 2

p h y s i c a l t e r n p e r a t w e (" 50' C) a t a f l u e n c e o f a b o u t 6.5 x 10 n/cm . Above t h i s v a l u e , t h e s w e l l i n g r a t e i n c r e a s e d more r a p i d l y (he l ium bubb le

f o r n a t i o n on rain b o u n d a r i e s ) ; a n i n c r e a s e o f t h c t r i t i u m c o n t e n t i n w a t e r

was obse rved i n p a r a l l e l . Helium c o n c e n t r a t i o n measurements i n s m a l l

samples of the unloaded pieces will allow to confirm the fast neutron

dose distribution. Fresh beryllium samples were previously irradiated

together with neutron dosimeters for calibration purpose.

The new matrix was loaded in January 1980 [lo] and the reactor was made

critical again at low power on May 12, 1980.

FAST REACTORS - 1. Critical experiments

The analysis of the experiments carried out in ZEBRA at Winfrith in the

frame of th-e BIZET programme under AEEW/KfK agreement was pursued.

BELGONUCLEAIRE, as DeBeNe partner, took part to this work :

- the correction factors for the modified source multiplication method applied to the conventional 2-zone core loading BZA were recalculated

using SNR methods and data

- the calculations of the control rod experiments in BZA were reanalysed in order to compare the UK and DeBeNe results; nine symmetrical B4C

control rod configurations were considered. On the DeBeNe side, diffu-

sion theory is used with cross-sections condensed from the

26-group KFK/INR 001 set; the calculations were made in two and three

dimensions

- the gamma-ray energy deposition measurements in the heterogeneous loading BZC/I [II] were also calculated with the SNR design methods and photon

libraries. Differences in the fissile zone are attributed to the photon I( source library and in the fertile zones to the diffusion/transport

effects

- the relative reactivity of pins and plate cells was calculated using the FGL5/MURAL cross-sections and the pin sector substitution measurements in

BZD/3 were extrapolated to a fuel pin core

- the effect of insertion of a Na/steel channel in the central fertile

island of BZD/3 was evaluated in function of the channel diameter.

On the experimental side, an intercomparison of gamma dose measurements

by means of thermoluminescent detectors was undertaken. The reference 60

gamma fields ( Co source) at Harwell and Mol, which were used for cali-

bration of the TLD's in the BIZET programme, were first intercompared by

means of the "MOL" ionization chamber exposed in both facilities; the

absolute value measured at Harwell agrees with the "HARWELL" ionization

chamber result within the 3 to 4 % systematical plus statistical uncer-

tainty. Moreover, TLD's from a batch calibrated at Mol were exposed at

Harlwell and a similar agreement was obtained. An improved proce6ure of

selection and calibration of TLD's is now 'being developed at Mol in order

to achieve a better control of the uncertainty on individual measurement

results in a critical facility.

2. Safety studies

The feasibility study of PAHR (Post Accide:nt Heat Removal) irradiation

experiments in BR2 is investigated in order to evaluate the heat removal

capabilities in the case of a core meltdown in a fast reactor, when the

core debris are collected in the core catcher. The irradiation device

would be located in the 20 cm diameter central channel of the reactor;

the fufl particle bed diameter simulating the core debris would amount

8 to 10 cm.

The neutronic calculations were carried out with the SCK/CEN version of

the neutron transport code DTF-IV and the BR2 fourty energy group library.

All calculations were made in R-geometry. The gamma calculations were

also performed with the DTF-IV code, together with the EURLIB ( Y , Y )

library with twenty energy groups. Preliminary two-dimensional (R,Z)

ganna heating calculations wzre performed with the aid of the DOT 3.5 code.

The possibility of increasing the diameter of the central channel up to

40 cm was also investigated with a view to accepting larger PAHR experi-

ments, with a fuel particle bed diameter of about 20 cm. Neutronic cal-

culations indicate that such n major modification is acceptable from the

reactor operation point of view and that irradiations of standard devices

as those used in the past cnn be carried on.

I n each c a s e , a r e l a t i v e l y f l a t r a d i a l f i s s i o n d e n s i t y d i s t r i b u t i o n could

be ob ta ined , t h e gamma h e a t i n g being of minor importance a s t h e r e a c t o r

would ope ra t e a t a reduced power l e v e l .

I n p a r a l l e l was s t a r t e d a d e t a i l e d modelling of t h e phenomena occur r ing i n

a d e b r i s bed made of oxide f u e l and s t a i n l e s s s t e e l p a r t i c l e s s a t u r a t e d

with l i q u i d sodium, wi th i n t e r n a l hea t genera t ion . One- and two-dimensional

approaches of t h e problem a r e be ing developed t o e v a l u a t e t h e u se fu lnes s

of PAHR experiments i n BR2 [12].

3. Neutron dosimetry f o r f u e l and m a t e r i a l i r r a d i a t i o n s

The c o l l a b o r a t i o n with t h e Max Planck I n s t i t u t e (MPI) i n S t u t t g a r t was

continued with a view t o t h e u t i l i z a t i o n of niobium a s monitor f o r long-

term i r r a d i a t i o n s . Very pure niobium m a t e r i a l , f a b r i c a t e d at t h i s

I n s t i t u t e , was i r r a d i a t e d t o g e t h e r with commercially a v a i l a b l e niobium

i n BR2. During t h e f i r s t weeks o r months a f t e r t h e i r r a d i a t i o n , t he

9 3 m ~ b a c t i v i t y ( h a l f - l i f e 16.4 ears) i s d i s t u r b e d by t h e 1 8 3 ~ a , 9 5 m ~ b ,

1 8 2 ~ a and 9 5 ~ b a c t i v i t i e s . A f t e r h a l f a y e a r , the. 93"~b a c t i v i t y is only

in f luenced by t h e IG2Ta a c t i v i t y ( h a l f - l i f e 115 days) . Th i s i n f luence

is p r a c t i c a l l y n e g l i e i b l e i n t h e MPI niobium whereas i t can c o n t r i b u t e . t o 20 ... 100 9: of t h e measured a c t i v i t y i n commercial niobium, depending

on t h e f o i l t h i c k n e s s i n t h e range 1 0 t o 100 pm. 22 -2

Niobium f o i l s i r r a d i a r e d i n EBR-I1 a t a t o t a l f l uence of 1 0 n.cm and 2 0 -2

i n BR2 a t a f i s s i o n equ iva l en t f luence of 5 x 10 n.cm were measured by

s i x l a b o r a t o r i e s . The measured 93m~b a c t i v i t i e s ag ree wi th in a few

pe rcen t s when t h e same d a t a a r e used f o r t h e h a l f - l i f e and t h e K(X)-ray 7

emission p r o b a b i l i t y .

The niobium measurements f o r EBR-I1 were p a r t o f m u l t i f o i l dosimetry

measurements ( T i , Fe, Co, N i , Cu, Nb, 235iJ 9 2381J, 2 3 7 ~ p ) a t d i f f e r e n t

a x i a l l e v e l s i n a c e n t r a l core channel. These measurements have been

completed; t h e s p e c t r a l unfo ld ing is performed a t HEDL.

REFERENCES

[ I ] L.W. WESTON and J . H . TODD (ORNL)

Neutron Cap tu re and F i s s i o n Cross S e c t i o n s of ~ l u t o n i u m - 2 4 1

Nucl. Sc. and Eng., 65, 4.54-463, 1978

[2] M. MELICE ( E l e c t r o b e l )

HERCATOR-2, a P e r t u r b a t i o n Approach t o PWR Core S i m u l a t i o n

P r o c e e d i n g s of a S p e c i a l i s t s ' Meeting on " C a l c u l a t i o n of 3-Dimensional

R a t i n g D i s t r i b u t i o n s i n O p e r a t i n g R e a c t o r s " ,

P a r i s , November 26-28, 1979

[3] M. HELICE ( E l e c t r o b e l )

A Nodal-Modal Coarse-Mesh Method f o r S o l v i n g t h e Two-Group D i f f u s i o n

E q u a t i o n

Repor t NEACRP-L-228, November 1978

[4] A. CIIARLIER, A. MOCKEI, and J.P. TESCH ( B e l g o n u c 1 6 a i r e )

TRILUX Nodal Sys tem f o r In-Core F u e l Manngement S t u d i e s

P r o c e e d i n g s of a S p e c i a l i s t s ' Meeting on " C a l c u l a t i o n of 3-Dimensional

R a t i n g D i s t r i b u t i o n s i n O p e r a t i n g R e a c t o r s 1 ' ,

P a r i s , November 26-28, 1979

[5] C. VANDENBERG ( B e l g o n u c l b a i r e )

Gamma and Neu t ron Dose R a t e s i n t h e Handl ing and t h e S t o r a g e of

P lu tonium F u e l A s s e m b l i e s

X i s $ , Oc tobe r 2 5 , 1979

[6] W.N. Mc ELROY (HEDL) e t a l .

LWR P r e s s u r e V e s s e l S u r v e i l l a n c e Dosimetry Improvement Program.

1979 Annual R e p o r t

NUREG/CR-1291, HEDL-SA 1949

[ 7 ] G. MINSART (SCK/CEN) ...

N e u t r o n i c Computa t ions o f t h e Poo l C r i t i c a l Assemt,ly P r e s s u r e V e s s e l

F a c i l i t y (PCA-PVF)

"PCA B1ip.d T e s t 1 ' Mee t ing , Washington, Nay 22-23 , 1 9 8 0

?2 , , , ., <..., ' . > , ., (.: , ,: .~,,'! . ' , . ' . :

:,, ..: C I..: 1: .'. ,.

[8] G. DE LEEUW-GIERTS, D. LANGELA (SCK/CEN) 6 Li Spectrometry Results in PCA

ItPCA Blind Test" Meeting, Washington, May 22-23, 1980

- [9] C.Z. SERPAN (NRC), M.N. Mc ELROY (HEDL), F.B.K. KAM (ORNL) and

A. FABRY (SCK/CEN)

Minutes of the May 22-23, 1980, Computational Blind Test on the

Pool Critical Assembly Pressure Vessel Mock-up Facility

'[lo] F. LEONARD, A. FALLA (SCK/CEN)

Remplacement de la Matrice de Beryllium du R6acteur BR2

Irradiation Devices Working Group (EURATOM), 26th Meeting,

Geesthacht, October 8-10, 1980

I ] A.D. KNIPE (AEEkl) , R. de WOUTERS (Belgonucl6aire)

Gamma-Ray Energy Deposition Measurements in a Heterogeneous Core

and their Analysis

International Symposium on Fast Reactor Physics (IAEA-SM 244).

Aix-en-Provence, September 24-28, 1979

1:12] C. BENOCCI, J.M. BUCHLIN (von KARMAN Institute, ~hode-Ste-Gensse,

Belgium), C. JOLY, A. SIEBERTZ (SCKICEN)

Pre-Boiling State in - Post - Accident - Heat gemoval Situation : 1 D and 2 D Theoretical Approaches Including

Natural Convection Effects

International Seminar on Nuclear Reactor Safety Heat Transfer,

. Dubrovnik, September 1980. ,

NEACRP-L-244 CANADA

REACTOR PHYSICS ACTIVITIES I N CANADA -

M.F. D u r e t

POWER REACTOR PROGRAM

The Bruce A and P i c k e r i n g c o n t i n u e t o o p e r a t e a t near maximum power. A g r e a t dea l o f i n t e r e s t has r e c e n t l y been expressed by O n t a r i o i n d u s t r i e s i n u s i n g steam produced i n t h e Bruce g e n e r a t i n g s t a t i o n . T h i s o b j e c t i v e i s b e i n g g i v e n h i g h p r i o r i t y by t h e government.

The r e a c t o r c o n s t r u c t i o n program i s on schedule w i t h abou t 15 Gw( e ) expected t o be i n s e r v i c e by 1990.

EXPERIMENTAL REACTOR PHYSICS

Work i n t h e ZED-I1 r e a c t o r d u r i n g t h e p a s t y e a r has g e n e r a l l y been s e r v i c e o r commercia l work t o s a t i s f y requ i rements o f o t h e r groups. T h i s i n c l u d e s work t o measure f l u x d i s t r i b u t i o n s and r e a c t i v i t y o f 36-e lement bund les o f PuD2/U0 f u e l w i t h a i r and D20 c o o l a n t s p r i o r to i r r a d i a t i o n t e s t i n g i n an NRU ? oop and f l u x measurements w i ' i h i n new des igns o f s e l f - powered f l u x d e t e c t o r s . The n e x t 1 a rge measurement program w i 11 i n v o l v e PuO /Th02.which i s p r e s e n t l y b e i n g f a b r i c a t e d a t CRNL and i s expected 6 t o e a v a i l a b l e e a r l y i n 1981.

D u r i n g t h e p a s t y e a r CRNL p a r t i c i p a t e d i n i in IAEA benchmark a c t i v i t y to i n t e r c o m p a r e measurements on s t a n d a r d gama-sources u s i n g Ge-Li d e t e c t o r s .

ANALYTICAL REACTOR PHYSICS

An approx imate method f o r s o l v i n g t h e B o l t m a n n e q u a t i o n i n t h e v i c i n i t y o f a p l a n e boundary between m a t e r i a l s w i t h d i f f e r e n t p r o p e r t i e s has been developed. A paper on t h i s t o p i c has been s u b m i t t e d t o t h e IAEAIENS s p e c i a l i s t s m e e t i n g i n A p r i l 1981.

The computer program most o f t e n used i n r e a c t o r p h y s i c s c a l c u l a t i o n s a t CRNL has been t h e c e l l code LATREP. T h i s program has been s u b s t a n t i a l l y r e v i s e d r e c e n t l y t o i n c l u d e a n u c l e a r da ta l i b r a r y , i n t h e W e s t c o t t fo rmal ism, based on ENDFIB. I n t h e process o f t e s t i n g t h i rogram, d e f i c i e n c i e s i n t h e ENDFIB c r o s s s e c t i o n s f o r L L I ~ ~ ~ and InT1! have been d i scovered .

ADVANCED FUEL CYCLES AND ASSESSMENT

The p o s s i b i l i t y o f u s i n g a once- through t h o r i u m c y c l e has been i n v e s t i g a t e d . T h i s concep t i n v o l v e s u s i n g s l i g h t l y e n r i c h e d uran ium " d r i v e r " f u e l i n c o n j u n c t i o n w i t h pu re t h o r i u m f u e l bund les . From t h e p o i n t o f v iew o f n e u t r o n ba lance and economics t h e c y c l e appears f e a s i b l e w i t h no c r e d i t t aken f o r t h e U-233 i n t h e spent f u e l . However, p r a c t i c a l i m p l i m e n t a t i o n r e q u i r e s f u r t h e r work on f u e l mariagement s t r a t e g i e s t o de te rm ine whether h e a t t r a n s f e r r e q u i r e m e n t s car1 be s a t i s f i e d .

Several concepts for advanced fuel cycles i n CANDU reactors proposed by A. Radkowski have been investigated and abandoned.

A study of the advantages of using s l i gh t ly enriched uranium in CANDU reactors has been completed. The optimum enrichment appears to be about 0.93%. A t t h i s enrichment, b u r n u p i s increased to about 15 MW.d/kg which reduces fuel consumption by about 25% and requires no change i n the reactor

* design. A t higher enrichments, fur ther improvements i n burnup and fuel consumption are possible b u t flux d i s tor t ions and hot spots begin to appear in the core and t h i s would probably require changes i n both the core and fuel design.

The concept of "spal la t ion breeding'' is being pursued a t CRNL and reactor physics calculations have been made for several accelerator-target- blanket combinations.

a A study of the s u i t a b i l i t y of several advanced fuel cycles for the Canadian s i tuat ion has been made. A moderate to ta l energy growth of 2.7%/ year was assumed together w i t h two growth ra tes of ins ta l led nuclear capacity i n the f i r s t 50 years of the next century. Thus annual growth ra tes are 3% and 4% between 2000 and 2050. I t i s expected tha t reactors operating on advanced cycles will not be introduced i n s ign i f ican t numbers before the year 2000. Over the period considered the P u / t h cycle conserves the most uranium, largely because i t uses the plutonium i n spent natural uranium fuel.

SMALL REACTOR DEVELOPMENT

A 2 MW reactor for low temperature heating ( 100•‹C) i s in the conceptual design stage. The core cons is t s of 200 CANDU-type fuel elements containing 5% enriched uranium oxide. Reactivity i s controlled by a motor- driven beryllium annulus surrounding the core. There are no other mechanical control devices.

Economic f e a s i b i l i t y a t such a low power level depends on achieving a high degree of inherent safety a t low cost , and eliminating the need for fu l l time sk i l led operators. These a t t r i b u t e s have already been demonstrated by the 20 kW research reactors over the past ten years. The 2 MW concept has similar inherent safety charac te r i s t ics , based on limited reac t iv i ty additions and a 1 arge negative void coeff ic ient . Theno- hydraulic experiments to study and demonstrate the inherent safety pr inciples are now i n progress. .

Ris@ Nat iona l Laboratory

Department of Reactor Technology

NEACRP-L-244 DENMARK

September 1980

Recent Reactor Phys ics A c t i v i t i e s i n Denmark

by

Hans Nel t rup

1. Fue l Box C a l c u l a t i o n wi th Response Ma t r i ce s

A program REPRO c a l c u l a t i n g t h e response ma t r ix f o r he te ro-

geneous square p i n c e l l s has been developed. The v a r i a b l e s

of t h e problem a r e t h e expansion components of t h e angula r

f l u x on t h e c e l l s u r f a c e .

The expansion, complete and independent f o r each h a l f space,

i s one of s e v e r a l p o s s i b l e f o r angula r f l u x e s which a r e sym-

me t r i c w i th r e s p e c t t o an xy-plane perpendicu la r t o t h e p lane

d e f i n i n g t h e two h a l f spaces .

The elements of t h e response m a t r i x r e p r e s e n t t h e coupl ing i n

s e v e r a l energy groups between i n - and outgoing components i n

a set o f p o i n t s on t h e c e l l s i d e s e.g. p o i n t s f o r g a u s s i a n

i n t e g r a t i o n a long each s i d e .

. C a l c u l a t i o n o f : t h e response m a t r i x i s performed w i t h c o l l i s i o n

p r o b a b i l i t i e s i n s i d e t h e ce l l which is rep re sen ted by a g r e a t

number o f subregions . The f l u x response i n t h e s e i s r e g i s t e r e d

a s e lements i n a r eg ion f l u x response mat r ix .

A four - fo ld symmetry reduces cons ide rab ly t h e necessary com-

a p u t a t i o n s and t h e number o f m a t r i x e lements t o be s t o r e d .

A second program FLUSO s o l v e s t h e e igen v a l u e equa t ions o f t h e

in - and outgoing components o f a r e c t a n g u l a r a r r a y of p i n c e l l s

w i t h s u i t a b l e (b lack or t r u l y r e f l e c t i n g ) boundar ies . . -- .- ..

The i n t e r n a l r eg ion f l u x e s i n a l l cells are found by o p e r a t i n g

t h e r eg ion f l u x response m a t r i c e s on t h e component e igen

v e c t o r .

C a l c u l a t i o n s w i t h two energy groups , two angu la r componenks

and up t o 10 g a u s s i a n p o i n t s p e r ce l l s i d e and 25 i n t e r n a l

r eg ions - 24 i n four - fo ld symmetry and one c e n t r a l c i r c u l a r

r eg ion - p e r c a l l have been performed w i t h r a p i d convergency

i n t h e FLUS0 r o u t i n e and y i e l d i n g reasonable f l u x d i s t r i b u , t i o n s . Unfor tuna te ly it is d i f f i c u l t t o f i n d comparable

measurements o r c a l c u l a t i o n s . However, t h e f l a t f l u x and t h e

- e x a c t c a l c u l a b l e keff and f a s t t o t h e thermal f l u x r a t i o i n

a t o t a l l y r e f l e c t e d homogenous s q u a r e c e l l is e x e l l e n t l y re-

produced, a l though t h i s ce l l i s n o t s p e c i a l l y s u i t e d f o r t h i s

c a l c u l a t i o n method. The f a s t f l u x is f l a t w i t h i n 0.5% from t h e

mean f l u x and t h e thermal w i t h i n 0.1%. The f a s t t o thermal -. .. i..:

f l u x r a t i o is c o r r e c t w i t h i n 0.08% and t h e ensu r ing keff w i t h i n 7.2

CS3 0.1%. (3

c-4 ~~ . i::.:r

2. Core Performance Eva lua t ion (core-s imula tor ) e ..f e\

Opera t iona l r e s t r i c t i o n s a r e imposed on l i g h t wate r r e a c t o r s

i n o r d e r t o avoid f u e l f a i l u r e s , o r a t l e a s t t o d imin i sh t h e

number of f a i l u r e s . A s such r e s t r i c t i o n s n e c e s s a r i l y r e s u l t i n

reduced power p roduc t ion from t h e r e a c t o r s they a r e u n d e s i r a b l e

from an economical p o i n t of view. Knowledge of t h e l o c a l power

ramps and t h e i r consequelices f o r t h e f u e l i s r e q u i r e d i n o r d e r

t o reduce t h e l e v e l of r e s t r i c t i o n s c o n s i s t e n t w i th s a f e t y

requi rement . : . . ., - . . .~

A comprehensive system f o r t h e c a l c u l a t i o n of t h e f a i l u r e proba-

b i l i t y f o r t h e i n d i v i d u a l f u e l r o d s throughout t h e r e a c t o r c o r e

i n l i g h t water r e a c t o r s i s be ing developed. The c a l c u l a t i o n a l

system i s s e t up a s a modular system. The modules t o be - inc luded

a r e : 3D-nodal n e u t r o n i c / h y d r a u l i c module f o r t h e c a l c u l a t i o n of

t h e 3D power d i s t r i b u t i o n (ANTI o r NOTAM), f u e l box module f o r

t h e c a l c u l a t i o n of homogenized c r o s s - s e o t i o n s f o r t h e i n d i v i d u a l

f u e l boxes (CDB), and f u e l r e l i a b i l i t y module f o r t h e c a l c u l a -

t i o n of t h e f a i l u r e p r o b a b i l i t y f o r t h e i n d i v i d u a l f u e l r o d s

(FRP). For t h e f u e l f a i l u r e c a l c u l a t i o n s , t h e power h i s t o r y f o r

each i n d i v i d u a l f u e l p i n i s r e q u i r e d ; a module f o r t h e c a l c u l a -

t i o n s of t h e s e h i s t o r i e s a r e l i k e w i s e t o be inc luded .

A s p a r t of t h e Ph.D. d i s s e r t a t i o n a number of methods of ca l cu -

l a t i n g t h e l o c a l p i n power i n a BWR has been i n v e s t i g a t e d .

1. A s a s imple approximat ion , t h e loca l . f l u x d i s t r i b u t i o n

found i n t h e f i r s t s t e p is renormali .zed; i n t h i s way t h e

assembly average power a g r e e s wi th the one ob ta ined from

t h e g l o b a l coarse-mesh s o l u t i o n .

2 . A more s o p h i s t i c a t e d method i s based on t h e modulat ion

model where t h e heterogeneous s o l u t i o n from t h e f i r s t s t e p

is m u l t i p l i e d w i t h a smooth f l u x - d i s t r i b u t i o n making use

of t h e boundary c o n d i t i o n s ob ta ined from t h e coarse-mesh

s o l u t i o n .

3 . A s u p e r p o s i t i o n of t h e two s o l u t i o n s , a s supposed t o t h e

modulat ion model, makes i s p o s s i b l e t o p r e s e r v e both t h e

average power and t h e e igenva lue from t h e g l o b a l coa r se -

mesh s o l u t i o n . T h i s method seems t o g i v e b e t t e r r e s u l t s

than t h a t u s ing t h e modulat ion model when t h e d i f f e r e n c e

between t h e two s o l u t i o n s can be regarded a s a sma l l p e r t u r -

b a t i o n of one of them.

Both t h e modulation and t he superposi t ion model a r e very sen-

s i t i v e with regard t o s t rong heterogeni t ies . To avoid some.of

t h e d i f f i c u l t i e s when deal ing with very heterogeneous regions,

a procedure based on t h e response matrix method has been

examined.

. 4 . When ca l cu l a t i ng t h e smooth f lux-d i s t f ibu t ion i n s ide t h e

f u e l boxes of a BWR t h e average fluxed and cu r r en t s a t t h e

boundaries a r e t r ans fe r red through t h e watergaps and im-

pressed d i r e c t l y on t h e homogenized f u e l region. In t h i s

way t h e inaccuracy involved when homogenizing very

a heterogeneous regions i s reduced.

A comparison i s made of t h e above mentioned four s t r a t e g i e s

f o r combining a heterogeneous box-solution and t he r e s u l t s from

the ove ra l l coarse-mesh solu t ion .

- . --. - - -- - - - - -

The inves t iga t ion shows t h a t t he b e s t way t o combine t h e t w o so lu t ions seems t o occur when t h e heterogeneous so lu t ion from

the box ca l cu l a t i on (with r e f l e c t i n g boundaries) is superposed

with a smooth f lux-d i s t r ibu t ion i n t h e homogenized f u e l region

of t h e f u e l box.

The t h r ee ca lc i i la t ion s t e p s then follow:

0 - A t f i r s t two sets of average c ro s s sec t ions are ca lcu la ted ,

one v a l i d f o r t h e whole f u e l box and t he o the r f o r t h e f u e l

region alone.

- I n a second s t e p t h e o v e r a l l f lux-d i s t r ibu t ion is found

from a coarse-mesh ca lcu la t ion .

- I n a t h i r d s t e p t he boundary condi t ions found i n t h e second

s t e p a r e t r ans fe r red through watergaps and con t ro l rods and

impressed d i r e c t l y on t h e f u e l region. A smooth f l ux -d i s t r i -

but ion i n t h e f u e l region of t h e box i s then ca lcu la ted and

t he so lu t ion obtained is superposed with t h e heterogeneous

so lu t ion from t h e f i r s t s t ep .

, BY use of t h i s method it i s assumed t h a t t h e l o c a l power ramps

t o be used l a t e r i n f u e l performance s t u d i e s can 5 e e s t ima ted

t o an accep tab le accuracy.

!

3 . Core s u r v e i l l a n c e

A programme f o r on-l ink s imu la t ion of t h e t h r e e dimensional

power distribution i n l i g h t wate r r e a c t o r s i s under c o n s t r u c t i o n .

The programme u s e s d e t e c t o r s i g n a l s an6 nuc lea r and o p e r a t i n g

d a t a a s i n p u t . I t i s assumed t h a t t h e d e t e c t o r s i g n a l is pro-

p o r t i o n a l t o t h e average power d e n s i t y i n t h e f o u r f u e l e l e -

ments surrounding t h e d e t e c t o r . Pseudo-s ignals a t each d e t e c t o r

l e v e l a r e c a l c u l a t e d by so lv ing a two-dimensional nodal equa t ion

i n which t h e f i s s i o n source o f each ins t rumented c e l l is norma-

l i z e d s o t h a t c a l c u l a t e d and measured power d e n s i t y agree .

Havlng determined d e t e c t o r s i g n a l s , pseudo o r r e a l , f o r each

c e l l i n t h e r e a c t o r , t h e a x i a l power d i s t r i b u t i o n of a l l c e l l s

i s c a l c u l a t e d by a d j u s t i n g r a d i a l i n t e r a c t i o n . F i n a l l y i n d i v i d u a l

segment powers a r e determined us ing a v o i d , exposure and c o n t r o l

rod dependent mismatch f a c t o r . I n doing s o a s imple thermo-

hydrau l i c model is a p p l i e d .

4 . ANTI

For t h e c a l c u l a t i o n of t r a n s i e n t s i n a PWR c o r e , t h e t h ree -

dimensional computer program ANTI wi th coupled n e u t r o n i c s and

thermal -hydrau l ics is under development, The program combines

t h e n e u t r o n i c s p a r t of t h e BWR program ANDYCAP wi th t h e sub- - channel h y d r a u l i c s program TINA. It is in tended f o r t r a n s i e n t s - where t h e s p a t i a l d i s t r i b u t i o n of power and c o o l a n t f low i n

t h e c o r e i s important , p a r t i c u l a r l y c a s e s where a l o c a l power

i n c r e a s e occurs . The s t eady s t a t e p a r t of t h e program is used .... .- . ,..~ i n connect ion wi th t h e Core-Sinulator work f o r c a l c u l a t i o n of . ,. . . ~,

, _.I

t h e o v e r a l l power d i s t r i b u t i o n . ...- * .-I . : ,.: .. c

.. , . ., The program i s now i n t h e running- in phase where t e s t i n g i s . ,

going on i n p a r a l l e l wi th mod i f i ca t ions and improvements. A 6:";

testcase, simulating a control rod ejection from a small reac-

tor core, has been calculated and is reported in Ris@-M-2209.

The report also contains a brief program description.

While such initial test calculations have demonstrated that

the ANTI program is able to carry out transient calculations

they are not very useful with regard to the verification of

the results. Therefore, calculations are needed either for

more realistic cases with comparison to measured data or, at

least, cases which can be compared to the results of other

computer programs.

A test case (also control rod ejection) which has previously

been calculated by the ANDYCAP program has been repeated by

ANTI. Rather large differences were found between the ANDYCAP

and the ANTI results, and the main reason seems to be the

different fuel rod models. The results indicate that it is

important to describe the heat conduction in the fuel rod

cladding, which is done in ANTI. In the ANDYCAP fuel rod model

the cladding is described as a simple resistance to the heat

transfer from fuel to coolant.

A more realistic study of the Westinghouse 3000 M W t reactor has

been initiated. So far only static calculations have been per-

formed using data from safety analysis reports. For verification

of ANTI, power shapes have been calculated with the finite

difference program TWODIM and successfully compared to Westing-

house results. ANTI is a nodal programme involving internodal

coupling parameters with a significant influence on the results.

However, for a given nodal configuration it is an easy task to find a set of parameters which results in an acceptable solution

for very different power shapes. For each nodal configuration

a new set of parameters should be found, otherwise serious errors

may be introduced. The study has been carried out with two

different nodal configurations of either one or four almost

cubic nodes per horizontal layer of a fuel assembly.

5. A Model for a Westinghouse PWR-P0we.r Plant

The old model PWR/PLASIM from 1975 has been revised and im-

proved for calculation of more severe transients. The model is

now developed in a similar way as BWR/PLASIM for the Barseback

plant and based upon data from the Westinghouse RESAR 31.

The main features are as follows:

The reactor is described in one dimension with 14 core nodes

for neutron kinetic and hydraulic calculation. The diffusion

equation is used with one energy group and prompt-jump approxi-

mation. Neutron cross sections are taken as functions of cool-

ant density, fuel temperature, control rod density and boron

concentration. Three groups of delayed neutrons are used and

six source groups for delayed heat release. The calculation of

delayed heat is done in a global manner disregarding the local

variation. The two cooling channels are used: a mean power

channel with calculation of coolant temperature and a hot

channel with calculation of both coolant temperature and void.

A fixed hot channel factor is used. The void in the mean power

channel is found from the hot channel void using a fixed

weighing factor.

The primary circuit has only one loop with steam generator,

pump and pressuriser. The propagation of temperature variation

is simulated with pure time delays for the tubes and pure

mixing in reactor and steam generator volumes and in the pump.

The heat transfer section in the steam generator is divided

into 3 nodes for the secondary side and 6 for the primary side.

The steam load circuit is not included in RESAR 31 so the model

for the turbine and feedwater heaters is only provisional with

one HP and one LP section for both turbine and feedwater heaters.

The description of the control circuit:s in RESAR 31 does not

give sufficient information for simulation, so only a very simple

control algorithm. with estimate? parameters have been used to

close the main control loops. -, ! .: ..; :... : -.~, .. , .* : , ; . . ' , ,., .,; . '.ye ,. .<.I . , ' :

- I

Before t h e model can be used f o r c a l c u l a t i o n o f t r a n s i e n t s

w i th a reasonable p r e c i s i o n a l o t o f d a t a must be provided,

n o t on ly f o r t h e steam and c o n t r o l c i r c u i t s , b u t a l s o f o r t h e

primary c i r c u i t and t h e r e a c t o r .

NKACRP-L-244. FINLAND

STATUS REPORT FOR TNE NEACRP 1 9 8 0

REACTOR PHYSICS ACTIVITIES I N FINLAND

R e s e a r c h a n d d e v e l o p m e n t i n a p p l i e d r e a c t o r p h y s i c s are

c o n c e n t r a t e d a t t h e T e c h n i c a l R e s e a r c h C e n t r e o f F i n l a n d

w h e r e most o f t h e s t u d i e s c a r r i e d o u t h a v e b e e n c o n n e c t e d

b o t h w i t h core f o l l o w a n d f u e l management c a l c u l a t i o n s and

w i t h t r a n s i e n t a n a l y s e s f o r L o v i i s a (PWR) a n d O l k i l u o t o

( B ! i R ) r e a c t o r s .

C e l l c a l c u l a t i o n s

T h e c e l l c a l c u l a t i o n p r o g r a m s have b e e n u s e d i n t h e

e v a l u a t i o n o f t h e n e u t r o n d o s e s a b o v e 0 . 4 MeV i n v a r i o u s

l o c a t i o n s i n s i d e t h e r e a c t o r p r e s s u r e v e s s e l w i t h

d i f f e r e n t core c o n f i g u r a t i o n s .

Work h a s s t a r t e d o n c e r t a i n c a l c u l a t i o n s c o n c e r n i n g t h e

s a f e t y o f s p e n t f u e l t r z n s p o r t , s p e c i a l l y o n c a l c u l a t i o n

c f t h e n e u t r o n d o s e r a t e o u t s i d e a d r y s p e n t f u e l

c o n t a i n e r a n d o n t h e c r i t i c a l i t y b e n c h m a r k c a l c u l a t i o n s

f o r s p e n t f u e l t r a n s p o r t c a s e s a g r e e d on a t a CSNI

w o r k s h o p i n May.

Core c a l c u l a t i o n s

I n t h e BWR f i e l d t h e c o d e d e v e l o p m e n t work h a s m a i n l y b e e n

r e s t r i c t e d t o m o d i f i c a t i o n s r e q u i r e d by t h e c h a n g e o f

c o m p u t e r s y s t e m a n d t o a u x i l i a r y c o d e s n e e d e d t o p r e p a r e

i n p u t d a t a f o r t h e c o r e s i m u l a t o r BOREAS. U t i l i z i n g two-

g r o u p d a t a c o m p u t e d b y t h e CASMO c o d e , a number o f s u r v e y

s t u d i e s h a v e b e e n made c o n c e r n i n g t h e f i r s t f o u r c y c l e s o f

t h e TVO I a n d t h e f i r s t two o f t h e TVO I1 r e a c t o r

i n c l u d i n g i n v e s t i g a t i o n s r e g a r d i n g d i f f e r e n t c y c l e l e n g t h s

and burnable absorber contents in the fuel. A separate

simulation study on the first year of operation of the

TVO I reactor has also been performed which has made

possible some comparisons between the computed results of

BOREAS and the real state of the reactor core represented

by power distributions obtained by combining the results

of measurements and calculations made by the power station

computer. These comparisons have, on the whole, turned out

quite satisfactory with differences of 1-2 %, typically,

in the horizontal power distributions. Vertically, the

differences vary more; sometimes the code gives very

accurate predictions, sometimes it is less successful.

However, these differences are, on an average, usually

smaller than 5 %.

The testing of the three-dimensional PWR-simulator

HEXBU-3D has continued with operating data of the second

and third cycles of the Loviisa 1 reactor. The

comparisons between calculations and measurements for the

first cycle have been reported in the NEACRP Meeting in

Paris, November 1979. Results for the two subsequent

cycles are also good and mostly similar to earlier

a comparisons. Due to a slight overestimation of the lenght

of the first cycle a correction factor for reactivity,

which adjusts the energy release per fission, was

introduced into the program. Reducing the energy release

by 1 % the simulation of operating history of the Loviisa

1 reactor gives the lenghts of the first three cycles with

an accuracy of about 2 %.

The treatment of thermal flux spatial transients between

neighbouring fuel assemblies has been modified in HEXBU-3D

to include an input coefficient for multiplying of the

transients. It was observed that, especially in the first

a n d s e c o n d c y c l e s o f t h e reactor, when f u e l e n r i c h m e n t

r a n g e d f r o m 1 . 6 % t o 3 . 6 8 , t h e power was s y s t e m a t i c a l l y

o v e r e s t i m a t e d i n a s s e ~ n b l i e s o f h i g h e r e n r i c h m e n t . The

r e d u c t i o n o.f t r a n s i e n t s d o e s i m p r o v e t h e s i t u a t i o n , b u t

e v e n a r e d u c t i o n o f 50 % w i l l n o t e n t i r e l y r emove t h e

d e v i a t i o n s f r o m m e a s u r e d a s s e m b l y p o w e r s . I t seems t h a t

a t l e a s t p a r t o f t h e d i s c r e p a n c y is c a u s e d by t h e

h o m o g e n i z a t i o n O F f u e l a s s e m b l i e s w h i c h i n a t w o - g r o u p

c a l c u l a t i o n c r e a t e s t o o b i g t r a n s i e n t s o f f l u x b e t w e e n

n e i y h b o t i r i n g a s s e m b l i e s . T h i s p r o b l e m is known f r o m

B N R - r e a c t o r s w h e r e , as i n W E R - 4 4 0 r e a c t o r s , w a t e r g a p s

a n d s h r o u d s s u r r o u n d i n g f u e l a s s e m b . L i e s make t h e s e

n e u t r o n i c a l l y i s o l a t e d f r o m e a c h o t h e r . The p r o b l e i n w i l l

b e s t u d i e d f u r t h e r .

Dynamic c a l c u l a t i o n s

T h e r e a c t o r d y n a m i c s p r o g r a m s TRAWA, TAPP a n d TRAB h a v e

b e e n u s e d i n t h e t r a n s i e n t a n a l y s e s o f d i f f e r e n t s i t u t i o n s

o n t h e L o v i i s a a n d O l k i l u o t o r e a c t o r s .

F o r t h e d y n a m i c c a l c u l a t i o n s a new p r o g r a m ODD h a s b e e n

made t o c r ea t e a x i a l l y o n e - d i m e n s i o n a l t w o - g r o u p

d i f f u s i o n p a r a m e t e r s a n d t h e r m o h y d r a u l i c f e e d b a c k

c o e f f i c i e n t s o n t h e b a s i s o f t h r e e - d i m e n s i o n a l s t a t i c

c a l c u l a t i o n s . N o w e v e n d e t a i l e d core m o d i f i c a t i o n s c a n b e

a c c o u n t e d i n t h e t r a n s i e n t a n a l y s e s . T h e r a d i a l

d i s t r i b u t i o n s o f t h e f a s t a n d t h e r m a l n e u t r o n f l u x e s , t h e

power d e n s i t y a n d t h e t h e r m o h y d r a u l i c v a r i a b l e s c a l c u l a t e d

by t h e c o a r s e mesh P W R - s i m u l a t o r HEXBU-3D a r e u t i l i z e d i n

t h e p r o g r a m ODD. T h e 1 - D g r o u p . c o n s t a n t s g i v e same

a v e r a g e a x i a l d i s t r i b u t i o n s as t h e t h r e e -

d i m e n s i o n a l c a l c u l a t i o n a n d a l so t h e c o r r e c t d y n a m i c a l

b e h a v i o r , i f t h e r a d i a l s h a p e s r e m a i n e s s e n t i a l l y

u n c h a n g e d .

In figure 1 the relative axial power distribution calculated by HEXBU-3D with 10 nodes axially and 349 nodes

radially in the end of fuel cycle is compared with the

stationary distribution calculated by TRAMA with 41 mesh points in 10 axial fuel regions. In spite of even the

dissimilar thermohydraulic models, the agreement is good,

the differences in the relative axial distributions are

usually smaller than 1 % and the maximum is below 2 %.

The development work, the goal of which is to eliminate

a the most of the existing thermohydraulic restrictions in

the presant dynamic programs, e.g. flow reversals, has

been continued.

The number of available thermohydraulic correlations

describing slip between phases and evaporation or

condensation has been increased in the dynamic programs.

The correlations are comprehensively compared and with

them it is possible to cover a wide variety of situations

in different reactor types.

D 0

0 I 1 I 1 I I I I 1 I I I--

OaO0 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0 -80 0.90 1 a00

H E I G H T F K R C T I O N F R O M B O T T O M O F C O R E

DRRW VTT/YOI 290280

- 37 - CCI4MISSARIAT A L'ENERGIE ATOMIQUE

Reactor Physics Act iv i t i es i n FRANCE

October 1979 - September 1980

23D NEACRP Meeting

September 22-26, 1980 - IDAHO

I. BOUCHARD - Ph. HAMMER

1 - GENERAL. - The f r ench nuclear programme

meeting i n October 1979 f i v e new PWR's g r i d : TRICASTIN 1 and 2, GRAVELINES 1

NEACRP-L - 244 FRANCE

September 1980

i s going on s a t i s f a c t o r i l y . Since t h e l a s t NEACRP have been s t a r t e d up and covpled t o t h e EDF and 2 and DAMPIERRE I . They a r e 920 tW(e) a s -

t h e two FESSENHEIM and fbu r BUGEY u n i t s a l r e a d y i n ope ra t ion and i s twenty more u n i t s under cons t ruc t ion . The f i r s t 1300 MW(e) u n i t i s expected t o s t a r t up i n 1984. The load f a c t o r s of ope ra t ing p l a n t s a r e r a t h e r b e t t e r than expected and t h e time between t h e f i r s t s t a r t up and t h e f u l l power o p e r a t i o n bas been cons iderably reduced.

The c o n s t r u c t i o n of SUPER PHENIX 1 i s going on according t o t h e expected time schedule. The v e s s e l i s a r r i v e d a t CREYS-MALVILLE by J u l y 1980 and i s now i n t h e r e a c t o r bui ld ing .

PHENIX is operated s a t i s f a c t o r i l y and has a l r e a d y de l ive red more than 7 b i l - l i o n s K W ~ .

To prepare a foreseen o r d e r of two 1500 MWe f a s t breeders (SUPER PHENIX 2) important s t u d i e s a r e i n progress f o r decreas ing t h e c o s t of such u n i t s .

. The EURODIF p l a n t a t TRICASTIN has reached an o p e r a t i o n l e v e l corresponding t o 6 Mi l l ions of UTS pe r year . More than 100 T of LWR f u e l have been reprocessed a t LA HAGUE dur ing t h e l a s t s i x months campaign.

2 - FAST REACTOR 1'IIYSICS

During the period elapsed between October 1979 and September 1980, the major points of the French Fast Reactor Physics programme have concerned :

. the KACINE programme which is devoted to the study of neutronic charnctcristics of commercial breeders and in particular to problems related to heterogeneous core concept ;

. experiments performed on PHENIX in order to determine the feed back coefficients (Doppler, power, temperature) ;

fuel cycle studies ;

neutron shielding studies ;

. development of experimental technics for critical experiments and for Power plant operation.

It must be underlined that most of the experimental programmes concerning the fast reactor physics and shielding are now prepared and performed in the framework of the CEA-CNEN-DEBENE cooperation.

The present status of the French studies in fast reactor physics and shielding areas are described in detail in t:he paper presented at the ANS Sun Valley meeting (1). Therefore onlya summary of thisstatus will be given here.

An increasi~ig effort is going on to use PHENTX operation results for testing and improving the multigroup data sets and design calculational methods for various neutronic parameters :

- The discrepancy between the reactivity loss per day calculated with the CARNAVAL pseudo fission-product and the measured one has been completely analyzed : this discrepancy (9 5 ) was due to the fact that the axial fuel dilatation was not fully taken into account in the reactivity loss calculations. Presently che predicted value of this reactivity loss

and the experimental value are consistent by less that 5 %. This confirms the validity of Lhe pseudo fission product capture multigroup cross section adjusteil wi~li the integral experiments (including ERMINE and the PROFIL irradiation in PIENIX).

- Measurements of the PHENIX residual power after a reactor shut down have been performed in order to check and improve the data and calculational methods used forpredicting the residual power . For times going from 0 t o e 7 0 hours after the reactor shut-down, the residual power is presently .

overestimated by 3210 - + 5 % by the design calculational method.

- The discrepancy between the calculated and measured compositions of the Pu included in the unloaded blanket subassemblies is presently investigated.

- Systematic measurements of the feedback coefficients (temperature, powcr, Doppler) have been undertaken to check and improve the safety analysis of fast breeders. A companion paper at this meeting gives the preliminary results obtained (2).

2.3 - CRITICAL FACILITIES

After the PRE-RACINE programme, mainly devoted to the physics study of the heterogeneous core concept and performed within the framework of the CEA-CNEN cooperation (3), the RACIKE programme has started on September 1979.It is performed in the framework of the CEA-CNEN-DEEENE cooperation on fast breeders and involves the use of fuel provided by the three partners (4).

Presently one investigates the reference configuration which reached criticality on the 24th of March 1980. This configuration includes a central fertile zone (15 cm radius) and one fertile ring (10 cm thick).

DEBENE

280 (platlets)

450 (platlets)

Pu (kg)

U235 (kg)

.

CEA

220 (rodlets)

750 (rodlets)

CNEN

370 (rodlets)

/

Soiiic preliminary results obtained up to now are presented at the Sun Valley ANS ~~icctin]: (1) : the coinplete results will be compared to the previous rcsul ts obtained tliirinj; the P - C I N E programme(concerningthe clcan core and the configuraLion with one central fertile zone) in order to check and improve the calculational methods for heterogeneous cores.

The JASON programme devoted to new shielding concepts for fast breeders ( 1 ) has started at the beginning of 1900. This programme aims at improving the PROPANE formulaire, devoted to fast breeder shielding design calculations, for :

. new materials such as BqC special steels including high contents of Idi, materials including hydrogen (such as ZrIi2) ;

. new shielding concepts (e.g. localized shields). 0 Preliminary results of this programme will be presented at the

f~recomi%l~E~CRP specialists'meeting on shielding (Paris, October 27-29, 1960).

In order to improve the stuctural material nuclear data fo the version V or the CARNAVAL cross section Set two specific experimental programnes have been undertaken on the RB2 (Ci?314 - BOLOGNA) and ERXINE (CEA - CADAlUCtIE) fast thermal coupled facilities.

Both experiments are of the k o o = 1 type and the investigated media have been selected in order to fit to commercial breeder spectra.

For ERMINE the six month programme a:Llows to study three media (5) :

OUlO : The reference medium basic cell includes one enrichided U02 (27 %) MASURCA rodlet and three natural uranium oxyde rodlets.

OAlO and ON10 : The two basic cells include one enriched U02 rodlet, one natural uranium oxyde rodlet and respectively two steel or two nickel rodlets.

The measurements performd concern :

- reaction rate ratios (fission chambers and detectors), - reactivities using the oscillation technique.

2. 4 - IRRADIATED 1'UI:LS The irradiated fuel analysis in progress concerns :

- fertile pins irradiated in the core 1 of PHENIX,

- U02-pu02 pins including a high content of higher plutonium isotopes .. (TRAPU experiment),

- PROFIL I1 experiment (mainly actinide sample irradiation).

2. 5 - THEORETICAL WORK - The major topics presently under study are :

- blanket calculational method development (6) - this development uses the results of a specific experimental programme (NEFERTITI) which started on TAPIR0 within the framework of a CEA-CNEN cooperation ;

- sensitivity code development for time dependant problems such as actinide and F.P build up (7) ;

- anisotropic.diffusion : the method developed for the treatment of the interface problemhas been extended to 2D problems (8) ;

- finite element method : this method is now applied to hexagonal 3D calculations and the corresponding code is being tested on a SUPERPHENIX type core.

2. 6 - DEVELOPMENT OF EXPERIMENTAL TECHNIQUES FOR CRITICAL FACILITIES AND POWER REACTORS

0 - The effort concerning the reactivity absolute measurement using the rod-drop technique is going on.

- Within the RACINE programme, systematic comparisons of the fission or U238 capture rate measurements performed with different technics are made. The same inter comparison is made for Y heating measurement using either different technics or different TLD detectors.

Thc 'development of t h e NEPTUNE system of codes and i t s a s s e s s m e n t a r e g o i n g on and d u r i n g t h e l a s t y e a r t h e main e f f o r c s were devo ted t o improve t h e sys tcm f o r p r a c t i c a l a p p l i c a t i o n s and Lo check i t on power r e a c t o r f o l l o w i n g ( 9 , 1 0 , 1 1 ) Exper imenta l s t u d i e s were concern ing t e m p e r a t u r e c o e f f i c i e n t s , s t o r a g e c r i t i c a l i t y and s p e n t f u e l a n a l y s e s .

3 .1 - T h e o r e t i c a l s t u d i e s

3 .1 .1 - -- Improvements of NEPTUNE

CRONOS, f o r t r a n s i e n t 3 D c a l c u l a t i o n s A new module named CROFOS h a s been added t o t h e sys tem i n o r d e r t o

p r o v i d e a n e u t r o n k i n e t i c s c a l c u l a t i o n with. t h e t h e r m o h y d r a u l i c f eedback t a k e n i n t o accoun t d u r i n g t h e t r a n s i e n t s . In CRONOS t h e same f i n i t e e l ement t r e a t m e n t a s i n ELECTRE, t h e p a r t of NEPTUNE d e s i g n e d f o r t h e 3 D power d i s - t r i b u t i o n c a l c u l a t i o n s , i s used f o r t h e s p a t i a l r e p r e s e n t a t i o n of t h e f l u x . The expans ion c o e f f i c i e n t s of t h e f l u x depend on t i m e , t h e y a r e r e p r e s e n t e d by a one-s tep s c h m e .

CRONOS a l l o w s t h e c a l c u l a t i o n s of any 3 D t r a n s i e n t problem, a c c o u n t i n g f o r c o u p l i n g between t h e sub-channels of t h e thermol iydraul ic r e p r e s e n t a t i o n . I t h a s been p r i m a r l y d e s i g n e d f o r computa t ion of rod e j e c t i o n e f f e c t i n PIJR.

Micro APOLI.0 model f o r t h e e v a l u a t i o n of f e e d b a c k e f f e c t s . For PILQs c a l c u l a t i o n s t h e feedback e f f e c t s between n e u t r o n i c s and

t h e r m o h p d r a u l i c s have t o be t aken i n t o accoun t . T h e r e f o r e t h e r e a c t o r code must c o n t a i n a s u b r o u t i n e a l l o w i n g t h e c a l c u l a t i o n of t h e macroscop ic c r o s s - s e c t i o n s v e r s u s t h e l o c a l t e m p e r a t u r e s . I n t h e c l a s s i c a l p r o c e d u r e p r e l i m i - n a r y t a b u l a t i o n s a r e made by t h e t r a n s p o r t code ( i . e , APOLLO ) and t h e s u b r o u t i n e performs o n l y a n i n t e r p o l a t i o n i n t h e s e t a b l e s . The drawback of t h i s p r o c e d u r e i s ' the g r e a t number of s . tandard APOLLO c a l c u l a t i o n s which a r e n e c e s s a r y ; even i f t h e g roup s t r u c t u r e i s s i m p l i f i e d ( e . g 30 g roups ) t h e runn ing c o s t i s q u i t e i m p o r t a n t .

The b a s i c i d e a of t h e Nicro-APOLLO .nodel i s t o r e p l a c e t h e i n t e r - p o l a t i o n s made i n t h e r e a c t o r code by a v e r y s i m p l i f i e d and t h e r e f o r e n o t e x p e n s i v e APOLLO c a l c u l a t i o n . The t e s t s we performed show t h a t t h e f o l l o - wing model shou ld b e s u f f i c i e n t : . Geometry : c o m p l e t e l y homogeneized . Spcctruni : 8 groups . Loca l c o n d i t i o n s : f o r t h e n u c l i d e s depending on t h e l o c a l c o n d i t i o n s t h e

t r e a t m e n t i s made by a m i x t u r e . The n u c l i d e whose c o n c e n t r a t i o n i s N i n c o n d i t i o n s C is r e p l a c e d by a m i x t u r e of t h e n u c l i d e 1 ( c o n c e n t r a t i o n N c o n d i t i o n s C ) and t h e n u c l i d e 2 ( c o n c e n t r a t i o n N c o n d i t i o n s C ) w i t h

1 '

N ] + N 2 = N I 2 ' 2

The Func t ion f c a n be chosen i n o r d e r t o make t h e model a s p r e c i s e a s possible.

I n t h i s new p r o c c d ~ i r e , o n l y a v c r y s m a l l number of s t a n d a r d APOLLO c a l c u - i d t i o n s woilld b c r u n ; t h e o u t p u t would be t h e 8 g roup l i b r a r y i n s t e a d of t i ~ c fcw group macroscopic c r o s s - s e c t i o n s ; a l ? t h e c a l c u l a t i o n s of t h e few group nlacl.oscopic c r o s s - s e c t i n n s would be niatle , i n a n a u t o m a t i c way, by tlic r c a c t o r code.

3.1.2 : Assessment of NEPTUNE

For 2 3 3 ~ and 2 3 2 ~ h q u a l i f i c a t i o n a sea rch of tendencies has becn performed from Rabcock and Wilcox Brookhaven and Oak Ridge benchmark e x p e r i m e n ~ s involv ing 2 3 2 ~ h and 2 1 2 ~ o r 2 3 5 ~ l a t t i c e s moderated by HZ0 o r D20 (12) . The b a s i c l i b r a r i e s a r e ENDFIB4 f o r 2 3 3 ~ and Der r i en ' s eva lua t ion (modified above 3 keV according t o de Saussu re ' s sugges t ions) f o r 2 3 2 ~ h .

No tcndency t o modify t h e 2 3 2 ~ h da ta has been obtained. For 2 3 3 ~ a decreas ing of 1 . I % of "of ( i . e . 0.9% lowcr than t h e recommended value of t h e Vienna Agency) i s suggested by t h i s research . The s l i g h t mo- d i f i c a t i o n s suggested f o r H20 age (+I .7%) and D20 age (-1.8%) a re consis- t e n t with the preceeding r e s u l t s . However we emphasize t h a t these benchmark e x p e r i w n t s a t our d i sposa l seem not s u f f i c i e n t t o draw p r e c i s e and d e f i - n i t i v e conclusions.

a.2 - Experimental s t u d i e s

3.2.1 : MINERVE ------- Apart from the ERMINE experiments ( see 2.3.3) t h e main programme

dur ing t h e l a s t year was concerning the Doppler e f f e c t . A new s e t of measurements has been r e a l i z e d t o complete the r e s u l t s obta ined i n t h e f i r s t h a i f of 1979. Some more U02 samples wi th d i f f e r e n t enrichments and a n uranium metal sample were o s c i l l a t e d i n t h e same l a t t i c e a s i n t h e previous campaign and more a c c u r a t e temperature de terminat ions were performed t o reduce the corresponding u n c e r t a i n t y . Pre l iminary r e s u l t s have bezn presented a t SUN VALLEY (13). The accuracy reached i s s a t i s f a c t o r y (- 5 %) and a n a l y s i s of the r e s u l t s o u t l i n e s mainly the problem of c r y s t a l l i n e biudings.

The nex t l i g h t water r e a c t o r programme i n MINERVE w i l l be devoted t o sample worth measurements by t h e o s c i l l a t i o n technique, mainly gadol i - nium samples and some thorium f u e l s imula t ions .

a 3.2.2:: EOLE --- The f i r s t campaign of measurements r e l a t e d t o t h e moderator tem-

p e r a t u r e c o e f f i c i e n t s took p l a c e from August 1979 t o March 1980. A s a l r e a d y ind ica t ed l a s t year the CREOLE loop was opera ted i n i t s nominal cond i t ions , i . e . 300•‹C and 120 ba r s . Four d i f f e r e n t loadings of t h e loop inc lud ing mixed oxide f u e l s were s tud ied . Some d e t a i l s a r e given i n a paper presented a t SUN VALLEY (1 3 )

The experimental accuracy i s s u f f i c i e n t t o g e t ve ry p r e c i s e v a l u e s of t h e temperature e f f e c t even tak ing account of t h e low r e a c t i v i t y we igh t of t h e t e s t zone, i . e . t h e l a t t i c e included i n t o t h e loop. The main problems a r i s e with t h e a n a l y s i s of t h e r e s u l t s which r e q u i r e s 2D (R, Z) t r a n s p o r t c a l c u l a t i o n s . Modif ica t ions of t h e loop temperature induced secondary e f f e c t s , such a s , changes ' i n the leakages, i n t h e s t r u c t u r e c a p t u r e s o r i n the spectrum i n t e r a c t i o n between the t e s t and d r i v e r zones, which a r e important a s compared to the main e f f e c t and thus r e q u i r e an a c c u r a t e c a l c u l a t i o n .

From the f i r s t r e s u l t s i t appears a gene ra l t rend t o havc d i r r p - r e n t discrepancies according t o the temperature range and on the contr i l ry q u i t c s i m i l a r d i sc repanc ie s f o r va r ious l a t t i c e s inc lud ing mixcd oxide f u e l

and b o r a t e d wnrcr s i t u a t i o n s . Some more t e s t s wi l l . b e performed by c a l - c u l a t i o n i n o r d e r t o e l i m i n a t e t h e r i s k of systematic e r r o r s i n t h e a n a l y s i s method. I n a d d i t i o n t h c s e r e s u l t s w i l l be u s e d , t o g e t h e r w i t h o t i ~ c r s coming from c r i t i c a l exper iments f o r t h e low t e m p e r a t u r e r a n g e and from power r e a c t o r s f o r t h e h i g h e r one, t o perform a s e a r c h of ten- dcncy and t r y L O s e p a r a t e among t h e v a r i o u s p o s s i b l e c a u s e s of d i s c r e - pancy.

Tak ing i n t o accoun t t h e t ime needed f o r t h e s e t h e o r e t i c a l works and 1-hc p r i o r i t y of CRISTO e x p e r i m e n t s , t h e n e x t p r o g r a m e u s i n g CREOLE, which was i n t e n d e d t o b r i n g more i n f o r m a t i o n s on uranium l a t t i c e s by v a r y i n g t h e p i t c h , t h e enr ichment and may be t h e geometry ( p l a t e f u e l s ) i s d e l a t e d t i l l t h e b e g i n n i n g of 1982.

A s soon a s t h e CRX0T.E f i r s t campaign h a s been completed a new s e t of CRISTO c x p e r i m e n ~ s was i n i t i a t e d i n t h e EOLE r e a c t o r . F i r s t two months were devoted t o CRISTO 111, a s t u d y of h i g h d e n s i t y l a t t i c e s w i t h gado l in ium i n t h e wa te r which aims a t a b e t t e r knowledge of m u l t i p l i c a - t i o n f a c t o r s i n p o i s o r e d t a n k s f o r r e p r o c e s s i n g p l a n t s . F o r t h i s s t u d y t h e t e s t zone has a c y l i n d r i c a l geometry and i n c l u d e s a b o u t 800 f u e l r o d s (O.D. =9.4 nun, p e l l e t d i a m e t e r = 8 nun, a c t i v e h e i g h t = 500 mm) loaded i n a v e r y conipact l a t t i c e w i t h a p i t c h of 9.6 mm c o r r e s p o n d i n g a p p r o x i m a t e l y t o t h e e x t e r n a l d i a m e t e r of t h e r o d s .

F i v e c o n f i g u r a t i o n s were s t u d i e d , d i f f e r i n g by t h e modera to r p o i s o n i n g :

Number Moderator p o i s o n i n g

no 0.8 g / l g a d o l i n i u m 1.8 g / l g a d o l i n i u m 2. g / S n a t u r a l boron 4. g / l n a t u r a l boron

Buck l ing measurements were performed i n A and B i n o r d e r t o g e t a b s o l u t e v a l u e s of t h e m u l t i p l i c a t i o n f a c t o r s and when t h e l a t t i c e s have a p p r o x i m a t e l y t h e same n e u t r o n b a l a n c e , t h a t i s B and D o r C and E, r e a c t i v i t y measurements a l l o w u s t o have r e l a t i v e v a l u e s of t h e m u l t i - p l i c a t i o n f a c t o r s . F I thermore s p e c t r a l i n d i c e s , m a i n l y t h e f i s s i o n r a t i o of 2 3 % ~ and '"U and t h e r a t i o of 2 3 8 ~ c a p t u r e t o 2 3 5 f i s s i o n , were measured i n each l a t t i c e . R e s u l t s a r e now b e i n g a n a l y s e d and w i l l b e used t o check t h e c r i t i c a l i t y c a l c u l a t i o n s .

I t i s expected t o pe r fo rm a new se t of CRISTO I11 e x p e r i m e n t s w i t h a d i f f e r e n t p i t c h , and t h u s a n h i g h e r m o d e r a t i o n r a t i o , i n 1981.

Now t h e CRISTO I1 campaign i s i n p r o g r e s s ( 1 f b ) . I t d i f f e r s from CRISTO I by t h e f a c t t h a t more compact s t o r a g e l a t t i c e s a r e s t u d i e d , t h e r e f o r e i n c l u d i n g s t r o n g e r a b s o r b e r s , b u t n o t f o r t h e p r a c t i c a l r e a - l i z a t i o n and t h e p r i n c i p l e of measurements whj ch remain t h e same (1 5 ) . T h i s programme i s in tended t o be completed by March 1981. F u r t h e r expe- r i n i e n t a l s t u d i e s i n t h i s f i c l d wi l l . be devocecl t o t h e low d e n s i t y mode- r a t o r c o n d i t i o n s i n d r y s t o r a g e s ("fog c o n d i t i o n s " ) b u t i t i s n o t y e t d c l i n e d how w i l l be t h e p r a c t i c a l r e a l i z a t i o n , use of low d e n s i t y po ly - e thylcnc? a s i n CRISTO I o r of iicavy w a t e r , t a k i n g i n t o a c c o u n t some d i f - f i c u l t i c 3 s botil i n t h e e x p c r i m e n t a l setxii lg-up and i n t h e t i i f o r e t i c a l arla- l y s i s of t h e r e s u l t s .

*. . . . . . . ! : : , L..! *,,(-, *-. . L:L]:.: ! . .,

The important progranimc which i s planned on TIHANGE ( a 15x15 I ,

type PWR) f u c l rods has been de laged , due t o the conjunct ion of t rans- p o r t a t i o n prol)lems and of m e t a l l u r g i c a l examination requirements . Never- t h e l e s s the samples f o r ana lyses a r e now being taken from t h e rods i n the SACLAY hot l a b s and we hope t o g e t the main r e s u l t s by t h e end of t h i s year . Po in t burn-up a r e ranging from 6000 t o 38000 MWd/T.

The SHBRWOOD experiment , a n i r r a d i a t i o n of pure i s o t o p e s d i - l u t e d i n uranium oxide p e l l e t s performed i n t h e MELUSINE r e a c t o r a t GRENOR1.E i n s i d e a 25 rods s imula t ion of a PWR f u e i assembly, h a s been completed a t the end of 1979 and a n a l y s i s of t h e samples took p lace i n the f i r s t h a l f of t h i s year . Real ized i n t h e framework of a n EEC con- t r a c t r e l a t e d t o the plutonium r e c y c l e t h i s experiment aims t o a n a c c u r a t e de terminat ion of sonw plutonium, americium and curium i s o t o p e cap tu res . The burn-up was low i n o r d e r t o l i m i t t h e secondary e f f e c t s , b u t thus r e q u i r i n g ve ry a c c u r a t e measurements which were performed us ing bo th mass spectrometry and a lpha spectrometry.

4 - OTHER TOPICS -

4-1 : Fuel c y c l e s t u d i e s ----- ----------- The s t u d i e s on uranium r e c y c l e i n PWR's were pursued. Experimental

checks of t h e 23% con ten t confirm t h e t h e o r e t i c a l l y p red ic t ed v a l u e of about 100 ppb f o r t h e r a t i o 2 3 2 ~ / 2 3 5 ~ i n a n uranium coming from rep rocess ing of LWR f u e l w i th 30000 Wd/T and a cool ing time of 3 years . The consequences f o r t h e conversion, enrichment and f a b r i c a t i o n p l a n t s a re be ing i n v e s t i g a t e d .

The s e l f s h i e l d i n g of 2 3 6 ~ c a p t u r e has t o be ' taken i n t o account f o r a n e s t ima t ion of the overenrichment r equ i red i n u s i n g recycled uranium and t h e plutonium pena l ty . A pre l iminary e v a l u a t i o n l eads t o a r e l a t i v e overenrich- men f 10 % f o r an i n i t i a l 23% c o n t e n t equal t o 1 % and t o a 5 % con ten t of ')'Pu i n t h e plutonium coming from a f u e l f a b r i c a t e d w i t h t h i s r e c c l ed uranium and burnt up t o 33000 MWd/T. T i l l a c o r r e c t t rea tment of t h e 23% cap- t u r e s e l f s h i e l d i n g be implemented i n NEPTUNE these f i g u r e s a r e on ly rough e s t ima tes . It i s planned t o check t h e i n t e g r a l 2% cap tu re c r o s s s e c t i o n i n t h e range of concen t ra t ion 0.5 % - 1.5 %, by r e a c t i v i t y and i r r a d i a t i o n measurements a s soon a s r e p r e s e n t a t i v e samples can be suppl ied .

4-2 : Reactor c o n t r o l and s u r v e i l l a n c e ................................ The t o p i c s mentioned i n the l a s t progress r e p o r t a r e be ing c a r r i e d

out . I n the f i e l d of no i se t echn ics an informal meeting h a s been h e l d a t

CADARACHE l a s t May with about e i g h t y p a r t i c i p a n t s from var ious coun t r i e s . A grea t emphasis has been put on p r a c t i c a l a p p l i c a t i o n s .

- REFERENCES - .-

I t . HIAMMER, " F a s t R e a c t o r P h y s i c s a t CEA, P r e s e n t s t u d i e s and F u t u r e P r o s p e c t s " ANS T o p i c a l Mceting on R e a c t o r P h y s i c s - SUN VALLEY (1980) .

J . C . GAU'HIIER, " R e a c ~ i v i t y Ba lance m e t e r : Feed Rack E f f e c t s Measurements a t PIIENIX". Communication t o t h i s Mee t ing ,

YH.BOUGET e t a l , " P h y s i c s S t u d i e s of Neutironic Problems r e l a t e d t o t h e he te rogeneous f a s t Reac to r c o r e concep t : Exper imenta l Progranune PRE- RACINE performcd on MASUIICA" - G a t l i n b u r g (1979) .

YH. BOURGET e t a l , "Main C h a r a c t e r i s t i c s of t h e RACINE Programme deve lo - ped by UI:BENE, CNEN and CEA on MASURCA f o r t h e Heterogeneous c o r e Con- c e p t S t u d i e s " . I A E A Symposium on F a s t R e a c t o r P h y s i c s - Aix e n Provence , 1979, IAEA-SM 244129

a M. DARROUZET e t a l , "Exper imenta l S tudy of S t r u c t u r a l M a t e r i a l s C a p t u r e i n F a s t Breeder S p e c t r a " - NEACRPIL

M. SALVATORES, "Blancket S t u d i e s a t CEA , S t a t u s of t h e A r t " - Communica- t i o n t o t h i s mee t ing .

J . C . ESTIOT, G . PALMIOTTI, M. SALVATORES, " A p p l i c a t i o n o f S e n s i t i v i t i e s S t u d i e s t o A c t i n i d e Build-up Problems" - Communication t o t h i s Mee t ing .

M. COSIMI e t a l , "A ~ e w A n i s o t r o p i c D i f f u s i o n Approximat ion t o T r e a t Hcte- r o g e n e i t i e s i n F a s t Power Reac to r s" , ANS T o p i c a l Meet ing on R e a c t o r P h y s i c s , SUN VALLEY (1980).

A. M V E N O K Y e t a l , "NEPTUNE : Les modules ELECTRF e t CRONOS pour l e s c a l c u l s s t a t i o n n a i r e s e t t r a n s i t o i r e s d e d i s t r i b u t i o n d e p u i s s a n c e a v e c p r i s e e n compte d e s c o n t r e - r g a c t i o n s thermohydraul iques" - NEACRP S p e c i a l i s t Meet ing on 3D C a l c u l a t i o n s , P m I s (1979). 0 B. NOEL, P.REUSS, " S u i v i d ' u n PWR d e 900 MNe" - NEACRP S p e c i a l i s t Meet ing on 3D C a l c u l a t i o n s , PARIS (1979) .

C. ANDRIEUX, J . KREBS, M. LEBARS, " C a l c u l s d e s u i v i 5 t r o i s d imens ions d ' u n r 6 a c t e u r PWR e x p e r i m e n t a l CAP 1800" - NEACRP S p e c k l i s t Meet ing on 3D Cal- c u l a t i o n s , PARIS (1979) .

H. TELLIER e t a l , "Uranium 233 - Thorium c y c l e : C r i t i c a l A n a l y s i s and Q u a l i f i c a t i o n of Nuclear Data and L i g h t Water R e a c t o r s F u e l Cycle S t u d i e s ' ' ANS T o p i c a l Meeting on Reac to r P h y s i c s , SUN VALLEY (1980) .

C. GOLINELLI e t a l , " ~ e m p c r a t u r e C o e f f i c i e n t and Doppler E f f e c t Measurements' ' ANS T o p i c a l Meeting on R e a c t o r P h y s i c s , SUN VALLEY (1980) .

J. BOUCHAKD e t a l , "some Recent S t u d i e s on L i g h t Water R e a c t o r Phy.?ics" - ANS T o p i c a l ~ c e t i n g on Reac to r P h y s i c s , SUN VALLEY (1980) .

A. DAItRAUD, C . GOLINEI.LI, Ph. MARSAULT, "CR'ISTO : Exper iment f o r F u e l S to - r ag? C r i t i c a l i t y E v a l u a t i o n s " , T r a n s . Am. Uucl. Soc. 31, 270 ( 1 9 7 9 ) . - .., ..., ., ,

: , , . , . . . , . *. , . . ,i i~,: . . ..: i!. .. .

NbACKP-L-244 F.R. G E W N Y

RECENT REACTOR PHYSICS AND SHIELDING INVESTIGATIONS IN THE FEDERAL REPUBLIC OF GERMANY

H. Kiisters

Nuclear Research Center Karlsruhe Institute for Neutron Physics and Reactor Technology

Postfach 3640, D-75CO Karlsruhe 1 Federal Republic of Germany

ABSTRACT

In the present paper reactor physics investigations done recently in 197911980 for light water reactors, high temperature reactors, and fast reactors are summarized. Shielding activities are briefly mentioned. Though only loosely related to reactor physics, some measurements on secondary neutron spectra in spallation reactions are reported.

1. PHYSICS INVESTIGATIONS FOR COMMERCIAL LIGHT WATER REACTORS

This chapter describes improved calculational methods, their validation in benchmark tests and their application to experiments performed in commercial PWRs and BWRs. Special problems in fuel cycle analysis are addressed, such as the treatment of burnable poisons, comparison of the nuclear characteristics of spent fuel with results from post-irradiation exoeriments. and the a~olication of isotooic correlation techniaues in -. safeguarding fissile material.

1.1. IMPROVED CALCULATIONAL TECHNIQUES

During the past few years substantial progress has been made in the development of efficient numerical methods for solving multi-dimensional neutron diffusion problems. At Kraftwerk Union (KWU), a coarse mesh nodal expansion method (NEM) has been successfully tested for two- and three- dimensional benchmark cases and also with experiments.l When the global NEM solution has converged, detailed local results as pin power distributions can be evaluated in consistency with the NEM solution by performing a high order polynomial expansion of the scalar flux in each node such that node-interface conditions for the neutron flux and the neutron current are f~lfilled.~ The accuracy obtained is completely satisfactory and the savings in computer time are remarkable when compared to a finite difference solution in three dimensions.

In order to describe off-nominal core conditions in LWRs, the code IQSBOX has been developed by KWJ3 to solve the two-group space- and time- dependent diffusion equation. The solution of the neutronic part is based

on NEX. It can be used t o s o l v e r e g u l a r and a d j o i n t e igenva lue problems and e x t e r n a l source problems. By u s i n g a f u l l y i m p l i c i t t ime d i f f e r e n c i n g scheme t h e s o l u t i o n p rocedure f o r t h e time--dep;?ndent problem i s about , t h e same a s f o r a s t a t i c s o u r c e problem. T r u n c a t i o ! ~ e r r o r s due t o t h e t empora l d i f f e r e n c i n g scheme a r e reduced by t h e we l l -knom e x p o n e n t l a l t r ans fo rma- t i o n t echn ique , Coarse mesh r e b a l a n c i n g is used t o a c c e l e r a t e convergence of t h e i t e r a t i v e s o l u t i o n p r o c e s s .

Two d i f f e r e n t the rmal -hydrau l i c modules can be coupled w i t h t h e neu- t r o n i c p a r t of IQSBOX. A t p r e s e n t t h e model used f o r PWR c a l c u l a t i o n s i s a s i m p l e c l o s e d channe l model which w i l l be r e p l a c e d by a more s o p h i s t i c a t e d model a l l o w i n g f o r c r o s s f low. For BWR a p p l i c a t i o n s a mul t i -channel the rmal - h y d r a u l i c model i s used. Th i s v e r s i o n of IQSBOX i s c a l l e d IQSBWR. The the rmal -hydrau l i c module of IQSBI.IR was q u a l i f i e d by t h e p o s t c a l c u l a t i o n of a subcoo l ing p e r t u r b a t i o n caused by a preheate;: t r i p and a mass f l o w t r a n - s i e n t induced by a pump t r i p . 4

I n o r d e r t o v a l i d a t e t h e performance of t h e s e dynamic codes , e x p e r i - ments have been performed a t l a r g e KWU r e a c t o r s by i n t r o d u c i n g n u c l e a r and therrnohydraul ic p e r t u r b a t i o n s . F u l l and s i n g l e ( o f f c e n t e r ) rod scrams were chosen t o v e r i f y t h e n e u t r o n i c c a p a b i l i t i e s of IQSBOX.

A s i n g l e rod d rop i s a s e v e r e t e s t f o r t h e n e u t r o n i c module of space- t ime k i n e t i c s codes , e s p e c i a l l y d u r i n g t h e f i r s t p a r t of t h e t r a n s i e n t c o v e r i n g t h e i n s e r t i o n of t h e c o n t r o l rod. Some r e s u l t s f o r t h i s t r a n s i e n t 5 a r e r e p r e s e n t e d i n Fig. 1. It shows t h e measured and c a l c u l a t e d a c t i v a t i o n s of t h e d e t e c t o r s nex t t o t h e dropping rod. A l l c u r v e s a r e normal ized t o one a t t h e beg inn ing of t h e t r a n s i e n t . The c u r v e s from t h e l e f t s i d e t o t h e r i g h t co r respond t o t h e sequence of d e t e c t o r s from top t o botrom. The d e t e c - t o r s i g n a l s demons t ra te t h e time-dependent inf l -uence of t h e dropping rod. The gap between t h e two groups of c u r v e s cor responds t o t h e d i s t a n c e between t h e upper and lower t r i p l e t of d e t e c t o r s . The c a l c u l a t e d and measured c u r v e s a g r e e f a i r l y w e l l f o r t h e upper d e t e c t o r s of b o t h groups . The maximum e r r o r is about 2 % of t h e i n i t i a l va lue . The o t h e r f o u r c u r v e s show l a r g e r d i s c r e p a n c i e s i n c r e a s i n g up t o 5 % of t h e i n i t i a l v a l u e s . The eva lua - t i o n of t h e f u l l scram r e s u l t s h a s shown t h a t t h e s e d e v i a t i o n s can p a r t l y be exp la ined by t h e s o c a l l e d z e r o d i sp lacement s h i f t of t h e ( n , D ) - d e t e c t o r s which a r e c a l i b r a t e d a t f u l l power. A f t e r t h e €!nd of t h e c o n t r o l rod movement a t 2.2 seconds the rmal -hydrau l i c feedback a n d , t o a l e s s e r e x t e n t , d e - l a y e d n e u t r o n holdback e f f e c t s de te rmine t h e c o u r s e of t h e t r a n s i e n t .

I n Fig. 2 t h e d e c r e a s e of t h e c o o l a n t o u t l e t t e m p e r a t u r e of t h e most a f f e c t e d f u e l e lement i s r e p r e s e n t e d . The agreement between measurement and c a l c u l a t i o n i s s a t i s f a c t o r y and i n d i c a t e s t h a t t h e use of a s imple c l o s e d channe l model i s s u f f i c i e n t f o r t h e d e s c r i p t i o c . of near-nominal c o n d i t i o n s . The a b i l i t y of IQSBOX t o f o l l o w t h e c o u r s e of 2. t r a n s i e n t caused by a s i n g l e rod drop conf i rms i t s f l e x i b i l i t y and accuracy .

A t t h e U n i v e r s i t y of S t u t t g a r t a d i f f e r e n t a - p o s t e r i o r i e v a l u a t i o n of t h e c o a r s e mesh r e s u l t s h a s been performed i n cmrder t o p r e d i c t t h e e f f e c t of a smal l c o n t r o l rod on t h e f l u x and power d i s t r i b u t i o n i n t h e env i ron- ment of t h e topmost end of t h e c o n t r o l r o d e 6 M t e r t h e convergence of t h e g l o b a l c o a r s e mesh s o l u t i o n of t h e d i f f u s i o n e q u a t i o n a new f i n e mesh i s imposed on t h e s p e c i f i e d r e g i o n . In d e a l i n g w i t h time-dependent problems

(e.g. control rod withdrawal), the discretized set of the reactor dynamics equations connects the values at the surface with the inner grid-points of the fine mesh, so that essentially a boundary value problem is to be solved at every time level.

The absorber is treated in a finite difference diffusion theory calculation with the standard technique of a logarithmic derivative boundary condition at the surface of the absorber.

In order to involve the strong dependence of the water density on the local power density into the analysis, the neutron diffusion calculation is coupled with a time-dependent thermohydraulic multi-channel analysis.

Experience has shown that mild heterogeneities caused by partially inserted small control rods ("fingers") in a LWR can be treated successfully with NEM. In a reactor with extended strong absorber regions or marked anisotropic neutron transport (large cavities in HTRs, water gaps), a treatment of these heterogeneities has to be performed more elaborately.

At the University of Stuttgart these macro heterogeneities have been investigated by splitting the whole reactor core into diffusion regions and transport regions.' The regions are coupled via currents along the surface between the transport and diffusion region. The inward directed currents are determined iteratively by applying response matrix theory, the response matrix for the transport region being pre-calculated. The inward current into the diffusion region is taken as an internal boundary condition for the next iteration step to solve the diffusion equation.

The method has been applied to describe a BWR fuel lattice with a cruciform control rod, which is treated together with the surrounding water gap as a transport region, while diffusion theory is used for the region of homogenized fuel rods. The results were compared to SN-solutions for the whole system. The agreement is satisfactory, the computing time being much less.

An unconventional approach to solve the homogenisation and condensation problem for a coarse mesh solution has been proposed by WU.* Consist- ent spatial homogenisation of the heterogeneous transport regions is done by a method, KWU called "equivalence theory".

Assuming that the exact global reactor solution is known, one can derive formally a set of homogenized differential equations whose solution preserves simultaneously all integral physics parameters of interest: keff. group reaction rates, group node-surface currents. Homogenized surface currents are related to the homogenized group scalar fluxes by a generalized Fick's law, in which the diffusion coefficients are assumed to be directionally dependent. These coefficients are treated as totally artifi- cial quantities which are determined by adjustment to the nodal leakage terms. Within this scheme, it is not possible to determine the appropriate interface relationship between nodal homogenized and heterogeneous surface fluxes. A solution of this problem is found by giving up the continuity condition for homogenized surface fluxes between adjacent nodes and introducing heterogeneity factors as the ratio of the heterogeneous surface fluxes to the homogenized surface fluxes. It is assumed that the hetero-

h : 9- fi :,*, :- I-' ,'? YULjL&j$ ,.J

..-

g e n e i t y f a c t o r s a r e e q u a l on both o p p o s i t e s u r f a c e s of a g iven node. The magnitude of t h e d i s c o n t i n u i t y i s t r e a t e d a s an a d d i t i o n a l hoinogenization parameter . By t h i s t r e a t m e n t a l l r e q u e s t e d i n t e g r a l q u a n t i t i e s , mentioned above, can be p rese rved .

The method h a s been a p p l i e d s u c c e s s f u l l y t o t h e well-known h e t e r o - geneous BWR-benchmark problem pos ted by Worley and Henrys9 A t y p i c a l d i s t r i - b u t i o n f o r t h e the rmal f l u x , averaged over t h e y - d i r e c t i o n of t h e assem- b l i e s a l o n g t h e main a x i s , i s shown i n Fig. 3 , t o g e t h e r w i t h t h e correspond- i n g h e t e r o g e n e i t y f a c t o r s ( d o t t e d l i n e ) . Only i n t h e r e f l e c t o r r e g i o n (peak on t h e r i g h t hand s i d e i n Fig. 3 ) both t h e hoioogeneous and he te rogeneous s o l u t i o n s c o i n c i d e approx imate ly .

Efeanwhile KIJU h a s t r e a t e d PWR configurat: .ons, d e a l i n g w i t h i n s e r t e d c o n t r o l r o d s , sh rouds and r e f l e c t o r r e g i o n s . The performance of t h i s "equ iva lence theory" on s t r o n g h e t e r o g e n e i t i e s i s v e r y encouraging. F u r t h e r improvements and r e s u l t s w i l l be p r e s e n t e d at t h e fo r thcoming 1981 ANSJENS- Conference on numer ica l methods i n Munich. lo

A coarse-mesh method h a s a l s o been developed a t KWU f o r t h e approx i - mate s o l u t i o n of t h e two-dimensional n e u t r o n t r a n s p o r t e q u a t i o n i n r ec tangu- l a r ( x , y ) I n c o n t r a s t t o c o l l i s i o n p r o b a b i l i t y methods w i t h p a r t i a l - c u r r e n t coup l ing a c r o s s t h e i n t e r f a c e s , s p a t i a l c o u p l i n g i s d e t e r - mined by a h i g h o r d e r approx imat ion of t h e f l u x based on a p a r t i a l l y i n t e - g r a t e d form of t h e t r a n s p o r t e q u a t i o n r e t h e r than on i n t e g r a l t r a n s p o r t t h e o r y i n a low o r d e r approx imat ion . Various sample problems have been t e s t e d and were compared t o c o n v e n t i o n a l 2d-SN methods. The r e s u l t s a r e encourag ing both w i t h r e s p e c t t o accuracy and computing time. In s e c t i o n 4.2 t h e method i s a p p l i e d t o a f a s t r e a c t o r benchmark.

1.2. BURN-UP PHYSICS ASPECTS OF LIGHT WATER REACTORS

For h i g h f u e l burn-up, r e a c t i v i t y compens.?tion a t s t a r t - u p can be a c h i e v e d by t h e use of burnab le po i sons such a s Gd o r Hf. In a d d i t i o n , f o r t h e d e s i g n work of t h e German n u c l e a r s h i p r e a c t o r OTTO HAHN t h e r e q u i r e - ment t o permanent r e a c t o r shutdown by a c o n t r o l rod sys tem wi thou t t h e use of s o l u b l e po i son had t o be f u l f i l l e d .

A t t h e " G e s e l l s c h a f t f i i r Kernenergieverwer tung i n S c h i f f b a u und S c h i f f a h r t " (GKSS) a t Geesthacht n e a r Hamburg e x t e n s i v e s t u d i e s on t h i s t o p i c have been c a r r i e d o u t f o r s e v e r a l y e a r s . S p e c i a l l y developed theore - t i c a l models i n SN-approximation w i t h a c c u r a t e t r e a t m e n t of t h e space-t ime behaviour of po i sons and f u e l have been v e r i f i e d b exper iments a t t h e Gees thach t r e s e a r c h r e a c t o r and t h e s h i p r e a c t o r . 1 3 In t h e second c o r e of t h e s h i p r e a c t o r a r e a c t i v i t y e q u i v a l e n t of abou t 10 % A f h a d t o be compen- s a t e d by burnab le po i son r o d s . For t h i s purpose Z r B 2 i n Zr02 p e l l e t s was used. By measur ing t h e c o n t r o l rod p o s i t i o n s a t f u l l power a s a f u n c t i o n of burn-up, t h e p r e d i c t e d r e a c t i v i t y e f f e c t of t h e burnab le po i sons can be t e s t e d . Fig. 4 shows t h e comparison between c a l c u l a t e d and measured c o n t r o l rod p o s i t i o n s u n t i l decommission of t h e n u c l e a r s h i p . E x c e l l e n t agreement i s o b t a i n e d .

For t h e p r e d i c t i o n of t h e ou t -o f -p i l e n u c l e a r c h a r a c t e r i s t i c s of s p e n t f u e l such a s a c t i v i t y , h e a t g e n e r a t i o n , neu t ron and gamma r a d i a t i o n , and

criticality in compact reactor storage facilities or in interim storage ponds outside the reactor plant, in transport flasks and in reprocessing units a great many nuclides have to be considered over long periods of

13 time.

It is essential that e.g. the inventory of all the nuclides can be predicted with sufficient accuracy at each stage of the fuel cycle. To achieve this goal, the in-pile description especially of the time behaviour of the fuel and of the fission products has to take into account (space- dependent) spectral changes during irradiation and, for some nuclei, mutual shielding in s ace and energy especially for the low lying resonances of heavy nuclei (t/39Pu, 2 4 0 ~ ~ , 2 4 1 ~ ~ , 241Am) in LWRs. 241Am is practically . shielded by 239~u and 2 4 1 ~ ~ around 0.3 eV and by 2 4 0 ~ ~ around 1 eV. There- fore the build-up of those isotopes, which lead to intense neutron emitters as 242~m and 244Cm, and 238~u (as major isotope for radiolysis in reprocessing), can be destinctively influenced.

In burn-up calculations usually only the major constituents of the reactor fuel, which influence the reactivity balance are treated. However, for the description of the out-of-pile fuel characteristics it is essential to include all nuclides of interest (more than 1000) already in the in-pile burn-up calculations, though they may not affect the neutron spectrum.

At Karlsruhe, a code system is being established by couplin a modified version of the ORNL-ORIGEN code14, which is called KORIGEN15 within the KfK-code system K A P R O S ~ ~ both to the nuclear data file KEDAK~? and to multidimensional diffusion or transport codes. Part of the system is working for fast reactor fuel investigations1*, the development for LWRs is underway. Additional code packages to treat the various paths of nuclides in the fuel cycle including waste storage will be integrated.

In order to validate the KfK data and methods for out-of-pile investigations extensive comparisons have been performed with results from postirradiation investigations on PWR and BWR fuel. Because the power density in time for an irradiated fuel subassembly was given experimentally, only cell burn-up calculations had to be performed. This was done within a coupled system of the modified version of HAMMER and KORIGEN, called HAYKOR.~~ Cross-sections were taken from KEDAK as far as available, and from other sources (mainly ENDFIB IV). As an example, Fig. 5 compares results between theory and experiments for the analysis of batches from the PWR Obrigheim (KWO) plant.20 The obtained agreement is quite satisfactory.

As already mentioned previouslyz1, the experimental uncertainty especially in the difficult measurements for e.g. 241Am and 242Cm has to be assessed more reliably. Often experimental results show discrepancies in interlaboratory comparisons by about a factor of two. To test the reliability of the theoretical predictions for 241Am the capture data in the resonance range, recently evaluated for KEDAK~~, have been compared to the measured resonance integralz3; excellent agreement was found. Sensitivity checks on a few methodical uncertainties indicate that some of the experimental values for 241~m concentrations after irradiation, must be in error.

The theoretical decay heat generation has also been compared with experiments; impressive agreement is found. 20

comes from t h e v e s s e l l o a d i n g problem. The u n c e r t a i n t i e s p r e s e n t l y g iven rn the h i g h energy p a r t of t h e neu t ron f i s s i o n spect rum a s w e l l a s f o r t h e i n e l a s t i c c r o s s - s e c t i o n s of s t a i n l e s s s t e e l and f o r some f o i l a c t i v a t i o n d a t a do not meet t h e h igh s a f e t y requ i rements of p r e s s u r e v e s s e l l o a d i n g .

R e c a l c u l a t i o n s of n e u t r o n s p e c t r a on t h e b a s i s of t h e l a t e s t EURLIE-I?- w l t i g r o u p l i b r a r y were performed by KWU/Erlac.gen i n c o o p e r a t i o n w i t h IKE/ s t u t t g a r t , and t h e f a s t n e u t r o n f l u e n c e was rede te rmined w i t h h i g h p r e c i - ; ion a t t h e p o s i t i o n s of t h e s u r v e i l l a n c e c a p s u l e s a s w e l l a s a t l o c a l : .d in ts i n t h e p r e s s u r e v e s s e l of LWRs. S e n s i t i v i t y s t u d i e s and e r r o r . i : ~ a l y s i s were done f o r t h e n u c l e a r d a t a a p p l i e d , One- and two-dimensional ;s-codes were used. There i s a s t r o n g i n c i t e m e n t t o improve t h e p o s s i - - i l i t i e s of th ree -d imens iona l c a l c u l a t i o n s , p r i m a r i l y t o g e t t h e n e u t r o n i x c t r u m of the a c t i v a t i o n f o i l s i n t h e s u r v e i l l a n c e c a p s u l e s bu t a l s o f o r .?11 l o c a l c o r r e l a t i o n s of t h e f a s t n e u t r o n f l u x and i t s spect rum between :,:.e p o s i t i o n of t h e s u r v e i l l a n c e c a p s u l e s and t h e d i f f e r e n t p o i n t s of l n t e r e s t i n t h e p r e s s u r e v e s s e l .

Within t h e German r e s e a r c h p r o j e c t , " s t r u c t u r a l i n t e g r i t y of r e a c t o r ~ c ~ ~ o n e n t s " , p r e c i s e neu t ron spect rum c a l c u l a t i o n s were performed i n t h e i r r a d i a t i o n r i g s of t h e m a t e r i a l t e s t i n g r e a c t o r s D I D O and ERLIN of KFA- .;Clich and GKSS-Geesthacht a s w e l l a s f o r t h e power r e a c t o r VAK a t Kahl, t i t e re v a r i o u s t e s t specimens of p r e s s u r e v e s s e l m a t e r i a l a r e i r r a d i a t e d $1:32r d i f f e r e n t n e u t r o n environments bu t o t h e r v i s e i d e n t i c a l c o n d i t i o n s . 7 , e c a l c u l a t e d s p e c t r a a r e checked by f o i l d e t e c t o r measurements. The s p e c t r a a r e used t o de te rmine t h e l o c a l neu t ron f l u e n c e and s e r v e a s b a s i c in f3 rmat ion f o r damage f u n c t i o n unfo ld ing w i t h d e t a i l e d e r r o r a n a l y s i s .

S ince t h e v a r i a t i o n of t h e neu t ron spect rum a v a i l a b l e f o r i r r a d i a t i o n c c s t i is somehow r e s t r i c t e d , because most of t h e damage i s produced by neu- t r o a ~ i n t h e upper keV- o r lower MeV-energy r e g i o n s , a b e t t e r guess func- t i t > ? is needed f o r h i g h e r and lower neu t ron e n e r g i e s , t o compensate f o r t h e I n s t ~ f f i c i e n c y of t h e u n f o l d i n g procedure . Normally one s t a r t s w i t h t h e ;ito:,.ic d i sp lacement c r o s s - s e c t i o n a s a measure of t h e pr imary damage pro- c c s ~ ~ : i . A procedure h a s been developed t o improve t h e guess f u n c t i o n by U 1 f f r : r e n t i a t i n g t h e pr imary damage p r o c e s s e s a c c o r d i n g t o d i f f e r e n t cascade encr;:ies. The cascade energy p r o v i d e s a b e t t e r c o r r e l a t i o n b a s i s t o t h e co:n;,ilcated a n n e a l i n g p r o c e s s e s t h a n t h e n e u t r o n energy does. With t h i s unf ' , ld ing procedure i t h a s been p o s s i b l e t o t r a n s f e r t h e compl ica ted resonance s t r u c t u r e of t h e i r o n c r o s s - s e c t i o n s t o t h e damage f u n c t i o n s of pre!jr;ure v e s s e l s t e e l i n f u l l d e t a i l .

2. PHYSICS INVESTIGATIONS FOR ADVANCED LIGHT WATER REACTORS

In t h i s c h a p t e r , v a r i o u s concep t s f o r advanced LWRs a r e d i s c u s s e d . Expcrf~:nce i n commercial power p l a n t s h a s i n d i c a t e d t h a t LWR f u e l may be i r r a d i ; ~ t e d t o h i g h e r burn-up above abou t 40 GWd/t. Besides t h e e x p e r i m e n t a l P I t o demons t ra te t h e f u e l behav iour ~ ~ n d e r h i g h burn-up c o n d i t i o n s ,

slim has performed many t e s t s i n r e c y c l i n g Pu i n German r e a c t o r s . J o i n t l y KfK-Karlsruhe and KWU, suppor ted by t h e T e c h n i c a l U n i v e r s i t y Rraunschweig ( T U B ) , a r e s t u d y i n g t h e f e a s i b i l i t y of h i g h c o n v e r t i n g PWR- concepts .

When the capabilities of the YJK data and methods had been checked, further important topics could be investigated from a stronger theoretical basis. . .

For some time it has been questioned whether there exists a possibility to support effectively by theory the more or less empirical deductions on the control of fissile material in the fuel cycle. An acceptable theoretical procedure has to fulfill the following conditions:

(1 ) It has to be shown, that experimental errors or booking errors can be identified without repeating the measurement.

+ (2) If, prior to dissolution, an incorrect subassembly (having, for example, different burn-up and enrichment characteristics) is acci- dentally (or deliberately) included in a batch it is important that the error should be detectable. Moreover, it should also be possible to identify reliably the nature of the unwanted subassembly (e.g. its

a enrichment and burn-up).

(3) If requirements (1) and (2) above can be fulfilled and the predictions verified experimentally then it seems possible to predict an inten- tional diversion of fuel out of the fuel cycle with some degree of confidence.

Clearly only those fuel diversions, which have a measurable effect on parameters, chosen for identification, can be located. For instance, pellet- wise accumulation of fresh fuel from a fabrication plant over longer time intervals can not be detected by measurements or theory in the process stages of the fuel cycle. This can be achieved only by other procedures.

At K f K it was looked at the possibility that the technique of isotopic correlations (i.e. a systematic analysis of many ratios of certain nuclides) could be such a tool. For that reason sensitivity calculations for a wide variety of isotopic ratios on e.g. enrichment, burn-up, boron concentration, water density etc. have been performed to find out which ratios are sensitive or insensitive to a given parameter. A recipe is now being worked out to use this information effectively according to the above mentioned condition^.^^ A systematic analysis of these isotopic ratios could indeed successfully demonstrate the capability of this method in a realistic situation, i.e. the theoretical conclusions are in agreement with the results of repeated measurements.

1.3. SOME REMARKS ON SHIELDING ACTIVITIES FOR LIGHT WATER REACTORS

Most of the shielding work in Germany is done by industry and at the Universities of Hannover and Stuttgart. Here only brief consideration is given mainly to the work associated with the University of ~ t u t t ~ a r t . ~ ~

A systematic study on requirements of nuclear data has been completed for a BWR-power plant with 1300 laJel. This, together with a previous investigation of a shielding benchmark problem for a PWR power plant, has enabled a priority list of data improvements needed for LWR shielding to be established. The uncertainty and error analysis of target quantities of interest showed, that for some reactors the most stringent data requirement

2.1. PLUTONIUi.1 KECYCLiNG I X LIGHT WATER REACTORS

The c o n s i d e r a t i o n s g iven h e r e a r e based on r e c e n t p u b l i c a t i o n s by KiJU, 26

Small s c a l e Pu r e c y c l i n g was i n i t i a t e d b:7 t h e i n s e r t i o n of mixed U- and l'u-oxide f u e l (MOX) i n t o t h e s m a l l p r o t o t y p e r e a c t o r i.n Kahl (VAX) a s e a r l y a s 1966. In 1970 and 1972 t e s t a s s e m b l i e s had been i n s e r t e d i n t h e r e a c t o r s i n Lingen (KIJL) and KarZsruhe (MZFR). Large s c a l e Pu r e c y c l i n g i n t h e commercial German LWRs a t Obrigheim (KWO) and Gundremmingen (XRB) s t a r t e d i n 1972 and 1974. Up t o now more t h a n 200 MOX f u e l a s s e m b l i e s (MOX- FA^) w i t h a l t o g e t h e r abou t 10 000 I'u-bearing r o d s des igned and s p e c i f i e d by KWU and f a b r i c a t e d by ALKEM have been suppl.ied t o German r e a c t o r s . S u f f i - c i e n t r e a c t o r performance e x p e r i e n c e has been ga ined t o d e s i g n and t o war ran t l a r g e s c a l e use of r e c y c l e f u e l i n conmerc ia l TJdR c o r e s .

The d e s i g n s t u d i e s cover a r a t h e r wide range of models f o r t h e Pu r e - u s e , f i r s t of a l l t h e s o - c a l l e d s e l f - g e n e r a t e d r e c y c l i n g mode (SGR). I n t h i s c a s e i t i s assumed t h a t on ly t h e p r e v i o u s l y g e n e r a t e d Pu of t h e same power s t a t i o n w i l l be r e c y c l e d w i t h a d e l a y f o r r e p r o c e s s i n g and r e f a b r i c a - t i o n of two o r t h r e e y e a r s , which probably h a s t o be i n c r e a s e d . The e a r l i e s t i n s e r t i o n of Pu c o n t a i n i n g a s s e m b l i e s cou ld s t a r t i n t h e 4 t h o r 5 t h cyc le . The amount of MOX p i n s i n c r e a s e s from 15 % t o abou t 30 % p e r r e l o a d d u r i n g a few c y c l e s .

It was found t h a t t h e economy, o p e r a t i o n .and s a f e g u a r d i n g i s improved i f t h e Pu bred i n a number of r e a c t o r s i s burned on ly i n some of them. Th i s "Open market r e c y c l i n g " (OMR) may be accomplished i n t h r e e ways:

(1) a c o n s t a n t number of MOX-FAs p e r r e l o a d i n an amount e q u i v a l e n t t o approx. SGR, s t a r t i n g w i t h t h e f i r s t r e l o a d ;

(2) a c o n s t a n t MOX-FA r e l o a d w i t h an e s s e n t i a l l y l a r g e r number than i n t h e l a t t e r c a s e ;

( 3 ) a thermal Pu-burner w i t h o u t U-FAs.

Because of t h e c o s t i n c r e a s e a s s o c i a t e d w j t h t h e f a b r i c a t i o n of MOX f u e l , t h e r e i s an i n c e n t i v e t o c o n c e n t r a t e t h e Pu-content i n p a r t of t h e r e l o a d f u e l and t o s e l e c t Pu c o n c e n t r a t i o n s a s h i g h a s p o s s i b l e under t h e g i v e n power peaking l i m i t a t i o n s . Thus, i n t h e SGR mode, o n l y 114 t o 113 of a l l f u e l rods c o n t a i n plutonium. This governs t h e s i z e of t h e c e n t r a l MOX- r e g i o n i n t h e c a s e of t h e i s l a n d type f u e l a s s e m b l i e s where e v e r y FA con- t a i n s a few MOX-rods. I n c a s e of MOX-assemblies where every p i n c o n t a i n s Pu (a l l -Pu d e s i g n ) t h e r e l o a d b a t c h c o n s i s t s of a p p r o p r i a t e f r a c t i o n s of U- and HOX-FAs .

The n u c l e a r c h a r a c t e r i s t i c s t h a t a r e i m p o r t a n t w i t h r e s p e c t t o s a f e t y and performance of an LWR f u e l e d w i t h mixed ox ide i n d i s t i n c t a r e a s i n c l u d e power peaking e f f e c t s , c o n t r o l r equ i rements , c o n t r o l rod wor ths and t r a n -

s i e n t behav io r d u r i n g a n t i c i p a t e d o p e r a t i o n a l o c c u r r e n c e s and p o s t u l a t e d a c c i d e n t s . Ln p a r t i c u l a r , Pu b e a r i n g f u e l r o d s and a s s e m b l i e s have t o uieet t h e same t l iermohydraul ic , t l iermal and mechanical d e s i g n l i m i t s as uranium f u e l . In a d d i t i o n , MOX-FAs shou ld p rov ide a r e a c t i v i t y e q u i v a l e n t t o t h a t of uranium FAs and shou ld a c h i e v e a d i s c h a r g e burnup n o t lower than U-Fhs,

Fuel elements with a MOX-island have emerged for the BWR but the concept of the all-Pu design can also be applied under special conditions. It was found that the use of an all-Pu design is always advantageous in PWRs. Flattening of the power distribution is achieved by employing graded Pu concentrations. Lower Pu content is applied in those MOX fuel pins adjacent to fuel regions enriched by U235 or to water gaps.

In order to verify and to improve the accuracy of calculational methods for MOX-FA and Pu-recycle core design, hot critical experiments have been carried out by AB Atomenergi (~tudsvik) under a KWU contract in the critical facility KRITZ. They were supplemented by power reactor neutron physics experiments at KWO and KRB. .

Theoretical investigations led to the conclusion that only fresh MOX- FAs in rod cluster control (RCC) positions of PWRs would result in a significant loss of control rod worth, whereas once or twice burnt Pu assemblies can be located at any core position including RCC positions, without marked deterioration of control rod worth. This theoretical expectation has been verified experimentally by the KRITZ experiments and at KWO.

Measurements have been performed during start-up after refuelling at KRB. These demonstrated that the reduction of control rod worth remains within the allowable range (if only one MOX-FA of the all-Pu-type is introduced per control rod cell). Furthermore, no indication of any systematic change in reactivity coefficients of the overall core has been found.

Isotopic analyses have been carried out on several KWO MOX-fuel samples with an initial enrichment of 2 % and 3.2 % mass fraction Pufiss covering a range of burnups up to 40 GWd/t. The comparison of predicted and measured Pu higher isotopic compositions shows very satisfactory agreement including measured and predicted local burnup.

Extensive operating and performance experience has been gained during the demonstration programs for plutonium recycling in light-water reactors. The operational performance of the Pu-bearing fuel has been very satisfactory and is almost comparable to U02 fuel. The success of the mechanical and neutronic design of MOX fuel and fuel assemblies is demonstrated by neutronic core follow calculations, in-core measurements and radiochemical analyses.

The demonstration programs furnish sufficient evidence that large scale use of MOX fuel is technically feasible in large commercial LWR cores. Shuffling proposals exist for all recycling options of SGR and OMR satisfying the restrictions on maximum local power including the fine structure power shape. The nuclear characteristics of a PWR Pu-burner core, designed for economy and high conversion, do not differ essentially from those of recycle cores.

The following Table I summarizes design data for typical mixed oxide fuel assemblies.

Fuel eleinents w i t h a MOX-island have eiiierged f o r t h e BWR b u t t h e con- c e p t of t h e a l l - P u d e s i g n can a l s o be a p p l i e d under s p e c i a l c o n d i t i o n s , It was Found t h a t t h e use of an a l l -Pu d e s i g n i s always advaniageous i n PWRs. F l a t t e n i n g of t h e power d i s t r i b u t i o n is ach ieved by employing graded Pi1 con- c e n t r a t i o n s , Lower Pu c o n t e n t i s a p p l i e d i n t h o s e MOX f u e l p i n s a d j a c e n t t o f u e l r e g i o n s e n r i c h e d by U235 o r t o w a t e r gaps.

In o r d e r t o v e r i f y and t o improve t h e accuracy of c a l c u l a t i o n a l methods f o r NOX-FA and Pu-recycle c o r e d e s i g n , ho t c r i t i c a l exper iments have been c a r r i e d ou t by AB Atomenergi ( S t u d s v i k ) under a h?.JU c o n t r a c t i n t h e c r i t i c a l f a c i l i t y KRITZ. They were supplemented by power r e a c t o r n e u t r o n phys ics exper iments a t KNO and KRB,

l h e o r e t i c a l i n v e s t i g a t i o n s l e d t o t h e c o n c l u s i o n t h a t on ly f r e s h MOX- FAs i n rod c l u s t e r c o n t r o l (RCC) p o s i t i o n s of PdRs would r e s u l t i n a s i g n i - f i c a n t l o s s of c o n t r o l rod worth , whereas once o r twice b u r n t Pu a s s e m b l i e s can be l o c a t e d a t any c o r e p o s i t i o n i n c l u d i n g RCC p o s i t i o n s , w i t h o u t marked d e t e r i o r a t i o n of c o n t r o l rod worth. Th i s t h e o r e t i c a l e x p e c t a t i o n h a s been v e r i f i e d e x p e r i m e n t a l l y by t h e KRITZ exper iments and a t WO. a

Heasurements have been performed d u r i n g s t a r t - u p a f t e r r e f u e l l i n g a t K R R , These demonstra ted t h a t t h e r e d u c t i o n of c o n t r o l rod worth remaics w i t h i n t h e a l l o w a b l e range ( i f o n l y one MOX-FA of t h e a l l -Pu t y p e i s i n t r o - duced p e r c o n t r o l rod c e l l ) . Fur thermore , no i n d i c a t i o n of any s y s t e m a t i c change i n r e a c t i v i t y c o e f f i c i e n t s of t h e o v e r a l l c o r e h a s been found.

I s o t o p i c a n a l y s e s have been c a r r i e d o u t on s e v e r a l KNO MOX-fuel samples w i t h an i n i t i a l enr ichment of 2 % and 3.2 % mass f r a c t i o n Pufiss c o v e r i n g a range of burnups up t o 40 GWd/t. The comparison of p r e d i c t e d and measured Pu h i g h e r i s o t o p i c compos i t ions shows v e r y s a t i s f a c t o r y agreement i n c l u d i n g measured and p r e d i c t e d l o c a l burnup.

E x t e n s i v e o p e r a t i n g and performance e x p e r i e n c e has been ga ined d u r i n g t h e demons t ra t ion programs f o r p lutonium r e c y c l i n g i n l i g h t - w a t e r r e a c t o r s . The o p e r a t i o n a l performance of t h e Pu-bearing f u e l h a s been v e r y s a t i s f a c - t o r y and i s a lmost comparable t o U02 f u e l . The s u c c e s s of t h e mechanical and n e u t r o n i c d e s i g n of NOX f u e l and f u e l a s s e m b l i e s i s demonstra ted by n e u t r o n i c c o r e f o l l o w c a l c u l a t i o n s , in -core measurements and rad iochemica l a n a l y s e s .

The demons t ra t ion programs f u r n i s h s u f f i c i e n t ev idence t h a t l a r g e s c a l e use of PlOX f u e l i s t e c h n i c a l l y f e a s i b l e i n l a r g e commercial LWR c o r e s . S h u f f l i n g p r o p o s a l s e x i s t f o r a l l r e c y c l i n g o p t i o n s of SGR and OXR s a t i s f y i n g t h e r e s t r i c t i o n s on maximum l o c a l power i n c l u d i n g t h e f i n e s t r u c t u r e power shape. The n u c l e a r c h a r a c t e r i s t i c s of a PWR Pu-burner c o r e , des igned f o r economy and h i g h c o n v e r s i o n , do n o t d i f f e r e s s e n t i a l l y from t h o s e of r e c y c l e c o r e s .

The f o l l o w i n g Tab le I summarizes d e s i g n d a t a f o r t y p i c a l mixed ox ide f u e l a s s e m b l i e s .

Table I. Typical Designs of Mixed Oxide Fuel Assemblies (KWU)

Reactor type P WR BWR

FA designed for: Average mass fraction of Pufiss in %

Recycle Pu-burner SGR OMR Pu-burner 2.83 2.70 0.79 1.54 2.20

Mixed oxide rods: Mass fraction of U235 in % Total number Number;mass fraction of Pufiss-enrichment in % 164;3.2 236;2.7 12; 2.5 8;4.1 22; 3.5

72;2.0 10;l.g 18;2.7 28;1.5

Uranium oxide rods: Total number;different enrichments

Rods containing Gd Water-rods or control rod positions

Outer diameter of fuel rod in mm 10.75 10.75 12.5 12.5 11.8

2.2. INVESTIGATIONS ON H I G H CONVERTING PWRs I N THE U/PU FUEL CYCLE

A s thermal r e c y c l i n g has been shown t o 'be t e c h n i c a l l y f e a s i b l e ; and, i f MOX-fuel f a b r i c a t i o n is no t too expens ive , a l s o economica l ly advantageous , i t i s i n t e r e s t i n g t o f o l l o w t h e i d e a of a Pu- fue l l ed LWX, b u t a iming t o a c h i e v e a h i g h convers ion r a t i o . Th i s i n p r i n c i p l e always can be r e a l - i z e d by hardening t h e neu t ron spect rum e.g. by t i g h t e n i n g t h e r e a c t o r l a t t i c e . In t h e Sh i p i n g p o r t ~ e a c t o r ~ ~ , a t i g h t l a t t i c e movable seed and b l a n k e t c o r e f o r 231;U/Th f u e l is i n o p e r a t i o n t o prove t h e f e a s i b i l i t y of l i g h t wa te r b r e e d e r r e a c t o r s . However, i t i s no t prudent t o deve lop h i g h c o n v e r t i n g r e a c t o r s wi thou t a l s o deve lop ing r e p r o c e s s i n g c a p a b i l i t i e s on an i n d u s t r i a l s c a l e . There fo re s t u d i e s i n Germany, j o i n t l y pursued a t KfK and KWU, sup o r t e d by t h e T e c h n i c a l U n i v e r s i t y of Braunschweig (TUB), and a l s o a t GKSSz8, a r e examining t h e UiPu c y c l e f o r WR-high-conver ters . P a r t of Radkowsky's ( I s r a e l ) work i s done i n c o o p e r a t i o n w i t h KfK. It h a s been shownz9 t h a t , i f f a s t b r e e d e r r e a c t o r s w i l l be commercia l ly a v a i l a b l e o n l y w i t h some d e l a y , t h e r e i s an i n c e n t i v e t o deve lop such a h i g h c o n v e r t i n g r e a c t o r t o t a k e an i n t e r m e d i a t e r o l e u n t i l f a s t r e a c t o r s a r e f u l l y a c c e p t - ed.

The p r e s e n t i n v e s t i g a t i o n s f o r deve lop ing a t i g h t l a t t i c e l i g h t wa te r cooled h igh c o n v e r t e r r e a c t o r a r e performed under t h e f o l l o w i n g con- s t r a i n t s :

(1) A h i g h c o n v e r t i n g t i g h t l a t t i c e c o r e sho'sld d i r e c t l y r e p l a c e a normal PCIR c o r e , such t h a t t h e e s s e n t i a l out-of-core components remain t h e same f o r t h e same r e a c t o r power.

( 2 ) The convers ion r a t i o should reach a v a l u e of C R r 0.95.

( 3 ) The f i s s i l e i n v e n t o r y should be comparable wi th t h a t of a f a s t r e a c t o r i n v e n t o r y of e q u a l s i z e .

( 4 ) The s a f e t y f e a t u r e s of t h e r e a c t o r must be such t h a t t h e y can be accep ted f o r l i c e n s i n g . That i s : - n e g a t i v e v o i d - c o e f f i c i e n t i n a l l burnu,? s t a t e s - a c c e p t a b l e p in de fo rmat ion behaviour i n t h e t i g h t l y packed l a t t i c e - proven emergency c o r e c o o l i n g f o r off-nominal c o n d i t i o n s .

The neu t ron p h y s i c s i n v e s t i g a t i o n h a s begun w i t h an assessment of t h e v a l i d i t y of t h e a v a i l a b l e c a l c u l a t i o n a l methods and d a t a b a s e s f o r t i g h t l a t t i c e (MOX and uranium f u e l ) a n a l y s i s . WIMS-D, KfK and TUB d a t a and methods have been used t o a n a l y s e s e l e c t e d exper iments .30 A comparison of t h e b a s i c d a t a used by t h e d i f f e r e n t (LWR) a n a l y s i s codes shows major d i s - c r e p a n c i e s i n t h e h i g h e r Pu i s o t o p e s and a r e be ing improved now. F a s t r e a c t o r methods, i f improved by a p p r o p r i a t e h e t e r o g e n e i t y c o r r e c t i o n s (e .g . by K A P E R ~ ~ o r by even s i m p l e r background cor :cect ions a c c o r d i n g t o e u iva - l e n c e r e l a t i o n s ) appear t o be s a t i s f a c t o r y f o r d e s i g n c a l c u l a t i o n s . " Using such d a t a and methods, a s sessment of t h e f o l l o w i n g f o u r c o n c e p t s , which a r e b e i n g s t u d i e d by KfK and KWU w i t h s u p p o r t from TUB, i s p o s s i b l e :

( a ) homogeneous concept ; t h e p r e f e r r e d s o l u t i o n f o r i n d u s t r y ( b ) he te rogeneous s o l u t i o n ; s e e d and b l a n k e t concept wi th movable seed f o r

c o n t r o l and s h u t down

(c) heterogeneous module concept with movable seed for control and control rods for shut down

(d) zonewise heterogeneous concept with rods for control and shut down (similar to FBR heterogeneous concepts).

In concept (a) the critical physics task is the clear identification of a negative void coefficient for an enrichment of about 7.0 % Pu fissile. (This allows a conversion ratio of about 0.95 to be obtained.) In all the heterogeneous concepts the fissile Pu is about 12 % to 14 %, with low enriched Pu ($3 5 %) breeder regions. In these concepts the void coefficient is definitely sufficiently negative to allow licensing from this particular point of view, but the overall concept clearly is more complicated from the engineering aspect. Safety experiments are being started now at KfK. If first results are encouraging, it will be possible to.identify the most promising concept. For that concept physics experiments are intended.

For the homogeneous concept (a), Fig.6 presents a possible solution.28 Table I1 summarizes the correspondmg reactor data. 33

Table 11. Typical Design Data for a Homogeneous APWR and PWR (KWU)

Thermal Power Mw Core Radius cm Core Height cm Core Volume m3 Number of Subassemblies Number of Fuel Pins Pellet Diameter mm Pin Diameter mm Pitch mm Effective Fuel Volume Fraction Effective WaterfFuel Volume Fraction Linear Power Rating W/cm Power Density MW/m3 Average Enrichment %

. Average Conversion Ratio Intended Burn-up Mwdfkg Full Power Days per Cycle Doppler Coefficient dk/dT oc-1

r Coolant Density Coefficient dk/dp cm3/G

Consideration is also be given to the possibility of building a prototype reactor, should the practical problems involved in straightforward core replacement be greater than presently anticipated.

3. PHYSICS INVESTIGATAIONS FOR HIGH TEXPEIUTUKE REACTORS

The gas cooled High Temperature Pebble Bed Reactor has been developed in the Federal Republic of Germany for more than a decade. The Federal Government sponsors several large power plant projects such as the AVR, the THTR, the PNP and the HHT.*

The construction of the prototype plant at Uentrop in Westfalia is delayed by licensing procedures. The following summary on the verification of the theoretical models and minimization of layout and safety margins for a pebble bed reactor is based on a private coinmunication by the Kern- forschungsanlage Jiilich (KFA) . 34

'In the frame of the physics program for pebble bed reactors the critical facility KAHTER", shown in FigR 7, has been established at KFA. KAHTER was first critical in summer 1973 and since then a detailed experimental and theoretical nuclear program has taken place. Emphasis has been placed on questions of control rod efficiency, macroscopic and microscopic neutron flux distribution, graphite damage by fast neutrons and effects of diverse neutron poisons.

The present experimental program deals with the investigation of multizone OTTO-cores (OTTO stands for "once through then out"). The aim is to study the axial asymmetric flux and power profile, the reactivity and flux flattening effect of the upper cavity, and the control rod efficiency in this cavity and the top reflector. Additionally investigations will be made on the reactivity worth of small pebbles as a secondary shut-down system, on water ingress in the core, and on a flux mapping system by out of core instrumentation to detect even differential perturbations. The latter aspect is essential, because in a pebble bed reactor the monitoring of the power distribution by incore instrumentation is prohibited by the high coolant temperatures within the core and the nature of the pebble bed. One relies therefore on neutron detectors distributed around and above the core. Diagnosis of local perturbations from signals in the upper reflector is however rendered difficult by the cavity above the pebble bed, which tends to smear out the radial profile. For the same reason the flattening of power peaks by specific rod,insertion is not obvious. Detailed studies at INTEIUTOH~~ on this problem show, that the combined signals of neutron detectors in the upper and radial reflectors together with measurements of the gas outlet temperature profile allows a sufficienrly accurate detection of such a disturbance in the core, which is large enough to cause unwanted changes in the gas outlet temperature profile. Furthermore it is shown, that these perturbations can be counterbalanced by specific absorber rod movements.

At KFA, theoretical calculations to interpret KAIiTER experiments are based mainly on transport methods (SN, Monte Carlo) and a coupled

* AVR = Arbeitsgemeinschaft Versuchsreaktor; THTR = - Thorium-Hochtemperatur- ~eaktor, PNP = @ototypreartor ~ukieare xrozeDv2rme, HBT =-~ochtem~eratur- - reaktor mit Helium-Turbine

" KAHTER = - Kritische - Anlage - Hochtemperatur-Realctor - -

d i f f u s i o n / t r a n s p o r t method7, which was mentioned i n s e c t i o n 1.1. S impl i f ied nlethods to be included i n the d i f f u s i o n code CITATION have been developed (based on d i r e c t i o n a l l y dependent d i f f u s i o n c o e f f i c i e n t s ) t o avoid l a r g e

computi;% times i n c a l c u l a t i n g the neut ronic p rope r t i e s i n the upper

cav i ty .

A l l the r e s u l t s of the e a r l i e r measurements f o r ke f f , r e a c t i o n r a t e s , c e n t r a l con t ro l rod worth were ca l cu la t ed with s a t i s f a c t o r y agreement wi th experiments, well i n s i d e the p resen t ly accepted l i m i t s . The r e fe rence core with THTR f u e l elements as well a s two boron poisoned cores conta in ing 435 and 955 s t a t i s t i c a l l y d i s t r i b u t e d boron elements have been ca l cu la t ed a s

, t h e e a r l i e r cores with AVR f u e l element loading. To see whether hafnium poisoned cores , l i k e the i n i t i a l THTR core , can be ca l cu la t ed with t h e same accuracy th ree hafnium experiments were performed. To compensate geometri- c a l e f f e c t s the boron and hafnium absorpt ion was equal ized i n equiva lent cases by ad jus t ing the poison content . The c a l c u l a t i o n f o r hafnium poisoned co res shows good agreement with experiments.

For High Tempirature Gascooled Reactors it is e s s e n t i a l t o know t h e f a s t neutron dose i n the g raph i t e r e f l e c t o r s . Therefore, i n the c r i t i c a l f a c i l i t y KAHTER benchmark experiments f o r f a s t neutron f luxes and doses were performed inc luding an accura te p red ic t ion and evalua t ion of the f a s t neutron f luxes . Rhodium was used t o t e s t the accuracy of g raph i t e damage r a t e predic t ions . Measurements (w) show a good agreement with the ca l cu la t ed f l u x e s i n the core region, but i n t h e r e f l e c t o r region t h e measured f a s t f l u x is overestimated by ca l cu la t ion . So the r a d i a t i o n damage of HTGR r e f l e c t o r s seems t o be lower than predic ted .

Experiments with the OTTO f l u x d i s t r i b u t i o n , which is a x i a l l y asymmetric a s shown i n Fig. 9 have been s t a r t e d . F i r s t a re ference core with an upper cav i ty of 50 cm and a top r e f l e c t o r of 50 cm height has been inves t iga t ed . In the next s t ep , e i g h t top r e f l e c t o r rods were i n s e r t e d i n t o the upper r e f l e c t o r and cav i ty and then a number of f u e l and g raph i t e e l e - ments were added to make keff equal one. These a d d i t i o n a l elements were a measure of t h e r e a c t i v i t y worth of the top r e f l e c t o r rods. Table I11 shows the r e s u l t s . The accuracy of the t h e o r e t i c a l models f o r these s t rong ly heterogeneous co res i s s i m i l a r to' t h a t obtained f o r the homogeneous loading.

. Table 111. keff Values of KAHTER-OTTO-Cores and Reac t iv i ty Worth of the Top Ref lec tor Rods (KFA)

Core

C-E Akeff 7 (%) Worth of

Ref l ec to r Rods 515 5/5a A k (%)

515 5/5a

Number of Fuel Elements 21805 23061 CITATION -0.023 -0.216 2.81 DIFF-H 0.468 0.535 2.80 MORSE-K 0.25 0.35 2.88

F u t u r e work a t KAHTER w i l l look i n d e t a i l t o t h e s t a b i l i z i n g e f f e c t of t h e OTTO-profile on l a r g e s c a l e f l u x i r r e g u l a r i t i e s and t o t h e n u c l e a r d a t a of U-233. Q u e s t i o n s d e a l i n g w i t h t h e low e n r i c h e d uranium f u e l c y c l e , de- s i r e d t o reduce p r o l i f e r a t i o n r i s k s , may a l s o be addressed .

It shou ld be kep t i n mind, t h a t some q u e s t i o n s remain unanswered because i t i s no t p o s s i b l e t o h e a t up t h e KAHTER f a c i l i t y . I n f o r m a t i o n abou t t e m p e r a t u r e e f f e c t s , on Doppler resonance broadening and t h e g r a p h i t e s c a t t e r i n g m a t r i x t h e r e f o r e have had t o be t aken from o t h e r exper iments o r from the (more than 10 y e a r s ) e x p e r i e n c e w i t h t h e AVR-power p l a n t .

4. PHYSICS INVESTIGATIONS FOR FAST REACTORS

Almost a l l t h e f a s t r e a c t o r p h y s i c s work i n t h e FRG i s done w i t h i n t h e frame of a common program i n v o l v i n g Germany, Belgium and t h e N e t h e r l a n d s (DEBENE) and , s i n c e 1977, w i t h France. The i n v e s t i g a t i o n s on he te rogeneous f a s t c r i t i c a l a s s e m b l i e s a r e performed a l s o i n c o o p e r a t i o n w i t h t h e Uni ted Kingdom. P r e s e n t l y t h e main emphasis of t h e p h y s i c s i n v e s t i g a t i o n s l i e s on f a s t r e a c t o r s a f e t y a s p e c t s ; t h i s i s d i s c u s s e d i n a s e p a r a t e paper t o t h i s

a c o n f e r e n c e . 37

4.1. THE FAST TEST REACTOR KNK AT KARLSRUHE

S t a r t - u p exper iments i n t h e 5 8 mqth f a s t t e s t r e a c t o r a t t h e Nuclear Research Cen te r Kar l s ruhe have shown s a t i s f a c t o r y agreement between t h e o r e t - i c a l p r e d i c t i o n s and exper imenta l r e s u l t s . 3 8 T h i s i s e s p e c i a l l y noteworthy because of t h e ex t remely heterogeneous c o r e - c o n f i g u r a t i o n wi th z i rconium- h y d r i d e i n p a r t of t h e d r i v e r zone. A f t e r a p in f a i l u r e t h e s p a t i a l d i s t r i - b u t i o n of t h e i s o t o p i c r a t i o 1 3 1 ~ e / 1 3 4 ~ e was s u c c e s s f u l l y used by INTEKATOM and K F K f o r an e s t i m a t e of t h e f a i l u r e l o c a t i o n .

Due t o s p e c i f i c f e a t u r e s of t h e pr imary c o o l a n t sys tem cover g a s bubbles a r e c a r r i e d through t h e c o r e by t h e c o o l a n t . I f t h e c o o l a n t f l o w r a t e is above 50 % of t h e nominal r a t e , i t a p p e a r s t h a t accumulated g a s bubbles a r e r e l e a s e d which may l e a d t o a r e a c t o r t r i p due t o t h e n e g a t i v e vo id c o e f f i c i e n t of KNK. The mechanism and p r e c i s e l o c a t i o n of t h e g a s accumula t ion a r e no t y e t comple te ly unders tood. S i g n a l s from a n e u t r o n d e t e c t o r , thermocouples a t t h e subassembly o u t l e t s and from s e i s m i c t r a n s -

a d u c e r s were ana lysed u s i n g n o i s e a n a l y s i s t e c h n i ues a t KfK i n o r d e r t o

3'',48 I t was found t h a t below i d e n t i f y t h e o r i g i n of t h e observed e f f e c t s . 50 % nominal f low r a t e cover gas i s e n t r a i n e d i n t h e c o o l a n t even when reac - t i v i t y peaks cannot be d e t e c t e d . T y p i c a l sampie r e c o r d s f o r a g a s bubble passage th rough t h e c o r e a r e shown i n Fig . 10. As can be s e e n , t h e r e a c t i - v i t y and cor respond ing power d i p i s preceded by a mechanical shock, proba- b l y o r i g i n a t i n g from the r e l e a s e of a l a r g e r bubble . F u r t h e r I n v e s t i g a t i o n s a r e underway t o c l a r i f y t h e s i t u a t i o n .

I n o r d e r t o o b t a i n b e t t e r c o n d i t i o n s f o r m a t e r i a l i r r a d i a t i o n s ( b y i n c r e a s i n g t h e r e a c t o r power) and a l s o f o r exper iments by i n s t a l l i n g a n i n t e g r a t e d smal l loop f o r performing o p e r a t i o n a l t r a n s i e n t s ( p o s s i b l y u n t i l 41 c lad-breach) studies are performed to modify the IQXCOI-e. =ese &dies a r e based on a c o r e , which w i l l c o n t a i n 20 % 2 3 5 ~ enr ichment , a uniform Pu enr ichment ( 5 3 0 % ) w i l l keep t h e r e a c t o r c r i l r i c a l . The f u e l is s i m i l a r t o

I f t r a n s p o r t e f f e c t s i n f a s t r e a c t o r c o n f i g u r a t i o n s a r e i m p o r t a n t , a s f o r i n s t a n c e i n r e l a t i v e l y smal l t e s t r e a c t o r s , i n s p e c i a l he te rogeneous d e s i g n s o r i n a c c i d e n t s i t u a t i o n s , w i t h g r e a t l y d i s t o r t e d r e g i o n s , a mul t i - d imens iona l t r a n s p o r t code w i t h r e a s o n a b l e computing t ime i s n e c e s s a r y . The noda l t r a n s p o r t method, a s mentioned i n s e c t i o n 1.1, was a p p l i e d by ~ l j ~ 4 7 t o a LMFBR benchmark48 i n ( x , y ) geometry. I n F u a n example f o r f l u x t r a v e r s e s a long t h e main a x i s i s shown. The d < v i a t i o n s from d i f f u s i o n t h e o r y i n t h e f o l l o w e r r e g i o n s f o r f a s t n e u t r o n s can be recogn ized c l e a r l y . The computing time f o r t h e nodal t r a n s p o r t method i s abou t 3 t imes l a r g e r than t h a t f o r noda l d i f f u s i o n theory.

5. SOXX PHYSICS ASPECTS OF SPALLATION REACTIONS

Though on ly l o o s e l y r e l a t e d t o t h e t o p i c of t h i s paper , a b r i e f com- ment i s g iven on e s s e n t i a l d i s c r e p a n c i e s between ( p r e l i m i n a r y ) t h e o r e t i c a l p r e d i c t i o n s of secondary neu t ron s p e c t r a i n s p a l l a t i o n r e a c t i o n s . The exper- i m e n t a l methods had been developed f o r c r o s s - s e c t i o n measurements i n t h e f a s t r e a c t o r p r o j e c t a t Kar l s ruhe .

a I n c o o p e r a t i o n between KfK-Karlsruhe and CFA J u l i c h p r e s e n t l y a f e a s i -

b i l i t y s t u d y i s being performed on an i n t e n s e s p a l l a t i o n n e u t r o n source f o r a p p l i c a t i o n i n fundamental r e s e a r c h mainly. A l i n a c of 1.1 GeV p r o t o n energy and a pulse-peak c u r r e n t i n t e n s i t y of abou t 100 mA i s env i saged ( a v e r a g e i n t e n s i t y 5 111.4, p u l s e width 550 psec , p u l s e f requency s 100 Hz). A pos- s i b i l i t y being cons ide red i s t o couple t h e l i n a c beam t o a s t o r a g e r i n g . For t h e t a r g e t a s o l i d m a t e r i a l is p r e f e r r e d , i . e . Pb-Bi (uranium is an o p t i o n ) .

The p h y s i c s i n v e s t i g a t i o n s have r e c e n t l y <:oncentra ted on t h e d e t e r - ~ n i n a t i o n o f neu t ron and on secondary p ro ton y i e l d s , and on n e u t r o n and secondary p ro ton s p e c t r a . A t SIN* a b s o l u t e measurements have been performed w i t h 600 MeV p r o t o n s on a b a r e c y l i n d r i c a l l ead t a r g e t . 4 9 Most exper iments today a r e made f o r t a r g e t s i n su r round ing moderator m a t e r i a l . The bare- t a r g e t exper iments g i v e d i r e c t i n f o r m a t i o n on h i g h energy n e u t r o n y i e l d s , an e s s e n t i a l parameter f o r an a p p r o p r i a t e d e s i g n of s h i e l d i n g s t r u c t u r e s between the v a r i o u s a r e a s f i n a l l y used i n s p a l l a t i o n s o u r c e exper iments . The y i e l d d a t a o b t a i n e d a r e somewhat h i g h e r than t h o s e g i v e n i n t h e l i t e r a -

a t u r e .

With r e s p e c t t o t h e s p e c t r a of e m i t t e d n e u t r o n s , Fig . 14 g i v e s a n e x c e l l e n t example of t h e p r e s e n t s t a t e of t h e a r t . The measurements of KfK and LASL, though p r e l i m i n a r y , show good agreement ( n o t e t h a t t h e LASL- exper iments a r e normal ized t o a 1.4 h i g h e r t o t a l n /p v a l u e , due t o t h e h i g h e r n e u t r o n y i e l d o b t a i n e d wi th 800 MeV p r o t o n s ) . P r e v i o u s c a l c u l a t i o n s done a t LASL and p r e l i m i n a r y a t KFA, however, show remarkable d i s c r e p a n - c i e s : b o t h c a l c u l a t i o n s r e s u l t i n c o n s i d e r a b l y s o f t e r s p e c t r a . The r e a s o n s f o r t h i s d i s c r e p a n c y a r e no t y e t w e l l unders tood . F u r t h e r work i s n e c e s s a r y on n u c l e a r d a t a and on methods assessment t o c l s r i f y t h e s i t u a t i o n .

* Swiss I n s t i t u t e of Nuclear Research

ACKNOWLEDGMENTS

Contributions to this paper were provided by Dr. H. Finnemann, Dr. K. Koebke, Dr. M. R. Wagner, W U Erlangen, and Dr. P. Strohbach, KWU Offenbach Dr. H. Henssen, INTERATOM, Bergisch Gladbach Prof. D. Emendorfer, Dr. G. Hehn, University of Stuttgart Dr. R. D. Neef, KFA Jiilich Prof. D. Biinnemann, GKSS Geesthacht

The author is grateful to their assistance in preparing this review. . Helpful discussions with Dr. B. Goel, U. Fischer, M. Marzo and Dr. H. W. Wiese (all KfK Karlsruhe) are gratefully acknowledged, also the careful reading of the manuscript by Dr. A. Rowe. He would also like to thank Mrs. G. Bunz for the efficient typing of the manuscript.

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22. F. FROHNER, B. GOEL, "Diskrepanz bei reaktorphysikal i schwicht igen Kerndaten - Stand und neuere KfK-Aktivitaten," in Proc. Jahrest.agung Kerntechnik, Kerntechnische Gesellschaft, Berlin (1980), p. 47.

B. GOEL, "Neutronen-Daten von schweren Aktiniden fiir KEDAK-4." in Proc. Jahrestagung Kerntechnik, Kerntechnische Gesellschaft, Berlin (1980), p. 51.

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Reactors, Part 1: Demonstration Programs," Siemens Forsch.- u. Entwick1.-Ber. Bd. 8 (1979) Nr. 2, Springer Verlag, p. 108.

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28. K. PENNDORF, F. SCHULT and D. BtfNNEMANN, "Absch2tzung der wesentlichen neutronenphysikalischen Eigenschaften hochkonvertierender DWRs im U/Pu- Zyklus," in Proc. Jahrestagung Kerntechnik, Kerntechnische Gesellschaft, Berlin (19801, p. 961.

29. H. H. HENNIES and H. MARE, "Uberlegungen zur Modifizierbarkeit eines LWR im Hinblick auf eine bessere Uranausnutzung," in - Proc. Jahrestagung Kerntechnik, Kerntechnische Gesellschaft, Berlin (1980), p. 953.

30. A. R. BOYNTON et al., "High Conversion Critical Experiments," ANL-7203 (1967).

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31. P. E. MCGRATH and E. A. FISCHER, "KAPER - A Computer program for the Analysis of Experiments Performed in Heterogeneous Critical Facilities," in Proc. Mathematical Models and Computational Techniques for the Analysis of Nuclear Systems, Am. Nucl. Soc., Ann Arbor (1974).

32. A. D. ROWE et al., "Final Report on High Converting PWR (APWR) Benchmark Activities," to be published as KfK report.

P. RAU et al., "Ergebnisse der Vorstudie fur einen Leichtwasserreaktor mit besserer Uranausnutzung," in Proc. Jahrestagung Kerntechnik, Kerntechnische' Gesellschaft, Berlin (1980), p. 957.

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A. STROMICH and M. KHAMIS, "Erkennbarkeit von Stsrungen in der Leistungs- und Temperaturverteilung bei HTR-Kernen," in Proc. Jahrestagung Kerntechnik, Kerntechnische Gesellschaft, Berlin (1980), p. 755.

W. SCHERER and H. GERWIN, "~iff~sionstheoretisches Simulations- verfahren zur Behandlung des oberen Hohlraums in Kugelhaufen-HTRs," Jul-1599 (1979).

R. FRGHLICH, F. HELM, G. KESSLER, E. KIEFHABER, "Safety Related Physics Activities in the DEBENE-FBR-Project," this conference.

A. STANCULESCU, U. WEHMANN ec al., "Nuclear Design of the Fast Test- Reactor KNK-I1 and its Experimental Verification," in a International Symposium on Fast Reactor :Physics, Aix-en-Provence (1979).

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M. EDELMNN et al., "Two-phase Flow Effects observed in a Sodium Cooled Reactor," Trans. Am. Nucl. Soc., iE, 798 (1980).

I. BROEDEKS, B. KRIEG, H. KUSTERS and E. STEIN, "Neutronen- physikalische Untersuchungen fur Nachladekerne von KNK-11," to be published.

U. WEHMANN, H. SPENKE and S. PILATE, "Design Aspects and Problems of Heterogeneous Cores for SNR-2," in -International Symposium on Fast Reactor Physics, AixLen-provence (1979).

E. KIEFHABER and A. POLCH, "Neutronenphysikalische Untersuchungen zu groBen modularen Cores," to be published.

44. E. A. FISCHER, "Neutron Streaming in Fast Reactor Slab Lattices, and in Cylindrical Channels," Nucl. Sci. Eng.. (submitted for publication).

45. P. KOHLER and J. LIGOU, "Axial Neutron Streaming in Gas-Cooled Fast Reactors," Nucl. Sci. Eng., 54, 357 (1974).

46. J. L. ROVLANDS and C. EATON, "The Effective Axial Diffusion Coefficient for a Low-Density Channel, &cl. Sci. Eng., to be published.

47. M. R. WAGNER, KWU Erlangen, Private Communication (1980).

48, G. BUCKEL, K. KUFNER and B. STEHLE, "Benchmark Calcula t ions f o r a Sodium-Cooled Breeder Reactor by Two- and Three-Dimensional Dif fus ion Methods," Nucl. Sc i . En&, 66, 75 (1977).

4 9 . S. CIERJACKS, M. T. RAINBOW e t a l . , "Neutron Yields and Spectra from 590 MeV (p ,n) React ions on Lead Targets ," i n Proc. Symposium on Neutron Cross Sec t ions from 10 t o 15 MeV, US-Department of Energy, Brookhaven (1980).

50. T. YOSHIDA. "Im~roved Treatment of the Neutron Strearnine. Through - - Control Rod Followers i n a Sodiurn-Cooled Fas t Reactor," Nucl. Sc i . w, 72 , 361 (1979).

1.0 2.0 3.0 Single rod drop In B 10 Detector posltlon B 09

Fig. 1 ;.sy ., ,,,..

Single Rod Drop Experiments and Theaiy (KWU) 1

I 1

Fig. 2 Decrease i n Coolant Outlet Temperature [KWU

-

Fig. 3 a 1 N j p

Equivalence Theory Solution for !he Henry-Worley Benchmark (KWU]

4 - calculaled measured

600

P 500

400

300 0 ZOO 400 600 BOO 1000 1200

Full Power Days ( IFPD 2 5 5 MWDITUI :14 ,I&

Fie. 4 Comparison ol Calculated and Measured Control Bank [GKSS)

- Korigen ~ u m - u p

--- Faser lKWUl MWdlkg HM . measuremenl [ICE)

Fig. 5 KWO-Postirradiation Tests I Fig. 6 APWR Core Arrangement in a Standard PWR Vessel I

I

- - 4 ? 9 s a - - - 4 m m I

FIG, 7. THE CRITICAL FACILITY KAHTER (KFA)

. - Selsmlc Transducer. a,,, larb. units)

LY Neulron Flux [ I %lunlll

Oullel Temperalure { IKluni l l Cenlnl Subassembly. AT = 2 4 0 K

kj11i Fig. 10 Signals in KNK during a Gas Bubble's Passage

0 Core Zone 5 3 4 O Ouler Blanket 2 1 0 O Inner Blankel 2 2 9 0 Control Rods 3 0 0 Secondary Shut Oown 2 4

1027 Fig 1 1 -.- - r h - - * Example for a Modular Fast Reactor Core

I 82 = I Z . l 8 m z

E - 2 0 2 - 5 m z - = s 3 8 2 ; 2 m t .- - .- - - 4 B ' = 0

5 5 Rowlands Model = - .- 1 - = - - .B 4 <

3

Total Cross Section Z, l c m ' l

Fig. 1 2 .,.. q p 1 : Ih&.

Elfeclive Axial Channel Diffusion Coefficients

Fig. 1 3 . . . . , . . q;,, . ' , , I .

Flux Traverses i n Nodal Transport Theory along Main Axis [KWU)

0.1 I 10 loo E, (MeV]

Fig. 1 4 ;@ Measured and Calculated Neutron Spectra (Lead Targets] normalised to respective n /p Values)

F.R. GERMANY S ~ E T Y RELATED PHYSICS ACTIVITIES IN THE DEBENE-FBR-PROJECT (contd. )

Present status, needs and perspectives+)

R.Frohlich, P. Helm, G. Kessler, E. Kiefhaber

Kernforschungszentrum Karlsruhe Postfach 3640, D-7500 Karlsruhe, FRG

1. INTRODUCTION

Construction and licensing of the sodium cooled prototype fast reactor SNR-300 is the main basis for a licensable safety concept of a future commercial size LMFBR plant SNRL1300 in the FRG. This safety design and licensing concept strongly influences the related R&D-programmes. Any discussion of the present status needs and future trends of safety R&D therefore must be seen in the frame of the overall plant safety concept. This paper deals only with safety and physics R&D activities related to core disruptive accidents with extremely low probability of occurence.

2. LMFBR SAFETY DESIGN CONCEPT

The safety concept of SNR-300 and a future commercial size plant SNR-1300 is based on the multiple barrier concept and on several levels of safety de~ignl.~r~. The first level of safety design prevents the occurence of fault initiation by application of inherent design characteristics, the design principles of diversity and redundancy as well as inservice inspection of important safety components and quality assurance during design and construction of the plant. This assures low probability of occurence for accident initiations. Thc second level of safety design prevents the propagation of faults into serious core accidents by maintaining integrity of the core and providing reliable reactivity control. The third level assures adequate mechanical response of the primary coolant systems, as well as the inner and outer containment against extremely unlikely severe accidents, thereby limiting the release of radioactivity to acceptable levels (barrier concept).

2.1 Accident Prevention

On the first and second level of safety design a criterion of less than per year for loss of coolable core geometry is the design objective.

b s s of coolable core geometry would lead to core melt down or core destruction. This may occur if the plant protection system together with the

+)sun Valley Topical Meeting on Advances in Reactor Physics and Shielding, September 14-17, 1980

redundant s h u t down sys tems o r t h e decay h e a t removal sys tems f a i l on d e ~ n a n d ~ , ~ . It has been shown f o r SNR-300 t h a t t h i s d e s i g n o b j e c t i v e can be met by us ing a p l a n t p r o t e c t i o n system w i t h two d i v e r s e l y and i n d e p e n d e n t l y a c t i n g s h u t down system^^,^. The same s h u t down sys tem d e s i g n w i l l a l s o be used f o r SNR-1300.

The pr imary s h u t down sys tem drops a b s o r b e r r o d s i n t o t h e c o r e from above by g r a v i t y f o r c e s . The secondary s h u t docm system p u l l s by u s e of a precompressed s p r i n g a f l e x i b l e a b s o r b e r c h a i n i n t o t h e c o r e from below. Both systems have d i v e r s e e l e c t r o n i c channe l s .

In c a s e of a commercial s i z e LMFAR p l a n t SNR-1300 f o u r main p r imary and secondary h e a t t r a n s f e r c h a i n s p rov ide f o r s a f e decay h e a t removal. I n a d d i t i o n a second d i v e r s e decay h e a t removal sys tem is prov ided by a sodium a i r emergency c o r e c o o l i n g sys tem which e x t r a c t s h e a t from t h e r e a c t o r o u t l e t plenum by immersed h e a t exchangers i n t h e r e a c t o r tank. Immersed h e a t exchangers w i l l a l s o be i n s t a l l e d i n t h e i n t e r m e d i a t e h e a t exchangers . I n c a s e of SNR-300 s i m i l a r d i v e r s e and redundant decay h e a t removal sys tems a t t a i n an u n a v a i l a b i l i t y of l e s s than p e r demand i n c a s e of t h e s team

a g e n e r a t o r f a i l u r e and , t o g e t h e r w i t h t h e f a c t , t h a t t h e steam g e n e r a t o r f a i l u r e r a t e is l e s s t h a n 1 e r y e a r , one o b t a i n s a n u n a v a i l a b i l i t y of less t h a n p e r r e a c t o r yea r6?$ . Even t h e complete l o s s of a l l a c t i v e sys tems would n o t l e a d t o c o r e damages. The decay h e a t would be t r a n s f e r e d t o t h e sodium and w i l l p e n e t r a t e through t h e i n s u l a t i o n of t h e s t i l l i n t a c t pr imary c o o l a n t p ipe i n t o t h e i n n e r ~ o n t a i n m e n t ' , ~ .

2.2 P r o t e c t i o n Agains t F a u l t P ropaga t ion

I n a d d i t i o n t o t h e r e q u i r e d h i g h r e l i a b i l i t y of t h e s a f e t y s h u t down sys tems i t must a l s o be a s s u r e d t h a t i n i t i a t i o n of f a u l t s which cou ld deve lop f r o ~ u l o c a l b lockages i n f u e l e l ements , a r e c o u n t e r a c t e d d i r e c t l y by t h e p l a n t p r o t e c t i o n sys tem t o avo id t h e p o s s i b i l i t y of f a u l t p r o p a g a t i o n and g l o b a l d i s t u r b a n c e s . E.g. l o c a l b lockages i n f u e l e l ements and f u e l p i n f a i l u r e s a r e d e t e c t e d by i n d i v i d u a l subassembly i n s t r u m e n t a t i o n

5 ( t h e m o c o u p l e s ) and de layed n e u t r o n moni to r s . a

2.3 Primary System and Containment Design B a s i s

Under t h e above c o n d i t i o n s s e v e r e c o r e m e l t down o r c o r e d i s a s s e m b l y a c c i d e n t s f o r t h e SNR-300 c a n o n l y be i n i t i a t e d i f t h e p l a n t p r o t e c t i o n sys tem w i t h both s h u t down sys tems f a i l on demand. Consequent ly t h e i r p r o b a b i l i t y of occurence would be less than p e r y e a r f o r t h e SNR-300; and t h i s w i l l a l s o be a d e s i g n requ i rement f o r t h e SNR-1300. D e s p i t e of t h i s low p r o b a b i l i t y of occurence such c o r e m e l t down a c c i d e n t s have t o be c o n s i d e r e d a s a d e s i n b a s i s f o r t h e r e a c t o r t a n k and t h e i n n e r and o u t e r conta inment ~ y s t e m l , $ , ~ . The r e a c t o r t ank , t h e pr imary c o o l a n t and i n n e r conta inment sys tem must w i t h s t a n d c e r t a i n p r e s s u r e s and h e a t l o a d s d e f i n e d from t h e a n a l y s i s of such s e v e r e c o r e a c c i d e n t s . The s t r u c t u r a l r e q u i r e m e n t s

f o r the ou'ter containment bui ld ing a r e given by the d e s i n bas i s f o r 1 3 earthquakes, a i r p l a n e crashes and gas cloud explosions . The l eak '

t i gh tness of the inne r and o u t e r containments i s determined by t h e r ad ioac t ive source term c h a r a c t e r i s t i c s of the low p r o b a b i l i t y severe core melt down accident . The o u t e r containment c o n s i s t s t he re fo re of a r e l a t i v e l y l e a k proof s t e e l containment surrounded by a re inforced concrete containment to p ro tec t a g a i n s t ex te rna l impacts. In case of SNR-300 i t is equipped with a v e n t i l a t i o n system t o keep a c e r t a i n underpressure i n the s t e e l

3 8 containment . P r i o r t o a d iscuss ion of t h e present s t a t u s of t h e core design and

s a f e t y a n a l y s i s of a commercial s i z e p l an t SNR-1300 and a subsequent d i scuss ion of key s a f e t y phenomena i t seems appropr i a t e t o present f i r s t the approach taken f o r t h e CDA a n a l y s i s of SNR-300.

3. CORE DISRUPTIVE ACCIDENT ANALYSIS PROCEDURE FOR THE SNR-300

A d e t a i l e d d i scuss ion of p o t e n t i a l i n i t i a t o r s f o r core d i s r u p t i v e acc iden t s of the SNR-300 l e a d s t o the conclusion t h a t primary pump coastdown with simultaneous f a i l u r e of both independent shutdown systems - the so c a l l e d LOF ( l o s s of flow) acc ident - r e q u i r e s the most c a r e f u l a t t e n t i o n . The removal of a l l 9 con t ro l rods (of t h e primary shutdown system) with maximum speed ( r e a c t i v i t y ramp of l e s s than 4 d l s e c ) and the subsequent f a i l u r e of both independent shutdown systems - a s o c a l l e d low ramp r a t e TOP ( t r a n s i e n t overpower) acc ident - has a much lower p robab i l i t y of occurance than t h e LOF-accident. The following d iscuss ion w i l l t he re fo re focus a t t e n t i o n on the LOF-accident and only a few remarks w i l l be made about the low ramp r a t e TOP-accident.

3.1 Accident Path S t ruc tu re and Key Phenomena

The poss ib l e acc ident pa ths and acc ident phases a r e ind ica t ed i n Fig. 1. They a r e defined i n a s i m i l a r way a s discussed by Jackson e t a lZ6. It w i l l not be poss ib l e he re t o d i scuss i n d e t a i l t h e numerous phenomena of t h e various phases which a r e r e spons ib le f o r p re sen t ly e x i s t i n g u n c e r t a i n t i e s f o r LMFBR acc ident a n a l y s i s and which r e q u i r e a concentrated R&D e f f o r t i n the fu tu re . These key phenomena determine whether the acc ident develops a , mild o r ene rge t i c behavior und whether t h e core d e b r i s can be contained wi th in the primary system.

3.2 Analysis of the Loss of Flow Accident f o r t h e SNR-300

The a n a l y s i s concent ra tes on t h e end of l i f e core of the Mark 1A r e a c t o r core design of t h e SNR-300, which has a l a r g e r ene rge t i c s p o t e n t i a l than the f r e s h core2'.

Best Es t ima te LOF Accident Pa th

By us ing b e s t e s t i m a t e pa ramete r s and model l ing f o r t h e LOF a n a l y s i s o f t h e SNR-300 (e.g:fuel a x i a l expans ion , f u e l d i s p e r s a l by f i s s i o n g a s and sodium vapor ) t h e a c c i d e n t proceeds ( s e e t h e b i g b l a c k arrows i n Fi . 1 ) th rough a r e l a t i v e l y mi ld i n i t i a t i n g phase i n t o a t r a n s i t o r y phase2$. During t h e t r a n s i t o r y phase and t h e phase of i n t e g r a l m a t e r i a l movement secondary e x c u r s i o n s cannot be excluded bu t i f t h e y occur t h e y a r e expec ted t o be mi ld (based on d i s p e r s i v e p r o p e r t i e s of c o r e m a t e r i a l and on incoherency arguments f o r f u e l movement d u r i n g t h e t r a n s i t o r y phase) . They w i l l promote a mild d i s c h a r g e p r o c e s s of c o r e m a t e r i a l r edomina t ly i n t o t h e upper plenum of t h e r e a c t o r It has been showng9 t h a t t h e c o r e m a t e r i a l d e b r i s i s permanently c o o l a b l e w i t h i n t h e r e a c t o r v e s s e l . The mechanical l o a d of t h e r e a c t o r v e s s e l and of t h e pr imary c i r c u i t s i s n e g l i g i b l e . The same c o n c l u s i o n ho lds f o r a r ange of pa ramete r s around t h e b e s t e s t i m a t e v a l u e s .

E n e r g e t i c LOF Accident P a t h s

Because of s t i l l e x i s t i n g u n c e r t a i n t i e s f o r some of t h e key phenomena i n t h e o r e t i c a l u n d e r s t a n d i n g , t h e o r e t i c a l mode l l ing , and e x p e r i m e n t a l v e r i f i c a t i o n of t h e s e models i t h a s been c o n s i d e r e d p ruden t and n e c e s s a r y t o use p e s s i m i s t i c a s sumpt ions f o r t h e a c c i d e n t development ( e n e r g e t i c s ) d u r i n g t h e v a r i o u s a c c i d e n t phases . For t h e i n i t i a t i n g phase a l i m i t i n g e n e r g e t i c a c c i d e n t p a t h ( s e e t h e s m a l l b l a c k a r row 1 i n Fig. 1) h a s been d e r i v e d by making s e v e r a l r a t h e r p e s s i m i s t i c a s sumpt ions 1;e.g. no a x i a l f u e l expans ion r e a c t i v i t y f e e d b a c k ; no f u e l d i s p e r s a l by f i s s i o n g a s e s , sodium vapor , and s t e e l vapor ; i n s t a n t a n e o u s f r a g m e n t a t i o n and mixing f o r fue l / sod ium h e a t t r a n s f e r w i t h a s m a l l f u e l p a r t i c l e r a d i u s ) . This a c c i d e n t p a t h l e a d s i n t o a LOF d r i v e n TOP which was m e c h a n i s t i c a l l y s i m u l a t e d by u s i n g t h e SAS3D code systern31. In a d d i t i o n e n e r g e t i c l i m i t i n g a c c i d e n t p a t h s due t o secondary e x c u r s i o n s have been d e r i v e d o u t of t:he t r a n s i t o r y phase and t h e phase of i n t e g r a l c o r e m a t e r i a l movement ( s e e t h e s m a l l b l a c k a r r o w s 2 & 3 i n Fig. 1) by u s i n g p e s s i i u i s t i c assumpt ions . For t h e t r a n s i t o r y phase r a t h e r c o h e r e n t slumping of i n t a c t f u e l p i n segments i n t o non b o i l i n g r e g i o n s h a s been assumed t o o b t a i n a l i m i t i n g e n e r g e t i c secondary e x c u r s i o n , which was m e c h a n i s t i c a l l y s i m u l a t e d by u s i n g t h e SAS3D code system31. For t h e phase of i n t e g r a l c o r e m a t e r i a l movement c o h e r e n t recompact ion p r o c e s s e s of b o i l i n g p o o l s and c o h e r e n t slumping of r emol ten f u e l s l u g s (from upper c o r e s t r u c t u r e b lockages ) i n t o a b o i l i n g poo l have been c o r ~ s i d e r e d ~ ~ , ~ O t o g e t l i m i t i n g e n e r g e t i c secondary e x c u r s i o n s . The i n i t i a l and boundary c o n d i t i o n s f o r t h e s e l a t t e r c a s e s were d e r i v e d from b e s t e s t i m a t e SAS3D c a l c u l a t i o n s and t h e e n e r g e t i c e x c u r s i o n s were s i m u l a t e d by K A D I S ~ ~ . The SIMMER-I1 code system33 a l t h o u g h a v a i l a b l e i n K a r l s r u h e f o r a m e c h a n i s t i c a n a l y s i s of t h e i n t e g r a l c o r e m a t e r i a l movement phase34 was i n t e n t i o n a l l y n o t a p p l i e d f o r SNR-300 l i c e n s i n g purposes because e x p e r i m e n t a l t e s t i n g of SIMMER-I1 f o r t h e complex phenomena i n v o l v e d i s s t i l l n o t s u f f i c i e n t .

A t t h e end of t h e s e e n e r g e t i c n e u t r o n i c e x c u r s i o n s t h e rhermal e n e r g y c o n t e n t of t h e mol ten f u e l ( ene rgy above t h e l i q u i d u s ' p o i n t ) i s up t o 5500 MJ. But one h a s t o r e a l i z e , t h a t t h e o t h e r c o n d i t i o n s (sodium c o n t e n t of t h e c o r e , s t e e l t e m p e r a t u r e s , e t c . ) a r e q u i t e d i f f e r e n t f o r t h e d i f f e r e n t c a s e s . The p r o c e s s of t r a n s f o r m i n g the rmal i n t o mechanical ene rgy i s s t r o n g l y

dependent 'on these conditions. The mechanical energy potential of the hot molten fuel masses has been conservatively estimated by taking these . conditions into account and by considering the effect of other (partly more effective) working fluids.

In fact, several combinations of hot fuel and one additional working fluid (sodium, steel, fission products) have been evaluated with respect to their mechanical energy potential30. In addition first mechanistic simulations of the ener etic expansion and discharge phase for SNR-type reactors with SIHMER-II'5, taking into account the complex processes involved, show that the mechanical energy (kinetic energy) generated is much less (almost one order of magnitude) than the isentropic fuel work potential. But due to insufficient experimental testing of the SIMMER-I1 code for some of the phenomena involved, these SIMMER results have not been introduced into the licensing procedure of SNR-300. One can conclude that the mechanical energy potential for all of these energetic accident paths is reasonably well bounded by the isentropic fuel work potential, which yields maximum values around 100 MJ, which is well below the design load limit of 370 MJ for the SNR-300, which was set by the licensing authorities as early as 1972.

In spite of the conclusion that the primary vessel and the circuits will withstand the mechanical load generated by these energetic excursions, two other problems need careful1 evaluation:

1.) Core debris distribution and long time coolability within the reactor vessel,

2.) Containment,loading by leaking core material (through un-tight seals and valves) . The first problem has been discussed in some detail2* and it has been

shown that long time coolability can be established within the reactor vessel. Thus, the external core catcher of the SNR-330 has therefore only a backup function.

The analysis of the second problem shows that even for these severe accidents the release of radioactive material will lead to dose values below the maximum values permitted by law9r10.

3.2 Some Remarks about the Low Ramp Rate Transient Overpower Accident

The low ramp rate TOP accident has been analysed taking into account incoherence effects of pin failure in different fuel suba~semblies~~. The best estimate path leads through a relatively mild initiating phase into a phase of early shutdown with in-place coolability of the reator core (see the big hatched arrows in Fig. 1). Some uncertainties related to fuel freezing processes and blockage forruation make it necessary to consider also an accident path leading into whole core involvement (see arrow 4 in Fig. 1). The conditions reached have some similarity with the transitory phase of - - an LOF accident.

4.1 ~ a l c u i a t i o n a l Too l s f o r Heterogeneous Cores

Improved c a l c u l a t i o n a l t o o l s have t o be used f o r t h e a n a l y s i s o f he te rogeneous c o r e s . Because of t h e i r s p e c i f i c compl ica ted geometry l e a d i n g t o t h e presence of many i n t e r n a l boundar ies w i t h i n t h e c o r e r e g i o n t h e a p p l i c a t i o n of d i f f u s i o n approx imat ion i s sometimes n o t v e r y s u i t a b l e and may be a t o o c r u d e method. Thus, t r a n s p o r t t h e o r y may be t h e p roper method t o choose and t h i s c h o i c e may sometimes be mandatory. I n a d d i t i o n , t h e 2-dimensional c y c l i n d r i c a l R-Zmodels which a r e u s u a l l y s u f f i c i e n t f o r t h e c a l c u l a t i o n of c o n v e n t i o n a l c o r e s a r e o n l y a d e q u a t e f o r s c r e e n i n g s t u d i e s o f he te rogeneous c o r e s and p o s s i b l y f o r t h e d e t e r n i n a t i o n of t h e i r f i s s i l e i n v e n t o r y , t h e i r b reed ing p r o p e r t i e s and rough c a l c u l a t i o n s of t h e sodium vo id r e a c t i v i t y , They have t o be r e p l a c e d by 3-dimensional hexagonal-z models t o o b t a i n more a c c u r a t e v a l u e s e s p e c i a l l y f o r t h e power shape , c o n t r o l rod worths and sodium-void c o e f f i c i e n t : ; . But p r e s e n t l y t h e a p p l i c a t i o n of 3-d methods i s u s u a l l y r e s t r i c t e d t o d i f f u s i o n theory . Even then t h e computer t ime r e q u i r e d f o r a c c u r a t e Na-void c a l c u l a t i o n s u s i n g a s u f f i c i e n t number of ene rgy groups may be q u i t e c o n s i d e r a b l e s i n c e t h e convergence behav io r f o r he te rogeneous c o r e s i s u s u a l l y worse t h a n f o r homogeneous ones. Due t o t h e i n c r e a s e d occurenae of i n t e r f a c e s between r e g i o n s of d i f f e r e n t m a t e r i a l composi t ion and t h e p resence of t h e s e boundary s u r f a c e s i n r e g i o n s of h i g h n e u t r o n f l u x and importance, t h e v a l i d i t y o f t h e u s u a l c o n v e n t i o n a l homogenization p rocedures has t o be c o n s i d e r e d more c a r e f u l l y f o r a p roper t r a n s i t i o n from t h e heterogeneous c e l l a r rangement t o a homogenized m a t e r i a l composi t ion which i s oft:en used i n g l o b a l r e a c t o r c a l c u l a t i o n s .

4 . 2 S a f e t y Ana lys i s of SNR-1300 Cores

P r e l i m i n a r y r e s u l t s of s a f e t y a n a l y s i s f o r t h e c o n v e n t i o n a l c o r e d e s i g n h a s been o b t a i n e d com l e t e d r e c e n t l y . It was performed i n a manner s i m i l a r t o e a r l i e r s t u d i e s 16, P 7 f o r a 2000 MW(e) LMFBR. P r e l i m i n a r y r e s u l t s a r e a v a i l a b l e l a f o r t h e CDA i n i t i a t e d by a pump c o a s t down fol lowed by a f a i l u r e of both s h u t down systems. S i m i l a r s t u d i e s f o r t h e he te rogeneous c o r e d e s i g n have been s t a r t e d but r e s u l t s a r e n o t a v a i l a b l e y e t . A f a i r copmparison i s t h e r e f o r e no t p o s s i b l e y e t .

S ince t h e s a f e t y a s p e c t s of a r e a c t o r d e s i g n depend i n a f a i r l y complex and compl ica ted way on i t s n e u t r o n i c , thennohydrau l i c and mechanical c h a r a c t e r i s t i c s i t i s n e c e s s a r y t o perform a d e e a i l e d s a f e t y a n a l y s i s f o r both a l t e r n a t i v e c o r e d e s i g n s . Only a q u a n t i t a t i v e s a f e t y a n a l y s i s of t h e complete a c c i d e n t p r o g r e s s i o n f o r r e p r e s e n t a t i v e a c c i d e n t s can show t h e m e r i t s of d e s i g n d i f f e r e n c e s and t h u s a l l o w s r e l i a b l e c o n c l u s i o n s on t h e s a f e t y performance of t h e r e a c t o r . Fur thermore , one should be aware t h a t a n a c c i d e n t a n a l y s i s approach which uses p e s s i m i s t i c a s sumpt ions t o cover t h e u n c e r t a i n t i e s i n t h e d a t a and t h e o r e t i c a l model l ing might n o t p rov ide a r e a l i s t i c comparison of homogeneous v e r s u s he te rogeneous c o r e a c c i d e n t behav io r and t h i s approach i s , t h e r e f o r e , of l i m i t e d u s e f u l n e s s .

4.3 Role bf Sodium Reactivity Effect in the Accident Analysis

Based on the fact that the sodium void reactivity is one of the most important reactivity coefficients which influence the accident sequence - at least during the initial phase of LOF inltiated CDA - there is a tendency to reduce its magnitude by design modifications. One aims at coming down from

+ about 4-5 $ for conventional large LMFBR cores to values close to 1 or 2 $ for heterogeneous cores. It is difficult to achieve this restriction with reasonable design modifications for radially heterogeneous cores12. However, . there exist possibilities for heterogeneous core arrangements to attain values between 2 and 3 $. Such sodium void reactivities should also be acceptable, because the importance of the sodium void reactivity for the core safety behavior should not be overestimated. Other effects, e.g. positive reactivity effects by voiding of internal fertile subassemblies, steel relocation, fuel slumping and the associated phenomena of freezing and blockage of coolant channels have a dominant influence in later stages of the accident transient. These predominantly positive reactivity effects will still play their dominant role even if the core will be designed for very small positive - or even zero - sodium void reactivity coefficients. In addition, the small positive sodium void coefficient of heterogeneous cores is usually combined with a reduced Doppler reactivity of the fissile region. The Doppler feedback reactivity associated with the temperature increase of the fuel, however, makes a strong, negative contribution to the power-coefficient and mainly determines (for a given reactivity ramp) the energy release during the disassembly phase.

5. ON SOME KEY ISSUES OF SAFETY RESEARCH FOR CORE DISRUPTIVE ACCIDENTS

In section 3 and 4 the current accident analysis procedure for the SNR-300 and the present status of analyses for SNR-1300 have been explained. Uncertainties for some of the key phenomena with respect to accident energetics in the SNR-300 safety analysis have been treated by using pessimistic assumptions for the accident simulation. This procedure is practicable for the SNR-300. For large commercial LMFBR's like SNR-1300 a

. similar procedure could lead to a very expensive plant design. Therefore it is highly desirable to develop a better understanding of the key phenomena.

In this section the influence of some of the key phenomena on accident progression will be briefly discussed and it will be explained which R&D ~

efforts are required for a better understanding of these phenomena.

5.1 Neutron Physics Evaluation of Accident Relevant Material Configurations

In the course of the CDA analysis the reactivity and power distribution must be determined for material configurations which deviate significantly from that of the reactor in normal operation. These configurations are characterized by

cavities and streaming channels resulting from melting, relocation and expulsion of some or all materials from portions of the core compacted fuel and structural material, resulting from the movement of molten materials into neighbouring parts of t:he reactor or from lateral compaction of fuel elements.

Experiments to verify the calculational met:hods for this type of configurations and in particular for the reactivity effects of fuel slum ing were performed in ZPPR 5 - a mock-up of the Clinch River Breeder Reactor f 9 -, and in an assembly in ZPR 9 which was especielly designed to test safety related physics parameters20. Evaluation with both diffusion and transport theory have shown considerable discrepancies between measurements and calculations by non negligible amounts.

In SNEAK (critical facility at the nuclear research center Karlsruhe, Germany), an extended series of experiments on the effect of material movemects in accidents is presently being performed (SNEAK 121, covering a wide variety of configurations for the first time, this type of experiments will be performed in both plate and pin geometry (Fig. 2).

The core of SNEAK 12A is constructed of plates and uses 20% and 35% enriched uranium metal plates as fuel. The experimental program is being performed mainly in the central 16 elements of the assembly. It comprises the following steps:

1) Introduction of cavities of varying sizes up to 8000 cm 3 2) Introduction of streaming channels through the core and through core and

blanket in one and four SNEAK-elements. 3) Slumping and compaction of steel in core and blanket 4) Slumping and compaction of fuel in core and blanket 5) Simulation of molten pools in which fuel and steel are mixed or

vertically separated.

The use of SNEAK plates in the test zone allows independent changes of concentration for fissile and fertile material, structural material and sodium. The streaming effects for plate cells are not typical for fast breeder pin lattices, their order of magnitude, however, can be tested by using horizontal and vertical plate orientations.

The heterogeneity and also the composition of a fast breeder lattice is better simulated in SNEAK 12B where the testzone consists of a Pu02U02 pin lattice (15% Pu enrichment, 8.2 m pin diameter). As of present planning the rods are placed into calandrias, assuming a condition during which the sodium voiding has already taken place. In order to allow a simulation of various types of material movement, the pins in the 16 central elements are vertically subdivided into 7 core sections and a top and bottom blanket section. Pin sections can be taken out of selected regions and be compacted in other ones, reaching 1.5 times the original material concentrations thus simulating material movements in vertical and hocizontal directions during slumping processes and the formation of cavitite!;. According to preliminary estimates, changes of neutron streaming through the pin lattice caused by vertical dislocation of pin sections in the envisaged experiments will bring about only a relatively small fraction of the total reactivity effect. Ifowever, corresponding changes of the streaming properties direction are expected to constitute the main component of

for the radial the effect for

horizontal' fuel movement. Therefore the use of pins for this type of experiments is essential.

The evaluation of the experiments asks for the application of specific transport theoretical methods. A number of methods which are inportant in this context were developed at KFK in recent ears. These include: One and two dimensional transport perturbation codes2Y, collision probability methods for plate and pin cellsz2, a cell code taking into account the conditions at region interface^^^, and a new method to treat streaming through empty channelsz4. These methods will be further developed, particularly in order to allow the treatment of pin geometry and of anisotropic diffusion in all types of calculations which so far is possible only partially. In addition, the currently used KFKINR group constant setz5 will be replaced by an updated, improved version taking into account presently available evaluations of nuclear cross section data.

The test and validation of the applied methods and data in the analysis of the critical experiments described here will establish a more accurate and reliable basis for the nuclear part of LMFBR safety analysis and help to identify the needs where improvements in data and methods could reduce the uncertainty margins in the prediction of the LMFBR core safety behaviour.

5.2 Other key phenomena requiring more R&D-effort

Apart from the mainly neutronic aspects of future safety related research activities mentioned in section 5.1, Table 3 shows a list of several physical key phenomena which require intense future R&D efforts.

The axial fuel expansion due to fuel temperature increase has - if fully effective - a large negative reactivity effect. The CABRI is expected to provide the required experimental data.

Clad material melting an$ movement, occuring separately from fuel melting and movement, is especially important for reactor core designs with a low maximum positive sodium void reactivity effect (e.g. heterogeneous core designs). This clad material movement could cause positive reactivity effects and/or blockages above or.below the active core region. The latter effect could lead to a bottled-up pool configuration and thus complicate - inspite of a relatively mild primary excursion - this later phase of the accident (potential of secondar excursions). Out of pile experiments39* 40, at KfK and in-pile experiment^^^,^^,^^ have already and will be extremely useful. In addition, improved theoretical simulation models have been developed44 at KfK and were incorporated into the SAS3D code ~ y s t e m ~ ~ , ~ ~ .

Early fuel dispersal in sodium-voided fuel-elements is extremely important for assuring a mild primary excursion i.e. for avoiding a LOF driven TOP situation. In-pile experiment^^^,^^,^^ contributed already to a better understanding of this phenomenon. Additional experiments are planned within a cooperative effort between SANDIA and KfK. Theoretical simulation methods for fission gas behavior and fuel breakuplmotion in sodium voided subassemblies are under development at KfK 47,48.

Fue l p i n f a i l u r e , f a i l u r e p ropaga t ion and f u e l movement i n p a r t l ; sodium voided f u e l e l ements t lu r ing LO17 d r i v e n 'COPS w i l l s t r o n g l y i n f l u e n c e t h e e n e r g e t i c s of LOF d r i v e n TOP events49. I n - p i l e exper iments i n CABRI should p rov ide more r e l e v a n t i n f o r m a t i o n f o r t h e s e phenomena i n t h e f u t u r e .

F u e l j s t e e l p e n e t r a t i o n i n t o a x i a l and r a d i a l s t r u c t u r e s t o g e t h e r w i t h a p o s s i b l e i n c o h e r e n t slumping of upper c o r e and b l a n k e t subasembly s t u b s de te rmine t h e a c c i d e n t development d u r i n g t h e t r a n s i t o r y phase; e s p e c i a l l y , t h e s e phenomena de te rmine whether o r n o t a bo t t l ed -up pool s i t u a t i o n w i l l develop. The r e s u l t s of r e a c t o r m a t e r i a l f r e e z i n g t e s t s 5 0 i n d i c a t e t h a t t h e d a t a base i s incomple te and t h a t a more u n i f i e d t h e o r e t i c a l modeling method i s needed. A s t r o n g e x p e r i m e n t a l and t h e o r e t i c a l e f f o r t i s planned a t KfK.

Melt through o r mechanical f a i l u r e of hexcan-walls a r e a l s o i m p o r t a n t f o r t h e e v o l u t i o n of t h e a c c i d e n t d u r i n g t h e t r a n s i t o r y phase because blockage f o r m a t i o n , f u e l d i s p e r s a l phenomena, and f a i l u r e of hexcan-walls ( a l s o f o r c o n t r o l rod subdssembl ies ) a r e competing e f f e c t s and t h e p a t h of t h e a c c i d e n t depends on t h e outceme of t h i s c o n p e t i t i o n . Exper imenta l and t h e o r e t i c a l s t u d i e s w i t h r e a c t o r - p r o t o t y p i c boundary c o n d i t i o n s a r e needed.

T r a n s i e n t b e h a v i o r of b o i l i n g f u e l l s t e e l poo l s i s a n o t h e r key phenomenon. Only r e c e n t l y a p o t e n t i a l l y adequa te t h e o r e t i c a l t o o l , i . e . SIMMER-11, was developed a t Los Namos S c i e n t i f i c ~ a b o r a t o r ~ ~ ~ and was adop ted a t KfK. Scoping c a l c u l a t i o n s by ~ o h 1 3 4 have shown an e n e r g e t i c s p o t e n t i a l f o r bo t t l ed -up pool s i t u a t i o n s . It i s n e c e s s a r y t o e v a l u a t e c r i t i c a l l y whether t h e s e - p a r t l y assumed - i n i t i a l and boundary c o n d i t i o n s a r e p o s s i b l e d u r i n g a c c i d e n t e v o l u t i o n of a s p e c i f i c r e a c t o r des ign . These c o n d i t i o n s a r e c h a r a c t e r i z e d by a bo t t l ed -up pool s i t u a t i o n w i t h a h i g h i n v e n t o r y of mobile f u e l and t h e p o t e n t i a l of c o h e r e n t m a t e r i a l movement. I f such s i t u a t i o n s cannot be avo ided they have t o be i n v e s t i g a t e d e x p e r i m e n t a l l y and t h e o r e t i c a l l y .

SIMMER a n a l y s i s of e n e r g e t i c expans ion and d i s c h a r g e p rocesses35 , 51 i n d i c a t e t h a t t h e t r a n s f o r m a t i o n of the rmal i n t o mechanical ene rgy y i e l d s d r a m a t i c a l l y reduced mechanical e n e r g i e s (by a n o r d e r of magni tude) i f compared t o t h e i s e n t r o p i c f u e l work p o t e n t i a l ( i s e n t r o p i c expansion t o t h e cover g a s volume). Th i s i s v e r y encourag ing , but t h e p resence of sodium and t h e q u e s t i o n whether t h e upper c o r e s t r u c t u r e s s t a y i n t a c t d u r i n g t h e expans ion p r o c e s s need more a t t e n t i o n . An e x p e r i m e n t a l t e s t i n g program was r e c e n t l y i n i t i a t e d a t KfK.

I f i n - v e s s e l c o o l i n g of f u e l d e b r i s shou ld be accomplished f o r l a r g e commercial LMFBRts one needs more i n f o r m a t i o n on p e n e t r a t i o n and m e l t t h r o u g h of s t r u c t u r e s and on t h e behav io r of f u e l / s t e e l p a r t i c u l a t e beds i n sodium. This i s s o because t h e f u e l d e b r i s d i s t r i b u t i o n p r o c e s s w i t h i n t h e r e a c t o r v e s s e l depends v e r y much on t h e t ime needed f o r p e n e t r a t i o n and m e l t t h r o u g h of s t r u c t u r e s and on t h e c o o l a b i l i t y of p a r t i c u l . a t e beds i n sodium. Some remarkable p r o g r e s s h a s been ach ieved on t h e p a r t i c u l a t e bed c o o l i n g i s s u e by some i n - p i l e exper iments a t S A N D I A ~ * . A combined i n - p i l e and ou t -o f -p i l e e f f o r t was r e c e n t l y i n i t i a t e d by SANDIA and KfK.

In ail accident phases a good knowledge of thermodynamic data for reactor mterials (fuel, steel, sodium, fissio; products) is necessary, partly up to relatively high temperatures (e.g. 5000 K). At KfK much effort has been put into experimental and theoretical work for equation of state data of fuel, steel and sodium53* 54, 55.

6. RESULTS AND CONCLUSIONS

. 6.1 The accident analysis must consider the complete path of an accident starting with an initiating event, leading through several phases up to mechanical and/or thermal loading of the vessel, piping and containment systems and up to permanent cooling of core-debris and possibly release of radioactivity. Separate limiting cases must be considered for mechanical loading and for thermal loading of the primary vessel and piping system.

6.2 Some key phenomena of the later accident phases especially phenomena which are crucial for the occurance of secondary excursions need more attention of the experimental and theoretical safety analysists in order to have a more balanced understanding of the whole accident path (a benign belavior for the primary excursion could lead to a more complicated and energetic situation for the later accident phases).

6.3 Cores with a low maximum void reactivity effect (e.g. heterogeneous cores) should not be judged on the basis of limiting energetics accident behavior of the primary excursion alone. The merits of different designs can be judged reliably only on the basis of a complete, thorough and realistic safety analysis. Especially, a heterogeneous core with a low void reactivity worth might encounter a relatively mild primary excursion, but clad material movement, which in low void worth cores precedes the fuel movement, could form axial blockages which might in the later accident phases lead to more complicated and possibly more energetic situations (secondary criticalit; potential of bottled-up pools).

6.4 Structural behavior is not only an issue for the mechanical loading of the vessel and piping system,at the end of an excursion, it is important in all accident phases and should receive more attention during the accident progression period. . 6.5 There is still a large number of challenging problems to be solved for comercially sized LMFBRs, which include for example: neutron physics evaluation of accident relevant configurations, effective equations of state for irradiated fuel, multifield and multicomponent fluiddynamics coupled with space dependent neutron kinetics for analysis of secondary excursions, structural behavior during expansion of extremely hot core-materials, entrainment of sodium into a hot expanding two phase bubble, aerosol behavior and development of filter techniques.

6.6 A strong international cooperation in the experimental and theoretical solution of design and safety problems for commercial-size LMFBRs would be extremely valuable. Besides the BIZET- and RACINE-programs, another

r e c e n t example i s SIMMER e x p e r i m e n t a l t e s t i n g and a p p l i c a t i o n t o t h e l a t e r a c c i d e n t phases where i n t e r n a t i o n a l c o o p e r a t i o n r e c e i v e d more momentum d u r i n g t h e l a s t yea r . SIMMER i s c o n s i d e r e d i n Germany a s a n ex t remely f l e x i b l e and u s e f u l l t o o l which should be f u r t h e r developed and v a l i d a t e d t o improve our unders tand ing of key phenomena and t o r e s o l v e some of t h e remaining impor tan t i s s u e s of c o r e d i s r u p t i v e a c c i d e n t s .

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20. S.K. BHATTACHARWA e t a l . , "A C r i t i c a l Experiment Study of I n t e g r a l Physics Parameters i n Simulated Meltdown Cores", Proc. I n t . Meeting on Nuclear Power Reactor Safe ty , October 1978, Brussels (1978)

21. K. KOBAYASHI, "Description of t h e one- and two-dimensional pe r tu rba t ion t r a n s p o r t code TP 1 and TP2", KfK 2738 and KFK 2787 (1979)

22. P. McGRATH and E.A. FISCHER, "KAPER - A Computer Program f o r t h e Analysis of Experiments performed i n Heterogeneous C r i t i c a l F a c i l i t i e s " , ANS Topical Meeting on Mathematical Models and Computational Techniques f o r Analysis of Nuclear Systems, CONF 730414, Ann Arbor Michigan (1973)

A Recent developments of KAPER: R. BBhme and E.A. F ischer , pr iv. communication

23. R. BOHME, " In t eg ra l e Transpor t theor ie n i t l i n e a r e r Anisotropic zur . " Berechnung der Neutronenverteilung i n endlichen Plattenanordnungen s c h n e l l e r Reaktoren", KfK 2501 (1977)

24. T. YOSHIDA, "Analysis of Sodium Void Experiments i n ZPPR 3 Modified Phase 3 Core", YdK 2668 (1978)

25. E. KIEFHABER, "The KFKINR-Set of Group Constants: Nuclear Data Basis and F i r s t Resul ts of i t s Applicat ion t o t h e Recalculat ion of Fast Zero-Power Reactors", KfK 1572 (1972)

26. J. F. JACKSON e t a l . , "Trends i n LMFBR Hypothet ical Accident Analysis", CONF 740401 (1974), Beverly H i l l s

27. P. ROYL, e t a l . , "Untersuchungen zu K i i h l m i t t e l d u r c h s a t z s t o r f a l l e n ~ i m abgebrann ten Mark 1A Kern des Kernkraf twerkes Kalkar", K•’K 2845 (1979); o r P. ROYL e t a l . , "Analys is of H y p o t h e t i c a l Loss-of-Flow Acc iden t s Without Scram i n t h e SNR-300 End-of-Life Mark 1A Core Using t h e SAS3D Code system", Proc, of t h e I n t e r n a t i o n a l Meeting on F a s t Reac to r S a f e t y Technology, S e a t t l e , August 1979, Vol. .XI, pages: 624-634

28. 1.1. MASCHECK e t a l . , " R e c r i t i c a l i t y C o n s i d e r a t i o n s and Core M a t e r i a l D i s t r i b u t i o n i n t h e Reactor Vesse l of SNR-300 a s Consequence of Unprotected Loss-of-Flow T r a n s i e n t s i n t h e Mark 1 A Core", Proc. of t h e I n t e r n a t i o n a l Meeting on F a s t Reac to r S a f e t y Technology, S e a t t L e , August - 1979, Vol. 11, pages: 721-722

29. VOSSEBRECKER e t a l . , " I n h e r e n t S a f e t y C h a r a c t e r i s t i c s of Loop-Type LMFUR's", Proc. of t h e I n t e r n a t i o n a l Meeting on F a s t Reac to r S a f e t y Technol*, S e a t t l e , August 1979, Vol. 11, pages: -554-566

30. R. FROEHLICH e t a l . , " S t a t u s of Ana lys i s of H y p o t h e t i c a l Core D i s r u p t i v e Accident E n e r g e t i c s f o r L icens ing of SNR-330", a c c e p t e d f o r p u b l i c a t i o n i n Trans. of t h e American Nuclear S o c i e t y , November 1980

31. J.E. CAHALAN e t a l . , "A P r e l i m i n a r y User ' s Guide t o Vers ion 1.0 of t h e SAS3D LMFBR Accident Ana lys i s Computer Code", SR 239831 ( J u l y 1977)

32. P. SCHMUCK e t a l . , "KADIS - Ein Coinputerprogramm zur Analyse d e r Kernzer legungsphase b e i h y p o t h e t i s c h e n S t o r f a l l e n i n s c h n e l l e n , nat r iumgeki ih l ten Reaktoren", KfK 2497 (November 1977)

33. L.L. SMITH e t a l . , "SIMMER-11: A Computer Program f o r t h e Disrupted Core Ana lys i s" , LA-7515-M, NUREGfCR-0453 (October 1978)

34. W. R. BOHL, "Some R e c r i t i c a l i t y S t u d i e s w i t h SIMMER-II", Proc. of t h e I n t e r n a t i o n a l Meeting on F a s t Reac to r S a f e t y Technology, S e a t t l e , August 1979, Vol. 111, pages: 1415-1424

35. P. SCHXUCK, "The P o s t Disassembly Expansion Phase f o r SNR-type Reactors", I s p r a Course on Mul t iphase Processes i n LMFBR S a f e t y A n a l y s i s , May 5-7, 1980

36. P. ROYL e t a l . , "Analys is of H y p o t h e t i c a l (Overpower Acc iden t s i n t h e SNR-300 Mark 1A Core w i t h Modell ing of F a i l u r e Incoherence", Nucl.Eng.. and Design, 43, (1977) , pages: 239-248

. 37. A.B. ROTHMANN e t a l . , "TREAT Experiments w i t h I r r a d i a t e d Fuel S i m u l a t i o n

H y p o t h e t i c a l Loss-of-Flow Acc iden t s i n Large LMFBK's", Proc . Int .Meet ing on F a s t Reac to r S a f e t y Technology, S e a t t l e , August 1979

38. J. DADILLON e t a l . , "The CABRI F a c i l i t y Experiment Fue l P i n Program - I t s O b j e c t i v e s and P r e s e n t R e s u l t s Obtained U n t i l Now", proc.lng.Meeting on F a s t Reac to r S a f e t y Technology, S e a t t l e , August 1979

39. R.E. HENRY e t a l . , "Cladding R e l o c a t i o n Experiments", Proc. I n t . Meeting on F a s t Reac to r S a f e t y and R e l a t e d P h y s i c s , Chicago (Oktober 1976) , CONE-761001-P4

40. V. CASAL, "Messung von Wechselwirkung zwischen Rieself i lmen und e i n e r Gasgegenstromungt', PSB-Bericht 197911, 43

41. B.W. SPENCER e t a l . , "Summary and Evaluat ion of R-Series Loss-of-Flow Safe ty Tes t s i n TREAT", Proc.Int.Meeting on Fast Reactor Sa fe ty and Related Phys ics , Chicago (Oktober 1976) CONF-761001

42. B.W. SPENCER e t a l , , "Cladding Motion and Blockages i n R-Series Safe ty Experiments", Trans. Am. Nucl. Soc. - 19, 238 (1974)

43. G. HOPPNER e t a l . , "TREAT-R5 Loss-of-Flow Experiment i n Comparison wi th v SAS P r e t e s t Analysis", Trans. Am. Nucl. Soc. - 18, 213 (1974)

44. G. ANGERER; "Modelltheoretische Untersuchungen des Abschmelz- und Wiedererstarrungsvorgangs von Brennstabhiillen wahrend S t o r f a l l e n i n schne l l en natriumgekiihlten Reaktoreu", KFK 2662 (August 1978)

45. G. ANGERER, G. ARNECKE, A. POLCH, "SAS3DC - Ein Programmsystem zur Analyse von Stor fXl len i n schnel len natriumgekiihlten Brutreaktoren" KfK-report t o be published

46. D.H. WORLEDGE, G.L. CANO, "Study of t h e Dispersive P o t e n t i a l of I r r a d i a t e d Fuel Using In-core Experiments", Proc.Int. Meeting on F a s t Reactor Sa fe ty Technology, S e a t t l e (August 1979)

47. E.A. FISCHER, L. VXTH, "The Karlsruhe Approach t o Modeling F i s s ion Gas Behavior f o r Fast Reactor Accident Analysis", Transact ions of t h e ANSIENS Meeting i n Hamburg, May 1979, . TRANSAO -t 31 P. 367

48. K. THURNAY, "KANDY - Entwurf e i n e s Rechenmodells zur Beschreibung d e r Materialbewegungen i n einem e n t l e e r t e n Kiihlkanal des Schnellen Briiters", unpublished

49. H.U. WIDER, "The PLUTO-2 Overpower Excursion Code and a Comparison wi th EPIC", Proc.Int.Meeting on Fas t Reactor Sa fe ty Technology, S e a t t l e

a (August 1979)

50. B.W. SPENCER e t al.. "Summarv and Evaluat ion of Reactor Mate r i a l Fuel Freezing Tests" , Proc. 1nt.Meeting on F a s t Reactor Sa fe ty Technology,

A S e a t t l e (1979), Vol. I V , p. 1766

51. C.R. BELL e t a l . , "Advances i n t h e Mechanistic Assessment of Postdisassembly &erget ics1 ' , Proc.Int.Meeting on Fas t Reactor Sa fe ty . Technology, S e a t t l e (August 1979); Vol. 1, pages 207-218

52. R.J. LIPINSKI, J.B. RIVARD, "Debris Bed Heat Removal Models: Boi l ing and Dryout with Top and Bottom Cooling", ANSIENS 1nt.Meeting on F a s t Reactor Sa fe ty Technology, S e a t t l e (August 1979)

53. M. BOBER et a l . , " Inves t iga t ions of thermodynamic d a t a of s t a t e of f a s t r e a c t o r co re m a t e r i a l s f o r hypo the t i ca l acc ident ana lys i s . - Theore t i ca l and experimental Work a t Karlsruhe -", IWGFR Meeting on Equations of S t a t e of Mate r i a l s of Relevance t o the Analysis of Hypothet ical Fas t Reactor Accidents, Harwell, UK, June 19-23, 1978

54. M. BOBER, E.A. FISCHER, "Out l ine of t h e p r e s e n t p o s i t i o n on t h e Thermophysical Data used a t Kar l s ruhe i n H y p o t h e t i c a l Accident Analys is" , IWGFR Meet ing on Equationlj of S t a t e of Mate r i a l s ' of Relevance t o t h e ~ n a i y s i s of H y p o t h e t i c a l Fas t Breeder Reactor Acc iden t s , ~ a r w e l l j U K , June 19-23, 1978

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PHASE OF EARLY SHUTDOWN

ENERGETIC D l S A S S E f i B L Y

DISCHARGE

ACCIDENT HEAT REXOVAL

BEST E S T I P A T E LOF ACCI l lENT PATH

PHASE OF CONTA1Nb:ENT -3 BEST EST I i l A T E

TOP ACCIDENT PATH

-I, ENERGETIC PRIMARY EXCURSIONS

PHASE OF RADIOLOGICAL ENERGETIC SECONDARY EXCURSIONS

ENERGETIC SECONDARY EXCURSIONS

-5.- F A I L U R E OF IN -P IACE-COOLING

F I G U R E 1: PHASE DIAGRAM FOR HYPOTHETICAL CORE D l S R l l P T l V E ACCIDENTS (SNR-300)

SNEAK 12 A MOLTEN POOL CONFIGURATION

.empty tubes

.37 or 38 pins per cl

.25 pins per clement

cment

(normal

SNEAK 128 TESTZONE AFTER RADIAL DISLOCATION OF CENTRAL PINS

Fig . 2

loading )

--- .. IUdl . I'ower d is l r ibu l ion u i l h i n the core 1 .0 : O D 4 . .- nreedlnq ra t io 1 .02 z 0 . 0 6

Absorber rod

in core cen ler t ...... 0 . 7 5 . 1 . 2 5

D n ~ p l e r ~ e l l e c l 0 8 2 :: 0 03

T a b . 2

4. I T A L Y

Suxx~ary of Reactor Physics Activities in I t a l ~

(Octobsr 1959 - Scpterbei- 1980)

R. Mzrtinelli

The Italian activities on fast reactors are a part of the French-

!::alian prograx, both as concerns the research activities carried

out by CNEN and :he industrial activities carried out by NIXA.

-. :.ntegral Zx?eriments. The implementation of the common CEA-CNEX- - IZ;"K/IA "iL4CIX /I/ has been c~mpleted. First criticality for a

configuzation incor2orating fvro blanket regions (central island + 1 ring) :.'as achieved in March 1980.

?he pzoqrmne, started with measurenears of reaction rats traver-

ses and of spectral indices in fissile and fertile regions, is

progressing with sodiuin void reactivity measurements in various

core recjions (including power distribution measurements in corre-

s?occ?ence of local voidingsl.

1'--,-,roved calculation methods for the interpretation of negative

reactivity measuremelts (by modified source mdltiplic2tion and

rsd-drop techniques) have been inplemented.

- dlankets. The first measurement caqaign of the "NEFERTITI" Frs- - <:rz?rne - a parmetric study of the ~le~tronic parameters of blan- ksts - has been compl~ted. The 35 crr, thick blanket region is fed

by a l a r g e f i s s i l e i n t e r 2 a c e d r iven hy t h e l a s t source r e z c t o r Th

PIKO.

A p r e l i a i n a r y a n a l y s i s of t h e resul .ks o:f r e c c t i o n r a t s and spec.^..

t r a l i n d i c e s d i s t r i b i l t i o n s h ~ i i s t h . e przsence of s i c j n i f i c s n c d i -

s c r s p a n c i e s between ca l cu l aked ( 0 . . 5; S ; C N I N 2 V A L : - . i V 25-qyniip 4

c r o s s s e c t i o n s ) 2nd measured (pr-oton--recoi.l c o u n t e r s j s p e c t r a , 2 : ~

both buf f e r - b l a n k e t and b lanke t - , re f lec t i i r r i x h e r f a c e s . Such discri.: ...

panc ie s a r e c o n s i s t e n t w i t h t hose o b s e i - ~ ; - ~ d f o r f i s s i o n r a t e r a t i o

d i s t r i b u t i o n s i n s i d e the b l a n k e t r e g i o n . F u r t h e r work t o i d e n t i f y

e r r o r s o u r c e s is i n p r o g r e s s .

Sh ie ld iny . The propaga t ion experimemts i n sod iun - . s t a i c l e s s s t e e l .

heterogeneous m i x t u r e s made i n TAPIR@ /Z/ have been r e - i n t e r p r e t e d

u s i n g t h e common CZA-CXEX sh ie ld- in? "formill.aire" P3OT3lliZ 7 / 3 / . :::'hi<

g e n e r a l t r e n 6 i s an unc is res t ixa te of t h e exper-imental d e t e c t . 0 ~ z.cti.

v a t i o n s , c o k e r s n t l y w i t n . t h e r e s u l t s

exper iments i n pure Sodiuin

A des ign -o r i en t ed wcrk has

s i m p l i f i e l r e c i p e s f o r t h e

t i o n i n a pool- type DWBR.

b locks .

been completed

calcl;?.a'iion of

of a p r e ~ i o u s s e r i e s of

I n t h i s work a complete two d i x ~ e n s t o n a l d e s c r i p t i o n of t h e sys?:enr

i n c l u d i n g c o r e , s h i e l d i n g and sodi~zm up ::o h e a t ~ i ~ c h a n g e r s , was

coupled t o Local h e a t exchanger Monte--Carlo c a i c u l a t i o m . T h i s r e

f i n e d c z l c u l a t i o n w e r e used t o deduce slimpl.:ifl.ed exp res s ions trbicii~

t a k e i n t o account t h e coup l ing o f r a d i a l propagakicn i n t.he BE znc;

i ts f i n i t e c y l i n d r i c a l s t r u c t u r e . The res-ul.ts of t h e work i n d i c a t e

t h a t a s u b s t a n t i a l r e d u c t i o n ( a f a c t o r of 1 . 4 2) of t h e c a l c u l a t e d

a c t i v a t i o n can be ob ta ined u s i c g the proposed methods i n s t e a d of

t h e s t anda rd one-6iinensional models. Moreover, t h e accuzacy of the

s i m p l i f i e d f o m l a t i o n proposed is such t h a t t h e u n c e r t a i n t y due t o

t h e c ~ l c . u l a t i o n approximat ion - k s reduced from a t y p i c a l 1 0 0 8 t o a

2 0 % va lue . . .. . r '. : . . > ; , ; :; ;

r !. : . ,', ,,< ,, : : , .'

Strcctural Materials. The measurement caxpaign of integral czptu - - re cross sections of Fe, Ni, Cr and SS in hard spectra - represen - tative of a conmarcia1 LIQRX /5/ - is in progress at the coupled critical facility RB-2/TV; ics completion is expected for Novem-

- ber 1980. A very good internal consistency among measurements has

been obtained so far.

Experiments are carried out by using the null-reactivity oscilla-

tion tech~ique, complemented by microchamber measurements of 5-10

capture and U-235 fission rates.'

.Cora Dynamics. The 2-D core dynamics code NKGYP-3 is being imple-

mented at CNEN. NADYP-3 is a new version of NADYP-2 / 6 / , allowing

some di f fe ren top t ic rnsmcalcu la t ion strategies: the modular sche-

me adopted permits the utilization of new phenomenological models.

NADYP-3 is also presently utilized in order to develop the new 3-D

code EDIPO in the near future.

Some modules are now available:

a) the program GICCASTA /7/ which calculates the peutron and

thermoydraulic steady-state conditions before the accident;

b) the program GIZLUO /8/ for void fraction calculations during

0 two-phase boiling;

C) a "spring-model'' module for bubble boiling calculations /5/

d) a nodule for fuel and cladding time-dependent temperature cal - culations, using different numerical methods.

Design Activities for the PEC Reactor. A re-calculation of the - critical enrichment has been made, which takes the last changes

in c:ore design and Pu isotopic composition into account /9/. Bias

fact.ors have been determined for radial power distributions /lo/

by t.ransposing ".PECORE " results with SS and Ni reflectors, to

the design calculation method.

2 . LIGET WATER FGACTORS

The Caorso BWii power p l a n t has reached t h e norninal powex (840 ~i~,\.i.c) .

S t a r t u ~ aad s p e c i a l t e s t s have been completeci i n 1 3 7 9 : t h e r e o u l t s

o f c o a t r o l rod c a l i b r a t i o n , tmpera t ?z re c o e f f i c i e n t , 1.ocal cri.tici.i . ..

l i t y and 2r;utron f l u p r o f i l e neasurements a r e r e p o r t s d i n /:2,';

conpared t o t h e r e s u l t s obtainei! from t h e c o r e s i z u l a t o r .

Post - i r ; raci ia t ion e x a ~ i n a t i o n s of f u e l p i n 2nd corl:rol ro-7 sanY L -.C3 .~ ..

r 'rc3 Gari i j l iano BWR and Tr ino P:C3 a r s iil p r o g r e s s : t h e r e s u l t s of

t h e s e sezsurements w i l i complete t h e e x t e n s i v e measurexent campziq:!c

c;ev.'FseZ by E E L i n o r d e r t o v z l i d a t e t h e per fomzr lces of i t s c a l c c

l z c i o z a l methods / 13 /

A 3-9 coarse-mesh code f o r s t eady- s t ake PWR s L ~ u i a ; i i o n , d e v e l c ~ i - L

by CISS / I ? / , i s i n t h e v a l i d a t i o n phase. The n e u t r o ~ nniodel i.s ;:,?--.

s e d on 2 p.olynornialmode1 which a l lows t h e f i n e f l u x di.stxibutior:e

i n s i d e a s i n g l e node t o be r e c o n s t r u c t e d . By app ly ing the ove r l ap - .

p i n y e f f e c t t e z h n i ~ e , one can c a l c u l a t e s i n g l e p i n powers.

3 . HEAT/? FiATZ2 REACTORS

Ths problem of c a l c u l a t i n g t h e r e a c t i v i t y w o r t h & a c o n t r o l rod

perpenci icular t o t h e c o r e axis (CANCU--.tY'pe "ac ; ju s t e r4 ' and "zone

c o n t r o l l e r " roi!s) has been d e a l t w i th . A c a l c u l a t i o n method / 1 5 /

based on the code chain GAM-GATHER-COMBINE-ANISNP and of HETROIS

(heterogeneous diffusion) has been set up. These contro1,rod ty-

pes have been sinulated at different axial positions in experi-

ments made at the BR-3 critical facility. The comparison of cal-

. culated andmeasureddata shows an agreement generally better than

5% except for the regions where the rodworthis very low.

4. - FUEL CYCLE STUDIES

In cooperation with J.R.C. Ispra, detailed calculation of the neu - tron transmutation of actinides in a commercial LMFBR have been

completed /16/. Both homogeneous and heterogeneous recycling ha-

ve been considered, and their influence on the local and integral

characteristics of the host reactor has been calculated.

By following the indications in the existing literature on the ar - gument, about the R an D needs in neutron transmutation, a feasibi - l.ity study of experimental irradiation of actinide containing fuel

pins in a fast •’1- facility is in progress /17/.

/ I Y.H. BOUGET, P. H-, R . CONYETISANO, R. WTINELLI, F.HEW, . W. SCHLOLTYSSEK: "Main characteristics of the Racine Program me developed by DEBEN, CNEN and CEA on ~ S U R C A for the hetez rogeneous core concept studies", IAEA-SM-244/29, Aix-En Pro- vence, Sept. 24-28 (3979)

/ 2/ M. CARTA, A. DE CARLI, y. RADO, M. SUYATORES, J . P . TRAPP: "Propagation Experiments in Sodium -Stainless Steel Mixtu- res at Tapiro", to be presented at a Specialists' Meeting on Nuclear Data and Benchmark for Reactor Shielding, Paris, Oct. 27-29 (1980)

J.C. ESTICT, X . SAL-JATORES, J.P. TRAP?: "Basic Nuclear Datz and the Fast Reactor Shieldirig Des:.g.n Formulzire PROPANE Do"; ?roc. 1nt.l Conf. on Nuclear Cross Sections 2nd Technology, Knoxwilie (1 979)

G. PAulIOTTI, V. R3D0, M. SALVATOXS: "Nonte-car1.o Valzea- tion of Secondary So?.imn Activation in a Pool - Type L > l F 3 R U , presented ac '1980 Advances in Reac:ror ?hysics and Shielsing', Sun Vailey, Sept. 14-18 ( 3 980)

P.M. AZZONI, A. SUONONI, P.L. CFIICIDI, C. GIULIANI, R. XAR- VASI: "Measurcvents of structuxai, rrlateriais capture to U-235.- Fission rate ratios in intermediate and fast spctra", SM-244/64, Aix-en-Provence, Septa 2i!-,28 (1979)

A. GALATI, A. MUSCO: "Dinamica dei reattori veloci.. 3) 11 codice NP3YD-2". RT/FI (79) 24 CNEN Report

A. GALATI: "The code GIOCASTA" , W-FI (80) . CNEN report in press

A. GAWTI, F. NO-LLI: C~~munication at the Liquid Metal Bailing Working Group 4-6/6/80 - Rome

/ 9 / G. DCMINICI, 3. TAVONI: "Massa critia del PEC", CNEN TIT-CD- 00053, April 1980

/lo/ G. DCMINICI, R. TAVONI: "Correzioni sistenatiche ai valori di poterza del PEC", CNEN-VT-CD-00053, FARCE 1980

/ I I / J.C. CABRILWT, R. CONVXTISANO: "Optimization of a Radially Eeterogezeous Core As a Function of Penetration Strategy", presented at '1980 Advances in Reactor Physics and Shielding', Sun Valley, Sept. 'I 4-18 ( 3 980)

/12/ S. GRIFONI: "Special Tests at CACRSO Startup for BWR simulator Methods Veri5icationn, paper presented at this meeting

I S. E'ELICT, P. PEROXI: "aperients-T3-Theory Feesback in the Developent of Core Performance i;?ialuatj.on Ne~thods", paper presente?. at this meeting

/14/ G. 3OTTONI et al.: Paper presented 3t '1930 Advances in 3eac .- tar Physics and Shielding', Sun Valley, Sept. 14-38 (1980)

/Ij/ B. CAF24IGNANI: "Messa a pnto di un ri~etodi di calcolo per bz r - re di controllo perpendicolari agli elementi di combustibile", CNEX Xeport RIT/FIS (LFS2) 79-AI 1

/ 1 6 / G. OLIVA: "Neutronic Transmutation of Transuranium Isdtopes in a Fast Breeder Power Reactor", presented at the Second Technical Meeting on the Nuclear Transmutation of Actinides, Ispra, (Apr. 1980)

/17/ L. TONDINELLI: "Irraggiamento di capsule contenenti barrette con attinidi in un reattore veloce sperimentale", CNEN inter - nal report in press.

Reactor Physics Activities in J-

October 1979 to August 1980 --

J. Hirota

Fast and Thermal Reactor Physics

1. Preliminary Benchmark Tests of JENDL-2 (.I)

Compilation of JENDL-2 is now under way. At its first step the highest

priority was put to evaluation of most important nuclides for fast reactors; 235", 238", 23gPu,

240~u, 241~u, Cr, Fe and Ni, responding to an urgent

request to use JENDL-2 for analyses in the JUP1:TER project, joint USA-Japan

mock-up experiments of large fasr reacrors in ZPPR facility. The evaluation

of the 8 nuclides above was completed in Novemt'er 1979. The benchmark tests

were made on these evaluated data. The data of JENDL-1 were used for other

nuclides. This combined library is called JENDL-2B library.

The following results were obtained:

1) JENDL-2B gives keff = 0.998 with standard deviation of 0.65 %. The

discrepancy between the Pu cores and the U cores is 0.2 %.

2) The ratio of 2 3 8 ~ fission to 2 3 5 ~ fission is overestimated bv 6 % on an

average. The ratio of 239~u fission to 2 3 5 ~ fission is underestimated

by 1.2 % but is 1.5 % higher than JENDL-1. The C/E values are satis-

factory for both the ratios of 2 3 8 ~ capture to 2 3 5 ~ fission (0.99) and

of 2 3 8 ~ capture to 239~u fission (1.00).

3) As to central reactivity worths, the C/E values of JENDL-2B are satis-

factory for 2 3 5 ~ (1.01) and 2 3 8 ~ (1.03). ::he worth of 'OB obtained

from JENDL-2B is underestimated by 9 % and becomes 3.5 % lower than

that from JENDL-1. This difference is caused by the core neutron spec- .

trum, because the cross sections of JENDL-1 were used for 'OB.

4 ) Doppler coefficients are very satisfactorily predicted with JENDL--2B,

while JENDL-1 overestimates them by 10 % and JAERI-Fast-I1 underesti-

mates them by 10 %.

From these tests, it can be concluded that JENDL-2B predicts various

characteristics of fast reactors better than JENDL-1 as a whole. However,

the following problems should be further investigated; the overestimate

in the ratio of 2 3 8 ~ fission to 2 3 5 ~ fission, 2) the slight underestimate

in the fission rate ratio of 239~u to 2 3 5 ~ and 3) the underestimate of 'OB

worth.

(1) Kikuchi, Y., Narita, T. and Takano, 1%. : J. Nuc. Sci. Technol., z, 567 (1980)

2. Self-shielding Effect of Inelastic Scattering Cross Section of Iron on

Neutron Spectrum

In the multigroup constants set such as the ABBN, JFS-I1 or LIB-IV,

the self-shielding effect for inelastic scattering cross sections is neg-

lected by assuming small energy variation of the cross sections. This

assumption may be valid, except for iron of which the energy fluctuations

for total and inelastic scattering cross sections are remarkable below 2

MeV, where is the dominant part of neutron spectrum in fast reactor.

. Furthermore, iron is a very important structual andlor reflector material

in fast reactor.

The self-shielding factor of inelastic scattering cross sections of

iron were calculated from the evaluated nuclear data of JENDL-2 and ENDFIB-

I". The self-shielding effect on the neutron spectra was studied for the

MZB and ZPR-3-54 assemblies''). The self-shielding effect gives a harder

neutron spectrum because of decreasing the neutron slowing down power, and

is distinguished especially in steel reflector. The radial reaction rates

calculated considering the self-shielding effect become smaller for 235 b f and 232f, by about 8 % but larger for 23%f by about 5 % in the reflector

region while the effect on reaction rates is very small in the blanket

region . Hence, the inelastic scattering self-shielding effect should be

considered in the reactor shielding calculation and also in the analysis

of neutron spectrum experiments on the iron assembly.

(1) H. Takano, and K. Kaneko, : to be published in Nucl. Sci. Eng.

3. Analysis of Fuel Slumping Experiment on FCA Assemble VIII-2

A series of experiments have been made on FCA Assembly VIII-2 to

examine the calculation method for the reactivity effect due to axial . displacement of fuel/cladding'l). In this study, emphasis was placed on

the systematic measurement of reactivity changes in simple configurations

rather than the simulation of a possible accident sequence. The analysis

has been made using JFS-11. In order to check the adoptability of calcula-

tion methods used in analysing core disruptive accidents, the results obtained

by the three methods, transport (Sn), conventional and modified diffusion '2) methods, were compared with the measured ones .

The calculated reactivity change with the conventional diffusion

theory underestimates considerably the experimental one when the neutron

streaming effect is large. The conventional diffusion calculation

mispredicts 2 3 8 ~ fission rates both in the fuel compacted and voided region.

No s i g n i f i c a n t improvement i s achieved by t h e use of the modified d i f f u s i o n

c o e f f i c i e n t i n t h e voided region , al though i t cons iderably improves t h e

p r e d i c t i o n of r e a c t i v i t y change. The r e s u l t s obta ined by the t r a n s p o r t

theory wi th the S P approximation f a i r l y we l l zgree with the measured va lues 4 0 f o r r e a c t i v i t y change a s we l l a s r e a c t i o n r a t e s . However, t h e r e s t i l . 1 remains

the tendency of underest imation of r e a c t i v i t y change which inc reases wi th ex-

tens ion of the f u e l slumping region t o t h e core edge.

(1) Nakano, M. , Tsunoda, H. and Hi ro ta , J . : F i s s i o n Rate and Sample Worth

Eleasurement i n Simulated LMFBR Meltdown Cores, JAERI-M ( t o be published)

(2) Tsunoda, H . , Nakano, M. and Hiro ta , J . : Analysis of F i s s i o n Rate and

Sample Worth measured i n Simulated LMFBR M e l t d o n Cores, JAERI-M ( t o

be published)

4 . SRAC : A Standard Computer Code System f o r L a t t i c e C e l l and Core

Calcula t ions on Reactor Design and Analysis

A code system SRAC i s being developed a s the nuclear des ign and a n a l y s i s

p a r t of t h e JAERI s tandard thermal r e a c t o r code system, which aims t o supply

the s tandard d a t a and methods f o r r e a c t o r des ign and a n a l y s i s and t o p resen t

the r e fe rence va lues with high accuracy t o conventional methods.

The SRAC system is designed t o a l low a v a r i e t y of usages descr ibed a s

fo l lows :

1) The foundamental group cons tant l i b r a r y produced from t h e ENDF/B-IV

nuclear d a t a f i l e has t h e energy group s t r u c t u r e of 107 groups ( 4 5

groups f o r thermal and 74 f o r f a s t energy rsnges , r e s p e c t i v e l y , wi th

12 groups over lapping) . The l i b r a r y holds t h e Bondarenko type t a b l e

f o r resonance s h i e l d i n g f a c t o r s and keeps, a d d i t i o n a l l y , t h e resonance

a parameters on the energy range between 130.7 t o 0.4 eV f o r t h e o p t i o n a l . use of t h e in t e rmed ia t e resonance approximation ('") o r t h e r igorous

c a l c u l a t i o n using the c o l l i s i o n p r o b a b i l i t y method(3). The thermal

group-to-group t r a n s f e r ma t r i ces a r e compiled on t h e t e n ass igned tem-

pe ra tu res .

2 ) The user c o n s t r u c t s h i s own l i b r a r y from t h e foundamental l i b r a r y f o r

t h e d i s i r e d n u c l e i and temperatures wi th h i s own-energy group s t r u c t u r e ,

where t h e thermal cu to f f energy i s a l s o chosen from t h e overlapping energy

range.

3) The SRAC system can so lve t h e f ixed source problem i n t h e p a r t i a l energy

range, which is mainly used f o r c e l l c a l c u l a t i o n , and can a l s o so lve t h e

eigenvalue problem i n t h e whole energy range which is used i n c e l l and/or

core calculation.

A variety of the transport codes are available for cell calculations

(collision probability method with 14 type of geometries4), ID-SN, 2D-SN).

The one space-point solution by B or P1 approximation is also available 1 after smearing the cross sections.

For core calculations, the above mentioned transport codes and the I-, 2-

and 3-D diffusion codes are available.

Smearing and/or collapsing macroscopic cross sections is done separately by

the user's selection.

Double heterogeneity such as the grain effect in the burnable poison rod

or the pin rod in a channel box can be treated by utilizing smeared cross

sections obtained from homogenizing the microscopic heterogeneity into the

macroscopic heterogeneity.

The several PDS (Partitioned Data Set) files are used for the data storage.

The built-in FACOM utility program is ready for the file control (LIST

DIRECTORY, DELETE, CONDENSE, COPY, RENAME), and a few service programs are

ready to readlwrite the contents of these files.

Ishiguro. Y, and Takano. H, : J.Nucl.Sci. Technol., 6,380 (1969)

Tsuchihashi.K., Ishiguro.Y, and Kaneko.K, : Nucl. Sci. Eng., 3, 164 (1979) Ishiguro. Y, : PEACO-I1 : A Code for Calculation of Effective Cross Sections

in Heterogeneous Systems, JAERI-M 5527 (1974)

Tsuchihashi.K, : LAMP-B : A Fortran Program Set for the Lattice Cell Analysis

by Collision Probability Method. JAERI 1259 (1979)

5 . Double Finite Element Method to Solve the Three-Dimensional Neutron Transport

Equation

A double finite element method algorithm, in which both angular and spatial *

variable domains are approximated by the finite element subdomains, has been

formulated to solve the three-dimensional multi-group neutron transport equation . in the framework of a Vladimirov-Ritz-Galerkin type approximation. The spatial

domains are treated in the same way as in the FEM-BABEL code"), in which com-

binations of prism and box shaped spatial elements can be used. The application

of the finite element method to the angular domains mitigetes the ray effect.

In order to make the total number of meshes(energyx spacex angle) as small as

possible, the bicubic spline polynomials are employed as the angular bases, -, -+-

following the arguements given by Kaper et a1"). The angular flux u(r,n )

symmetrized in the angle 8 is expressed as

The angular bases are decomposed further to

n ./ .@i = A fC(w -TP ( 3 where 0 5 (p5 7 and C 5 < 1 , 14 and 72 being the cubic spline functions.

The programming work of the DFE-3 code based on this algorithm is now in

progress.

(1) Ise T., Yamazaki T. and Nakahara Y. : FEFI-BABEL: A Computer Program for

Solving Three-Dimensional Neutron Diffusion Equation by the Finite Element

Method, JAERI 1256 (1978)

(2) Kaper H.G, Leaf G.K. and Lindeman A.J. : Application of Finite Element

Methods in Reactor Mathematics. Numerical Solution of Neutron Transport

Equation, ANL-8126 (1974)

6. Measurement of Reactivity Worths of Burnable Poison Rods in SHE-14 Core

As a core design for the Experimental Very High Temperature Gas Cooled

Reactor (VHTR) progresses, evaluation of the design precision has become increas-

ingly important. A Simulation core for VHTR, SUE-14 was assembled using the

graphite moderated 20 % enriched uranium Semi-Homogeneous Experimental Critical

Facility (SHE), in order to obtain experimental data useful for evaluating

the design precision. The VHTR is designed to accommodate burnable poison and

control rods for reactivity compensation. Experimental burnable poison rods

which are slmilar to those to be used in the experimental reactor were prepared,

and their reactivity values were measured in SHE-14 core. The experimental

burnable poison rods were made by inserting 114 of absorbing pellets in hollow

graphite rod of 1180 mm in length. The absorbing pellets with 8 mm diameter

and 10 mm height were formed by sintering B4C particles with graphite powder.

One to three rods of the above experimental burnable poison rods were

inserted into the central column of the SHE-14 core, and the reactivity values

were measured by the period and fuel rod subsitution method. It is clearly

shown that due to the self-shielding effect of B4C particles the reactivity

value decreases with increasing particle diamenter. For the given particle

diameter, the reactivity value is found to increase linearly with the logarithm

of boron content. The measured values and those calculated are found to agree

with each other within 5 %.

7. Probabilistic Method for Evaluation of Reactivity Margin of Experimental

VHT R

A probabilistic method is proposed

possibility that the safety criteria in

into consideration the uncertainties in

to evaluate in the core design stage the

reactivity margin are satisfied, taking

design calculation. In application of

the method to the design study of Experimental VHTR, investigations were made

on the relation between the design accuracy and the probability that the safety

criteria in both one rod stuck shut-down and operation margins are satisfied.

In conclusion, with the correlation disregarded, the ratio of the standard

deviations to the design values of the MARK-I11 core must be less than 0.79 %

and. 5.3 % respectively for the cold clean effective multiplication factor and

each of the reactivity worths of control rods, burnable poisons and core tem-

perature rise, in order that the probability is larger than 99.7 % (three times

the. sigma limit). In the analysis, the uncertainties of the reactivity worths

are. assumed to be equal each other. With the correlation regarded, the ratios

must be considerably smaller.

(1) Kaneko, Y. : Pfobabilistic Method for Evaluation of Reactivity Margin of

Experimental VHTR, JAERI-M 8847 (1980)

8. VHTR-GCFR Symbiotic Energy System

A symbiosis energy system between GCFR and VHTR apperas to have substantial

promise as an energy system self-sufficient in fuels which can produce electici-

ty and high temperature process heat. In this system, the GCFR has to supply

sufficient amount of plutomium to keep the reactor going, and in addition

produce 2 3 3 ~ necessary to the associated Th-233~ fuelled VHTR. In this context,

a strategy analysis for the symbiosis system has been made using the computer

programme NUCRES on the basis of a presumption of growth in energy supply and

nuclear power contribution in Japan in the time range from 1980 to 2050. The

purpose of the analysis is to clarify energy sufficiency, breeder capacity, ,

acc:umulated natural uranium ore, separation work and reprocessing requirements

in the energy production system with GCFR-VHTR symbiosis. The VHTR considered *

in the symbiosis is of 3000 MWt grade and is charaterised by high conversion

ratio of 0.9. The analysis indicated that the GCFR-VHTR symniosis has a very

promising feature.

An optimisation study is under progress for finding out such a GCFR with

thorium blanket that has good plutonium and U-233 breeding performance necessary

to keep the above mentioned symbiosis working well, through a neutronic and thermo-

hyclraulic analysis. The neutronic analysis is made using two dimensional dif-

fusion codes, CITATION and PHENIX, together with a set of 26-group cross sections

generated from JENDL-2 and ENDFIB-IV.

9. Application of Exponential Experiment to High Subcriticality Determination

From the viewpoint of criticality safety, there arises a need to develop

becter methods to evaluate high subcriticality of fuel systems moderated with

light water. For this purpose, an exponential experiment technique was exam-

ined on H 0 moderated 2.6 w/o enriched U02 1att:ices at the Critical Assembly TCA. 2 The fuel rods were arrayed in nxn, where n=17, 15, ..., 3, with square lattice pitch of 19.56 mm. The central fuel rod was replaced by an A1 tube of 18 mm

outer diameter, in which a BF counter was traversed to measure the vertical 3 neutron flux distribution and 252~f of 5.5 mCi was used as a neutron source.

The active height of the lattices was fixed at 122.5 cm which was the critical

water level of the n=17 lattice. The multiplication factor k was determined eff

under the one-group diffusion model with continuously slowing-down sources.

For comparison, k was also measured by a pulsed neutron source tecnique. eff The criticality calculation was performed using KENO-IV with cross sections

of 137-energy groups prepared from ENDFIB-IV. The results of exponential experiment agreed well with those of KENC-IV in the wide range of k from 0.94 to 0.45.

eff

a On the otherhand, the results of the pulsed neurron source experimenrs show a

marked tendency to saturate in the subcritical state of k (0.95. eff indicates that the exponential experiment is an effective technique

keff of a highly subcritical system, when M~ of the system is known

methods and there is an sufficient region of uniform composition in

direction.

This

to determine

by some

at least one

10. Dynamics of Coupled-Core Thermal System

To verify the validity of the moderator region response function method (1,2)

measurements were carried out with a pile oscillator on the Kyoto University

Critical Assembly (KUCA), which is a light water moderated and reflected assem-

bly. Varying the distance of the two rectangular cores the kinetic parameters (3) of the system were measured and compared with the theoretical values , as a

joint activity of Nagoya University with Research Reactor Institute, Kyoto

University.

The coupling coefficient, the delay time, a,; well as the frequency response . agreed well with the values calculated by the one-group, one-dimensional version

of the above method. The generation time of individual cores proved to be 30 %

smaller than the theoretical values. The fluctuation measurements and pulse

decay measurements are underway on KUCA to obtain the same kinetic parameters.

The concept of local core reactivity as well as t:he local core generation time

appears useful in investigating nuclear criticality safety of fuel installations.

A joint research activity of Nagoya University wi.rh JAERI is underway in this

direction.

(1) Shinkawa, M., Yamane, Y., Nishina, K. and Tamagawa, H.: Nucl. Sci. Eng.

(2) Yamane, Y., Shinkawa, M. and Nishina, K: Nucl. Sci. Eng. - 72, 244 (1979).

(3) Yamane, Y., Tanaka, K., Nishina, K., Tamagawa, H. and Shiroya, S:: Nucl.

Sci. Eng., to be published (1980).

11. Study of the Stochastic Point Reactor Kinetic Equation

Recently, the effect of fluctuating reactivity noise in point kinetic '

equation was investigated by Quabili and ~arasulu") using Bouret's approxi-

mation and logarithmic linearization. Once ~otoh'~) had atempted to solve the

stochastic point reactor kinetic equation using a diagrammatic method. The

method has been applied to a white noise for which the exact result was obtained (4) by Akcasu and ~arasulu(~), and also to a non-white noise .

It is shown easily that the condition the mean power does not diverge is

a t h e same one obtained by Akcasu and Karasulu for a white noise. The variance

to mean-squared ratio can be obtained by integrating the power spectral density.

It is analitically integrated for a white noise and the result is exact. To

evaluate the variance to mean-squared ratio for a non-white noise, the approxi-

mate solution was integrated.. The following noise correlation function was used

to evaluate the variance to mean-squared ratio.

GK A - 4 @(t) = e t

2 t

It is found that the diagramatic method applied to the stochastic point

react:or kinetic equation is successful to yield the exact result known for a

white noise. For a non-white noise, the variance to mean-squared ratio

decreases as the correlation time of noise increases, and for the short corre-

e a t i o n time of noise the numerically evaluated value approaches to the exact

one for a white noise.

(1) Quabili, E.R. and Karasulu M. : Ann. Nucl. Energy - 6, 133 (1979) (2) Gotoh, Y.: Ann. Nucl. Energy 2, 119 (1975) (3) Akcasu, A.Z. and Karasulu, M.: Ann.nuc1. Energy 3, 11 (1976) (4) Gotoh, Y.: 13th Informal Meeting on Reactor Noise, Cadarache, 7-9 May,

Fusion -- 12. Measurement of ( n , ~ y ) and (n, in) Cross Sections for 15-Mev Neutrons (1)

Cross section measurements were made for (n,iy) and (n,Zn) reactions of

Mo, Ti and Ni at 120" with respect to the incident neutron at Kyoto University.

The uonodirectional source neutrons were obtained by the associated particle

method using D-T reaction. The (n,i y) and (n, ;C n) data were simultaneously

measured by an NE-213 scintillator using a pulse-shape discrimination circuit.

The photon s e l f abso rp t ion i n t h e sample was cor rec ted by t h e r e s u l t s of Monte

Carlo c a l c u l a t i o n s .

The (n,Xy) c r o s s s e c t i o n s obtained i n t h i s experiment agreed w e l l with

o t h e r experimental d a t a . Followings were pointed o u t f o r t h e eva lua ted d a t a

i n the ENDFIB-IV f i l e :

1 ) The ( n , % ~ ) c ross s e c t i o n f o r N i and t h e (11, L n ) d a t a f o r Mo a r e considered

t o be adequate.

2) The (n,ZY) c r o s s s e c t i o n s f o r Mo and T i a r e not adequate.

3) The ( n , X n ) c r o s s s e c t i o n s f o r T i and N i seem t o have l a r g e e r r o r s .

The more d e t a i l e d measurements f o r t e s t i n g (n, Ln) c r o s s s e c t i o n a r e neces-

s a r y , because t h e est imated e r r o r s f o r t hese d a t a i n t h i s work a r e too l a r g e t o

l ead t h e f i n a l conclusions.

(1) Shin, k. e t a l . : J. Nucl. Sc i . Technol. 17 , 531 (1980) 0

-

13. Angular Flux Spect ra from Lithium and Graphite Slabs

Angular f l u x s p e c t r a from l i t h i u m and g r a p h i t e s l a b s were measured t o t e s t

nuc lear da t a and c a l c u l a t i o n a l methods f o r D-T fus ion r e a c t o r neu t ron ic s i n

Osaka ~ n i v e r s i t ~ " ) . For t h e numerical c a l c u l a t i o n , one-dimensional d i s c r e t e

o r d i n a t e s t r a n s p o r t codes (ANLSN and NITRAN (')) were used. The multi-group

c ross s e c t i o n s processed wi th SPTG4Z from ENDFIB-IV were used a s common nuclear

d a t a base. Comparison wi th the experimental s p e c t r a showed t h a t t h e angular-

dependent c r o s s s e c t i o n s f o r non-e las t ic s c a t t e r i n g a v a i l a b l e i n ENDFIB-IV were

q u i t e i n s u f f i c i e n t and t h a t t h e an i so t ropy of the s c a t t e r i n g could not be calcu-

l a t e d wi th ANISN which u t i l i z e d t h e s c a t t e r i n g ke rne l s generated by i n c o r r e c t

t reatment of s c a t t e r i n g kinematics i n t h e process ing codes. F a i r l y good agree-

ment between measurements and c a l c u l a t i o n s was a t t a i n e d by t h e use of t h e NITRAN

system wi th t h e appropr i a t e process ing codes of i n e l a s t i c s c a t t e r i n g a n i s o t r o p i e s , - a s t h e a d d i t i o n a l d a t a f o r d i f f e r e n t i a l i n e l a s t r c c r o s s s e c t i o n s were taken i n t o

the ENDFIB-IV da ta . However, more d i f f e r e n t i a l i n e l a s t i c d a t a a r e needed.

Genera l iza t ion of the NITRAN system i s under way t o inc lude S codes and a Monte n

Carlo code.

(1) Yamamoto, J . , e t a l . : J. Nucl. Sc i . Technol. , - 1 7 , 255 (1980)

( 2 ) Takahashi, A. and Rusch, D . : KfK-283211, I1 (1979)

14. S e n s i t i v i t y Analysis of F i s s ion Rates i n a Graphite Ref lec ted Lithium

Oxide Assembly.

I n the concept ional des ign of JAERI's experimental f u s i o n r e a c t o r JXFR,

l i t h i u m oxide (Li 0) i s adopted f o r a s o l i d s t a t e b lanket breeding m a t e r i a l s . 2 235", 238,, 232

The abso lu te f i s s i o n - r a t e d i s t r i b u t i o n of Th and 2 3 7 ~ p were

measured wi th micro- f i ss ion chambers i n a n assembly which s imulated a g r a p h i t e

r e f l e c t e d l i t h i u m oxide blanket") . I n o r d e r t o g e t a c l e a r e r understanding

i n t h e i n t e r p r e t a t i o n of t h e r e s u l t s of t h e measurement and c a l c u l a t i o n , t h e

s e n s i t i v i t y c a l c u l a t i o n of 2 3 5 ~ and 2 3 8 ~ f i s s i o n r a t e s on each c o n s t i t u e n t

c r o s s s e c t i o n a t each measured p o s i t i o n were performed by t h e computer code (2) SWANLAKE .

t The c r o s s s e c t i o n s e t was processed from ENDFIB-IV d a t a us ing NJOY code.

The c a l c u l a t i o n a l model of t h e L i 0-C assembly was assumed t o be s p h e r i c a l 2

symmetry, uniform composition i n each r eg ion wi th an i s o t r o p i c D-T neut ron

source a t t h e c e n t e r . The forward and a d j o i n t angular f l u x e s i n t h e assembly

a r e c a l c u l a t e d by ANISN code wi th t h e P5-S64 approximation. These angular

f l u x e s were app l i ed t o SWANLAKE code based on t h e p e r t u r b a t i o n theory . The

r e s u l t s of s e n s i t i v i t y c a l c u l a t i o n were examined through a i n t e g r a t e d sens i -

It i n d i c a t e s t h a t t h e discrepancy of t h e C/E va lues may be p a r t l y a t t r i b u t e d

t o t h e u n c e r t a i n t y of 12c c r o s s s e c t i o n da ta .

(1) Maekawa H., e t a l . : 3. Nucl. Sc i . Tecnol., 16, 377 (1979).

(2) Oyama Y . , et a l . : "Cross Sec t ion S e n s i t i v i t y Analysis of 235 U and 238 U

F i s s i o n r a t e s i n a Graphite Ref lec ted Lithium Oxide Assembly", JAERI-M

8870 (1980) ( i n Japanese) .

15. Discrepancy between Ca lcu la t ion and Experiment of F a s t Neutron Spec t r a

The neut ron spectrum from a g r a p h i t e assembly bombarded wi th 14-MeV

ne t rons was measured us ing an NE213 s c i n t i l l a t i o n spec t rometer system i n t h e (1) Research Laboratory f o r Nuclear Reactors , Tokyo I n s t i t u t e of Technology .

It w a s compared w i t h t h e s p e c t r a c a l c u l a t e d wi th ANISN and MORSE-GG us ing

ENDFIB-IV and r ecen t measured c r o s s s e c t i o n s . The r e s u l t s of t h i s s tudy i n d i c a t e

that. t h e e f f e c t of t h e an i so t ropy of t h e i n e l a s t i c s c a t t e r i n g c r o s s s e c t i o n is - very l a r g e , and t h e c o n t r i b u t i o n of t h e he igher- leve l i n e l a s t i c s c a t t e r i n g is

cons iderable t o t h e spectrum below 7 HeV.

Regarding t h e f a s t neut ron spectrum i n a g r a p h i t e r e f l e c t e d l i t h i u m assembly,

t h e examination on t h e sources of d iscrepancy between c a l c u l a t i o n and experiment

a r e continued by Nagoya Univers i ty . A s t o t h e ana lyses of experiments, t h e

unfolded energy spectrum o u t of NE-213 s c i n t i l l a t i o n pu l se h e i g h t spectrum i s

t h e s u b j e c t of i n v e s t i g a t i o n . The in f luence of nuclear d a t a and l i g h t emission

data on t h e response mat r ix used i n t h e unfold ing procedure was analyzed. An

a t tempt is being made t o develop an unfold ing method based on a n inconvent ional

approach. A s t o t h e c a l c u l a t i o n s , a modified v e r s i o n of 05R r e v e a l s t h a t t h e

discrepancy near 9 MeV and 5 MeV can be a t t r i b u t e d t o t h e incomplete nuclear 7 . . data of L1 r n e l a s t i c s c a t t e r i n g .

(1) Sekimoto, H., et al. : Measurements and Calculations of the Neutron

Spectrum from a Graphite Assembly Bombarded with 14-MeV Neutrons, Bull.

Res. Lab. Nucl. Reactor, 5, 9 (1980)

16. Development of PALLAS-TS (1)

The code PALLAS-TS has been developed for analyzing fusion blanket experi-

ments. A 120-group library of neutron group constants was prepared using the

ENDFIB-IV file and the process code PROF-GROUCH-GII. The energy range is from

16.487 MeV to 0.32242 eV, and data of 29 nuclides were processed. Not only

elastic scattering but also inelastic scatrering at discrete levels were treated

in anisotropic model. mile inelastic scattering in the continuum level region

and other reactions like (n, 2n) were assumed to be isotropic. For the 121st

(thermal) group, constants are to be prepared for each problem by a cell calcu-

lation code like SRAC.

The double differential scattering kernels were calculated without

Legendre's polynomical expansion. As to the calculation of azimuthal angle

weight in the scattering kernel, it was found that the 1.-f~nciton'~) can be 1

directly derived by differentiation of a formula of spherical trigonometry. The

transport equation was solved with the direct integration method for one-

dimensional, plane or spherical, multi-regional geometry. In this step of pro-

gramming, the original PALLAS-PL, SP method (3) was modified, that is, the

ordinary concept of the group theory was engaged so that the group flux is repre-

sented by the integrated value in each energy range of the group. By this

modification, an iteration procedure was introduced using neutron balance in

every group.

A test calculation was performed for a lithium-carbon, four-region, spheri-

cal system. The results showed a satisfactory agreement with a P5-S64 ANISN

calculation and experimental datac4). CPU time was about 15 minutes on FACOM/

M200 computer. Further development of this code is planned to deal with 2-

and 3-dimensional geometries and to equip them with gamma-rays calculation.

Suzuki, T., Ishiguro, Y., et al. : PALLAS-TS, A One-Dimensional Neutron

Transport Code for Analyzing Fusion Blanket Neutronics, to be published

in JAERI-M report.

Takahashi, A., et al. : J. Nucl. Sci. Technol. - 16, 1 (1979).

Takeuchi, K. : PALLAS-PL, SP: A One-Dimensional Transport Code, Papers

of Ship Res. Inst., No. 42 (1973).

Seki, Y. : private communication.

Construction of FNS and OKTAVIAN

The accelerator for Fusion Neutronics Source (FNS) had finished installation

and adjustment at GIC factory after some delay in schedule due to design moditi-

cation of the telemetry system, bearing failure of the motor alternator and

reshaping of 90 degree analyzing magnets. The factory demonstration test was

performed in this February. For the combination of 740A high current ion source

and the A beam transport, 40 mA and 14.5 mA of H beam were obtained at insertable

Faraday cups just after the acceleration tube and in fromt of target respectively.

The corresponding numbers were 6 mA and 2.8 mA for the combination of 820 low

current ion source and the B beam line. The results were a little lower than

the specified values mainly because of insufficient beam axis alignment. The

assembling of the accelerator was begun at the end of April in ENS site and the

first beam passed to the target position in the A beam line at the end of

June. The beam adjustment to get the rated current and pulse specifications

is now in progress. So far, 12mA of H and 7mA of D target beam have been ob-

tained for the A beam line. A tritium removal system for vacuum exhaust was

completed and tested with satisfactory result. Experimental equipments such as

a pneumatic transport device for sample irradiation and a matrix assembly to

bui.Ld in simulated fusion blanket have been completed. A TOF neutron spectrum

measurement system is now under installation.

The installation of the accelerator OKTAVIAN is nearly completed, and the

final adjustments and the characterization of specific parameters for operation-

al performance are being carried out now at Osaka ~niversit~") . The neutron

production is scheduled at the end of 1980 fiscal year after the completion of

two type tritium collector systems. From the beginning of the next fiscal year,

it :is expected to start the co-operative experiments with research groups from

other national universities, and to keep the annual total operation period over

1800 hrs. - I L

On the way to attain the stable intense neutrons above 10 D-T neutrons/

sec, various technical difficulties must be overriden. The stable operation of

a high current duoplasmatron ion source, the beam profile monitoring for high

beara currents, especially for nano-sec pulse beam, diagnosis of target condi-

tion under beam irradiation and the development of long-life rotating targets

are still the typical R & D items. In addition to above items, the development

of suitable tritium recovering methods for such an accelerator system, and the

tritium monitoring instrumentation against unexpected contamination increase

and intense gamma ray backgrounds by induced activity are considered to be

essential.

(1) Sumita, K., et al. : Proc. 3rd Symposium on Accelerator Science and

Technology, pp.53 to 54, August 27 to 29, 1980, at Research Center for

Nuclear Physics, Osaka Univ., Japan

18. Nuclear Design of INTOR-J

Nuclear design of INTOR-J"), the J~ERI proposal for the International

Tokamak Reactor (INTOR) has been carried out using ANISN, DOT-3.5 and MORSE-I.

A 42 group-neutron 21 group-gamma coupled cross section library for 40 nuclides

GICX~O'~), which is based on ENDFIB-111 and IV for neutron data and POPOP4

Library for gamma-ray production data, was used.

The 1D calculations were used in preliminary calculations and parametric

surveys on tritium breeding(3) and magnet shielding design. Neutron and gama-

ray fluxes, radiation dmanage rates and nuclear heating rate distributions were

calculated with cylindrical models. Gamma dose rates after reactor shutdown

were also calculated using 1D model and THIDA code system(4). THIDA is a code

system which calculates - Transmutation, Hazard potential, lnduced activity, Dose rate and - Afterheat. It consists of the followings: Induced activity calculation

code; activation chain, activation cross section, radionuclide gamma-ray energy/

intensity and gamma-ray group constant files; and gamma-ray flux to exposure

dose rate conversion coefficients. The 2D calculations were carried out to

investigate the effect of neutron streaming through the divertor channel. As

a result, the streaming effect is found to cause about two times as much nuclear

heating rate in the magnet than the case withouc divertor channel.

Poloidal distributions of 14 MeV neutron flux, helium production, displace-

ment damage and nuclear heating rates in the first wall system were calculated ( 5 )

using MORSE-I. The peaking factors of 14 MeV neutron flux and helium production

rate distributions were found to be about 1.3 and those of displacement damage

and nuclear heating a little smaller. The peaks were always in the outboard

side of the first wall. An albedo Monte Carlo method was developed in order to ( 6 ) evaluate the neutron streaming through neutral beam injector (NBI) ports .

Preliminary results of the evaluation suggest that nuclear heating and radiation

damage in the NBI seems to be tolerable but the induced activation necessitates

remote operation for the repair and maintenance of NBI components around the

ion source.

(1) Sako, K., Tone, T., Seki, Y., Iida, H., Yamato, H. et al.: Engineering

Aspects of the JAERI Proposal for INTOR (I) and (II), JAERI-M 8503 and

8518 (1980)

(2) Seki, Y. and Iida, H.: Coupled 42-Group Neutron and 21-Group Gamma Ray

Cross Section Sets for Fusion Reactor Calculations, JAERI-M 8818 (1980)

(3) Iida, H. and Seki, Y.: Studies on Increasing Tritium Breeding Ratios of

JXFR and INTOR-J Blankets, JAERI-M 8896 (1980) ." . . ..'. . (in Japanese) ;.; $> ,.., .<- i.) ; . .

I i d a , H. and I g a r a s h i , P l . : THIDA-Code System f o r Calcuk: : ' . . ' - .

l a t i o n of t h e Exposure Dose Rate Around a Fusion Device-, JAERI-M

8019 (1978) ( i n Japanese)

I i d a , H . , Seki , Y . , Yamamoto, T. and Kawasaki, K.: P o l o i d a l Dis t r ibu-

t i o n s of Neutron Flux, Radia t ion Damage and Nuclear Heat ing Rate i n a

F i r s t Wall System of INTOR-J, JAERI-M 8517 (1979)

Yamauchi, M., I i d a , H. and Murata, T.: Ca lcu la t ion of Neutron and Gamma-

Ray Streaming Through t h e Neut ra l Beam I n j e c t o r P o r t of INTOR-J, t o be

published i n Proc. 1 1 t h Symposium on Fusion Technology, 0xford.Sept.

15-19 (1980)

Shic!Iding - 19. A Computational Method f o r An i so t rop ic Transmission Problems by SN-

Transport Code (1)

I n many deep p e n e t r a t i o n and s t reaming c a l c u l a t i o n s , t h e angular f l u x

d i s t r i b u t i o n i s a b a s i c q u a n t i t y f o r e v a l u a t i n g t h e r a d i a t i o n c u r r e n t i n s h i e l d

reg ions . I n o rde r t o eva lua te t h e a c c u r a t e angular f l u x i n a medium having an

a n i s o t r o p i c source o r a s t r o n g streaming, i t w i l l be necessary t o use t h e method

whic:h can c a l c u l a t e t h e r a d i a t i o n t r a n s p o r t wi thout us ing t h e Legendre expansion

f o r express ing t h e a n i s o t r o p i c components. To perform more p r e c i s e e v a l u a t i o n of

t h e angular f l u x e s , a new one-dimensional S - t r anspor t code, D I A C , has been N

developed. The method app l i ed i n DIAC makes use of t h e p r o b a b i l i t y func t ion

P . t o e l i m i n a t e t h e f i t t i n g e r r o r due t o t h e f i n i t e Legendre ex- n m , j f + j pamion . The s c a t t e r i n g sources a r e g iven by i n t e g r a t i n g d i r e c t l y t h e p r o b a b i l i t y

a func:tion about energy E., and s o l i d ang le V m f + . The p r o b a b i l i t y func t ion J +j

P i s generated from t h e group-to-group-transfer c r o s s s e c t i o n s a s a f u n c t i o n of

sca t : te r ing ang le us ing RADHUT-V4 code system.

I n o r d e r t o g ra sp t h e c h a r a c t e r i s t i c s of t h e method, some t e s t c a l c u l a t i o n s

were c a r r i e d out and compared wi th t h e e x a c t s o l u t i o n s and t h e r e s u l t s c a l c u l a t e d

by ANISN-JR. The exac t s o l u t i o n s were obta ined on t h e problems of one-group

sca1.ar f l u x e s i n an i n f i n i t e medium wi th a n i s o t r o p i c po in t source and a p lane

source. Good agreements a r e obta ined between t h e DIAC r e s u l t s and t h e exac t

s o l u t i o n s . The angular f l u x e s c a l c u l a t e d by ANISN-JR show nega t ive va lues i n

some ang les f o r t h e problems conta in ing t h e s t r o n g an i so t ropy of source. However,

t h e angular f l u x c a l c u l a t e d by DIAC does n o t show any nega t ive va lue f o r a l l

d i r e c t i o n s .

(1) Yamano, N., Koyama, K. and Minami,K. : J. Nucl. S c i . Techno., 16, 919 (1979).

20. Measurement and Ca lcu la t ion of Radia t ion Streaming through Annular Ducts (1)

Radia t ion s treaming through a s t r a i g h t annular duct i n an asymmetric

configuration and an annular duct with one bend were measured by activation

detectors and thermoluminescent dosimeters. Experimental results were obtained

in terms of reaction rate and dose rate. Thermal and epithemal neutrQn fluxes

were also obtained from the measured reaction rates under the assumption of 1/E

spectrum in the intermediate energy region.

A calculation was made for the former asymmetric configuration with the

aid of PALLAS ZD, a two-dimensional transport code, in which two steps of

calculation were taken, that is, firstly, the calculation for angular flux

distribution in water, and secondly, the calculation for the duct with sym- . metrical boundary condition obtained from the first step calculation.

Although there was a discrepancy in the attenuation rate between calculation

and measurement, agreement was quite fair for the fast neutron flux distri-

bution along the duct axis. In the case of lower energy neutron flux distri- e bution along the axis, the calculation overestimated within a factor of 3,

though fairly good agreement was obtained with respect to the attenuarion

rate.

(1) Miura, T., Takeuch, K. and Fuse, T.: Measurement and calcuLation o f

Radiation Streaming through Annular Ducts, Rep. Ship Res. Inst.,

16 [6], 17 (1979).

21. Neutron Skeaming Calculations Using Coupling Monte Carlo and Monte Carlo

Technique (1)

( 2 ) Two neutron steaming benchmark experiments were analyzed by using a

coupling Monte Carlo and Monte Carlo technique. These experimental configura-

tions have the characteristic that the distance between the neutron source

region(JRR-4 (Japan Research Reactor-4) core) and the mouth of the slit or the

annular duct in which the threshold detectors are installed is long and the

interval is filled up with water.

The coupling Monte Carlo and Monte Carlo technique is investigated so as

to enhance the probability of which neutrons come into the mouth and to in-

crease the collision density in and around the slit or the duct. In the first

Monte Carlo calculation of the coupling calculation, a pseudo disk is placed

as a detector in front of the mouth of the slit o-r the anuular duct. The

disk source serves as the plane source in the second Monte Carlo calculation. -, +-

The radial distributions of total flux I@ (r), energy flux @ (r) and angular + -+

g g flux @ (r,Q ) in the pseudo disk are reserved to make the probability density

functions for the second Wonte Carlo calculation. All the calculations were

performed with the Monte Carlo code MORSE. Fairly good agreement between

measured and calculated values was obtained for the thresold detectors which

detectors. However, the disagreement between measured and calculated values

was relatively large for the thresold detector which has the response to

epithermal neutrons, such as Cd covered Au(n, 'T) detector.

(1) Ueki, K.: Variance Reduction Techniques Using Adjoint Monte Carlo

Method and Monte Carlo-Monte Carlo Coupling in Deep Penetration Problems,

Monte Carlo Seminar-Workshop, ORNL/RSIC-44, (Apr. 1980).

(2) Tanaka, S. et al.: Shielding Benchmark Problems (II), JAERI-M 8686, (1980)

22. Fast Neutron Specta Transmitted through Iron and Sodium Slabs (1)

Transmitted neutron spectra were measured for Fe slabs of 9.6 and 19.2 cm

thick and 1 and 2 canned Na slabs of 15 cm thick (thickness of the stainless-

steel can was 3 mm). The neutrons from the core of fast neutron source

reactor "YAYOI" of the University of Tokyo were collimated and incident to

the slabs. The intensity and the profile of the incident neutron beam was

determined by a preliminary experiment. The transmitted neutrons were measured

on the collimated axis and at the polar angle of 30" and 45' to the axis. A

point-to-point Monte Carlo calculation was carried out using elastic and inelastic

scattering data taken from ENDF/B-IV. The results of the calculation showed

good agreement to the experimental spectra.

(1) Shin, K. et al.: 3 . Nucl. Sci. Technol., 17, 37 (1980)

23. Transport Calculation of Gamma Rays Including Bremsstrahlung

For transport calculation of gamma rays including bremsstrahlung, an

improvement of PALLAS-PL,SP was made. The electrons resulted from Compton

scattering, pair production and photoelectric effect are individually evaluated

using the data of primary gamma flux calculated with the code. Secondary

gamma-ray production due to radiative energy loss of electrons, bremsstrahlung,

is calculated by applying the continuous electron slowing-down model. For

this purpose both the electron stopping power and the differential cross

section of the bremsstrahlung production were evaluated.

Comparisons of PALLAS calculations with experiments were made to test

the validity of this code and method. As a result, it has been observed that

the PALLAS calculations result in fairly good agreement with measurements of

the energy spectra transmitted through lead, iron and concrete for a plane

isotropic 6.2 MeV gamma ray source and with another measurement of trans-

mission dose attenuations through lead for a plane monodirectional perpendicu-

lar incidence of 8 MeV gamma rays. Buildup factors and energy specta for

lead and tungsten including bremsstrahlung were also calculated for a plane

monodirectional perpendicular incidence of 8 MeV gamma rays, and they were

compared with the results obtained by the moment method, in which any

bremsstrahlung was not considered. I t was found t h a t t h e c o n t r i b u t i o n of

bremsstrahlung to t h e dose bui ldup f a c t o r i s aoout 44-55 % wi th in 15 m.f.p.

of lead and tungs ten .

:,L :%x.,L.~: ,-L-244 JAi3AN (Contd. j

PNC N241 80-13

Reactor Physics Activities Relevant to FBR and ATR Programs in PNC, Japan

October 1979 .L September 1980

T. INOUE Power Reactor and Nuclear Fuel Development Corporation, Tokyo

Since last meeting of NEACRP, reactor physics activities have been performed in PNC, Japan, to support FBR and ATR development programmes.

The experimental fast reactor JOYO is operated at the 75 MW normal operation cycle. On the prototype fast breeder reactor MONJU, the preparation of safety evaluation is now being progressed by the government.

A prototype reactor of heavy water-moderated, boiling light water-cooled, pressure tube type, FUGEN, is operated at the rated power of 165MWe.

2. Experimental Fast Reactor, JOYO 'I' - The power ascension program to 75 MW which was ceiling

capacity of the present core configuration (MK-I core) started at the besinnina of Julv 1979 and was completed toward the end -

a of ~u~ust-1979. - Concerning performance characteristics of the reactor,

the experimental results so far obtained through these - performance tests have been in general satisfactory, being in

good agreement with predicted values.

The 75MW normal operation began from January 1980, after the annual inspection by the regulatory body. The normal operation at 75 MW is scheduled for about one and half years in order to accumulate sufficient technical data with the present core.

3. Prototype Fast Breeder Reactor, MONJU - In fiscal years 1978 and 1979, the preparatory manufactur-

ing design was under way and in fiscal 1980 orders are placed to manufacturers in line with the design of some modifications and detailed specifications in order to be made to get ready for safety evaluation and subsequent start of construction.

The basic specifications have been established at present and the design including balance of plant is now being finalized incorporating the results of the relevant research and development.

The Shiraki area on the Peninsula of Tsuruga, Fukui Pre- fecture, approximately 250 miles west of Tokyo, was decided as the site of Monju and talks were held with the local authorities to obtain their understanding. With the support of various circles, the meteorological, seismic and geological surveys of the site have been completed and the enviornmental and national park evaluations are conducted by the government and also by the local government. The safety evaluation report has been provided.

4. ~esign Study of the Demonstration Fast Breeder Reactor

The phase I studies of the conceptual design have been performed for a 1000 MWe LMFBR Demonstration Plant, the start of whose construction will come after the start-up of MONJU. The principal objectives are to determine the concept of a demonstration power plant on the bases of the design techniques accumulated by the experience of MONJU and JOYO, and to review its appropriateness. Research and development required to develop the Demonstration Fast Breeder Reactor are also indicated.

5. Heavy Water Moderated Soiling Light Water Cooled Reactor,

FUGZN (165 &We), a prototype reactor, of heavy water moderated, boiling light water-cooled, pressure tube type, is operated at the rated power since its commercial operation. Core average ex;?osure of its 1st cycle was about 6,000 klwd/MT. 1651QSe HWEI-FUGZN generated 1.0 billion kilowatt hours, which corresponds to 265 full power day, with the load factor of 72.4 per cent in 1979 fiscal year. The anual inspection began on the 1st Februaxy 1980 and the 36 fuel assemblies were changed with new one.

The 600 MWe demonstzation reactor wi1l be the second step of the FUGEN project, and its conceptual. design is already completed, wherein the emphasis has been. put on such aspects as safety, operational reliability, maintenability and economy of the nuclear power plant. In reactor physics, the interest has been focused on flexible use of PUWX and U02 fuels.

. - 6. Fast Reactor Physics Activities

1) Evaluation of Nuclear Date ( 3 )

242m Am and 2 4 2gAm Evaluation of neutron nuclear data for

was ~erformed in the energy range of eV to 20 MeV. ' For 42mAm, resonance parameters were used up to 3.5 eV .to represent the cross sections. Experimental data for the fission cross section were reproduced by spline functions up to 1.5 keV, and by a semi-empiricalformula up to 20 MeV. Other cross sections were estimated by taking into account of structure of the fission cross section below 1.5 keV, and were calculated with the optical and statistical models from yL.5 keV to 20 MeV. Cross sections for the (n,2n) and (n,3n) reactions were obtained with Pearlstein's method. Optical potential parameters were determined so that they might reproduce the neutron strength function and the compound nucleus formation cross sections suitable for the fission cross section calculation. F ~ r ~ ~ ~ g A r n , there are no experi-. mental data except thermal energy. The fission and capture cross sections were assumed to be l/v form below 0.225 eV, and the elastic scattering cross section to be a constant. a Above 0.225 eV, the fission cross section was estimated from .that of 242mAm, and the (n,2n) and (n,3n) reaction cross sections were assumed to be the same as those of 242~Am. Other cross sections were calculated with the optical and statistical models. Angular distributions of elastically scattered neutrons were calculated with the optical model for both 242mIun and 2u9Am. Those of the inelastic scattering, (n,2n) and (n,3n) reactions were assumed to be isotropic in the center-of-mass system. Furthermore, ; was given for both states. The results of evaluation will be compiled in JENDL- 2.

2 ) Assessment of Nuclear Reactor Constants Sets

The group constants library for fast reactor calculation, JFS-342 was produced by using the processing codes PROF.GR0UCH.G-I1 and TIMS 1 from the evaluated nuclear data file JENDL-2. The ~resentlv ~roduced JFS-3-J2 differs from :he JAERI-Fast set ~JFS-2) ,*as to concept of group constants, as follows: - --

(1) Group structure: Seventy energy group = 69 groups for the energy range from 10 MeV to 0.414 eV with equal lethargy of 0.25 and one thermal energy group. (2) Weighting spectrum: REMO-correction by using collision density for a typical fast reactor core. (3) Self-shielding factor table: Temperature = 300, 800, 2100 and 4500•‹K; 00 = 0, 1, 10, loL, lo3, lo4, lo5 and lo6 barns; shielding factor, fin for inelastic scattering cross section; temperature dependent self-shielding factors for structural materials. (4) Scattering matrix: Scattering matrices for (n, n') and

(n, 2n). The benchmark test for the presently produced library,

JFS-342 was performed. The assessment for JFS-342 was of the same degree as that for JFS-2.

3) Development of Core Analytical Method

For the analysis of sodium-void experiments in fast critical assemblies direction-dependent diffusion coefficients are often utilized to accurately treat the neutron streaming effect. A unified diffusion coefficient is derived which can be applied to both fuel cells and control rod positions,(4) although the diffusfion coefficients derived by Benoist and Seki are currently used for fuel cells and control rod positions, respectively. This diffusion coefficient is obtained by applying transport theory to a super-cell containing few different cells, and therefore can treat the interference effect between different cells.

Theoretical and numerical comparj-sons among the unified diffusion coefficient and those by Benoist, Gelbard and Seki are performed for lattice cells contai-ned in the ZPPR-3 modified phase 3. For a sodium-voided fuel cell adjacent to a voided sodium follower the interference is large and the diffusion coefficient of the fuel is increased by about 30% relative to that from the Benoist formula.

To check the applicability of the unified diffusion coefficient sodium-void worths in one-dimensional slab test problems are calculated and compared with the reference results from transport theory treating heterogeneity of each plate using the ANISN code. Furthermore sodium-void worths for four void patterns selected from the ZPPR-3 modified phase 3 void experiment are evaluated from 16-group three-dimensional diffusion calculations using the unified diffusion coefficient and compared with the measured worths. From these comparisons it follows that:

: (1) The diffusion coefficients by Benoist and Seki introduce large errors in void worths compared with the reference heteroqeneous calculations for void patterns where sodium foliowers and surrounding fuel ceils are voided. The use of the unified diffusion coefficient removes

these errors. This implies that the present model treats the local streaming effect caused by the presence of sodium followers accurately by using transport theory in a super cell model. For sodium-void worths in ZPPR-3 diffusion calculations

using the unified diffusion coefficient produce only a few percent errors for void patterns except the cases of core center voided and outer core voided patterns. When the core center including a control rod is voided, the calculation overestimates the void worth by 2 5 % , while the calculation using the conventional diffusion coefficient produces an error of 40%. Thus further improvement is necessary for this pattern.

Reaction rate distribution measured in fast critical assemblies has been analysed to estimate the prediction accuracy and examine about the cause of the discrepancies

between calculations and measurements. The selected assemblies are ZPPR-2, ZPPR-3 phase 2, MZB (2) and MZB (3) . The analysis has been performed with use of the JAERI Fast set version 2. At first, the heterogeneity calculations were carried out to obtain cell averaged cross sections and they were used in two dimensional diffusion and transport calculations. To estimate the influence of the anisortropic elastic scattering effect, P ? group constants have been produced and used in Sn calculations.

. The results show that the reaction rate distribution could be predicted within an error of +3% in the core regions. On the other hand, it was almostly in the range of 25% for the radial blanket, and the discrepances between measurements and calculations seem to be systematic. It was shown that the contribution from below 1 keV was significant in the blanket region.

The effects of cell models for control rods with different 'OB enrichment and B,+C pin arrangements on their reactivity worths at the core centre of a fast reactor are studied. The fast reactor considered in the study is a simple cylindrical one which is provided modifying FCA VII-1 90'2, Effective cross-sections of various control rod cells (homogeneous cell, cylindrical cell and cluster cell) are produced by using a collision probability code PIGEON and JAERI-FAST V-I1 25 energy group cross-section set. Diffusion and transport calculations are made using CITATION and TWOTRAN-I1 codes, respectively.

From a simple comparison between measurement and calculaion of control rod reactivity worth, it can be expected that the difference in C/E between all the control rods with different '% enrichments would be within the expected experimental errors, when the effective cross-sections are produced so as to properly take into account of the cluster geometry of the control rods and also transport theory is applied to neutronic calculations.

4) Mockup Experiment and Analysis

A series of experiments has been made on FCA Assembly VIII-2 in order to refine the calculation method for the reactivity effect due to axial displacement of fuel/cladding in connection with a LMFBR core meltdown accident. FCA Assembly VIII-2 consisted of a central test region simulating the prototype FBR "MONJU" in axial dimensions and composition, and a uranium-fueled driver region as shown in Fig. 1.

Axially symmetric and asymmetric displacements of fuel were made in the central 3 x 3 drawers (equivalent radius 9.3 cm) with sodium expelled. Axial distribution of stainless steel reactivity worth was measured in the central 3 x 3 drawers of reference core. Stainless steel worth was also measured in the void region of each fuel displadement configuration. It was shown that reactivity change due to the fuel slumping is sensitive to the amount of stainless steel in the region where

the fuel displacement takes place. The distortion in 2 3 9 ~ ~ fission rate distribution was measured with a multiple chamber scanning system (MCSS), about 60 fission chambers distributed in the test and driver regions.

Reactivity change for the respective fuel displacements was calculated by diffusion theory using JAERI-Fast Set Varsion 11. The convensional calculation underestimates considerably the experimental values when the neutron streaming effect is large. Much improvement is achieved by using a modified diffusion coefficient for the void region. However, there still remains the tendency of underestimation which increases with expansion of the fuel slumping region to edge of the core.

The subsequent experiments on FCA have been carried out this year on the neutron penetration in the area of supporting grid and on the assessment of Keff in the ex-vessel storage e tank, in relation with the design of MOIJJU.

Analyses for ZPPR-9 as a part of JIJPITER program, which is the DOE-PNC joint program, have been completed, and analyses for ZPPR-10 are now continued.

5) Research on Shielding

Sensitivities of the atomic disp1ac:ement rate in the reactor vessel and the total dose rate at the outside of concrete layer of JOY0 were analyzed for microscopic partial cross sections of constituent elements, developing a sensitivity analysis code system for partial cross sections. Sensitivities were calculated for total, elastic scattering, inelastic scattering, (n, 2n), (n, Y) , (n, p) , (n, a) and the secondary gamma-ray production cross sections of 7 elements such as sodium, iron, chromium, nickel, carbon, silicon and oxygen. Effective cross sections for reactions of each element were generated based on JSDlOO set and POPOP4 library with RADHEAT-V3.5 code system considering resonance self-shielding. Forward and adjoint angular fluxes of neutron and gamma-ray were calculated with one-dimensional Sn transport code ANISN-T. Sensitivity calculations were performed with perturbation code SWANLAKE-N.

Main results as follows. The atomic displacement rate in the reactor vessel and the total dose rate at the outside of concrete layer are both remarkably sensi-tive to cross sections of sodium and iron. For example, they decrease by 2.0 and 3.4% respectively due to the 1% increase of neutron total cross section of iron. As for partial cross sections, sensitivities for elastic scattering cross section are more than 70% those for total cross sections and those for inelastic scattering cross section are at most 20%. Sensitivities for the other neutron reaction cross sections are comparatively small. But as regards total dose rate, sensitivities for capture cross section of iron in the carbon steel between the graphite and the concrete layer are fairly large, that is, 80 90% those for total cross section. Besides sensitivities of total dose

rate for secondary gamma-ray production cross section are also large in those regions. Hence it is deduced that the total dose rate at the outside of the concrete layer is mostly due to gamma-rays produced in those regions. (5) One example of sensitivity profile is shown in Fig. 2.

The benchmark experiment for the neutron streaming through two-bend ducts by the method of boot-strap calculations of DOT-3.5 has been analysed. Making a comparison between calculated and experimental results, the characteristics of calculations have been estimated. Moreover, the corner of the bend duct in the benchmark experiment by three-dimensional SN code, ENSEMBLE, has been analysed and the obtained informations have been reflected in the estimations of the two-dimensional calculations. From these results, the reasonable technique to estimate the radiation streaming through multi-bend ducts by using a two-dimensional SN code, has been found. The values of C/E for In(n,n1) reaction rate, P(n,p) reaction rate and Dy(n,y) reaction rate obtained by using this technique in the benchmark experiment were 2.41, 1.90 and 0.884, respectively.

A new technique of calculation has been developed which couples the discrete ordinate method (Sn method) with the albedo Monte Carlo technique to establish more effective and more accurate method for evaluating both neutron streaming and transmission. There are many penetration of complex geometry in the main shield which can be expressed with two-dimensional geometry. Therefore, neutron transmission in the main shield is obtainable with two-dimensional Sn forward and adjoint calculation, and neutrons streaming along the duct can be estimated by the albedo Monte Carlo technique by coupling with the forward Sn results. Finally, shielding characteristics can be estimated effectively by using adjoint Sn results, taking into account both the neutron streaming components and the transmission component through the bulk shield which cap not be counted with the albedo Monte Carlo calculation.

. The streaming analysis was performed with the newly developed method for the sodium duct in the iron shield, which could be modeled with two-dimensional geometry. Two-dimensional

n Sn calculation was also carried out for that model and its result was regarded as a reference data. The albedo Monte Carlo calculation was performed for the boundary crossing estimator with 20,000 histories, applying the importance sampling techniques such as source biasing and history splitting. Comparing with the reference data, not only the neutron flux distribution in the sodium duct agreed well, but also the other shielding characteristics agreed within 40% error. Computer time of the additive adjoint Sn calculation for small system including detector is much less than those of the reference forward Sn calculation and the albedo Monte Carlo calculation. It is concluded that this Sn-Monte Carlo-Sn (adjoint) coupling method is effectively applicable to evaluate the shielding characteristics in various systems having many penetrations.

For an efficient approach to a detailed analysis of ' radiation transport through an irregular shaped shield around the reactor core, a computer code FEM-3DP3 has been developed for solving the multigroup P3 transport equation in the three-dimensional (x, y, z ) coordinate system by means of the finite element method. The computer code uses the relaxation method for each (x, y) plane to solve the system equation for prism- and/or box-shaped elements with the bases parallel to the (x, y) plane. Furthermore, a special device is taken for saving the computer storage in order to solve the system equation with the matrix of a very large dimension. By applying FEM-3DP3 to an external source problem and to a keff calculation problem, it has been shown that the present code is useful for solving these problems and the mesh width can be coarse to be twice as large as for finite difference computer codes.

1; addition, nuclear group constants are produced from the JENDL-2 nuclear data file in the form of 100 neutron groups and 20 gamma-ray groups for the use of computer codes.

6 ) Study of Large Heterogeneous Core

Large heterogeneous core (1000 MWe) with the radial internal blanket has been studied to improve performance characteristics about the following points. The design margin of the fuel pin life time has been studied. Because the heterogeneous core has the smaller power peaking factor than homogeneous core. And the design margin has been utilized to improve the breeding performance. The heterogeneous core configuration has been studied again based on the previous study. ( 6 ) For the island heterogeneoue; core (Fig. 3) the whole core bowing behaviors has been analized, and the characteristics has been obtained.

The island heterogeneous core has the following characteristics compared with the homoqeneous core.

Core specification The equivalent radius at the radial blanket outer boundary is equal to that of the homogeneous core. The assembly dimension is the same as the homogeneous core. The clad of the fuel pins of the heterogeneous core is thinner by 0.05 mm. (The pellet diameter of the heterogeneous core is O.lrnrn larger) Advantageous performance Shorter doubling time (Homogeneous: 34.5 years, Heterogeneous: 23.5 years) Smaller sodium void reactivity Uniform plutonium enrichment of the whole core fuel Smaller power peaking factor (XY plane power peaking factor, Homogeneous: 1.24, Heterogeneous: 1.14) Disadvantageous performance Large plutonium inventory (13% larger) Larger coolant flow rate ( 7% larger)

(7) 7. Reactor Phvsics Activitv in DCA

In order to investigate nuclear characteristics of the commercial-scale Fugen-Type reactor that aims to increase channel power compared with that of the prototype, studies of reactor physics parameters have been continued by using uranium and plutonium fuel in DCA (Deuterium Critical Assembly).

The measurement items are : (1) lattice parameters, (2) microscopic and macroscopic power distributions in the core, and (3) loss-of-coolant activities. In the experiment 25.0 cm pitch square lattices loaded with 0.79 w/o enriched Pu02-U02 clusters were used. The cluster has 54/60 fuel pins of 12.5 mm in diameter. In preparation for the next measurement program 36-fuel-pin clusters are being assembled using the same fuel pins of 12.5 mm in diameter.

For confirmation of the possibility of load-following-type reactor operation, changes in local power distributions in the adjacent clusters to a control rod have been measured in detail corresponding to the small displacement of the control rod. Two types of control rods made of ordinary B4C and stainless streel are being tried and the analysis are also in progress.

REFERENCES

Yamamoto, H., Sekiguchi, Y., Inoue, T., and Nomoto, S., Reactor Physics Characteristics from Operational Testing of the JOYO Experimental Fast Reactor, ANS Topical Meeting on 1980 Advances in Reactor Physics and Shielding, Sun Valley, September 1980.

Miyawaki, Y., Private Communication.

Nakagawa, T. and Igarashi, S., Evaluation of Neutron Nuclear Data for 242mAm and 242gAm, JAERI-M8903, May 1980.

Takeda, T., Arai, K., Yamaoka, M., Tanimoto, K., and Sekiya, T., Unified Diffusion Coefficient for Analysis of Sodium-Void Worth in Critical Assembly ZPPR-3 Modified Phase 3 Core, to be published in J. Nucl. Sci. Technol.

Ohtani, N., Yamauchi, M., Itoh, J., Kawai, M., Application of Cross-section Sensitivity Analysis to JOYO Main Shield, Specialist Meeting, NEAsCRP October 1980.

Tsutsumi, K., et al., Study of Large Heterogeneous Reactor, Int. Symp. on Fast Reactor Physics, IAEA/OECD-NEA, Aix en Provence, France, 1979.

Miyawaki, Y., Private Communication.

&y FJ Control/Sofety rod Fl Counter

Fi g . 1 Vertical cross section of FCA Assembly \/I1 1-2

I. ':

h? E N E R G Y I E V 1 c . 2 Fig. 2 Sensitivity Profile of ' ~ o t a l Dose Rate for the Elastic Cross Section of Fe

- 128 --

R a d i a l B l a n k e t

I n t e r n a l B l a n k e t

@ P r i m a r y C o n t r o l Rod

@ S e c o n d a r y C o n t r o l Rod

F i g . 3 He te rogeneous C o r e Control R o d P a t t e r n ( Selected C o r e )

FU1:NACE needs d i l 1 e r e n f : i ; i l neu t ron - and pho ton -a lbcdo ' s f o r t h e b l a n k e t

c o n f i g u r a t i o n s u r r o u n d i n g t h e t o r o i d a l v e s s e l . . These a r e c a l c u l a t e d w i t h

a r e v i s e d v e r s i o n of t h e code ANISN.

Recent ly n e u t r o n i c s c a l c u l a t i o n s have been performed f o r t h e FINTOR-D

sys tem s t u d y . The aim was t o s t u d y t h e i n f l u e n c e of t h e r e a c t o r geometry

and the f u s i o n d e n s i t y d i s t r i b u t i o n s i n t h e plasma on t h e t r i t i u m b r e e d i n g

r a t i o and t o de t e rmine t h e p o l o i d n l d i s t r i b u t i o n of t h e n e u t r o n w a l l

l o a d i n g , t h e r a d i a t i o n damage i n t h e f i r s t w a l l and t h e the rma l power

p roduc t ion .

References

I I ( H . G ruppe laa r and B.P .J . van den Bos. The c o n t r i b u t i o n of (n ,p )

and (n ,a) r e a c t i o n s t o f i s s i o n - p r o d u c t c a p t u r e c r o s s s e c t i o n s ,

ECN-78 (1979), ( s e e a l s o NEANDC(E)209"L1', 209) .

(21 G. D e l f i n i and H . Gruppe laa r . Maximum-1ikeLihood a n a l y s i s of r e -

so lved r e sonance pararnc ters f o r 18 f i s s i o n - p r o d u c t n u c l i d e s ,

ECN-82 (1 98O), ( s e e a l s o NEANI)C(E)209"L", 169) .

131 R . J . H e i j b o e r and A . J . J a n s s e n . S t a t u s 05 pseudo f i s s i o n - p r o d u c t

c r o s s s e c t i o n s f o r f a s t reactors; s e n s i t i v i t y s t u d y f o r sodium

void e f f e c t . P roc . of t h e S p e c i a l i . s t s S Meeting on Neutron C r o s s

S e c t i o n s of F i s s i o n P roduc t Nuc le i , Dec. 1979, Bologna,

NEANDC(E)209"L", 375.

141 J .L . Rowlands. Needs and a c c u r a c y r e q u i r e m e n t s f o r f i s s i o n p r o d u c t

n u c l e a r d a t a i n t h e p h y s i c s d e s i g n of power r e a c t o r c o r e s .

Proc . Second IAEA Advisory Croup Meeting on F i s s i o n P r o d u c t Nuc lea r

Data (FPND), P e t t e n , 1977, IAEA-213, v o l . - I (1978) 41.

1.3. Power reactor noisc experiments at the Borssele PWK

~easurenients at full power during normal operation have been continued

at intervals of approximately three months. By using cross power

spectrum information of all combinations of signals from six ex-core

neutron detectors the amplitudes and directions of motion of the core

barrel at several frequencies could be detected with high resolution

and accuracy. Frequency, r.m.s. amplitude and direction of motion

appear to vary only slightly from measurement to measurement and from

core to core. A reactivity noise peak at.9.2 Hz appears to increase

linearly with decreasing boron concentration, in a way which is very

reproducible from core to core.

The new interest (after the TMI-2 accident) in noise measurements

under abnormal conditions and in post-accident situations has led

to the following sets of new measurements.

- At full power the signals from more than 100 reactor instrurnenta- tion signals, which are connccted to a patch panel in the control

room, have been measured. Mean and r.m.s. values w i ~ h autopower

spectra have been determined.

- During the stop procedure of the reactor, before refuelling, the system has been stabilized at several different power levels and

different cooling conditions. Each time an appropriate selection

of signals was recorded, analysed and reported.

These data may provide important base-line information for noise

measurements after unforeseen adverse events.

For multi-channel noise analysis an array processor (FPS model

AP120-11) has been connected Lo Lhe PDY-11/34. Up to now digital

input can be analysed, but work is in progress for ex~ension to

direct analog input, which will allow on-line real-~ime noise ana-

lysis of a large number of simultnneous signals.

The method of partial and multiple coherence has been applied to

multi-detector noise measurements in the Borssele reactor. It ap-

peared that the method can be useful for the separation of local

and global noise components, and also to separate several simulta-

ncous global noise sourccs to some extent.

By requcst of the organiscrs a m:lgnetic tape has bccn prepared with

Borssele reactor signals for the benchmark test in connection with

the forthcoming NEACRI' spccin 1 is t mee ti ng SEIORN-111.

1.4. International benchmark calculation project REAL-80

At the third ASTM-Euratom Symposium on Reactor Dosimetry, held in Ispra.

October 1-5, 1979, W.L. Zijp, chairman of the workshop session on ad-

justment codes, uncertainties and input needs, launched the idea for an

international project on a specified benchmark calculation.

The plenary final session adopted this idea, that a study should be made

on the uncertainty of integral parameters (such as displacement rates

or activation rates), derived from neutron flux density spectrum in-

formation based on experimental activation rates by an unfolding pro-

cedure. The outcome of such an international intercomparison should give

the state-of-the-art of that laboratory capabilities in deriving dis-

placement rate uncertainties using the existing practices. The study

should be carried out under well chosen and well defined conditions

with respect to input ncutron spectra, input activation rates, input

cross. section values, taking into account as much as possible realistic

values of the variances and covariances of the input data.

The IAEA was invited to assist in the organization by sponsoring the

planned activities, in the framework of the IAEA program on the Standard-

ization of Neutron Measurements, in close collaboration with the Nuclear

Data Section, and ECN, Petten.

The aim of the international intercomparison is to arrive at a realistic

value for the uncertainty in integral parameters (likc a displacement

rate, an activation rate), when such a value is derivrd by means of

existing unfolding procedures. For instance the exercise should give

answers to questions like:

1. What is the quality of the neutron spectrum derived by different

existing unfolding/adjustment procedures?

2. What is the quality of an integral damage parameter, like the number

of displacements per atom (d.p.a.), derived with aid of an adjusted

spectrum?

3. What is the quality of a predicted activation rate?

It was urged that the excrcisc should bc performed within a very tight

time schedule covcring a period of only two years, so that the results

will be available for discussion at the next ASTN-Euratoa~ Symposium.

The following typical ncutron spcclra were suggested for intercomparison

in the excrcisc: MTR, FER, LWK-PV and MIX-blanket.

A l l l abora tor ies should use tlic same ac t iva t ion rrosn sec t ion da ta s e t

(e.g. the ENDFIB-V dosimetry f i l c , o r the IRDF, which w i l l both include

covariance information) and. the same damage cross sec t ion s e t (e.g.

based on the recent ly published ASTM standard E693-79 on dpa determinat ion) .

In f i r s t instance t h i s p ro jec t was calledDREAM (Displacement Rate

A Evaluation from Activat ion Measurement). Hopefully t h i s dream may become

a REAL exerc ise i n a double meaning (Reaction Ratc Estimates. Evaluated

by Adjustment Analysis i n Leading Laboratories) . The p a r t i c i p a t i o n w i l l . be open t o a l l l abora to r i e s , wl~ich have the c a p a b i 1 i . t ~ t o perform an un-

ce r t a in ty ana lys is coupled Lo t h e i r uofoldiiip,/adjust.i~~c.iit code. I t i s

necessary t h a t the procedures t o be followed have been documented and

can be made ava i l ab le a t an e a r l y s tage of the p ro jec t .

XI. Reactor Physics a t the I n t e r u n i v e r s l ~ Reactor I n s t i t u t e ( IRI)

a t D e l f t

11.1. Power r e a c t o r n o i s e -

Noise measurements i n the 54 Mde BIm w i t h n a t u r a l c i r c u l a t i o n a t

Dodewaard have been cont inued. A new v e r s i o n of s e l f powered n e u t r o n

d e t e c t o r , the so -ca l l ed twin d e t e c t o r , was i n s t a l l e d and i n t e n s i v e l y

used f o r steam v e l o c i t y and coherence measurements. I n t h e upper p a r t

of t h e core the phase behaviour of t h e twin d e t e c t o r s i g n a l s e x h i b i t s

a d e v i a t i o n from what i s expcctcd i n c a s e of a p r o c e s s w i t h a s i n g l e

t r a n s i t time. This may be duc t o the r a d i a l steam v e l o c i t y d i s t r i b u -

t i o n i n s i d e the f u e l e lements . This phenomenon w i l l be s t u d i e d f u r t h e r

by us ing twin self-powered gamma d e t e c t o r s .

S p e c i a l a t t e n t i o n was paid t o t h e f requency dependent coherence of s i g -

n a l s from d e t e c t o r s a t d i f f e r e n t , r a d i a l p o s i t i o n s . A t f r e q u e n c i e s

about I Hz t h e n o i s e s i g n a l s a r e a lmost f u l l y c o h e r e n t , b u t w i t h de-

c r e a s i n g f requency and i n c r e a s i n g r a d i a l s e p a r a t i o n o f t h e d e t e c t o r s

the coherence d e c r e a s e s . This cannot be exp la ined by u s i n g t h e concept

of ' g l o b a l n o i s e ' caused by r e a c t i v i t y f l u c t u a t i o n s . A model h a s been

developed on the b a s i s of a d i s t r i b u t e d n o i s e source i n t h e c o r e ,

caused by r a d i a l l y incoherenL steam bubble p roduc t ion . This source i s

observed by t h e d e t e c t o r s wi th a ' f i e l d of view' d e s c r i b e d by a d j o i n t

neutron t r a n s p o r t e q u a t i o n , i n which t h e power feedback t o t h e neu t ron

c r o s s s e c t i o n s i s taken i n t o account . I n t h i s way t h e observed phenomena

could be exp la ined and an at-power r e a c t i v i t y t r a n s f e r f u n c t i o n could

be c a l c u l a t e d from the measured coherences . During t h e r e p o r t p e r i o d

two p u b l i c a t i o n s [ 1 , 2 ] wcre i s s u e d i n connect ion wi th t h e BWR-noise

work and a t h i r d one i s i n p r c p n r a t i o n .

I n the n o i s e work a t t h e 2 MW pool-type r e s e a r c h r e a c t o r HOR t h e a t t e n -

t i o n has been focussed on the e x p l a n a t i o n of t h e exper imenta l r e s u l t s

ob ta ined u n t i l now on the b a s i s of a dynamic model of t h e r e a c t o r and

i t s c o n t r o l systcln. 1Jith ch i s model the main n o i s e s o u r c e s could be

i d e n t i f i e d and some t r a n s f c r f u n c t i o n s wcre d e t e r n ~ i n c d . An exper imenta l

se t -up f o r stutiying b o i l i n g i n an MTK-type f u e l clement i s i n p repara -

t i o n . F i r s t s t u d i e s w i l l he made i n an out-of co re loop w i t h an e l e c t r i -

c a l l y hcnted s imula ted Eucl e l c ~ n c n t .

A s spin-off from the t h e o r c t i c s l noise work hnsed on ad jo in t neutron

t ransport equations, a publicat ion has been issued on the app l i ca t ion

of ad jo in t formulations t o hea t t r anspor t theory [31.

11.2. Monte Carlo Pro jec t

A

Final repor ts on the general purpose Monte Carlo computer code FOCUS

and codes f o r the preparat ion of a d j o i n t c ross sec t ions s t a r t i n g from

the ENDF-B f i l e have been issued and a r e ava i l ab le v i a NEA Computer

.? Program Library [ 4 , 5 , 6 1 .

11.3. Miscellaneous

The r eac to r physics s tud ie s on the gas core r eac to r concept have been

continued on a very modest sca le . A compilation has been made of a l l

the aspects s tudied u n t i l the beginning of 1980 [71 . Moreover pre-

liminary f u e l cycle s tud ie s have been performed, inc luding enrichment

and re-enrichment of the i u e l .

F ina l ly a computer-controlled r eac to r s imulator ahs been constructed

f o r operator and s tudent t r a in ing purposes.

References

[ I ] K l e i s s , E.B.J. and H. van Dam:

Analys is of Neutron D e t e c t o r Response t o Bubbles i n a Water Moderated

Reactor.

Ann.Nucl.En. - 6 (1979) 385-398.

[ 2 ] K l e i s s , E.B.J. and H. van Dam:

A Twin S e l f Powered Neutron D e t e c t o r f o r Steam V e l o c i t y De te rmina t ion

i n a BIJR.

To be publ . i n Nucl. Technology.

[ 3 ] Dam, H. van and J . E . Hoogenboom:

The a d j o i n t space i n h e a t t r a n s p o r t theory .

1 n t . J . H e a t and Mass T r a n s f e r - 23 (1980) 949-353.

[ 4 ] Hoogenboom, J .E. and P.F.A. de Leege:

ADX-A code t o c a l c u l a t e a d j o i n t n e u t r o n c r o s s s e c t i o n s from t h e ENDFIB

f i l e .

Report IRI-131-77-04 (1979).

151 Hoogenboom, J .E.:

ETOF - A program t o p r e p a r e a c r o s s s e c t i o n d a t a t a p e from t h e ENDFIB

f i l e f o r t h c a d j o i n t Monte Car lo code FOCUS.

Report IRI-131-77-05 (1979).

I61 Hoogenboom, J .E . :

FOCUS - A v e r s a t i l e non-multigroup a d j o i n t Monte Car lo n e u t r o n code.

Report IRI-131-77-06 (1979)

(71 Kistemaker, J . e t n l . :

Same Aspects of a Plasma Power Reac to r based on F i s s i o n .

Report of F O W I n s t i t u t c f o r Atomic and Molecular P l ~ y s i c s , Amsterdam.

111. Reactor physics a t KEMA Arnhem

m.1. Dodewaard r eac to r (50 MI&?, BWR, GE design)

The r e a c t i v i t y behaviour of n heavi ly Gd loaded core (cycle 1 1 ) in-

d i ca t e s t h a t the methods used t o ca l cu la t e the Gd-absorption give

a underestimation of 202. Improved r e s u l t s a r e expected from the

use of t ranspor t methods ins tead of d i f fus ion theory f o r the ca l -

cu la t ion of the power d i s t r i b u t i o n i n an element. A number of ex-

perimental power d i s t r i b u t i o n s of elements has been compared wi th

the LWR-WIMS code and gave good agreement [ I ] .

The o f f gas a c t i v i t y a t a sampling poin t has been modelled using

a s impl i f ied desc r ip t ion of gas r e l e a s e and gas t r anspor t . We hope

t h a t eventual ly we w i l l be ab le t o determine number and charac ter

of the leaking f u e l assemblies.

IlI.2. Borssele r eac to r (470 ElIJ, YWR, KW design)

No research a c t i v i t i e s .

[ I ] Veri f ica t ion of WIMS ia Dodcwaard.

Paper presented a t the IIlLMS uscr meeting Apr i l , 1980.

NEACRP-L-244 NORWAY

STATUS REPORT TO NEACRP ( 1 9 7 9 - 1980)

Resc to r p h y s i c s a c t i v i t i e s i n Norway, September 1979 - August 1980 --

Compiled by T . ~ k a r d h a m a r (1.F.E.)

1. LIGHT WATER REACTOR PHYSICS

The modular code system FMS (Fue l Management System) f o r l i g h t

wate r r e a c t o r calcu1.a t ions (developed a t I .F .E. /ScP)* h a s reached

a h igh deg ree of m a t u r i t y a f t e r many y e a r s of development and

con t inuous a p p l i c a t i o n s by power u t i l i t i e s and o t h e r o r g a n i z a t i o n s .

The system, t h e main programs of which a r e shown i n f i g . 1 , e n a b l e

r e a c t o r c a l c u l a t i o n s t o b e done a t d i f f e r e n t l e v e l s o f d e t a i l and

s p a t i a l r e p r e s e n t a t i o n , and i n c l u d e s a l s o 1 t o 3-dimensional

dynamic codes f o r t r a n s i e n t a n a l y s i s , and codes f o r f u e l p e r -

formance s t u d i e s . For core- fo l low a n a l y s i s and d e t a i l e d r e a c t o r

s i m u l a t i o n i n t h e p r e d i c t i v e m o d e , t h e h a s i c codes of FMS a r e t h e

RECORD and PRESTO codes . They have been a p p l i e d th rough many a

y a a r s t o t h e a n a l y s i s of a l a r g e number o f o p e r a t i n g c y c l e s , and

a summary o f some o f t h e main e x p e r i e n c e s o f t h e s e codes was

p re sen ted a t t h e MEACRP S p e c i a l i s t s ' Meeting, P a r i s , November

1979 (1).

The f u t u r e development and f u r t h e r r e f i n e m e n t s o f FMS i s now

bee ing s t e e r e d more and more by t h e a c t u a l u s e r s of t h e system.

Through r e g u l a r mee t ings o f a FMS User Group a forum i s provided

f o r exchange of r e s u l t s and e x p e r i e n c e s , and t o p r o v i d e d i r e c t i o n

f o r f u t u r e inlprovenients .

* I .F .E . I n s t i t u t e f o r Energy Technology ( f o r m e r l y I n s t i t u t t f o r Atomenergi) , K j e l l e r , Norway

ScP Scandpower A/S, Kjellex,Norway , . . . . . , .. . ,I > ' . l z ' s L ' , .; .. ! .,,. .. ' !

\

\'>

Previously FMS has been mainly applied in Eurcpe, but is now

available to power utilities in the USA through data links CO

the University Computing Company (UCC) in Dallas, Texas, where

the codes have been installed. The analytic methods of RECORD/

PRESTO and associated codes have been extensively evaluated during

the past year by a USA power utility (Carolina Power and Light

Company). Benchmark calculations were carried out on hot opera-

ting state conditions, cold critical states, and xenon transient

states. The very successful outcome of this study has led

the utility to use the FMS methods for their fuel management

work and plant operation support.

Recent model extensions to the PRESTO code include modifications

of the logic for xenon dynamics calculations, and the implementa-

tion of a samarium transient model. Many changes have also been

made which relate to improved input and cutput features of the

code. A study has shown the feasibility of running PRESTO on a

mini-computer, and studies of Core Surveillance Systems including

on-line versions of PRESTO are being performed.

1.2 RECORD/PRESTO analysis of y-scan measurements of Hatch-1

nuclear power plant

The y-scan measurements at end of cycle 1 of Edwin I Hatch

nuclear power plant unit 1 provide a very comprehensive set of

power distribution data on which BWR codes can be aualified..

The RECORD/PRESTO codes of FMS have been applied to a detailed-

core-follow study of cycle 1 of this reactor, with subsequent . comparison between measured and calculated data. The first

phase of this study has recently been completed. Very satis-

factory agreements between measured and calculated distributions

have been obtained, which confirm the efficiency and accuracy of

the analytic methods of RECORD/PRESTOfor predicting power distri-

butions in operating BWR1s. Examples of calculated and measured

axial y-scan results are shown in fig.2.

A core- fo l low s tudy has been under taken o f t h e f i r s t c y c l e o f

t h e Beaver Va l l ey power p l a n t i n t h e USA i n o r d e r t o e v a l u a t e f u r -

t h e r FMS f o r PWR a p p l i c a t i o n s . The average o f c a l c u l a t e d keff

v a l u e s through t h e c y c l e ( u p . t o ave rage burnup of 11400 Mwd/tu)

i s 0.9951 wi th a s t a n d a r d d e v i a t i o n o f 0.0019, and good ag ree -

ment i s o b t a i n e d between c a l c u l a t e d power d i s t r i b u t i o n s and

p r o c e s s computer r e s u l t s . F ig .3 shows exan~p le s o f comparisons of

a x i a l and r a d i a l power d i s t r i b u t i o n s . it i s concluded t h a t t h e

s y s t e m a t i c approach o f RECORD/PRESTO, w i t h a u x i l i a r y codes ,

a p p e a r s e q u a l l y s u i t a b l e f o r FPIR a p p l i c a t i o n s a s f o r BWR

a p p l i c a t i o n s .

2 . REACTOR DYNAMICS

J . Rasmussen ( I .F .E . )

2 . 1 Code D e v e l o ~ t s -

The work has been c o n c e n t r a t e d on developments and improvements

of t h e t h r e e di inensional r e a c t o r dynamics code RAMONA-111.

I n t h e n u c l e a r model, t h e c r o s s s e c t i o n r e p r e s e n t a t i o n h a s

been extended t o a l l o w f o r s e v e r a l polynomial d a t a sets f o r

each f u e l t y p e a t v a r i o u s burnup s t a t e s . The code w i l l s e l e c t

e t h e d a t a s e t c l o s e s t t o t h e a c t u a l burnup s t a t e o f each node. .

The n u c l e a r f u e l model has been extended t o accoun t f o r t h e h e a t

c a p a c i t y o f t h e c l a d d i n g . P r e v i o u s l y , t h e c l a d d i n g w a s t r e a t e d

a s a h e a t r e s i s t a n c e o n l y .

The thermo-hydrau l ics model has undergone s e v e r a l improvements

and e x t e n s i o n s .

I n a d d i t i o n t o t h e dynamic c a l c u l a t i o n o f t h e r e a c t o r system

p r e s s u r e , a dynamic c a l c u l a t i o n of t h e c o r e upper plenum p r e s s u r e

h a s been inc luded . T h i s makes t h e program more s u i t a b l e f o r

r e a c t o r t r a n s i e n t s w h e r e f a s t changes i n p r e s s u r e a r e involved .

A dynamic model of a d r i v e l o o p f o r a BWR of GE t y p e h a s been

added, assuming incompres s ib l e f l u i d f low i n t h e d r i v e l oop .

Also a s team s e p a r a t o r model has been inc luded , accoun t ing f o r . steam s e p a r a t o r i n e r t i a e f f e c t s and p o s s i b i l i t i e s f o r steam

c a r r y under i n t o t h e downcomer.

These l a s t two model changes a r e v e r y s i m i l a r t o co r r e spond ing

models d e s c r i b e d i n ( 2 ) .

2 . 2 Code A p p l i c a t i o n s

Tie RAMONA programs have been used t o a n a l y s e t h e Peach Bottoin

Turb ine T r i p T r a n s i e n t s ( 3 ) , ( 4 ) . An e x t e n s i v e comparison

between expe r imen ta l d a t a and c a l c u l a t i o n s were c a r r i e d o u t .

The c a l c u l a k e d r e s u l t s depended s t r o n g l y on t h e t r e a t m e n t o f

t:ne vo id dependence i n t h e n u c l e a r c r o s s s e c t i o n d a t a . The

i n f l u e n c e o f t h e water g a p on t h e vo id dependence i n t h e n u c l e a r

d a t a w a s improved. The r e s u l t s o b t a i n e d were q u i t e s a t i s f a c t o r y ,

c a l c u l a t e d n u c l e a r power t r a n s i e n t s were w i t h i n *6-8% from t h e

a expe r imen ta l d a t a . An example o f t h e t r a n s i e n t neu t ron f l u x

peak, a s c a l c u l a t e d by RRbIONA-111, i s shown i n f i g . 4 .

- Comparisons were a l s o made w i t h 1-dimensional k i n e t i c s , and

p o i n t s k i n e t i c s c a l c u l a t i o n s . R e s u l t s i n d i c a t e t h a t bo th

1-dimensional k i n e t i c s and p o i n t k i n e t i c s may g i v e good r e s u l t s

f o r t h i s t y p e o f t r a n s i e n t s as long a s t h e r a d i a l l y averaged c r o s s

s e c t i o n s (1-dimensional k i n e t i c s ) and r e a c t i v i t y c o e f f i c i e n t s

( p o i n t k i n e t i c s ) can be a c c u r a t e l y e s t i m a t e d . Tha t i s , however,

n o t a s i m p l e t a s k and i n many cases it w i l l be n e c e s s a r y t o check such c a l c u l a t i o n s a g a i n s t f u l l 3-dimensional c a l c u l a t i o n s .

The RAMONA-I11 program h a s been i n s t a l l e d a t U n i v e r s i t y Computing

Company (UCC) i n D a l l a s , Texas.

K . Ilaugset ( I . F . E . , rlalclen)

The r e s e a r c h programmes a t t h e OECD Aaltlen Reactor P r o j e c t a r e

a t p r e s e n t main ly concerned w i t h f u e l performance exper iments

and a n a l y s i s , and w i t h a p p l i c a t i o n s of p r o c e s s computers i n

p l a n t c o n t r o l . The l a t t e r i t em is r e c e i v i n g i n c r e a s i n g a t t e n t i o n

due t o t h e importance a t t a c h e d t o e x p l o i t i n g t h e c a p a b i l i t i e s o f . modern p r o c e s s computer t echnology t o improve e f f i c i e n c y and

o p e r a t i o n a l s a f e t y of n u c l e a r power p l a n t s .

Most o f t h e a c t i v i t y on r e a c t o r p h y s i c s a t t h e Halden P r o j e c t

i s connec ted t o t h e development o f a system f o r s u r v e i l l a n c e 0 and control . of c o r e power d i s t r i b u t i o n , SCORPIO. Requirements

t o speed and accu racy of t h e c o r e c a l c u l . a t i o n s needed i n such

an o n - l i n e system has l e d t o t h e development o f a new s i m u l a t o r .

Both t h e PWR and t h e BWR v e r s i o n s o f t h i s s i m u l a t o r have been

t e s t e d d u r i n g t h e l a s t y e a r .

Another f e a t u r e of SCORPIO i s i t s a b i l i t y t o p r e c a l c u l a t e i n an

o p t j m a l way t h e c o n t r o l maneuvers d u r i n g power changes such a s

r e s t a r t a f t e r shutdown o r r e g u l a r l oad c y c l i n g . The power d i s -

t r i b u t i o n c o n t r o l methods t a k e i n t o accoun t l i m i t a t i o n s on c o r e

v a r i a b 1 . e ~ such a s l o c a l power d e n s i t y . Such l i m i t a t i o n s w i l l

p a r t l y o r i g i n a t e from f u e l f a i l u r e models which a l s o a r e be ing

de- eloped a t t h e Halden P r o j e c t . .

REFERENCES

S.B@rresen, T.~kardhamar, S.Wennemo-Hanssen:

"Applications of FMS-RECORD/PRESTO for Analysis and

Simulation of Operating LWR Cores",

NEACRP Specialists' Meeting on Calculation of 3-Dimensional

Rating Distributions in Operating Reactors, Paris (Nov.1979).

R.B.Linford:

"Analytical Methods of Plant Transient Evaluations for the

General Electric Boiling Water Reactor",

NEDO-10802 (Peb. 1.973).

L.Moberg, J.A.Naser, J.Rasmussen:

"RAMONA-I11 Analysis of the Peach Bottom Turbine Trip Tests",

Trans. Am. Mucl. Soc. 34, 503 (1980).

L.Moberg, J.Rasmussen, T.O.Sauar, O.Oye:

"TCWlONA Analysis of the Peach Bottom Turbine Trip Transients",

EPRI Research Project 1119-2. To be published.

GENERATION

FIG, 1, THE FMS FODULAR CODE SYSTEP FOR LIGHT

WATER REPKTOR CALCULATIONS

RXIRL LR-ILIO INTENSITY HRTCH-1 EOC-1

0.0

8 : a 6 s . 12 ( 5 . $8 21 . ZY RXlRL HEIGUT. INOGESI

RXIRL LR-1YO INTENSITY ' UFITCII-1 EOC-1 . 4.5

. YCRCUREO UX2S7

2 0.0

c" a , 3 6 B ( 5 1 0 21 2'1

c .., RXlRL HEIGHT (NCOEZl e\_;

RXIRL Lfl-1 'I0 INTENSITY ' HRTCil-I EOC-I

CNCULQICO PRCSIO . YEASURCO L X ~ 3 0

0 3 6 B 5 2 . .> - ( 8 ?I zq I RXlRL HEIGHT (NOCES)

RXIRL LR-IL10 INTENSITY HRTCU-1 EOC-1 2 P

1 .s u! r CALCULRICO PRCSIO I . YCRSUREO U X Z S

t -2

5- .- " 3 f 0.9 - es = .- 3 0.6

.

W > ? ,- u. d o.a OL

$ . 3 6 RXlRL 9 HEIGHT '12 (NOGESI i s l a 1( 2~

. . . .

C> FIG. 2. EXAMPLES OF RECORD/PRESTO CALCULATED AND MEASURED Y-SCAN RESULTS -2%

.b. FOR EDWIN I HATCH (BWR) POWER PLANT UNIT 1.

,.. .,,. .,...'

Dewcr volley. Avrroae axia l >ewer density a: 78CO MWDITU.

Eo?!om Axial height (nodes) TOP

The numbers i n parenthesis are the d i f f e i e n c e between the highest and lowest measured value for r);ulerrIc p s i t i o n s i n the core.

MEAS.

CALC.

Measured ond calculated ossernbly Power. 7800 MWDITU.

Beover Volley.

FIG, 3, EXAMPLES OF RECORD/PRESTO CALCULATED AXIAL AND

R A D I A L POKER D I S T R I B U T I O N S I N COMPARISON W I T H

PROCESS COMPUTER RESULTS FOR THE BEAVER V A L L E Y

PEQCII BOTTOM TURBTME TRIP TRQNSIENT T T ~

FIG, 4 , EXAMPLE OF TRANSIENT NEUTRON FLUX PEAK AS

CALCULATED BY R/!%NA- 1 1 1

REACTOR PHYSICS ACTIVITIES I N SPAIN

(September 1979 t o September 1980)

C o m p i l e d b y G . VEL,ARDE

1. SWIMMING-POOL REACTOR DESIGN

NEACRP-L-244 SPAIN

J E N i s p r e s e n t l y e n g a g e d i n a f e w p r o j e c t s wi th - r e a c t o r s o f s u c h t y p e , m a i n l y t h e more- than-10-MW r e a c t o r

b e i n g b u i l t i n L o - A g u i r r e ( C h i l e ) , a n d t h e 3-MW r e a c t d p l a

n n e d i n A y c h a p i c h o ( E c u a d o r ) . M e t a l u r g y D i v i s i o n o f J E N w i l l

p r o b a b l y f a b r i c a t e t h e n e c c e s a r y p l a t e f u e l , a n d c o r r e s p o n

d i n g R e a c t o r P h y s i c s c a l c u l a t i o n s a r e b e i n g d o n e i n t h e R e a 5

t o r C o m p u t a t i o n S e c t i o n o f JEN, a s i n d i c a t e d b e l o w .

F o r t h e L o - A g u i r r e r e a c t o r , w o r k h a s p r o c e e d e d

i n s e v e r a l a r e a s , s p e c i f i c a l l y equ i l ib r ium-cyc le burnup, enrichment, re

d u c t i o n , h o t c h a n n e l f a c t o r s ( r e v i s e d ) , c o n t r o l r o d w o r t h s

( r e v i s e d ) , e t c . F o r t h e l a s t t w o i t e m s , e x t e n s i v e t w o - a n d

t h r e e - d i m e n s i o n a l d i f f u s i o n c a l c u l a t i o n s w i t h c o d e CITATION

h a v e b e e n d o n e , s e a r c h i n g f o r u n f a v o u r a b l e p a t t e r n s o f p a y

t i a l l y - i n s e r t e d c o n t r o l b l a d e s .

F o r t h e E c u a d o r i a n r e a c t o r , p r e l i m i n a r y c a l c u -

l a t i o n s a r e i n c o u r s e . Most o f t h e m 3 e a l e d w i t h t h e p r o b l e m

o f c o r e o p t i m i z a t i o n , p e r f o r m e d i n a f i r s t s t e p u s i n g c o d e s

WIMS-D (JEN-TRACA v e r s i o n ) a n d SOTHIS f o r t h e n e u t r o n i c a n d

e c o n o m i c a s p e c t s , a s a l r e a d y q u o t e d ( p a s t y e a r r e p o r t ) . F u r

t h e r X - Y c a l c u l a t i o n s w i t h c o d e CITATION-2/2 f a i r l y c o r r o b o

r a t e d t h e r o u g h m o d e l r e s u l t s f o r r e a c t i v i t y a n d f l u x e s , b u t

n o t f o r e q u i l i b r i u m - c y c l e l e n g t h o r r e a c t i v i t y ( a t l e a s t f o r

s c h e m e s h a v i n g f r e s h f u e l a t p e r i p h e ~ y ) . I m p r o v e m e n t s o n t h e

f u e l - m a n a g e m e n t l i n e a r m o d e l a r e c o n t e m p l a t e d . ?? , . , . .~

; ,' 9 ; , : .. .. L . 2 i ',!

Also, for the selected (optimum) core, pre-de - sign calculations have been iniciated. E.g., dynamic cdl-

culations (including neutron kinetics, core thermal-hydrau

lics and regulating-rod control) have been done using code

GASA, so obtaining conditions for reactor stability.

2. NEUTRONIC CALCULATIONS OF PWR CORES FOR SUPPORT OF DE-

a SIGN EVALUATION AND REACTOR OPERATION

The development of methods and computer codes

for neutronic calculations of PWR cores has been underta-

ken at the Reactor Theory and Calculations Section of JEN

in collaboration with the electric utility IBERDUERO, with

the goal of providing the utility a suitable capability of

neutronic calculations for support of the evaluation ofthe

nuclear core design and of the reactor operation. The first

applications to the analysis of the first core of a 900Mwe

a PWR under construction, includes also the experiencing and

validating of methods and procedures through the comparison

. of results with those provided by the reactor designer and

with those obtained using more detailed methods as referep

ce.

The methods in use include:

a) A detailed modelling of PWR fuel assemblies

of every type, including control and burnable poison rods,

for the generation of effective cross sections and neutro-

nic properties at the different operating conditions and

along burnup. The JEN version of the WIMS-D code ( 2 ) is

used in its cluster option for the explicit treatment in a

cylindrized equivalent bundle of every fuel rod, guide ard

instrumentation tube, and burnable poison or control rod in

each type of assembly. The calculating procedures have been

validated by comparison with explicit and detailed whole

assembly calculations.

b) A detailed modelling of PWR cores in 2Dwith

explicit treatment of every fuel, poison or control rod,

core and reflector components, using a 2 group diffusion

theory approximation on a fine mesh. The JEN version(3) of

the VENTURE code ( ' ) is used to achieve an efficient use of

computer memory for the large data storage required, while

keeping low computing times. Criticality conditions and d e

tailed power distributions are obtained directly by this m o

del.

C) A coarse mesh modelling of PWR cores in 3D

with 1 or 4 nodes per fuel assembly, using a FLARE type no-

dal method with empirical parameters adjusted consistently

for minimum rms deviation of assembly powers in a broad r a 2

ge of operating conditions.

d) A simplified core modelling of PWR cores for

approximated calculations of cycle length and burnup sharing

by fuel batch in succesive cycles. The CICLON code ( 5 ) is

used to obtain those variables needed for economic calcula

tions of reload cycles.

Application of these methods for neutronic an=

lysis of the first core of a 900 MWR PMR is under way as a

. cooperative JEN-IBERDUERO effort. Qualification of methods

has been completed in areas a) and b), and is well advanced

in areas c) and d) within the inherent limitations of those

models. The complete neutronic calculations for reviewing

the nuclear core design of the first core are being carried

out, with excellent agreement in the results at BOL-HZP con - ditions both unrodded and with different patterns of control

rod insertions.

References

1. Carol AHNERT, "Programa WIMS-TRACA para el cdlculo de elp

mentos combustibles. Manual de usuario y datos de entra-

da". JEN-461 (1980).

2. J.R. ASKEW et al., "A General Description of the Lattice

Code WIMS", J. Brit. Nucl. Energy Soc., Oct 1966, 56b.

3. Carol AHNERT, "Implementation of the VENTURE-TRACA pro-

gram in the UNIVAC 1100. Summary of modifications and

user's manual". To be submitted to the NEA Data Bank

(1980).

4 . D . R . V O N D Y e t a l , "VENTURE: A Code B l o c k f o r S o l v i n g Mu1 -

t i g r o u p N e u t r o n i c s P r o b l e m s A p p l y i n g t h e F i n i t e - D i f f e r e n -

c e D i f f u s i o n - T h e o r y A p p r o x i m a t i o n t o N e u t r o n T r a n s p o r t " ,

ORNL-5062 (1975).

5 . J.M. ARAGONES, "CICLON: A N e u t r o n i c : F u e l Management P r o -

g ram f o r PWR's C o n s e c u t i v e C y c l e s " , JEN-336 ( 1 9 7 7 ) .

3 . COUPLED NUCLEAR- THERMOHYDRODYNAMIC ANALYSIS OF LASER-

FUSION-FISSION MICROSPHERES -

A n u m b e r o f s t u d i e s h a v e b e e n p e r f o r m e d d u r i n g

t h e p r e s e n t y e a r o n i n e r t i a l c o n f i n e m e n t f u s i o n , u s i n g a s

d i v e r s l a s e r s o r h e a v y i o n s t o i m p l o d e m i c r o s p h e r e s o f D + T .

The i n n e r c a s e o f t h e m i c r o s p h e r e i s made u p o f

a c e n t r a l s o l i d DtT z o n e a n d a l a y e r o f P u , w h i c h a c t a s a

p u s h e r m a t e r i a l . The o u t e r c o r e i s made u p b y a b l a t o r a n d

o t h e r l a y e r s o f y e t u n d e f i n e d m a t e r i a l s .

The s

c r o s p h e r e u n d e r a

r a t e d i n t w o p a r t s

F i r s t

u d y o f t h e p e r f o r m a n c e o f t h e w h o l e m i -

a s e r p u l s e o r a h e a v y i o n s beam i s s e p a -

t h e p e r f o r m a n c e o f t h e i n n e r c o r e o f

t h e m i c r o s p h e r e ( ( D t T ) + P u ) u n d e r s u i t . s b l e b o u n d a r y c o n d i -

t i o n s ( p r e s s u r e p u l s e ) i s a n a l y z e d t o d e t e r m i n e i t s o p t i m a l

c o n f i g u r a t i o n a n d c o m p r e s s i o n .

S e c o n d , t h e r e q u i r e d c o n f i g u r a t i o n o f t h e o u t e r

core of the microsphere and the required profile of the in-

cident laser pulse or of the incident heavy ions beam to ob

rain that compression on the inner core will be calculated

by methods with less complexity and calculational effort

that if the whole target is considered (1, 2 ) .

Since analytical solutions to synchronise pre-

ssure pulses are very difficult to work out in multilayered

spheres, a numerical method has been developed named SINCRO.

This code searchs for a ladder of successive pressure steps

launched at time selected initially from the predicted pro-

pagation time to the target center at the sound speed through

the material conditions reaches after the previous shocks.

'Each pressure step is iteratively retined to achieve its si-

multaneous convergence at the target center with previous

shocks.

The NORMA and CLARA programs for modelling of

the thermohydrodynamics and nuclear processes, respectively,

in inertial confined fusion-fission microspheres, were pre-

sented at the Graz conference ( 3 , 4) and in the previous me2

ting of the NEACRP.

A summary of the description of the characteris

tics and calculated results for the cases considered is gi-

ven in the table. In this table, DtT and Pu burnups, inter-

nal energy (descomposed in the energy deposited by parti-

cles and neutrons in the DtT region and the energy deposited

in Pu by all the neutron reactions), total energy (internal

p l u s t h e e n e r g y c a r r i e d o u t b y t h e l e a k e d n e u t r o n s ) , a n d i n -

t e r n a l a n d t o t a l g a i n r e l a t i v e t o c o m p r e s s i o n work a r e a l s o

g i v e n .

A

R e f e r e n c e s

1. G . VELARDE, C . AHNERT, J.M. ARAGONES, G . LEIRA, R . LUQUI,

J.M. MARTINEZ-VAL, " C o u p l e d n u c l e a r , - t h e r m o h y d r o d y n a m i c a n 5 e

l y s i s o f l a s e r - f u s i o n - f i s s i o n m i c r o s p h e r e s " , S e c o n d I n t e r

n a t i o n a l C o n f e r e n c e o n E m e r g i n g N u c l e a a - E n e r g y S y s t e m s ,

A p r i l 1 9 8 0 , L a u s a n n e .

2 . J .M. ARAGGNES, C . AHNERT, " P r o c e s o s n u c l e a r e s e n l a f u s i 6 n

- f i s i 6 n d e m i c r o b o l a s d e D T , Pu y U p o r c o n f i n a m i e n t o i n e r -

c i a l " , S o c i e d a d N u c l e a r E s p a f i o l a , J u n i o 1 9 8 0 , M a d r i d .

3 . G . VELARDE, C . AHNERT, J.M. ARAGGNES, M . GGXEZ, G . LEIRA, R . LU- .

Q U I , J.M. MARTINEZ-VAL, J.M. PERLADO, "Analysis of l a s e r - f i s s i o n - f k

s i o n systems: Time-dependend coupled nuclear- t h e r m o h y d r o d y n a m i c

a n a l y s i s a n d a p p l i c a t i o n s " , A t o m k e r n e n e r g i e 3 2 , 5 8 - 7 2

( 1 9 7 8 ) .

4 . G . V E L A R D E , C . AHNERT, J.M. ARAGONES, M . GOMEZ, G . LEIRA, R .

L U Q U I , J.M. MAR'I'INEZ-VAL, J . M . PEKLADO, "Analys is o f l a s e r - f i s s i o n -

f u s i o n systems: Development of Methods f o r n u c l e a r a n d t h e r m o h y d r o -

d y n a m i c c a l c u l a t i o n s " , A t o m k e r n e n e r g i e 3 5 , 4 0 - 5 3 ( 1 9 8 0 ) .

4 . H Y B R I D REACTORS

D u r i n g t h e p a s t y e a r , some c a l c u l a t i o n s h a v e

. b e e n p e r f o r m e d on t h e f u s i o n - f i s s i o n s y s t e m s w h i c h a r e f u ~

d a m e n t a l l y b a s e d on t h e n u c l e a r a n a l y s i s o f t h e b l a n k e t s A

w i t h a c e n t r a l s o u r c e o f n e u t r o n s ( 1 4 Mev) o b t a i n e d f rom

n u c l e a r f u s i o n .

To o b t a i n r e s u l t s on e n e r g y m u l t i p l i c a t i o n ,

f i s s i l e b r e e d i n g r a t e and r a t e of p r o d u c t i o n of t r i t i u m h a s

b e e n t h e f i n a l o b j e c t i v e s i n o r d e r t o h a v e a c r i t e r i u m on

t h e g o o d n e s s of t h e d i f f e r e n t c o n f i g u r a t i o n s w h i c h h a v e

b e e n c o n s i d e r e d .

A g r o u p of b l a n k e t s f rom Lee (1, 2 ) a n d Law-

r e n c e L i v e r m o r e L a b o r a t o r y ( 3 ) h a v e b e e n used a s a b a s e t o

have a p a r a m e t r i c e v a l u a t i o n w i t h r e f e r e n c e t o t h e r e l a t i -

v e p r o p o r t i o n s be tween t h e f i s s i l zone and t h e t r i t i u m ge -

n e r a t i o n o n e . The c a l c u l a t i o n s w e r e done i n s t a t i o n a r y a n d

dynamic s i t u a t i o n , c o n s i d e r i n g c r i t i c a l a n d s o u r c e p r o b l e m s .

I n t h e s t u d y of t h e s e s y s t e m s i t h a s b e e n con-

. s i d e r e d two r e s o l u t i o n m e t h o d s : d i s c r e t e o r d i n a t e s a n d Mon

t e C a r l o . On t h e f a s t b l a n k e t s p r o b l e m t h e model t y p i c a l l y

u s e d i s :

- S t a t i o n a r y c r i t i c a l c a l c u l a t i o n s : XSDRN ( 4 )

and T I M O C ( 5 ) c o d e s .

- O b t a i n i n t e g r a l v a l u e s o f r e a c t i o n r a t e s a n d

e n e r g y g e n e r a t i o n : t h e a b o v e c o d e s w i t h s o u r

c e p r o b l e m w h e r e t h a t s o u r c e i s t h e f u s i o n o n e ,

o f 1 4 Mev n e u t r o n s c e n t e r e d i n t h e s y s t e m .

- B u r n u p e v o l u t i o n o f t h e s y s t e m : ORIGEN ( 6 ) .

The r e s o l u t i o n o f t h e p r o b l e m w i t h TIMOC c o d e

n e e d e d a p r e v i o u s l i b r a r y g e n e r a t i o n f r o m t h e E N D F B / - I I I /

I V w i t h t h e C O D A C ( 7 ) c o d e , o b t a i n i n g a 2 3 g r o u p l i b r a r y

f r o m 200 e v t o 1 6 Mev. U s i n g t h e TIMOC c o d e a s a c e n t r a l

p a r t i t h a s b e e n d e v e l o p e d some r o u t i . n e s t o o b t a i n t h e r e a c

t i o n r a t e s o f t h e d i f f e r e n t n u c l e a r r e a c t i o n s i n t h e s y s -

t e m , TIMOC-TASAS ( a ) , w h i c h c o u l d b e u s e d t o c a l c u l a t e t h e

b u r n u p of t h e b l a n k e t w i t h t h e h y p o t h e s i s o f t h e c o n s t a n -

c e o f t h e u n i t a r y r e a c t i o n r a t e o b t a i n e d i n t h e b e g i n o f

t h e t e m p o r a l s t e p .

A g o o d a g r e e m e n t h a s b e e n o b t a i n e d w i t h o t h e r

p r e c e d e n t r e s u l t s i n t h i s n u c l e a r b l a n k e t e v a l u a t i o n , a n d

i n t e r e s t i n g c o n c l u s i o n s w e r e l o o k e d f r o m t h e p a r a m e t r i c

s t u d y w i c h p e r m i t t o e v a l u a t e t h e f i r s t w a l l r a d i u s , a n d t h e

s i t u a t i o n a n d m a g n i t u d e o f t h e t w o p r i n c i p a l z o n e s w i c h a r e

t h e c o m p o n e n t s o f t h e b l a n k e t ( 9 , 1 0 ) .

On t h e t h e r m a l b l a n k e t s w i c h r e s u l t f r o m t h e

h o m o g e n e o u s s y s t e m s o f f u s i o n - f i s s i o n (9, l o ) , t h e WIMS-

TRACA (11) a n d DTF-TRACA (12) h a v e b e e n u s e d . The f i r s t o n e

c o d e p e r m i t s t o o b t a i n t h e l i b r a r y w i c h u s e s t h e DTF-TRACA

that calculates the typical source problem. The energy

group from 10 Mev to 14 Mev, necessary to consider the

fusion problem, that isn't in the WIMS code, has been in-

- troduced from the ENDF/B-IV, considering a slowing-down

weighting spectrum.

References

1. J.D. LEE, "Neutronics analysis of a 2500 Mwth Fast Fi-

ssion Natural Uranium Blanket for a DT Fusion Reactor",

2. J.D. LEE, "Mirror Fusion-Fission Hybrids'!, Atomerkener - pie 31 (1) 19-29 (1978).

3. Reference design for the Standard Mirror Hybrid Reactor,

Lawrence Livermore Lb. UCRL-52478.

a 4. N.M. GREENE, C.W. GRAVEN, Jr.,"XSDRN, A Discrete Ordina-

tes Spectral Averaging Code", ORNL-TM-2500,

5. H. KSCHWENDT, H. RIEF, "TIMW, A General Purpose Monte-

Carlo Code for Stationary and Time Dependent Neutron Trans

port", EUR 4519 e, 1970.

R. JAAF.SMA, H. RIEF, "TIMOC-72 Code Manual", EUR 5016 e,

(1973).

6. M.J. BELL, "ORIGEN, The ORNL Isotope Generation and De-

pletion Code", ORNL-4628.

7. H. KRAINER, "CODAC, A Fortran IV Prograwme to Process a

TIMOC Library from the ENDF/BU, EUR 4521 6 (1970).

8. J.M. PERLADO. "TIMOC-TASAS Code, to Calculate Typical

Blanket Reaction Rates in Hybrid Systems", TRCN (JEN)

to be published.

9. R. CARO, E. MINGUEZ, J.M. PERLADO, "Hybrid Reacrora An2

lysis; High Burnup and Transmutation Problem", JEN/TCR/

A-79-17. Submitted to the V Annual Meeting of the Nu-

clear Spanish Society (1979).

10. R. CARO, E. MINGUEZ, J.M. PERLADO, "Nuclear Blanket A n 2

lysis of a Hybrid Reactor", Progress Report. Report JEN

to be published. Submitted to the Monographic Meeting on

Fusion Problems of the Nuclear Spanish Society (June

1980).

11. C. AHNERT, "The Neutron Transport Code DTF-TRACA User's

Manual and Input Data", JEN-b48.

12. C. AHNERT, "Programa 'WIMS-TRACA' para el Cdlculo de E l e

mentos Combustibles", Manual de Usuario y Datos de Entra

da. JEN-h61.

CASE

d e n s i t y ( g . ~ r n - ~ )

rad ius (crn)

d e n s i t y ( g . ~ r n - ~ ) mass (mg) radius (cm)

Hax. pressure o n o u t e r b o ~ n d a l y ( b a r )

C x s r e s i o n work (NJ)

I riternal energy f i.(.l) Tot31 energy (MJ)

Ir ,ternal g a i n T z t a l ~ a i n

STUDSVIK ENERGITEKNIK AB

K i m Ekberg, e d

R e a c t o r P h y s i c s i n Sweden

S t a t u s R e p o r t f o r NEACRP

NEACRP-L-244 SWEDEN

1. INTRODUCTION

Al though t h e impasse w i t h r e s p e c t t o n u c l e a r

power i n Sweden now h a s been r e s o l v e d t h e a c t i -

v i t y i n r e a c t o r p h y s i c s r e m a i n s a t a b o u t t h e same

l e v e l . Work h a s c o n t i n u e d o n d i f f e r e n t f u e l c y c l e s ,

h i g h e r a c t i n i d e s a n d f i s s i o n p r o d u c t s and o n w a s t e ,

u t i l i z a t i o n and t r a n s m u t i o n . The e f f o r t f o r code

deve lopmen t i s s m a l l , and t h e a c t i v i t y i n t h e f a s t

r e a c t o r f i e l d is d e c r e a s i n g .

2 . FUEL CYCLE PROBLEMS

2.1 C a l c u l a t i o n s o n c o a t e d p a r t i c l e f u e l

CASMO c a l c u l a t i o n s have been pe r fo rmed o n a c o a t e d

p a r t i c l e f u e l e x p e r i m e n t g o i n g on a t t h e m a t e r i a l

t e s t i n g r e a c t o r R2 i n S t u d s v i k . T h i s e x p e r i m e n t

c o n t a i n s b o t h p e b b l e bed f u e l and f u e l o f t h e b l o c k

t y p e . The p u r p o s e was t o r e l a t e t h e r e a c t i o n r a t e s ,: i n s i d e t h e f u e l t o measured a c t i v i t y i n n e a r b y co-

b o l t d e t e c t o r s . An e x t e n d e d v e r s i o r o f CASMO sup-

p l i e d w i t h t h e capability t o h a n d l e t h e t h o r i u m

c y c l e was u s e d a s well a s a m o d l f i e d l i b r a r y t a k i n g

i n t o accoulat t h e e f f e c t s o f g r a l n s h i e l d i n g f o r t h e

r e s o n a n c e s Jn Th232 and U238.

2 . 2 C a l c b l a t i o n s o n heavy n u k l i d e s f c r

- Swede~:'s 1 2 1 . w ~ ~

The amount and n c t i v i t y o f heavy n u c l i u ~ . w i t h o r

w i t h o u t p l u t o n i u n r e c y c l i n g have been c a l c u l a t e d

w i t h t h e c o d e s ChSMO and BEGAFIP tor Sweden ' s

LWR-system c o n c i s t i n g of 9 I3I:Rs a n d 3 P'iJRs. I n

p a r t i c u l a r a f i n a l p h a s e w i t h o u t uran ium and a f t e r

a b o u t 3 0 y e a r ; of r e a c t o r operation h a s t e e n s t u d i e d .

The work also inc1,udes some investigations on the

CASblO code such as a comparison with burnup re-

sults from the Trino-Vercellese reactor.

2.3 Continued studies on the thorium cycle

Calculations have been performed with CASMO (ex-

tended version) on alternative operation of Sweden's

12 LWRs using the thorium cycle in normal or de-

natured versions.

2.4 Waste management

6 manmonths a year are allocated for studies of

waste utilization and transmutation. Important

foreign results have been reviewed. Calculations

using an extended CASMO version indicate that

waste actinides can be used as burnable absorbers

in LWR. Results were reported at "The Second Tech-

nical Meeting on the Nuclear Transmutation of

Actinides", which was held in Ispra, Italy

April 21-24, 1980.

3 . POWER REACTORS

After the referendum on the use of nuclear power

in Sweden loading permits have been given for two

reactors and are expected shortly for two more,

all belonging to the Swedish State Power Board.

Activities within the SSPB have been concentrated

on starting these reactors. In addition, SSPB and

the other two nuclear utilities in Sweden have a

rather high ambition what regards optimizing the

operation and fuel management of the six reactors

previously operating.

A study on the use of gadolinia in PWR has been

made at the SSPB. Results indicate that it is

possible to design a reload for a long cycle using

this concept. A s with other burnable absorber con-

cepts, however, it requires a large computational

effort.

The u s e o f t h e c e l l and a s sembly c o d e CASMO is i n -

c r e a s i n g , e s p e c i a l l y among US u t i l i t i e s . The n o d a l

r e a c t o r s i m u l a t o r code SIMULATE h a s been implemented

a t S t u d s v i k . F u r t h e r benchmark s t u d i e s i n v o l v i n g

CASMO and SIMULATE a r e u n d e r way i n S t u d s v i k and

e l s e w h e r e .

4 . CODE DEVELOPMENT WORK

A twod imens iona l d i f f u s i o n c o d e , MBS, i s b e i n g

d e v e l o p e d f o r PWR a p p l i c a t i o n s . The code w i l l t a k e

c r o s s s e c t i o n s produced by CASMO and w i l l b e u s e d

f o r core f o l l o w and power d i s t r i b u t i o n c a l c u l a t i o n s .

CASPIO is b e i n g p r o v i d e d w i t h v a r i o u s new o p t i o n s

e x t e n d i n g i t s a p p l i c a b i l i t y t o new p rob lems .

Development work h a s s t a r t e d o n a new one-dimen-

s i o n e d code f o r dynamic r e a c t o r problems. I t w i l l

be based o n a n e x i s t i n g c o d e , TRAC.,The new c o d e

w i l l be modu l i zed , a n d e f f o r t s w i l l be d i r e c t e d

t o w a r d s s p e e d i n g u p t.he N u c l e a r Steam S u p p l y mo-

d u l e a n d m o d e l l i n g t h e b a l a n c e o f p l a n t (BOP).

5. FAST BREEDERS

5 .1 He te rogeneous LMFBR cores

The development of h e t e r o g e n e o u s cores and t h e

more r e c e n t d e s i g n s a d o p t e d i n d i f f e r e n t c o u n t r i e s

have been r ev iewed . The a d v a n t a g e s i n b r e e d i n g p e r -

formance and i n g e n e r a l s a f e t y a s p e c t s r e l a t i v e t o

homogeneous c o r e s have been a n a l y s e d .

The l a y o u t o f h e t e r o g e n e o u s c o r e s h a s conve rged t o -

wards c o n f i q u r a t i o n s c h a r a c t e r i z e d by a c e n t e r b r e e -

d e r r e g i o n s u r r o u n d e d by a s i n g l e e n r i c h m e n t c o r e

r e g i o n p a r t l y d i v i d e d by t h i n , b roken b r e e d e r r i n g s .

However, a wide s p r e a d i n t h e r a t i o o f i n t e r n a l

b r e e d e r volume t o c o r e volume e x i s t s a t p r e s e n t

f rom a c u r r e n t F rench c h o i c e o f 0 .16 to t y p i c a l

US d e s i g n s a round 0.5. Compared t o e q u i v a l e n t

homogeneous c o r e s t h e b r e e d i n g r a t i o i s a l w a y s i m 4

p roved and t h e d o u b l i n g time a l s o ( t y p i c a l l y by 1 0 % )

p r o v i d e d t h e geometry a l l o w s a good n e u t r o n i c Coup-

l i n g . The sodium v o i d r e a c t i v i t y o f t h e c o r e zones

d e c r e a s e s r a p i d l y w i t h i n c r e a s i n g r a t i o o f b r e e d e r

volume t o c o r e volume. The r educed v o i d e f f e c t l e a d s

i n g e n e r a l to s m a l l e r e n e r g y r e l e a s e s i n t h e d i s -

a s sembly p h a s e o f HCDAs. However, a more c o h e r e n t

v o i d i n g o f t h e c o r e s u b a s s e m b l i e s may change t h e

c o u r s e o f t h e a c c i d e n t s t o more e n e r g e t i c e n d re-

s u l t s . A v a i l a b l e s t u d i e s d o n o t g i v e c o n c l u s i v e

r e s u l t s and f u r t h e r more d e t a i l e d s t u d i e s a r e

needed t o d e m o n s t r a t e c l e a r l y t h e s a f e t y a d v a n t a g e

o f h e t e r o g e n e o u s c o r e s .

5.2 Comparison o f n u c l e a r p a r a m e t e r s f o r a

LMFBR h e t e r o g e n e o u s benchmark c o r e .

A LElFBR h e t e r o g e n e o u s c o r e model was a few y e a r s

a g o p roposed by CEA a s a Benchmark c o r e f o r com-

p a r a t i v e c a l c u l a t i o n s . The g e o m e t r i c a l RZ model

c o n s i s t s o f t h r e e r a d i a l f i s s i l e z o n e s o f t h e

same e n r i c h m e n t d i v i d e d a t t h e m i d p l a n e by a n

a x i a l s l ice o f i n t e r n a l b r e e d e r m a t e r i a l . The

f i s s i l e zones a r e s e p a r a t e d by t h r e e i n t e r n a l

b r e e d e r z o n e s , o n e c e n t r a l zone and two b r e e d e r

r i n g s . They a r e a l s o s u r r o u n d e d by a n a x i a l and

r a d i a l b l a n k e t and a r a d i a l r e f l e c t o r .

The c o r e h a s been s t u d i e d w i t h 2D d i f f u s i o n c o d e s

i n 1 0 t o 2 6 e n e r g y g r o u p s . Compar isons have been

made between CEA (CARNAVAL 111) INTERATOM (KFK I N R )

a n d STUDSVJK (ENDF IV) s o l u t i o n s .

Neut ron c r o s s s e c t i o n s had been p r o c e s s e d f rom

ENDF/B-IV b o t h a t S t u d s v i k and a t INTERATO>l, which

e n a b l e d a compar ison o f t h e d i f f e r e n t p r o c e s s i n g

c o d e s . I t a p p e a r e d t h a t t h e p r o c e s s i n g methods i n -

f l u e n c e d t h e r e s u l t s t o a h i g h d e g r e e . F o r t h i s \ ... ,.., ., ,.

r e a s o n , t h e i n t e g r a l r e s u l t s a g r e e d b e t t e r be tween 9.t' . L ... . ',;d " i 6<'

CEA and Studsvik than between INTERATOM and

Studsvik.

Comparing keff and breeding ratio we find that

the spread in keff is 1.8. % with the lowest value

for Studsvik (ENDF IV) and the highest value for

INTERATOM (KFK INR). The spread in breeding ratio

is 0.03 with the highest value for Studsvik and

the lowest for INTERATOM.

The CEA keff and BR lie between but rather close

to the Studsvik values. The CEA breeding ratio is

comparatively high which is explained by the lower

capture rates in the structural materials.

6. DECAY HEAT

Two sets of experiments are under way in Studsvik

both aiming at improving the accuracy of available

data on fission product ( F P ) decay heat.

In one set of experiment a radiometric method is

used to study the decay of FP from small uranium

and plutonium specimens irradiated at the Studsvik

van de Graaff machine. Gamma measurements on 6 - 2 3 5

have been finished and reported. The results gave

an uncertainty of - + 7 %(la) for the total gamma ray energy in the decay range of 14 - 1500 seconds. Beta measurements on U-235 will soon be reported.

Studies of Pu-239 with the same technique as for

U-235 will continue. The measurement of beta energy

have been finished and the analysis is in progress.

Because of lack of man power the project has been

delayed.

The second set of experiment is part of a large

program for studying the decay properties of

fission products. 1.1-235 samples are used in the

ion source of a niass spectrometer allowing indivi-

dua l f i s s i o n produc ts t o be i s o l a t e d and measured.

The average be ta decay energy f o r a l a r g e p a r t of

t h e a v a i l a b l e e lements has been determined. Mea-

surements of average gamma e n e r g i e s a r e i n p rogress .

The t ime r e s o l u t i o n i s from a few seconds and up-

wards.

NEACRP-L-244 SWITZERLAND

REACTOR PHYSICS ACTIVITIES IN SWITZERLAND

October 1979 to September 1980

Compiled by P. Wydler

1. Introduction

In the research programmes of the Federal Institute for Reactor Research

(EIR) emphasis has shifted from reactor physics to safety problems. This

applies particularly to the fast reactor area, where the core physics and

alternative fuel cycle studies (theoretical and experimental) have

effectively been discontinued. On the other hand, the physics of conven-

tional and advanced LWRs has received additional interest. These changes

in the programmes reflect important facts such as the present (world-wide)

difficulties with the GCFR development programme and the increasing

dependence of the Swiss energy economy on LWRs. For instance, in 1979,

LWRs contributed 24.7 % to the total electricity production of the country

(With the full availability of the 1000 MWe Gasgen plant the 30 % limit

will be exceeded in 1980).

2. Conventional LWRs and Compact Storage Pools

A programme package consisting mainly of the fuel element calculation code

BOXER and the LWR simulator SILWER [l] has been further developed and

validated. Numerical improvements in the resonance calculation and

allowance for time-dependent microscopic one-group cross-sections and/or

the predictor/corrector method in the burnup calculation have enhanced

the accuracy of the codes. The 3-dimensional nodal code now offers an

analytical solution 121 with favourable computing times in addition to

the polynomial expansion solution.

The programmes have been applied to various aspects of BWR cycles using

data typical of the Mfihleberg plant. In particular, the role of Xe in power

peaking effects resulting from the retrieval of control rods was studied.

Under contract the code package was used for accurately determining the

reactivity of compact storage pools offered and built by Sulzer in Switzer-

land and abroad. For different PWR fuel element configurations, reactivity

measurements have been performed at the compact storage pool of G6sgen

using the pulsed neutron source technique. A comprehensive analysis of

these measurements is in progress.

With BOXER work is in progress on the problem of Gd burnup in BWR fuel

elements. A method for handling Gd pins in adjacent lattice positions is

under evaluation. Fuel elements of this type are used in the Milhleberg

reactor to enhance the performance of the fuel (increased burnup due to

reduced reactivity swing). For checking the predicted Gd depletion in

neighbouring Gd pins experimentally, it is proposed to perform irradiations

and neutron transmission measurements on Gd pins with high spatial reso-

lution using the reactor SAPHIR.

The BOXER/SILWER code package is also being applied to the CSNI benchmark

problem on spent fuel transport casks.

3. Experimental Studies on Tight Pitch LWR Lattices

4 The current rate of depletion of uranium by the use of uranium in standard

LWRs would lead to the exhaustion of workable reserves within a few

decades. There is therefore a considerable incentive to improve the

utilization of uranium and (since about 80 % of current nuclear power is

produced by LWRs) such an improvement must come, in the first instance,

from the LWR area itself and not from the advanced

Various proposals have been made for improving the

LWRs, generally on the lines of introducing a more

reactor area.

conversion ratio of

tightly-packed lattice

into a standard LWR. A wide range of assessment studies has been carried

out and these have included investigations of "homogeneous" (i.e. uniform)

lattices, heterogeneous lattices with separate breeding zones and seed-

breed configurations with reactor control effected by movement of the seed

elements. Both the U-Pu cycle and the cycle have been considered

although the latter is not yet of practical significance.

Assessment studies are currently being carried out at the Kernforschungs-

zentrum Karlsruhe with particular emphasis on the calculation of the void

coefficient which has important safety implications. Further studies are

in progress at KWU and also at MIT.

A major problem in assessing the validity of the various studies made in

the tight pitch area arises because essentially no benchmark experimental

measurements have been made on lattices of this type so that no integral

data is available to check the performance of data sets and calculation

methods used in these studies.

It is therefore proposed to set up tight pitch 1.attices in the central

zone of the reactor PROTEUS, measure their physi.cs properties and use the

results to check data sets and calculation methods.

Lattices covering a range of average enrichments. will be investigated and

various degrees of Hz0 voiding will be simulated.

In the meantime lay-out calculations have been carried out using WIMS-Dl.

Engineering work for converting the reactor is in progress. The measure-

ments are planned to begin in early 1981 and to continue for a period of

about one year.

4. GCR Physics and Fuel Cycle Studies

In this area also the emphasis is on safety studies and preparations for

the measurement of the physics properties of low-enriched HTR lattices at

elevated temperatures are continuing [3,4]. The main emphasis will be on

the determination of the temperature coefficient of reactivity and the

results will be used in support of the licensing submission for the HHT

demonstration plant. This work is being carried out in association with

the HTR-Entwicklungsgemeinschaft.

The GCFR core physics and fuel cycle studies have been terminated

[5,6,7].

5. Reactor Noise Analysis

The activities in this field mainly concentrated on the analysis of two-

phase flow characteristics in BWRs 181. In addition to this work, new appli-

cations of the noise analysis technique for problems such as the early

detection of the onset of boiling and the measurement of Xe poisoning in

a swimming pool reactor were examined.

In more detail, the work covered the following subjects:

- Comparison of fluid velocity profiles measured in the Miihleberg BWR

with thermal-hydraulic subchannel analysis results predicted by COBRA

computer calculations. The fluid velocity profiles were obtained by

cross-correlating the high frequency neutron noise (f > 4 Hz)

measured with axially-placed in-core detectors.

- Development of transient analysis methods for measuring nonstationary

fluid velocities in operating reactors and in ex-core test loops

[9, 101 . The methods appear to be applicable, for instance, to re-

flooding experiments.

- Detection of the onset of nuclear boiling in the reactor SAPHIR. Bubble

collapse was found to produce characteristic neutron noise signatures.

Using this principle, an on-line surveillance system is presently under

development.

- Measurement of the Xe poisoning of the reactor SAPHIR after shut-down

using an on-line reactivity meter developed at EIR and ex-core detectors

located near the core boundary in the reflector. The investigations

showed that an analysis using the point reactor model is adequate for

these experimental conditions.

6. LWR Pressure Vessel Surveillance Dosimetry

EIR is in charge of the pressure vessel surveillance programme on the

Swiss power reactors. Material samples are tested regularly for mechanical

property changes due to the radiation exposure. The neutron dosimeters 237 238"

NP r , Ni, Fe and Cu are used to determine integral neutron fluences in the energy ranges E > 0.1 MeV and E > 1.0 MeV by gamma activity measure-

ments.

It was found that calculated and measured detector reaction rate ratios

agree well, i.e. ANISN calculations adequately predict the shape of the

neutron spectrum in the capsule position. An important problem is the

accurate determination of the lead factor, i.e. the damage or fluence 1

ratio between capsule position and - T (thickness of the vessel). Measure- 4

ments were performed in a Swiss power reactor to determine the lead

factor experimentally.

Research on new dosimeters and the improvement of the existing methods

has been continued. Particular attention is presently focused on the gamma

sensitivity of the monitors Nb, 238U and Ni.

REFERENCES

I. C. Maeder et al. "Calculations with the EIR Light Water Reactor Code System", NEACRP Specialists' Meeting on Calculation of 3-Dimensional Rating Distributions in Operating Reactors, Paris (Nov. 1979)

2. M. Makai, C. Maeder "SEX1 - A Fast Diffusion Program Based on Exponential and Trigonometric Interpolation", EIR-Report 401 (1980)

3. R. Chawla et al. "Heterogeneity Effects in Temperature Coefficient Calculations for MEU/LEU Pebble-Bed HTRs", Trans. Am. Nucl. Soc., 3, 863-866 (1979)

4. R. Chawla and R. Richmond "A Comparison of MICROX and WIMS-Dl Calculational Results for HTR Lattices", Jahrestagung Kerntechnik 80, Berlin, 23-26 (1980)

5. U. Schmocker et al. "Reaction Rate Distributions Through a Central Thorium Metal Zone in a GCFR Lattice", Trans. Am. Nucl. Soc., 34, 761-763 (1980)

"A Detailed Study of the Steam Entry Effect of a Gas-Cooled Fast Reactor", IAEA Specialists' Meeting on Gas-Cooled Reactor Safety and Licensing Aspects, Lausanne (September 1980)

7. P. Wydler et al. "A Uranium-Plutonium-Neptunium Fuel Cycle to Produce Iso- topically Denatured Plutonium", Nuclear Technology, 49, 115 (1980) -

8. K. Behringer, R. Crowe "Practical Application of Neutron Noise Analysis at Boiling Water Reactors", EIR-Report 385 (1980) To be published in ATOMKERNENERGIE

9. D. Liibbesmeyer, R. Crowe "Time-Dependent Coolant Velocity Measurements in an Operating BWR", to be published in Annals Nucl. Energy

10. D. Lubbesmeyer, B. Leoni "Geschwindigkeits-Messungen und Identifikation der Stromungsformen von vertikalen Luft-Wasser StrBmungen mit Lichtschranken-Sonden", EIR-Report 400 (1980)

NEACRP-L-244 UNITED KINGDOM

REACTOR PHYSICS I N THE UNITED KINGDOM

J R Askew C G Campbell

GENERAL PROGRAMME

The c u r r e n t l y approved r e a c t o r bu i ld ing programme i s confined t o t h e AGR s t a t i o n s a t Torness (SSER) and iieysham (CEGB), both of 2 x 660 MW(e).

0 Licence agreements have been concludcd wi th Westinghouse With a view t o bu i ld ing PWRs t o t h e i r des ign. A p u b l i c enquiry is expected t o be he ld i n 1982.

THERMAL REACTORS

At t en t ion has been concent ra ted upon t h e problems of modelling s a f e t y r e l a t e d t r a n s i e n t s i n PWRs. I n p a r t i c u l a r t h e RETIUN code has been purchased and deve1opc:d t o g i v e it t h e c a p a b i l i t y t o model ATWS events . Good agreement has been

. . demonstrated between t h e r e s u l t s g i v e n i.n t h i s code f o r a s eve re t r a n s i e n t ( l o s s of feed water wi thout r e a c t o r o r t u r b i n e t r i p ) and those of t h e Westinghouse LOFTRAN and A1:tiebolaget Atomenergi TRANS-PWR codes. This g i v e s some confidence t h a t t h e b a s i c mechanics of t h e code is s a t i s f a c t o r y f o r t h i s app l i ca t ion . It must, however, he recognized t h a t key physics d a t a , f o r example on the . expec t ed flow through s a f e t y va lves , p r o b ~ b l y stems from t h e same source and t h e assessments a r e n o t wholly independent.

Three dimensional s t u d i e s o f rod eject . ion t r a n s i e n t s have ,

been made us ing t h e M E K I N code. I t was not p o s s i b l e t o niodel t h e most s eve re asymmetric t r a n s i e n t s on a f u l l s i z e r e a c t o r , and t h e

. problem had t o be s i m p l i f i e d t o permit a so lu t ion . The t imesca le and magnitude of t h c t r a n s i e n t is illustrated i n Figure 1.

Fur the r s t u d i e s of t h e a p p l i c a t i o n o f t h e MONK Monte Car l0 C code using group d a t a der ived from WIMS have demonstrated t h a t

t h i s i s an a t t r a c t i v e t o o l f o r a number of d i v e r s e app l i ca t ions . Ca lcu l a t i ons of shutdown and holddown margins, he igh t , . bo ron and temperature c o e f f i c i e n t s on t h e SGHWR r e a c t o r were c a r r i e d Out wi thout s i g n i f i c a n t geometric approximation, a l l t h e p i n s (up t o 60 i n each fuel. elemcnt) i n each of t h e 104 channels being represen ted exp l i c i . t l y . T h i s g r e a t l y i m ~ r o v c d niodell ing, e s p e c i a l l y i n t h e d i f f icu1 . t leakage regimes p o s s i b l e i n 'his r e a c t o r where pressure tubes containing f u e l and l i g h t watcr may extend wel l above t h e l e v e l of t h e d~nnped heavy water moderator.

lksuits for two of the quantities studied are shown in the ..?5les below. The sutdown margins show that whilst the changes becween different core configurations in the cold state are well predicted, there remains a significant hot to cold discrepancy which cannot be ascribed to reactor modelling. The comparison between different calculations of boron worth for the infinite lattice, on the other hand, demonstrated that this is sensitive to modelling, the Monte Carlo predictions being 10 + 5% greater than thase from deterministic calculations. This was reflected in better predictions for the reactor. The cost of such calculations was competitive with that from simpler models, the greater solution time being offset by the less 1.aborious lattice calculations required.

Comparison of Boron Worths Predicted by WIMS (DSN) and MONK

Concentration of B-10 Calculated Values of pm

in Moderator WIMS

4.15 11.97 Change

MONK Calculations of Shutdown Margin

Reactor

WSGHWR (cold) (Core 92 to Core 58)

WSGIiWR (Hot - Cold) (Core 164 to Core 58)

Discrepancy in

Shutdown Shutdown Worth Calculated by MONK

1.02 + 0.35 Niles 6.1tiiles 1 2.21111 1 -

5.7 Niles

In other applications the benefits of using the WIMS library stem from the ability to adduce a wide range of experimental validation evidcnce, obtained using deterministic methods, to support arguments about predictive accuracy. The idea of post calculati,onnl smoothing of rating predictions has continued to be of value in lattice cell calculations, and studies are in hand to explore the applicability of similar techniques to whole reactor rating assessments.

1.71 m -0.09 t 0.21 Niles

Use of the same data base also facilitates comparison of the Monte Carlo geomctric solution capabilities with deterministic methods in the sub-set of gcomctrics for which this is possible. Sbme verification studies have been carried out in simple x-y layouts, covering the new characteristics discrete ordinate code CACTUS as well as TWOTRAN and collision probability formulation (MINOS).

. A modest revision of nuclear data is in hand, aimed only at removing some significant discrepancies in the WIMS library identified by integral comparisons and re-evaluation of the differential data where appropriate. Out of date resonance data for capture in Plutonium-239 leads, in the current version, to significant reactivity errors in intermediate spectrum Systems. It is considered that thc 'hermal Uranium-235 cross sections can now be improved. Preliminary analysis of measurements from Harwell show eta to be significantly cnergy dependent below 0.1 eV, consistent with the earliest measurements but inconsistent with most current evaluations. Figure 2 shows alpha from this analysis. The change, together with smaller changes in energy dependent absorption cross-sections, will help to reduce modefator temperature coefficient discrepancies in all types of rcactor.

Validation of fuel management predictions continues for all reactor types. Figure 3 shows the reactivity prediction for a number of fuel cycles in the Beznau and Tihange PWRs using the lattice code LWR WIMS and reactor code JOSHUA. It has been established that the use of improved algorithms of a coarse mesh corrected kind, as opposed to the simple diffusion-like formalism used in the reported results, would give a slight improvement in observed rating distribution predictions (Figure 4) equivalent to approximately 2% in the ratio of power in the chequerboard of different irradiations of fuel in thc inner zone of the reactor.

NUCLEAR DATA

Work has continued on the analysis of capture data for Fe, Cr and Ni, and the analysis of natural metallic: iron capture data has been completed. Resonance parameters for all but a few minor resonances have been determined up to 40 keV. Below 10 keV, 8 resonances were observed wkjch have not been previously reported, and5kave been attributed to Pe. For the important 1.15 IceV resonance in Fe, the parameters are in excellentagreenent with those rqorted from transmission measurements by Geel. In the energy region from 40 to 200 keV, the aggeelnent with capture yie1.d~ calculated from the evaluation of Fe by E'. Perey is not good, and in some energy regions the calculated yields are as much as a factor of two higher. Evaluation of the data for iron is in progress.

Evaluation of data for the 2 4 3 ~ m has progressed; a set of resonance parameters has been chosen for neutron energies up to

a 250 eV and appropriate parameters determined for bound levels in order to agree with the shape of the total cross-section in the thermal region. The evaluation of the measured fission cross-section data in the fast region has been completed.

Uncertainties have been assessed, arising from the nuclear data, when the FISPIN code is used to compare such quantities as:

a the arisings of certain minor constituents of irradiated fuel

b the contribution to decay power due to the actinides

c the total neutron production from spent fuel.

Sensitivity coefficients obtained using FISPIN have been combined with assessments of the accuracies of cross-sections and other relevant data to see if the target accuracies required are achieved and to identify where better dat?&ght be required. Deficiencies found were that the arisings of Pu and items b and c could not be calculated with the required accuracy using current data. This has led to the specification of new nuclear data requests.

As part of the continuing application of ZEBRA to integral data . s&dies relevant to the fast reactor fuel cycle, the reactor has been operated under steady conditions to allow the irraqiatioq of a large number of samples to a peak integrated flux of-10 n/cm . The samples, which are being a n a ~ ~ ~ e d in cggjunc.clion with Chemistry Division, Harwell., included Am and Am for furt9~5 measu~fiyents of the respective capture cross-sections leading to 'Cm and Cm, typical fast reactor cladding and wrapper alloys (PE16, M31.6, FV548) to investi236e activ9tjAon of potential primary-circuit corrosion products, Pu and U for 0 and y-decay power determinations, and B4C pellets to examine the feasibility of using mass-spectrometric

kechniques to obtain the absolute 1•‹g(n,a) reaction-rate in control rods by helium production.

The 1980 versions of the U.K. Nuclear Data Library ape's NDLl (milin library) and NDL3 (dosimetry) have been completed and despatched to the NEA Databank in Paris.

The BIZET Programme

The joint UK/KfK BIZET experimental programme in ZEBRA was completed in August 1980. During the last year the experiments have concentrated on the study of a clean reference version of the annular core (with a central breeder island). Earlier work had studied two versions of the core with 12 partly-inserted control absorbers each with different sizes of central breeder, and a version with the control absorbers removed but with the sodium-filled auide tubes <

a remaining.

The following topics were covered in the clean reference version:

(i) critical loadings and associated corrections;

(ii) fuel element reactivity worths in pin and plate geometry, including sector replacement;

(iii) reactivity-scale calibration by kinetic and static (enrichment) techniques;

(iv) cell reaction-rates in fissile and fertile elements;

(v) sodium void reactivity distributions and the integral worth of a 97-element voided zone;

(vi) reaction-rate distributions including Pu-239 and U238 fission (for power), U238 capture (for breeding) and Rh (n,n') and Fe54 (n,p) (for damage).

(vii the inf luence of poisons (B,No) simulating fission products on sodium-void reactivity, and complementary small-sample perturbation measurements in normal and voided environments ;

(viii) worth distributions including structural and absorbing materials, and hydrocarbon at two densities;

(ix) reactivity and power distribution measurements following 1.0% radial fuel compactions and creation of sodium-filled and voidcd streaming slots;

(XI replacement of the inner part of the central breeder island by sodium/steel elements.

:. . j s ot BIZET Results . . -- . .. - - - - - The analysis of the BIZET experiments uses data and methods as

far as possible the same as those currently used for C D F R design. The analysis of the conventional two-zone design (BZA and BZB) is well advanced, and in general shows that 3D diffusion theory with FGL5 data predicts the major core properties to the desired accuracies. There are, however, exceptions t.o this which require further attention:

(i) As reported at the Aix meeting, introduction of a sector zone of pin fuel into the B Z B core produces an unexplained flux tilt, which suggests that the reactivity of pin cells is overpredicted relative to pldte cells by more than 1%.

(ii) Whilst the influence of both distributed and lumped B C on sodium-void worth is reasonably well pre$icted, that of the resonmt absorber molybdenum is appreciably overestimated. This suggests errors in the molybdenum data and in turn simjlar errors in the effects of the fission product term on sodium void worths.

The analysis of the heterogeneous B I Z E T cores ( B Z C , B Z D ) is being made initially using the same data and diffusion theory methods as applied to the conventional cores (BZA, B Z R ) , except that for the internal breeders the MUIWL cell calculations used group-dependent bucklings to allow for the rapidly varying spectra. Amongst the conclusions drawn so far from the analysis are:

(i) The diffusion theory k-valuo for the salt-and-pepper (distributed island) core Y Z C - 1 is 0.992, implying a correction of +0.009 for mesh, transport theory, and resonance interaction effects between fissile and fertile cells to bring the prediction into line with the value of 1.001 found in conventional cores. This correction is some three times greater than for conventional cores. For the BZD/ :3 annular core the correction from the standard mesh diffusion theory k-value to the zero-mesh size transport theory calculation is estimated from R-Z DOT calculations to be 4-0.006, bringing the C / E for the k-value to 0.998~0.002.

(ii) Diffusion theory predictions of the radial Pu-239 fission rate distributions are moderately well predicted in n%C-1 . There is a 4% dispersion in the C/E-values on the core mid-plane with higher values in the breeder islands and lower values in the fissile regions. However, much larger discrepancies are observed for U-238 fission, with calculation over-estimating the reaction rate by up to 18% in the breeder regions. For the annular cores (BZD-1) the Pu 239 distribution in the fissile annulus is well predicted by 337group X Y Z diffusion theory, hut is over-predicted by 7 to 9% in the large ccntral breeder island.

(iii) Good agreement is again found between calculation and experiment of the reactivity changes produced by enriching selected elements in BZC. The implied agreement between the kinetic reactivity scale based on calculated delayed neutron parameters and the static scale based on calculat.ed plutonium worths is perhaps surprising in view of the U-238 fission distribution discrepancies, because U-238 is an important contributor to the delayed neutron fraction.

(iv) In the salt-and-pepper core BZC, the small-zone sodium void worths in thc fissile regions are all underpredicted by first-order perturbation theory. Where the leakage is small, in the regions betwcen the central brecder and the surrounding breedcr i.slands, the under-estimate is typically 20% of the non-lcakage term (compared with+5% for conventional cores). On the other hand, agreement is close for the central breeder island and calculations over-estimate the voiding worths in the next group of islands closer to the core centre by 5 to 10% of the non-leakage terms.

(v) An experiment in the annular BZD/3 core progressively replaced plate e1ements.b~ pin elements; Extrapolation of these limited sector measurements suggests that an all-pin fuel-led version of this core would have a k-value some 0.012 lower than the all-plate version. The standard MURAL heterogenei.ty treatment of the pin and plate cells predi.cts reactivity difference of 0.002 in the opposite sense, implying a pin/pl-ate discrepancy of 1.4%. This provides further evidence of a discrepancy in the MUML treatment of heterogeneity of plate cells relative to pin cells.

Analysis of Breeder Reaction Rates

The analysis of the reaction-rate distributions in the radial breeder of the zebra PFRmork-up (Core 13k has been completed. Three conditions were represented in a 90 sector, viz. a clean uranium-oxide brecder, an irradiated UO breeder (2% Pu/U), and a - natural uranium-carbide breeder. The r2actions measured (by foil activation techniques) were fission of U235, U238 and pu239, and capture of U238. -

Cross-sections for the plate-geometry core and breeder elements were derived from FGLS/MURAL cell calculations. Comparisons were made first with diffusion-theory (TIGAR) in 16 groupsandone mesh per lattice pitch, this'being analogous to the standard method for PFR/CDFR calculation. Using this route, U23S and Pu239 fission and U238 capture rates in the breeder are overestimated (relative to those in the core) by several per cent. U238-fission is overestimated b y ~ l l % at the inner edge of thc! breeder,with C/E values decreasing to below unity at the outer boundary adjacent to the steel reflector. As a consequence, the weighted fission power from U235, U238 and Pu239 is overestimated by up to 9% adjacent to the core (where the

U2325 component is dominant) with bettw agreement at deeper penetrations. Improvements to the de~ivation of breeder cross- sections, by imposing a more realistic spectral environment in the MURAL cell-heterogeneity calculation vil the use of energy-depencent bucklings, reduce the calculated .U235, 3~233 fission and U238 capture rates by about 1 .5%, and the U238 fissica rate by up to 4 % , although these changes are partially off-set by t:.e effect$ of modifying the ccefficients to take h t o account preferential streaming in the sodiwn plates.

Transport-theory calcuations were mad6 using the DOT code in S4 and S8. Compared with the diffusion-themy solution, the U235t Pu239 fission and U238 capture-rates are rechced by a furtherk1.5%t with considerably larger changes (of up to 1.0%) for U238 fission. Most reaction-rates are predicted to within 5$., while the C / x weighted fission rate in breeder elements (relative to the core) is 1.03 i 0.03.

The increasc in breeder subassembly power due to the introd- ntween uction of plutonium in the burnt-up simulation (a factor of b-

2 and 3) is predicted' to within 2% by both methods, although transport-theory gives slightly better consistency. In the Zebra sinulations the amount of plutonium present is precisely known whereas in the power reactor case the uncert.ainty in plutonium production must also be allowed for. Predictions of the U238 capture rate are found to be relatively insensitive to the calculation method and a general tendency to overestimate by ( 4 ' 3%) relative to the core is deduced; this discrepancy is the most important in determining the power ratio.

Similar conclusions can be anticipated from the analysis of reaction-rates in the internal breeder regions of the heterogeneous cores. It has already been noted that diffusion-theory seriously overestimates the U238 fission ratc in the EZC fertile islands.

Future ZEBRA Progranune - The evidence for an error in the FC,L5/P?URAL calculations of

the relative reactivities of Zebra Pu02/ua fuelled pin cells and Pu- metal fuelled plate cells in lattices of d r y similar overall CoWositions (apart from thcir oxygen content) has already been discussed.

0

It is considered important to resolve this discrepancy prior to any further data adjustment work and As a first step a simple cylindrical core of 2 4 9 Pu/iu + U fuel is being built in Zebra. This core will initially be loaded with plate elements, which will then be progressively substituted by pin elements of which there are sufficient to provide about 802 of the critical mass. Measurements of excess reactivity and other diagnostic parameters will be made at each stage of this substitution and the results will be extrapolated to the wholly pin-fuelled situation. The sequence will be carried out in both normal and sodium-voided conditions to help to identify any errors due to treatment. of leakage. In parallel with this work, more exact theoretical treat- ments of the Zebra cell heterogeneity are being explored, in particular to improve the present one-dimensional representation of the plate-eel-1.

Methods Development - During the year, some of the more important development of

methods are as follows:

(i) Further work has taken place in design methods for control rods. The use of modified transport cross- sections for the control rod channel increases the calculated worth of a natural boron carbide control rod, relative to the channel, by some 4% which is significant in relation to the target accuracy of 5% for rod worths. -

(ii) Considerable effort has been applied to the development of metliods for determining the radiation- induced distortions of fast reactor cores and the consequent interactions between core components. The methods will allow for the irradiation history

0 of the reactor and changes in core loadings.

(iii) Effort is applied to the development of faster solutions of the transport equations, with a view to the possible development of 3D methods.

(iv) A new version of the Montc Carlo code MONK, used for critically purposes, is being developed with

' improved physics modelling and completely revised cross-section data.

(v) The techniques for assessing the effeots of data uncertaintics on shielding estimates have been appropriate only to configurations which can be modelled in one-dimension. A new method suitable for three-dimensional geometries has been developed as an option in the McBEND Monte Carlo code . A calculated reaction rate is expressed analytically as the probability of occurrenceof all contributing tracks and differentiation of these probabilities with respect to material cross-sections leads

. directly to appropriate sensitivities. The penalty in computing time is small as use is made of existing tracking information. The ASPIS iron benchmark experiment has been used to test this new approach, which has led to the adjustment of the elastic and non-elastic cross-sections. .

(vi) Nodal expansion solutions are being investigated for fast reactor applications. Acceleration techniques are being investigated in 2D rectangular geometries, and an hexagonal version is being specified. The objective of the work is to seek faster flux solutions than provided by TIGAR, the standard finite difference diffusion theory method.

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- Calculation - Experiment - Calculation- Experiment

NEACRP-L-244 UNITED STATES

Reactor Phys ics A c t i v i t i e s i n t h e U n i t e d Sta tes- -A Repor t t o t h e NEACRP

September 22-26, 1980 P. B. H e m i g & J. W. Lewe l len

U.S. Department o f Energy Washington, D.C. 2054.5

INTRODUCTIOA - Major-U.S. r e a c t o r p h y s i c s a c t i v i t i e s have con t inued t o p r o v i d e b a s i c d a t a and advanced techno logy r e q u i r e d f o r LMFBR d e s i g n and l i c e n s i n g a p p l i c a t i o n s . Improvements i n n e u t r o n i c comput ing c a p a b i l i t i e s f o r l a r g e heterogeneous f a s t r e a c t o r s a r e b e i n g assessed, a l t e r n a t i v e m a t e r i a l s and des igns f o r s h i e l d systems a r e b e i n g i n v e s t i g a t e d and a p p l i c a t i o n cmf t h e ENDFIB-V d a t a base f o r d e s i g n use has begun. @ The ZPPR c r i t i c a l f a c i l i t y i n Idaho has been used f o r s t u d i e s o f t h e n e u t r o n i c c h a r a c t e r i s t i c s o f CRBR. Source range c o u n t r a t e s and c o n t r o l r o d i n t e r a c t i o n measurements a r e among t h e r e s u l t s ob ta ined . The d i a g n o s t i c c o r e program u s i n g t h e ZPR-9 f a c i l i t y i n I l l i n o i s , i s p r o v i d i n g exper imen ta l s t u d i e s t o h e l p r e s o l v e l o n g s t a n d i n g d i s c r e p a n c i e s between c a l c u l a t i o n s and measurements o f sample wor ths , b r e e d i n g g a i n s , n e u t r o n spect rum and sodium v o i d c o e f f i c i e n t s . The Tower S h i e l d i n g F a c i l i t y has been used t o p e r f o r m exper iments f o r t h e Gas Cooled Fas t Reactor Program.

Phys ics measurements have been c a r r i e d o u t t o suppor t FFTF s t a r t u p and t o c h a r a c t e r i z e t h e FFTF i r r a d i a t i o n t e s t environments. Source range c o u n t r a t e s observed d u r i n g t h e approach t o i n i t i a l c r i t i c a l i t y were w e l l p r e d i c t e d . Severa l n e u t r o n and gamna s p e c t r a measurements were a l s o made i n FFTF a t v e r y l ow power l e v e l s .

CRITICAL EXPERIMENTS

ZPPR

C r i t i c a l exper iments have c o n t i n u e d t h r o u g h FY 1980 a t ZPPR on " z e r o " power r e a c t o r c o n f i g u r a t i o n s which s i m u l a t e t h e c u r r e n t CRBR n e u t r o n i c c h a r a c t e r i s t i c s and c o r e d e s i g n as c l o s e l y as p o s s i b l e . A broad range o f measurements have been conducted, i n c l u d i n g r e a c t i o n r a t e d i s t r i b u t i o n ; Doppler i n t h e i n n e r and o u t e r f u e l r i n g s ; f i s s i l e , f e r t i l e , s t r u c t u r a l , and po fson m a t e r i a l wor ths ; and r e a c t i v i t y changes due t o c o n t r o l r o d i n t e r a c t i o n s . The r e f e r e n c e d e s i g n i s a c e n t r a l b l a n k e t heterogeneous c o n f i g u r a t i o n w i t h t h r e e i n n e r r i n g and t w e l v e o u t e r r i n g c o n t r o l r o d s . The s t u d i e s which began i n December 1979 i n c l u d e f o u r phases:

A. A B e g i n n i n g - o f - L i f e (BOL) c o n f i g u r a t i o n w i t h s i x ( row 7 ) CRs one h a l f i n s e r t e d and t h e r e m a i n i n g r o d s parked.

B. A BOL c o n f i g u r a t i o n w i t h no i n s e r t e d o r parked CRs.

C. An E n d - o f - L i f e (EOL) c o n f i g u r a t i o n w i t h no parked CRs. Core enr ichment reduced and f i s s i l e b u i l d u p i n i n n e r and r a d i a l b l a n k e t .

D. An EOL configuration with six (row 7) CRs one-third inserted and the remaining rods parked. Core depletion and blanket buildup as in Phase C.

In the approach to critical for each configuration, count rates in the ex-vessel detector position are obta'ned. Such studies are made possible by including a rnockup of the ex-core reg'm in one section of the ZPPR matrix. . Under-consideration for future studies are licensing and safety-related measurements with an emphasis on sodium-void reactivities and studies of larger heterogeneous cores. .

The Diagnostic Core Program on ZPR-9 is outlined in Figure 1. The program involves six assemblies designed to isolate systematically the key components of critical measurement discrepancies. The program to date has concentrated on small, clean benchmark studies and on an evaluation of the small-sample reactivity worth experimental technique. These studies evaluated all phases of the experiments but concentrated on the local effects resulting from the interaction of the samples and the core. No experimental problems were discovered. The maximum possible error associated with the experimental techniques appears to be approximately 5:.

I\ benchmark core composed entirely of enriched uranium and steel was also built and evaluated. The C/E ratio for the small sample central worths and Beff were of particular interest in this core since it contained no Pu-239 and very little U-23E, yet it was dilute enough to have a neutron energy spectrum not too different from that of a typical FBR. The Beff C/E values were close to unity and typically about 52 higher than C/E values for the simulated mixed oxide criticals. This result, which is discussed in a separate paper, is in the direction to explain the small sample C/E bias between U and Pu cores. The central worth C/E measurements are still being evaluated.

Experiments are just beginning on another small, clean benchmark core which, except for the stainless steel in the ZPR matrix, is composed entirely of uranium (approximately 9% enrichment). The central worth and Beff values in this very-hard-spectrum all-uranium core are of great interest. Detailed reaction rate distribution and ratio data will be obtained in all of the Diagnostic Core Program assemblies. A benchmark core composed entirely of Pu-239 and steel is scheduled following the all-uranium core. The diagnostic core prograG will soon be transferred to the ZPR-6 facility which has been . modified to provide a 12'X12'X8' matrix.

ASSESSMENT PROGRAM -

An assessment program was initiated to define the state-of-the-art in the prediction of several key FBR parameters. The program involves a systematic reevaluation and recalculation of much of the ANL critical experiment data base using consistent methods and data. Initial reports address the sodium void reactivity, control rod worth measurements, criticality, and the reactivity scale conversion factor.

N U C L E A R DATA

Evaluation of t h e E N D F / B - V l i b r a r y has continued by ca l cu la t ions of s e l ec t ed benchmark assemblies . I t i s concluded t h a t E N D F / B - V provides moderate improvements over E N D F / B - I V f o r both thermal and f a s t r eac to r a p p l i c a t i o n s . New data f i l e s f o r c ros s sec t ion s tandards , a c t i n i d e s , and f i s s i o n products have been re leased . The major chanrjgs of 10-53: in 14 f i s s i o * product c ross sec t ions from ENDF/B-IV a r e noted t o increase t h e aggregate f a s t spectrum absorpt ion by about 75. The f i s s i o n y i e l d f i l e s have been expanded t o include 20 independent and cumulative y i e l d s e t s and t h e i r u n c e r t a i n t i e s . Calcula t ions of prompt and delayed neutrons per f i s s i o n a s well as computations of t h e t o t a l beta and gamma energy r e l eases a r e noted t o be in genera l ly good agreement with a v a i l a b l e in t eg ra l experiments. Delayed neutron spec t ra appear t o be r e l a t i v e l y independent of the inc ident neutron energy b u t i nd ica t e s t r u c t u r e t r a c e a b l e t o t b j - var ious neutron precursors , e . g . , Rb-94, Br-88, BR-90, e t c . S tudies of r e a c t o r parameter s e n s i t i v i t i e s t o t h e d e t a i l s of the neutron prompt and delayed f i s s i o n spec t ra a r e in progress.

Measurements a t O R E L A have inves t iga ted t h e low energy i n e l a s t i c s c a t t e r i n g l e v e l s of U-238 and the t ransmission and capture c ross sec t ions f o r Th-232 below 4 KeV, f i s s i o n cross sec t ions f o r Pa-231 and Th-232 u p t o 10 MeV, V f o r U-233 u p t o several MeV and f i s s i o n neutron enercy s p e c t r a . FNG measurements have included t o t a l and s c a t t e r i n g cross sec t ions f o r Ni, Cr, and Cu from 1 . 5 t o 4 MeV, t o t a l c ross sec t ions f o r Y , Zr, Mo, C d , Sn, Te, Ag, Nb, R h , Pd, In and Sb from 50 KeV t o 4 MeV, and capture c ross sec t ions f o r Rh, Pd, Nd, and Sm.

The e f f e c t s of resonance se l f - sh ie ld ing on microscopic measurements with s t r u c t u r a l a s well as a c t i n i d e nuclei a r e being s tud ied . I t appears t h a t many of t h e d i f f i c u l t i e s in model i n t e r p r e t a t i o n of neutron processes f o r t hese ma te r i a l s can be t raced t o inappropr ia te o r nonexistent t reatment of resonance s e l f - s h i e l d i n g e f f e c t s . Activat ion cross sec t ion measurements f o r U-238 from thermal t o 3 MeV r e l a t i v e t o U-235 ( n , f ) and A u ( n , y ) cross sec t ions tend t o confirm E N D F / B - V values within - + 4:.

SHIELDING

Shie ld ing s tud ie s have continued a t O R N L f o r t h e design of GCFRs and t h e CDS. The TSF was used t o v a l i d a t e the methods used f o r predic t ing neutron streaming i n G C F R g r id -p la t e and gas plenum regions and r ad ia l streaming through the blanket and sh ie ld . Angular d i s t r i b u t i o n s of neutrons leaving t h e GCFR gr id p l a t e were poorly predicted with an S10 quadrature d i sc re t e -o rd ina te s cal cu la- t i o n . Conservative r e s u l t s were obta ined , however, with a highly biased 150 d i r e c t i o n quadrature s e t .

Computations of t h e sh ie ld ing requirements f o r CDS have inves t iga ted t h e l e v e l s of Na a c t i v a t i o n in decay heat exchangers, neutron streaming l e v e l s through pipe chaseways, as well a s the weight and cos t saving incent ives f o r using a l t e r n a t e sh i e ld ing ma te r i a l s .

CORE DESIGN

Core design a c t i v i t i e s du r ing the past year have cont inued on t h e Conceptual Design Study (CDS). The r e s u l t s o f t h e CDS w i l l p rov ide a l a r g e (1000 NWe gross) breeder r e a c t o r design, support ing cost, and schedule i n fo rma t ion f o r sub- miss ion t o Congress by March 1981.

The physics support f o r CDS i s planned t o i nc lude c r i t i c a l experiments, s h i e l d i n g analyses and experiments, as we l l as a v a r i e t y o f core design, nuc lear instrumen- t a t i o n and safety-physics assessments.

The CDS core design t r a d e - o f f s tud ies performed t o da te have l e d t o t h e s e l e c t i o n o f a re fe rence c o n f i g u r a t i o n w i t h the f o l l o w i n g c h a r a c t e r i s t i c s .

Mixed U, Pu ox ide Heterogeneous Standard mechanical f ea tu res

Also, t h e t r a d e - o f f s tud ies demonstrated t h e d e s i r a b i l i t y o f core interchange- a b i l i t y between ox ide and carb ide f u e l and t h i s i s c u r r e n t l y a CDS design requirement.

Recent work h a s in-volved- more de ta , i led desfgn o f the reference c o n f i g u r a t i o n p lus .a smal.1- e f f o r t on t+he carbide. i n t e r c h a n g a b l e c'org. ' ,

METHODS AND CODE DEVELOPMENTS

At ANL, development o f c a l c u l a t i o n a l methods and codes have ma in l y focused on data processing, s t a t i c and t r a n s i e n t neutronics, burnup methods, and per turba- t i o n methods. The M C ~ - ~ / S D X and VIM codes were updated and modif ied t o enable processing o f the ENDFIB-V data f i l e s , and a new mu1 t i -d rawer c e l l method f o r d e t a i l e d c a l c u l a t i o n s o f ZPR r e a c t i o n r a t e s a t co re lb lanke t i n t e r f a c e s was added t o SDX. A d e t a i l e d i n v e s t i g a t i o n o f t h e resonance and slowing-down a lgor i thms i n use l e d t o an improved a lgo r i t hm f o r c a l c u l a t i o n o f t h e u l t r a - f i n e roup spectrum. Another delayed neutron data processing module was added t o MCj-2 and the RABANL o p t i o n i n t h i s code was rev ised t o improve accuracy o f t h e f l u x c a l c u l a t i o n .

The DIF3D code i s being f u r t h e r developed i n several ways corresponding t o t h e var ious uses o f a basic g loba l ana lys i s code. Add i t i ona l e d i t s and a c r i t i c a l i t y 5,earch have been programed. Two fo rmula t ions o f t h e t ranspor t equat ions i n an i so t rop i c media have been s tud ied and coded i n a spec ia l vers ion, OIF3D/transport. A pe r tu rba t i on method f o r c a l c u l a t i n g eigenvalue d i f f e rences between, e.g., d i f f u s i o n theory o r low-order Sn and h igh order Sn has been developed. To support t h e TREAT upgrade p ro jec t , a l i n e - o f - s i g h t neutron streaming c a p a b i l i t y has been added t o another vers ion o f DIF3D. This provides w i t h i n the d i f f u s i o n theory fo rmu la t i on an accurate t reatment of t h e hodoscope s l o t vo id. DIF3D has been adapted and pu t i n t o use a t several user s i t e s .

Methods were developed and implemented t o extend t h e REBUS-2 d e p l e t i o n code f o r burnup-dependent cross sec t ions and three-dimensional neutronics. Algor i thms which represent cross sec t ions as polynomial f unc t i ons o f burnup were developed and coded i n t o t h e d e p l e t i o n and homogenization modules. Special vers ions of REBUS were constructed which use SYN3D and DIF3D, respec t i ve l y , f o r t h e f l u x ca l cu la t i ons . SYN3D i s a f l u x synthesis code, and coup l ing i t w i t h REBUS

provides a powerful tool for fast-running, three-dimensional burnup. Coding continues on REBUS-3; several modules have been converted and tested.

The VARI3D sensitivity code has been in general use for some time for one-, two- and three-dimensional reactivity perturbation calculations. It is now beginning to be used for generalized perturbation calculations. Coding has been added to provide the options of linearized and exact forms of the leakage operator.

Studies of nodal formulations have begun at ANL. Promising early results in XYZ coordinates have been obtained, and development, including attention to HEX Z, continues. . Los Alamos is continuing development of an efficient two-dimensional transport code for core analysis. In 1980, supporting input and edit routines suitable for advanced codes were coded around the ONEDANT (supplanting "ONETWN-DA") code, and trials of an interim version of TWODANT (supplanting IWOTRAN-DA) will begin soon. A fully advanced version with search and triangular geometry capabilities is a goal of the following effort. Substantial method efforts to improve mu1 ti-dimensional space differencing schemes parallels this coding. Maintenance and updating of the LASL data processing methods in the NJDY, MAX, and CINDER codes continue.

ORNL work includes development and trial application of methods for data sensitivity analysis, methods for analysis of cores and shields, and data processing for shield analyses. Joint sensitivity analyses of the CFRMF with INEL (Ryskamp, Anderl, et al, ANS Transactions, Vol. 35, 1980) and of a standard neutron field with NBS (ANS Transactions, Vol. 33, p828, Nov 1979) have been reported. A book "Sensitivity and Uncertainty Analysis of Reactor Performance Parameters" by C. R. Weisbin, et a1 . provides a comprehensive summary of the utilization together of differential and integral data and will be published shortly by Plenum Press.

Development of three-dimensional, nonuniform triangular mesh and mesh edge geometry neutronics continues. Testing of mesh-edge capability has begun and smaller mesh errors than for the mesh centered approach were indicated for a heterogeneous core problem. Maintenance and upgrading of data processing continue. In one instance, the algorithms to shield the elastic scattering

a matrices in MINX have been tested. A comparison with independent calculations for 0-16, Na-23, and U-235 showed excellent agreement.

FFTF PHYSICS -

The Fast Flux Test Facility Reactor reached first divergence February 9, 1980. Trisector segments of the core were loaded in sequence and criticality was reached partway through the loading of the third trisector. Extensive analyses of the related FTR engineering mockup cri ticals (the "reverse approach to critical experiment") by HEDL, ORNL, and ANL supported the trisector approach to critical, and consequently the interpretation of counts logged by in-core detectors was straightforward. A sumnary of this work, "FFTF Initial Fuel Loading, Pre-Analysis, and Comparison with Preliminary Results," was presented by R. Rothrock, et al, at the ANS Physics Topical Meeting held September 14-17, 1980.

Compton r e c o i l gamma spectra, and proton r e c o i l and passive emulsion neutron spectra were measured near t h e center l i n e o f the nea r l y u n i r r a d i a t e d FFTF core. The data are expected t o p rov ide benchmark spec t ra l i n fo rma t ion f o r es t ima t ing cond i t i ons i n FFTF du r ing f u t u r e i r r a d i a t i o n tes ts .

Fur ther s t a r t u p physics t e s t s w i l l begin soon on complet ion o f t h e balance-of- p l a n t opera t ions now being conducted. Low and h i g h power measuremeats w i t h passive f o i l s and SSTRS are among t h e t e s t s t o be conducted t o ex t f ~d t h e c h a r a c t e r i z a t i o n o f the FFTF f i e l d s . -

KEY VARIABLES

1. FISSILE ISOTOPE

2. FERTIFISS. RLlTlO

3. SPECTRUM

Reactor Physics Activities at the JRC-Ispra

Compiled for the 23rd NEACRP by H.RIEF

I. RADIATION SHIELDING STUDIES

1.1 Nuclear Data for Shielding

Data Retrieval and Processing ............................ a In collaboration with RSIC an Energy Integral Library of

the Dosimetry Files of ENDF/B5 is under preparation by the use of RESEND and INTEND. The energy structure and the weighting function are that of the VITAMIN-E library. At present the energy integral library contains the following elements : He-4, B-10, Na-23, A1-27, Sc-45, ~i-46, Ti-47, Ti-48, Mn-55, Fe-54, ~e-56, Fe-58, Ni-58, Ni-60, Co-59, Cu-63, Cu-65, In-115, 1-127, Au-197, Th-232, Np-237, U-235, U-238, Pu-239.

The code CHAD, which transforms the differential data in- to Legendre expansion, has been converted for the IBM-360. The use of this code in conjunction with CRECT, which changes data in ENDF/B card image, allows to treat with AMPX the elements with scattering data in the form of tabulated functions of the cosine argument. CHAD has also been modified to include graphic subroutines to plot the angular distributions for elements with scattering data given either in the form of tabulated functions of the cosine or as tables of Legendre coefficients. Also the AMPX VASELINE module has been implemented to plot either the ENDF/B array or the weighted neutron and gamma cross-sections of the n-gamma coupled EURLIB-4 library. An approximate way of considering the bremsstrahlung effects in the gamma-ray interaction with matter is under preparation. . Data Retrieval and Testing ......................... Experience has shown that multigroup cross section libraries compiled from basic differential data have to be carefully tested on processing errors and physical consistency. For the retrieval and checking of coupled neutron-gamma libraries the computer code CAT (Choice and Test) was written

and debugged. CAT allows to handle large amounts of data, especially to pick up subassemblies of the data which then can be transferred to other output units. Moreover it can decompose coupled neutron-gamma data into the three submatrices n-n, n-gamma and gamma-gamma. Finally it permits a series of tests: finding and printing of extreme values, counting of non-zero energetic up- scatter elements, (integral) checks of particle and energy con- servation, and (discontinuous) plotting. In part CAT uses the same subroutines for reading and writing in fixed and free field FIDO convention as the code FRIX did. (FRIX ;allows for instance to convert one FIDO format into another).

Nuclear Data Evaluation .......................

To investigate the radiation damage on Pusion reactor materials an "intense neutron source" is planned. For a theoretical simulation of the (future) experiments, neuzron cross sections are needed up to an energy of 40 MeV for Fe, Co, Ni and Mn. These data are not available from 20 to 40 MeV and have to be evaluated by model calculations. The GNASH-Code, developed at Los Alamos allows the calculation of these data. In collaboration with LASL an input data set for final production runs with GNASH was established. The main work consisted inrthe generation of optimum neutron-optical model parameter sets for Co and Ni. The results of the cross-section calculation were presented to the BNL-conference on neutron cross sections (Neutron Cross sections for Co (and Ni) up to 40 MeV by Ph.Young, E.Arthur and W.Matthes, Brookhaven, 13-15 May, 1980).

Shielding Benchmark Measurements at EURACOS

At the EURACOS facility where in front of a 235U converter plate the neutron penetration in a 1.5 x 1.5 x 1.3 m iron block is determined (see previous PPR's) a long series of tests with foil activation and proton recoil counters had been concluded. They were followed by first measurements with NE213 counters, which have carefully been calibrated at the Van der Graff of the JRC Geel.

Measurements of the epithermal and fast spectra in the EURACOS iron benchmark have been continued. After resolving difficulties with an imperfect insulation of the high voltage condensers used with the proportional counters and spectra shifts in the liquid scintillation counter caused by temperature variations the final measurements at 22 cm, 49 cm and 56 cm penetration depth were concluded. (The results of the position 22 cm are shown in Fig.1). At 70 cm and 86 cm the measurements are now in course.

1.2 Code Assessment

The testing of the ADJUST code has been continued.Preliminary calculations showed that in addition to the cross-sections and instruments-normalization factors the source-strength of the individual benchmark-experiments must be adjusted too. This made it necessary to restructure some parts of the Code. A further coding had to be done to create cross-section libraries in group independent form for each individual benchmark experiment from one master-library containing the cross- a sections for all elements appearing in the set of experiments.

1.3 Code Assessment

Charged Particle Transport in Matter ---- ............................... The neutron production in the "intense neutron source" is achieved by stopping a high energy Deuterium beam in Li and stripping the neutrons off from the Deuterium. To verify the technical realization of this concept a very detailed calculation of the energy-deposition (and neutron production) of the Deuterium beam in the Li-target is necessary. For this purpose a Monte Carlo procedure was developed to simulate the transport of charged particles in matter with special emphasis on the low energy range where small angle scattering tends to broaden the D-beam. The procedure given differs from published methals in that it weakens the influence of the many approxima- tions under which the sampling distributions used in the Monte

d Carlo tracking of histories were derived.

1.4 Technical Support to Shield Design

Gamma Shielding -------------- a) Gamma Ray Buildup Factors in Lead-Iron and Iron-Lead Shields

The photon transport was calculated for 1-10 cm lead layers followed by 60 cm iron, and for 2-40 cm iron layers followed by 30 cm ,lead. Four source energies were treated, they are: 6; 2; 1.25 (~060) and 0.66 MeV(Cs). The source was assumed to be plane isotropic. The effects of bremsstrahlung and of different conversion factors are discussed (the single collision concept which leads to the exposure dose, or the multiple collisions in a phantom

leading to an equivalent dose). As in former works, the discontinuity of the energy deposition buildup factor at the interface of the two materials is detected and discussed. Moreover, for E = 6 MeV in the sequence iron-lead, the buildup factors in yead resulted higher, than both B-factors in pure iron and in pure lead at the same total penetration depth (the first few cm ex- cepted). A paper on this subject will be published.

List of Publications

- E. ARTHUR, W. MATTHES and Ph. YOUNG, "C:alculation of 59Co Neutron Cross Sections between 3 and 50 MeV", Symp. on Neutron Cross Sections from 10 to 50 MeV, May 12-14,80, Brookhaven National Laboratory - N.Y. - USA

- H.RIEF, A. DUBI, V. SUNDARARAMAN, Experience with Correlated Tracking in Deep Penetration Monte Carlo Sensitivity Analysis, "1980 Advances in Reactor Physics & Shielding", ANS-Conf. Sept. 15-17, 1980, Sun Valley - USA

- P. ROCCO, C. PONTI and A.CARETTA, "Activation and Decay Heat of an Aluminium Blanket for Experimental Fusion Reactors", 11th SOFT-Meeting, Sept. 15-14,80, Oxford - UK

- C.PONT1, R. VAN HEUSDEN, "Tritium Breeding Problems in FINTOR-D", 8th Symposium on Engineering Problems of Fusion Research, Nov. 13-16, 79, Lawrence Livermore Laboratory, San Francisco - UK

- H. PENKUHN, "Gamma Ray Build-up Factors in Lead-iron and Iron- lead Shields", accepted for publication in Atomkernenergie/ Kerntechnik

- V. SUNDARARAMAN, "Effect of Bremsstrahlung on Scattered Gamma Ray Spectra", Nucl.Instruments & Methods, 171 (1980) pp.561-565

- ESIS Newsletter No. 31 - Oct. 1979 ESIS Newsletter No. 32 - Jan. 1980 ESIS Newsletter No. 33 - Apr. 1980 ESIS Newsletter Special Issue No. 6 - March 80 (Specialists1 Meeting on Radiation Damage in Reactor Components- Hannover, March 27, 79)

11. ASSESSMENT STUDIES ON NUCLEAR TRANSMUTATION OF ACTINIDES

11.1 Reactor Physics Studies on Transmutation of By-product

Actinides (BPAs)

Introduction - - - - - - - - - - - - A

The scope of this activity is to improve the calculational results of the build-up of BPAs like Np, Am and Cm isotopes during the operation of various types of nuclear power reactors and to identify an optimum recycle scheme for BPAs

0 taking into account transmutation rates, reactivity effects, power perturbation, fuel element behaviour and cost as well as technological implications on the various fuel cycle ope- rations. The four reactor types investigated were the U-fed LWR, the Pu-fed LWR, the FBR, and the pebble-bed HTR. For each of them, new cross section libraries were established, an appropriate computation technique consisting of a chain of computer programs was set up, and a series of calculations for the evolution of the BPA isotopes, the spatial flux distribution in the case of heterogeneous recycling, and the reactivity were performed. The results were published in C11, C21, 131, C41.

11.1.1 Calculations for the LWR

Nuclear Data ------------ For zero-dimensional burn-up calculations, the library of RIBOT-5A was used. The results were checked by means of

9 experimental data of the bench Bark Experiment activity of the JRC. The comparison showed that the values calculated by RIBOT-5A were in good agreement with the majority of measured BPA concentrations. A three-group cross section set has been prepared for the two-dimension burn-up code EREBUS using the multi-group, one- dimension cell burn-up code LASER. Its library is based on data of the computer programs MUFT and THERMOS. - Computer Program Used .....................

Zero-dimensional burn-up calculations were performed by means of RIBOT-5A. The one-dimensional transport burn-up code LASER was utilized for the evaluation of homogenized parameters and cross sections needed in EREBUS. It takes into account the spatial burn-up distribution in the fuel rod and the non- linearity of the flux during the irradiation. The impacts of the recycling scheme on the spatial power distribution were

investigated by means of the two-dimensional diffusion burnup code EREBUS.

Results - - - - - - - Compared to ORIGEN results, discrepancies in the range between 10 and 50 per cent have been found for Np-237, AM-241, Am-242 and Cm-245. Discrepancies between 50 and 100 per cent appeared in the isotopic densities of Am-243, Cm-242 and Cm- 246. The Cm-243 isotopic concentrations differed by a factor of 50. The results of bi-dimensional calculations indicate that the introduction of 2% BPA in a plutonium-containing fuel pin results in local power peakings and average flux depressions in a fuel element of about 25%. This is considered as a limit with respect to the present fuel element design. Replacing 2% of equilibrium-recycled plutonium by 2% of BPA does hardly lead to a reactivity loss.

11.1.2 Calculations for the FBR

Nuclear Data ------------ For fuel isotopes, coolant, and structure materials, the data of the ENDF/B-4 file were used. Data for the BPA were collected from ENDF/B-5 and, partly, from the ENDL files.

Computer Programs Used ...................... The ENDF/B data were converted by mea:ns of the code ETOE to be used in the code MC~-2. The last one provides multi-group cross sections as well as spectra. A calculation utilizing W 2 - 2 was performed for each reactor zone and the resulting cross sections were collapsed to six-group cross sections by means of MCDATA. The space-dependent burn-up calculations were performed applying CITATION in (r, z )-lgeometry .

Results ------- With respect to ORIGEN, the generated library demonstrates a larger ratio of capture to fission cross sections for many BPA isotopes. The greatest discrepancies between isotopic concentrations of this calculation and the ORIGEN results were found for Cm-245 and higher Cm-isotopes. On the other hand the concentrations for Np-237, Am-241, Am-243 and Cm-243 agreed within 20%. The decrease in the breeding ratio due to recycling of BPA was found to be negligible ( 0.05%). Homogeneous re- clycle does not lead to any reactivity loss. At the end of cycle, a reactivity gain was obtained.

11.1.3 C a l c u l a t i o n f o r t h e HTR

. A s a b a s i s f o r a l l problem dependent group c o n s t a n t s , two problem-independent l i b r a r i e s a r e used: THERM-126 and G G C . A l l t h e s e d a t a a r e based on ENDF/B-4.

L

Computer --- -------- Programs ------ Applied --- Uni t - ce l l c a l c u l a t i o n s were performed by ISOSTO. The d e t e r - minat ion of t h e c r i t i c a l i t y , t h e f lux,and power d i s t r i b u t i o n

0 was done by means o f t h e system RSYST. It c o n t a i n s t h e two- dimensional d i f f u s i o n code DIFF-2D which p rov ides t h r e e - group f l u x e s . These f l u x e s have been used f o r burn-up ca l cu - l a t i o n s wi th O R I G E N .

R e s u l t s ------- Due t o t h e long in-core t ime and h igh f l u x l e v e l s , t h e HTR appears t o be a n e x c e l l e n t t r ansmuta t i on dev i ce . It h a s been shown t h a t homogeneous as w e l l as heterogeneous r e c y c l e s tra- t e g i e s a r e p o s s i b l e . Applying t h e U-233-Th c y c l e , t h e p roduc t ion of BPAs remains very sma l l compared t o o t h e r r e a c t o r t y p e s .

11.2 Eva lua t ion of t h e Impacts on t h e Nuclear Fuel Cycle

11 .2 .1 Cost Eva lua t ion

The i n c r e a s e o f t h e c o s t s i n t h e f u e l c y c l e r e c y c l i n g BPAs has been c a l c u l a t e d by means o f :

- c o l l e c t i o n and a n a l y s i s o f l i t e r a t u r e d a t a f o r u n i t c o s t s ( m a t e r i a l s and b a s i c o p e r a t i o n s )

- t r a n s f o r m a t i o n by a computer program o f t h e u n i t c o s t s i n i n c r e a s e o f c o s t s f o r t h e f u e l c y c l e .

Cons ider ing v a r i o u s t r a n s m u t a t i o n s c e n a r i o u s , based on t h e u se . o f LWR, FBR and HTR r e a c t o r s , i n c r e a s e s o f c o s t s of t h e f u e l c y c l e between 10 and 23% have been c a l c u l a t e d .

11.2.2 Risk Source Eva lua t ion

The r i s k source e v a l u a t i o n s have been c a r r i e d ou t mainly u t i - l i z i n g t h e MARYON 3 code developed a t t h e JRC-Ispra. The code c a l c u l a t e s :

- The r e l e a s e s from n u c l e a r p l a n t s

- accumulat ions i n waste r e p o s i t o r i e s f o r f u e l a c t i n i d e s , BPAs and long- l ived FPs. The i n v e n t o r i e s a r e expressed

i n v a r i o u s u n i t s (g r ams , C u r i e s , h a z a r d u n i t s ) n o r m a l i z e d on THM and GWey.

The e v a l u a t i o n s c o n c e r n t h e i n c r e a s e o f s h o r t - t e r m r i s k s owing t o t h e p r e s e n c e o f t h e r e c y c l e d BPAs i n t h e f u e l c y c l e a n d t h e d e c r e a s e o f t h e l ong - t e rm r i s k owing t o t h e r e d u c t i o n o f t h e BPA a c c u m u l a t i o n i n t h e w a s t e r e p o s i t o r i e s . The r i s k was app ro - x ima ted by t h e u s e o f ICRP t a b u l a t e d :hazard i n d i c e s ..

m 3 Chl =

C i g r D o s i s

S h o r t - t e r m r i s k --------------- The h a z a r d d u e t o t h e r e l e a s e s f rom n u c l e a r p l a n t s i s domina t ed by t h e f i s s i o n p r o d u c t s s o t h a t t h e BPA r e c y c l i n g h a s a n e g l i - g i b l e i n f l u e n c e . Conce rn ing r i s k o f a c c i d e n t s , a q u a n t i t a t i v e a n a l y s i s c a n n o t b e done : however, i n most o f t h e p l a n t s a n i n c r e a s e o f a c c i d e n t s c a n be e x p e c t e d d u e t o t h e c h a r a c t e r i s t i c s o f t h e BPAs ( s m a l l c r i t i c a l masses , h e a t r e l e a s e , n e u t r o n e m i s s i o n ) . A l so t h e c o n s e q u e n c e s o f t h e a c c i d e n t s w i l l be w o r s e t h a n i n t h e no rma l f u e l c y c l e d u e t o t h e h i g h t o x i c i t y o f t h e BPAs.

Long-term r i s k --------------

A s measure o f t h e l ong - t e rm r i s k r e d u c t i o n , t h e i n g e s t i o n h a z a r d r e d u c t i o n f a c t o r h a s been t a k e n which i s t h e r a t i o o f t h e i n g e s t i o n h a z a r d s o f U,Pu BPAs and FPs a c c u m u l a t e d i n t h e w a s t e w i t h o u t and w i t h r e c y c l e .

Us ing h a z a r d i n d e x e s , i t i s assumed t h a t a l l t h e n u c l i d e s s t o r e d i n t h e w a s t e r e p o s i t o r y have t h e same p r o b a b i l i t y t o r e a c h t h e man. T h i s a s s u m p t i o n i s n o t c o m p l e t e l y a c c e p t - a b l e . It would be more c o r r e c t t o i n t r o d u c e t h e w a s t e i n v e n - t o r i e s w i t h and w i t h o u t BPA r e c y c l i n g i n r i s k models o f g e o l o g i c a l d i s p o s a l which t a k e i n t o a c c o u n t t h e d i f f e r e n t pa thways o f t h e v a r i o u s r a d i o n u c l i d e s . However, due t o t h e f a c t t h a t a t p r e s e n t t h e s e models s u f f e r f rom i n a c c u r a t e i n p u t d a t a , t h e u s e o f h a z a r d s i n d i c e s ap- p e a r s t o be t h e most c o n v e n i e n t a p p r o a c h .

A w a s t e management s t r a t e g y b a s e d on BPA s e p a r a t i o n f rom HLW and t h e i r r e c y c l i n g i n n u c l e a r r e a c t o r s c a n i n t r o d u c e a sub- s t a n t i a l r e d u c t i o n i n t h e l o n g - t e r m r i s k o n l y i f t h e p l u t o n i u m l o s s e s t o t h e w a s t e a r e s m a l l . T h i s i s n o t t h e c a s e f o r t h e p r e s e n t f u e l c y c l e ( s e e T a b l e I) where p l u t o n i u m l o s s e s o f 0 . 5 % i n HLW and 1 . 5 % i n MLW c a n be assumed. I n t h i s s i t u a t i o n t h e l ong - t e rm r i s k o f t h e w a s t e i s m a i n l y c a u s e d by t h e p l u t o n i u m l o s s e s i n MLW s o t h a t t h e i n -

t r o d u c t i o n o f a s t r a t e g y based on chemical s e p a r a t i o n ' o f t h e a c t i n i d e s ( f u e l a c t i n i d e s and BPAs) and t r ansmuta t i on h a s no s ense .

Thus i t was assumed i n t h e J R C e v a l u a t i o n s t h a t t h e r e c y c l i n g . s t r a t e g y i s in t roduced i n a n advanced f u e l cyc l e where t h e plutonium l o s s e s t o t h e MLW a r e reduced by a f a c t o r 100 w i t h r e s p e c t t o t h e p r e s e n t f u e l c y c l e ( s e e t a b l e I 1 ) . The i n g e s t i o n hazard

I. r e d u c t i o n f a c t o r s , which can be ob t a ined by t h e r e c y c l e s t r a - t egy have been determined t a k i n g t h i s advanced f u e l c y c l e as r e f e r e n c e .

Recycl ing BPAs two c a s e s have been cons idered : a R e a l i s t i c

e EUR Recycle Case ( R R C ) and a n Op t imi s t i c EUR Recycle Case ( O R C ) which d i f f e r f o r t h e v a l u e s o f t h e l o s s e s t o t h e waste ( s e e Table I ) .

C a l c u l a t i o n s have been c a r r i e d o u t f o r v a r i o u s n u c l e a r r e a c t o r u t i l i z a t i o n s t r a t e g i e s u s i n g d a t a from r e a c t o r phys i c s ca lcu- l a t i o n s a s i n p u t t o t h e MARYON-3 code. Recycl ing o f s e l f -gene ra t ed BPAs i n FBRs has been cons ide red a s w e l l as complex r e c y c l i n g s t r a t e g i e s u s i n g LWRs, FBRs and HTRs on t h e b a s i s of s c e n a r i o s of i n s t a l l e d n u c l e a r g e n e r a t i n g c a p a c i t y i n t h e European Community.

Table I1 p r e s e n t s t h e i n g e s t i o n hazard r e d u c t i o n f a c t o r s determined f o r t h e r e c y c l i n g o f s e l f -gene ra t ed BPAs i n FBRs f o r t h e 2 r e c y c l e c a s e s . The hazard r e d u c t i o n appea r s t o be l i m i t e d mainly f o r l ong decay t imes . This i s due t o l o s s e s accumulation a s w e l l a s t o t h e i n c r e a s e wi th t ime o f t h e r e l a t i v e c o n t r i b u t i o n t o t h e waste hazard o f t h e long- l ived f i s s i o n p r d u c t s 1-129 and

7

A r e c y c l e s t r a t e g y based on a combined use o f d i f f e r e n t t y p e s of r e a c t o r s may permit a l i m i t e d improvement. A s u b s t a n t i a l improvement cou ld on ly be o b t a i n e d by r e c y c l i n g a l s o t h e long- l ived f i s s i o n produc ts : t h i s would i n t r o d u c e f u r t h e r major compl ica t ions i n a n a l r e a d y complex problem. -

111.3 Conclusions o f t h e J R C Assessment

On t h e b a s i s o f va r ious s t u d i e s c a r r i e d o u t by t h e J R C , t h e fo l lowing conc lus ions have been drawn:

- Chemical s e p a r a t i o n and n u c l e a r t r ansmuta t i on o f BPAs a r e f e a s i b l e i n t h e sense t h a t a s o l u t i o n t o overcome techno- l o g i c a l problems can be found w i t h a l a r g e R and D e f f o r t .

- The homogeneous r e c y c l i n g i n FBRs appears t o be t h e b e s t cho i ce both f o r t h e t r a n s m u t a t i o n r a t e s and f o r compat ibi -

lity with the FBR fuel facilities anticipating the availability of an advanced fuel cycle technology for reprocessing and fabrication.

- The incentives for chemical separation and nuclear transmutation appear to be small, irrespective of the transmutation option considered. The reduction in the long-term risk is limited by the presence of long-lived fission products. .t

- A reduction of the long-term risk appears to be more readily achievable by improving the techniques of alpha-waste reduction, conditioning and disposal.

References

S. GUARDINI, B.G.R. SMITH, "Calculations for the assessment of actinide transmutation in light water reactors (LWR)", Proceedings of the CEC/NEA Second Technical Meeting on the Nuclear Transmutation of Actinides, JRC-Ispra 21-24 April 1980 (in print)

B.G.R. SMITH, S.GUARDIN1, G.OLIVA, L.TONDINELL1, "Comparison between measured and calculated by-product actinide build-up within fuel assemblies in a large burn-up range", ibidem

G.OLIVA, "Neutron transmutation of transuranium isotopes in a fast breeder power reactor", ibidem

H.SCHAAL, D.FILGES, R.D. NEEF, I.VIDO'VSKY, "Production and transmutation of actinides in high temperature gas cooled pebble bed reactors (HTGR)", ibidem

E.ZAMORAN1, J.CAMETT1, E.SCHMIDT, F.C.ALIGARA, A.MOREN0, "Design of fuel pins containing by-product actinides for irradiation in LWR and FBR power plants", ibidem 0 E.ZAMORANI, J.CAMETT1, C.PONT1, E.SCHIVIIDT, "Impact of by- product actinides on fuel behaviour and fabrication chain", ibidem

A. FACCHINI, S.GALLONE, E.ZAMORAN1, "Considerations on a sol- gel type process as a fuel fabrication route for the acti- *

nide recycle", ibidem

W.HAGE, L.ANSELM1, K.CARUS0, "A comparative analysis of the nuclear fuel cycle with recycling of by-product actinides" ibidem

E.SCHMIDT, "Possible recycle strategies for by-product actinides in a reference scenario of installed nuclear generating capacity in the EC and their valuation", ibidem

.W.HAGE(ed), Proceedings: Second Technical Meeting on the Nuclear Transmutation of Actinides,JRC-Ispra, April 21-24,1980(in print)

Table I - Fuel and BPA Losses Assumed i n t h e Eva lua t ion of , I n g e s t i o n Hazard Reduction

HLW MLW HLW MLW HLW

Presen t Fue l Cycle 5 x 1 . 5 x 1 0 1

Advanced Reference 10-3 1,5 10-4 0 1 Fue l Cycle (ARFC)

R e a l i s t i c EUR- Hecycle Case (RRC) 5 x 1 .5 x 10-I lo -*

Op t imi s t i c EUR- Recycle Case (ORC) 5 x 1 . 5 x

MLW -

0 -

0

Table I1 - Reduction F a c t o r s f o r Waste I n g e s t i o n Hazard Recycl ing BPAs i n FBRs

. R e a l i s t i c EUR- Recycle Case (RRC

Opt imi s t i c EUR-. Recycle Case (ORC

3 1 0 y e a r s

10.5

42

4 10 y e a r s

4.5

10

5 1 0 y e a r s

3 .1

4.8

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