National Science Center “Kharkov Institute of … Science Center “Kharkov Institute of Physics...

National Science Center

“Kharkov Institute of Physics and Technology”

National Academy of Sciences of Ukraine




Fast Reactor Related Activity

in Ukraine

Sergii Fomin

Akhiezer Institute for Theoretical Physics

NSC “Kharkov Institute of Physics and Technology”

e-mail: [email protected]

46th annual meeting of TWG-FR 21-24 May 2013 IAEA, Vienna, Austria

There are 4 NPP working in Ukraine now, which used 13 WWER-1000

and 2 WWER-440 nuclear power units with total installed capacity of

13,835 MW.

Energy Strategy of Ukraine till 2030

Nuclear Power > 50 % !!!

210,2

251,0

307,0

420,1

101,2 110,5

158,9

219,0

185,2

88,76

0

50

100

150

200

250

300

350

400

450

2005 рік 2010 рік 2015 рік 2020 рік 2030 рік

Всього на АЕС

Annual electricity production in Ukraine in 2005-2030 years, TWh

Total Nuclear

5

Installed NPP capacity (optimistic scenario)

The Ukraine State Program of Fundamental and Applied Researches

on Nuclear Materials and Irradiative Technologies

INSTITUTE OF SOLID-STATE PHYSICS, MATERIALS SCIENCE AND TECHNOLOGIES

(Director - V.N. Voyevodin)

PHYSICS OF RADIATION EFFECTS AND TECHNOLOGIES

New physical notions of the processes of interaction of fast particles andradiations with solids, mechanisms of defect formation and defect structureevolution are originated and developed. Phase transitions in metals, steels andalloys, in semi-conductive and superconductive materials are investigatedunder irradiation in reactor and accelerator conditions.

The theory and methods of rapid simulation of radiation effects occurring inmaterials of nuclear and fusion reactor cores are developed. The efficiency ofuse of high-energy electrons and gamma-quanta (30 - 250 MeV) to simulate theinfluence of reactor irradiation on the mechanical properties of materials issubstantiated and confirmed by experiments.

The methods devised for promoted (100 - 1000 times) study and prediction of thebehavior of materials in fissile cores and fusion reactors through the use ofecologically safe for the environment of high-current charged particleaccelerators and computer simulation allow one to quickly select materials forin-reactor high-dose tests and to reduce by factors of 3 to 5 the time allotted forresearch and development of new materials.


THE MULTICHARGED ION ACCELERATOR

FOR MATERIALS SCIENCE STUDIES ("ESUVI")

Based on the studies into the influence of dissolved additives and combined effects of

microalloying, high-frequency treatment and radiation-induced processes of

component redistribution and segregation of elements on structural-phase states,

properties and radiation resistance of steels and alloys, the scientists were able:

• to reveal new aspects of a physical nature of radiation-induced phenomena in

materials (swelling, embrittlement, creep, surface erosion, etc.), and thus to direct

the way to increase their radiation stability;

• to justify the possibility for formation and existence of nonsaturated sinks for point

defects in crystal and amorphous materials under irradiation, i.e., the so-called

alternating-polarity point defect recombination centres;

• to discover a new phenomenon of anomalous recombination of radiation defects in

continuously decomposing solid solutions under irradiation;

• to develop methods for calculating and constructing radiation-modified phase

diagrams of alloys at different defect production rates;

• to establish regularities in the evolution of a structural-phase state under irradiation

to high doses in view of cascade mechanisms of dissolution and growth of new

phases, and other factors.


RADIATION SWELLING OF FERRITIC-MARTENSITIC STEELS

EP- 450 AND HT-9 UNDER IRRADIATION BY METALLIC IONS

TO SUPER-HIGHER DOSES

O.V. Borodin, V.V. Bryk, V.N. Voyevodin, A.S. Kalchenko, Yu.E. Kupriyanova,

V.V. Melnichenko, I.M. Neklyudov, A.V. Permyakov

Swelling of ferritic-martensitic steels EP-450 and HT-9 is studied under

irradiation by Cr ions up to the doses 300 dpa. In temperature range

430F550 ºС parameters of porosity, incubation dose, dose range where

the swelling range reaches the steady state were determined. It was shown

that swelling of ferritic steel may exceed 20 %. The obtained results show

that radiation swelling of ferritic-martensitic steels is the critical parameter

that may considerably limit the commercial use of reactors of Gen 3 and 4.

ЦПАЗННЦ ХФТИ

23nd Meeting of the IAEA Technical Working Group on

Gas Cooled Reactors (TWG-GCR-23)

March 5 – 7, 2013 , IAEA. Vienna, Austria

9

Dr. Mykola Odeychuk

THE DEVELOPMENT OF TECHNOLOGIES OF FUEL

AND COMPONENTS ОF HTGR CORES




NSC KIPT ADVANCED TECHNOLOGIES

FOR HTGR CORE COMPONENTS

The main technologies - mechanical spheroidization and

gas-phase pyrolytic impregnation - are transformed into

new:

1. Mechanics-vacuum spheroidizing.

2.Gas-phase technology for nitride production with

controlled composition of the gas medium.

3. Gas-phase technology for composites manufacture with

controlled composition of the gas medium.

4. Technology for express simulation radiation tests with

using of charged particle accelerators


Coated particle

NSC KIPT CP design


TRISO Coated Particle

0.96 mm diameter

15000 coated particles embedded

in GSP Graphite Matrix F

F protected by a 5mm thick

GSP Graphite layer

The final “Pebble” 60mm in

diameter containing thorium

and enriched uranium

NSC KIPT Spherical Fuel

13

Microspherical uranium nitride fuel

Experimental installation for

microspherical uranium carbonitride fuel

manufacturing

Experimental installation for

microspherical uranium nitride fuel

manufacturing

Recent Publications:

M. Odeychuk, Fabrication, properties, in-pile performance and pie of

carbo-nitride and nitride fuels. 1-st Nitride Fuel Workshop, AlbaNova

University Centre Stockholm, Sweden, 01-02 February, 2012, 46 p. -

http://new.neutron.kth.se/NitrideWorkshop/Presentations.

M. Odeychuk, The advanced nitride fuel for fast reactors. IAEA Technical

Meeting on “Design, Manufacturing and Irradiation Behaviour of Fast

Reactors Fuels”, 30 May-03 June 2011, Institute of Physics and Power

Engineering (IPPE), , 2011, 17 p.

M. Odeychuk, The advanced high-temperature fuel for HTGR. AEA TM

“High Temperature Gas Cooled Reactor Fuel and Fuel Cycle“, Vienna,

Austria, September 06-09, 2010, IAEA, Vienna, Austria, 2010, pp. 6; 11; 23. ,


Electron irradiation test facility

Construction of EITF-KIPT General view of EITF-KIPT

KIPT – ANL Neutron Source Facility

Facility Concept: Experimental Neutron Source Facility based on the use of an electron

accelerator for driving a subcritical assembly with low enriched uranium fuel

Facility Objectives: • Provide capabilities for performing basic and applied research utilizing the

radial neutron beam ports of the subcritical assembly

• Produce medical isotopes and provide neutron source for performing

neutron therapy procedures.

• Support the Ukraine nuclear power industry by providing the capabilities

to perform physics experiments and to train young specialists




KIPT - ANL Neutron Source Facility

Electron beam power - 100 kW

Electron beam energy – 100 MeV

Neutrons exit from target – 3.28⋅⋅⋅⋅1014 n⋅⋅⋅⋅s-1

Target material – U

Fuel enrichment - ≤ 20 %

Flux - 1.3⋅⋅⋅⋅1013 n⋅⋅⋅⋅cm-2⋅⋅⋅⋅s-1

Allocated capacity in assemblage – 350 kW

KIPT - ANL Neutron Source Facility

Main Components:

• Electron accelerator

• Electron transport channel

• Target assembly for generating neutrons

• Subcritical assembly with low enrichment fuel, carbon reflector,

and water coolant

• Heavy concrete biological shield

• Radial neutron channels for basic and applied research

• Auxiliary equipments including the target and

the subcritical assembly coolant loops

Electron transport channel

KIPT – ANL Neutron Source Facility

23

№ Parameter Value

1 Electron beam power ~ 100 kW

2 Electron beam energy 150 MeV

3 Target neutron yeld (U/W) 3.28⋅⋅⋅⋅1014/1.91⋅⋅⋅⋅1014 n⋅⋅⋅⋅с-1

4 Target material U / W

5 Fuel 235U enrichment ≤ 20 %

6 Neutron flux near the target ∼∼∼∼ 2.4⋅⋅⋅⋅1013 cm-2⋅⋅⋅⋅ s-1

7 Neutron flux near the reflector ∼∼∼∼ 2⋅⋅⋅⋅1013 cm-2⋅⋅⋅⋅ s-1

8 Maximum fast neutron flux near a fuel, Еn > 0.1 MэВ ∼∼∼∼1.3⋅⋅⋅⋅1013 cm-2⋅⋅⋅⋅ s-1

9 Moderator Н2О

10 Graphite reflector density - 2.3 g⋅⋅⋅⋅cm-3

11 Thermal power absorption in fuel rods ∼∼∼∼ 230 kW

12 Thermal power absorption in reflector ∼∼∼∼ 20 kW

13 Total thermal power ∼∼∼∼350 kW

Main Parameters of the Installation

KIPT - ANL Subcritical assembly (ADS)

Subcritical assembly configuration

indicating four irradiation locations

where the first is close to the center

10−8

10−7

10−6

10−5

10−4

10−3

10−2

10−1

100

101

3e11

4e11

5e11

6e117e118e119e111e12

2e12

3e12

4e12

5e12

Energy [MeV]N

eutr

on F

lux

/ Let

harg

y [n

/s⋅c

m2 ]

1st Location

2nd Location

3rd Location

4th Location

Neutron spectrum in the four-irradiation

locations where the first is close to the

assembly center

Physical Basis of Innovative Fast Reactor

Working in the Nuclear Burning Wave Regime

(“Traveling Wave Reactor”)

Sergii Fomin


Akhiezer Institute for Theoretical Physics, Kharkov

also

"Nuclear Fuel Cycle" , NSC KIPT, Kharkov

Institute for Nuclear Research of NAS of Ukraine, Kyiv

Taras Shevchenko National University of Kyiv, Kyiv

Odessa Politechnical State University, Odessa

46th meeting of TWG-FR 21 May 2013 IAEA, Vienna, Austria

L.P. Feoktistov, 1988. An analysis of a concept of a physically safe reactor. Preprint IAE-4605/4.

L.P. Feoktistov, 1989. Neutron-fission wave. Sov. Phys. Doklady, 34, 1071.

238U (n,γγγγ) →→→→ 239U (ββββ) →→→→ 239Np (ββββ) →→→→ 239Pu (n,fission) ...

T1/2 ≈≈≈≈ 2.35 days

( )( )2

8 8 Pu2 Pua a f

n nD vn N N

t zσ σ σ

∂ ∂= + − +

∂ ∂

88 8

;a

Nvn N

tσ

∂= −

∂9

8 8 9

1a

Nvn N N

t β

στ

∂= −

∂

( )Pu9 PuPu

1a f

NN vn N

t β

σ στ

∂= − +

∂

x z Vt= +

8 8Pu aeq Pu Pu

f a

NN

σσ σ

=+

( 1)

ai iPu icr Pu

f

NN

σ

ν σ=

−∑

Pu Pu

eq crN N>

Edward Teller (USA, 1997): Traveling Wave Reactor (Monte Carlo simulation: Th-U fuel)

E.Teller, 1997. Nuclear Energy for the Third Millennium. Preprint UCRL-JC-129547, LLNL.

Hiroshi Sekimoto (Japan, 2001): CANDLE (Determ. appr.: 2d stationary problem: U-Pu fuel)

H. Sekimoto et al., 2001. A New Burnup Strategy CANDLE. Nuclear Science & Engineering, 139, 306.

(Analytical approach: 1d stationary problem: U-Pu fuel)

Lev Feoktistov (USSR, 1988): Neutron - fission wave

Goldin & Anistratov (USSR, 1992): (Deterministic appr.: 1d non-stationary problem: U-Pu fuel)

V. Goldin, D. Anistratov, 1992. Math. modelling of neutron-nuclear processes in safe reactor, Preprint IMM RAS # 43.

Necessary condition

!!

1928 - 2002

Our publicationsOur publicationsOur publicationsOur publications: : : :

S. Fomin et al., Annals of Nuclear Energy, 32 (2005) 1435-1456.

S. Fomin et al., Problems of Atomic Science & Technology, 6 (2005) 106-113.

S. Fomin et al., Nuclear Science & Safety in Europe. Springer (2006) 239-251.

S. Fomin et al., Problems of Atomic Science & Technology, 3 (2007) 156–163.

S. Fomin, Reactor Physics and Technics. PINP WS, St-Perersburg, XL-XLI (2007) 154-198.

S. Fomin et al., Progress in Nuclear Energy, 50 (2008) 163-169.

Yu.Mel’nik et al., Atomic Energy, 107 (2009) 288-295.

S. Fomin et al., Progress in Nuclear Energy, 52 (2011) 800-805.

Conference activityConference activityConference activityConference activity: : : :

2005 - ICENES (Brussels, Belgium) IC058; NATO-ARW NSSE (Yalta, Ukraine); IAEA-RCM ADS (Minsk, Belarus)

2006 - ICAPP’06 (Reno, USA) paper 6157; NPAE (Kiev); QEDSP’06 (Kharkov); INES-2 (Yokohama, Japan)

2007 - ICAPP’07 (Nice, France) paper 7499; WS PINP (St-Perersburg, Russia); IAEA-RCM ADS (Roma, Italy)

2008 - Channeling’08 (Erice, Italy); NATO-ARW SNE (Yalta, Ukraine) | NPQCD (Dnepropetrovsk, Ukraine)

2009 - IAEA-RCM ADS (Vienna, Austria), ANIMMA (Marseille, France); Global 2009 (Paris, France) paper 9456

2010 - IAEA-RCM ADS (Mumbai, India); PINP WS (St-Perersburg, Russia); ICAPP (San Diego, USA) paper 10302

NPAE (Kiev); IAEA-TWG-FR (Brussels, Belgium); 19 ICPRPRMS (Alushta, Ukraine); INES-3 (Tokyo, Japan)

2011 - IAEA-TWG-FR (Beijing, China); NSC KIPT SS (Alushta, Crimea); QEDSP’11 (Kharkov, Ukraine);

IAEA-TWG-FR (Chinnai, India); IAEA-TWG-FR (Vienna, Austria),

L z

r

R

L

ign

j ext

Breeding zoneIgnition zone

238 U 100 %238 U

90 %

239 Pu

10 %

2D Non-Stationary Theory of Nuclear Burning Wave

S. Fomin, Yu. Mel’nik, V. Pilipenko, N. Shul’ga, A. Fomin (1st IC “Global 2009”, Paris, paper 9456)

Nuclear Burning Wave

( )/ / / / / /

/ / /

1 1

1

1 1 1

1 1

( ) ( )

g g gg g g g g g g g g g

a in mod in modg

gG Gg g g j j g g j j j g g g

f f f d l f f l d l l inj l j lg g g

rD Dv t r r r z z

Cχ ν χ β ν χ λ

→ − −

−→

= = =

∂Φ ∂ ∂Φ ∂ ∂Φ− − + Σ +Σ +Σ −Σ Φ − Σ Φ =

∂ ∂ ∂ ∂ ∂

= Σ Φ − Σ Φ + + Σ Φ∑ ∑ ∑ ∑ ∑ ∑ ∑

( 1) ( 1) ( 1)

g gg glal l l c l l l

g g

NN N

tσ σ − − −

∂= − Φ +Λ + Φ +Λ

∂ ∑ ∑

Non-Stationary Nonlinear Multi-Group Diffusion Equation of Neutron Transport

Together with Fuel Burn-up Equations and Equations of Nuclear Kinetics

of Precursor Nuclei of Delayed Neutrons

96 6

NN

t

∂= Λ

∂

10

1,4,5,6,7

gg

fl l

l g

NN

tσ

=

∂= Φ

∂ ∑ ∑

( )j

j j j g g gll l l f f l

g

CC

tλ β ν

∂= − + Σ Φ

∂ ∑

Metal fuel (44%)

Pb-Bi coolant (36%)

CM - Fe (20%)

jext ~ 1015 cm-2 s-1

toff = 400 days

( 1 8);, l = ÷

Reactor variant: R=117 cm, L= 500 cm (Lig= 71.17 сm), toff=950 days

,1017 cм-2 с-1

,1021 cм-3

232Th 233Th

233Pa

233U 234U 235U 236U 237U

T½ = 22.2 min

T½ = 27 days

237Np

T½ = 6.75 days

238U 239U

239Np

239Pu 240Pu 241Pu 242Pu 243Pu

T½ = 23.5 min

T½ = 2,35 days

243Am

T½ = 4.98 hours

241Am

T½ = 14,3 years

Th-U fuel cycle

U-Pu fuel cycle

2009: NBW reactor with mixed Th-U-Pu fuel cycle

Example: Metallic fuel 232Th (62%) + 238U (48%) volume fraction = 55%,

fuel porosity p = 0.35; Coolant (Pb-Bi eutectic) vol. frac. = 30%,

Constr. materials (Fe) vol. frac. = 15%; R = 390 cm

L z

r

R

L зап

Jext

Breeding zoneIgnition zone

238 U 50 %238 U

90 %

239 Pu

10 %232Th 50 %

Fuel burn-up for Th-U-Pu cycle

NBW reactor with mixed Th-U-Pu fuel cycle

Example: Metallic fuel 232Th (62%) + 238U (48%) volume fraction = 55%, fuel porosity p = 0.35;

Coolant (Pb-Bi eutectic) vol. frac. = 30%, Constr. materials (Fe) vol. frac. = 15%; R = 390 cm

0 100 200 300 400 500

1

2

3

4

5

6

t7

t6

t5

t4

t3

t2,×0.02

t1

z

Φ

a)

0 100 200 300 400 500

0.2

0.4

0.6

0.8

1.0

1.2

t7

t6

t5t

4t3

t2

t1

z

N

b)0 100 200 300 400 500

10

20

30

40

50

t7

t6

t5

t4

t3

t2

z

B

c)

FIG. 3. the axial distributions (z, cm) of the nbw

characteristics: (a) scalar neutron flux

Φ (×1015 cm-2 s-1); (b) concentration n (×1021 cm-3) for 239Pu (solid curves) and 233U (dots); (c) fuel burn-up

depth b (%) for the fuel components 238U–Pu (solid

curves) and 232Th (dots) for calculation variant 1 for

time moments t1 = 4, t2 = 100 days, t3 = 10, t4 = 30,

t5 = 45, t6 = 60 and t7 = 70 years.


Nuclear Burning Wave Reactor:

Smooth Start-Up Problem

Sergii Fomin, A. Fomin, Yu. Mel’nik, V. Pilipenko, N. Shul’ga




IAEA TWG-FR 29 Feb – 2 March 2012 IAEA Vienna

Startup problem of the NBW Reactor

(10-3 b-1cm-1) (10-3 b-1cm-1)

238U/8 238U/8

Pb-BiPb-Bi 239Pu

181Ta

239Pu

Neutron flux Φ, b-1day-1

Smooth Startup of the NBW Reactor

0 100 200 300 400 500 600 700

4

8

12

16

20

24

28

PI , GW

t, days

0 2 4 6 8 10 12 14 16 18 20 22 24

0,5

1,0

1,5

2,0

2,5

3,0

3,5

4,0

PI , GW

t , years



Mechanism of Negative Reactivity Feedback

in Nuclear Burning Wave Reactor

Sergii Fomin, A. Fomin, Yu. Mel’nik, V. Pilipenko, N. Shul’ga

Akhiezer Institute for Theoretical Physics, NSC KIPT, Kharkov, Ukraine


IAEA TWG-FR 2-7 March 2013 FR-13 Paris

ττττ ≈ 2.5 days R = 230 cm

Perturbation of integral neutron flux Fint (×1022 cm/s)

caused by an external neutron source via time t (days)

The source with intensity Qext = 2×1011 (cm-3 s-1)

starts at t0 = 3650 days, lasts during 1 hour

and is situated at а) 160 < z < 170 cm,

b) 200 < z < 210 cm, c) 55 < z < 65 cm.

Negative Reactivity Feedback of the NBW Reactor

-10 0 10 20 30 40 50 60 70 80

5

10

15

20

25

30 a)Φ

int

t-t0

-10 0 10 20 30 40 50 60 70 80

2

4

6

8

10

12

14

16

t-t0

c)

Φint

-10 0 10 20 30 40 50 60 70 80

2

4

6

8

10

12

14

16

t-t0

c)Φ

int

160 < z < 170 cm

200 < z < 210 cm

55 < z < 65 cm

Evolution of the volume-averaged neutron flux Fav (×1015 cм-2 с-1) and concentrations Nav (×1017 cm-3)

of the main fissile and intermediate nuclides in the fuel of mixed ThUPu cycle with time t (days) at the

initial stage of the neutron flux perturbation t0 = 3650 days. The averaged nuclide concentrations: NNp

is for 239Np, NPa = NPa - 53.1·1017 cm-3, is for 239Pu is for 233U.

-1 0 1 2 3 4 5 6 7 8

1

2

3

4

t - t0

Φav

0

1

2

4

5

6

7

8

~

~

NNp

NPa

NPu

NU

~

Nav

Φav

Negative Reactivity Feedback: Stability of the NBW Regime

0Pu Pu Pu 1t

N N N−

= −%

0U U U 1

,t

N N N−

= −%

ττττ ≈ 2.5 days

Variation of the reactivity ρ (dollars) with time t (days)

along the variation of the volume-averaged neutron flux Fav (×1015 cм-2 с-1)

Negative Reactivity Feedback: Stability of the NBW Regime

-1 0 1 2 3 4 5 6 7 8

1

2

3

4

t - t0

Φav

-0,20

-0,15

-0,10

-0,05

0,00

0,05

0,10

0,15

0,20

ρ

ρ Φav

Stability of the NBW Regime

Stability of the NBW Regime

0 10 20 30 40 50

2

4

6

8

10

12

14

16

18

Pint

, GW

t, years

62% Th, R=230 cm, 5% brick, z=300÷320 cm

62% Th, R=230 cm, 10% brick, z=300÷320 cm

Variation of the neutron flux Fav (××××1015 cм-2 с-1) with time t (days) due to the partial

loss of coolant (а) 1% fraction during 0.1 days in front of NBW (270<x<290 см) and

(г) 5% coolant fraction during 0.1 days after the NBW (150<x<170 см)

Coolant loosing NBW reactor behavior

-1 0 1 2 3 41.2

1.4

1.6

1.8

2.0

2.2

2.4a)

Φav

t-t0

-1 0 1 2 3 4

1.6

1.8

2.0

2.2

2.4

2.6

2.8г)

Φav

t-t0

- negative feedback on reactivity - intrinsic safety

- long-term (decades) operation without refueling and external control

- possibility of 232Th and 238U utilization as a fuel

- fuel burn-up depth for both 238U and 232Th ≈ 50%

- neutron flux in active zone ≈ 2·1015 n/сm2s

- neutron fluence during the whole reactor campaign ≈ 3·1024 n/сm2

- energy production density in active zone ≈ 200 W/сm3

- total power at the steady-state regime ≈ 1.2 GW

- wave velocity at the steady-state regime ≈ 2 сm/year

- possibility of nuclear waste burn out (expected)

Main features of NBW reactor with mixed Th-U-Pu fuel cycle

Reactor composition (vol. frac.):

Fuel = 55% (FTh = 62%, p = 0.20), Coolant = 30%, CM = 15%, R = 215 cm

1A-1-2: Sustainable Burning Reactors - Chairs: Kevan Weaver (TerraPower, USA)

Traveling-Wave Reactors: Challenges and Opportunities - Kevan Weaver et al. (TerraPower, USA)

Feasibility of LBE Cooled Breed and Burn Reactors - Ehud Greenspan (UC, Berkeley, USA)

Preliminary Engineering Design of Sodium-Cooled CANDLE Core - Hiroshi Sekimoto (TIT, Japan)

Nuclear Burning Wave in Fast Reactor with Mixed Th-U Fuel - Sergii Fomin et al (NSC KIPT, Ukraine)

Nuclear Traveling Wave in a Supercritical Water Cooled Fast Reactor – W. Maschek (KIT, Germany)

Development and Prospects of TWR Project in China - Zheng Mingguang (Shanghai NER&DI, China)

1A-3: Thorium Fuel Reactors - Chair: Sergii Fomin (KIPT, Ukraine)

(Th-U-Pu) - Mixed Fuel Cycle and Proliferation– E. Kryuchkov et al, (MEPhI, Russia)

Large Scale Utilization of Thorium in Gas Cooled Reactors - V. Jagannathan (Bhabha ARC, India)

INTERNATIONAL ATOMIC ENERGY AGENCYINTERNATIONAL ATOMIC ENERGY AGENCYINTERNATIONAL ATOMIC ENERGY AGENCYINTERNATIONAL ATOMIC ENERGY AGENCY

Nuclear Burning Wave Benchmark Specifications

for the IAEA Coordinated Research Projects

Analytical and Experimental Benchmark Analysis

on Accelerator Driven Systems

& Technical Working Group – Fast Reactors

Compiled By

S. Fomin, Yu. Melnik, V. Pilipenko, N. Shul’ga


National Science Center “Kharkov Institute of Physics and Technology”


[email protected]

Issued 2009-03-30

Спасибо за внимание

Thank you for attention!

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