Dynamic calculations of the IAEA safety MTR research ... calculations of the IAEA safety MTR...

18
Dynamic calculations of the IAEA safety MTR research reactor Benchmark problem using RELAP5/3.2 code Tewfik Hamidouche a , Anis Bousbia-Salah b, * , Martina Adorni b , Franscesco D’Auria b a Laboratoire des Analyses de S^ uret e, Centre de Recherche Nucl eaire d’Alger (CRNA), 02 Boulevard Frantz, Fanon, B.P. 399, 16000 Alger, Alg erie b Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolt a di Ingegneria, Universit a di Pisa, Via Diotisalvi, 2 – 56126, Pisa, Italy Received 19 November 2003; accepted 10 March 2004 Available online 23 April 2004 Abstract Nowadays, increased attention to safety issues for nuclear research reactors has emerged as a consequence of their enlarged commercial exploitation. Almost all of the research reactors safety analyses were, so far, performed using conservative computational tools. Currently, the application of Best-Estimate methods constitutes a real necessity in order to get a more re- alistic vision of the system behavior, and overcome constraining limits related to conservative approaches. The aim of the current work is an attempt to apply this technique using the system thermal-hydraulic RELAP5/Mod3.2 code. For this purpose, the IAEA 10 MW MTR pool type research reactors Benchmark problem is considered. The exercise consists in performing some fast transients related to typical reactivity induced accidents, and relatively slow tran- sients related to loss of flow accidents. The RELAP5 results were compared against previous data obtained by various conservative channel codes. Differences between the two modeling approaches are afterwards emphasized and discussed. Ó 2004 Elsevier Ltd. All rights reserved. * Corresponding author. E-mail addresses: [email protected] (T. Hamidouche), [email protected] (A. Bousbia- Salah). 0306-4549/$ - see front matter Ó 2004 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2004.03.008 www.elsevier.com/locate/anucene Annals of Nuclear Energy 31 (2004) 1385–1402 annals of NUCLEAR ENERGY

Transcript of Dynamic calculations of the IAEA safety MTR research ... calculations of the IAEA safety MTR...

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www.elsevier.com/locate/anucene

Annals of Nuclear Energy 31 (2004) 1385–1402

Dynamic calculations of the IAEA safetyMTR research reactor Benchmark problem using

RELAP5/3.2 code

Tewfik Hamidouche a, Anis Bousbia-Salah b,*,Martina Adorni b, Franscesco D’Auria b

a Laboratoire des Analyses de Suret�e, Centre de Recherche Nucl�eaire d’Alger (CRNA), 02 Boulevard

Frantz, Fanon, B.P. 399, 16000 Alger, Alg�erieb Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolt�a di Ingegneria, Universit�a

di Pisa, Via Diotisalvi, 2 – 56126, Pisa, Italy

Received 19 November 2003; accepted 10 March 2004

Available online 23 April 2004

Abstract

Nowadays, increased attention to safety issues for nuclear research reactors has emerged as

a consequence of their enlarged commercial exploitation. Almost all of the research reactors

safety analyses were, so far, performed using conservative computational tools. Currently, the

application of Best-Estimate methods constitutes a real necessity in order to get a more re-

alistic vision of the system behavior, and overcome constraining limits related to conservative

approaches. The aim of the current work is an attempt to apply this technique using the system

thermal-hydraulic RELAP5/Mod3.2 code. For this purpose, the IAEA 10 MW MTR pool

type research reactors Benchmark problem is considered. The exercise consists in performing

some fast transients related to typical reactivity induced accidents, and relatively slow tran-

sients related to loss of flow accidents. The RELAP5 results were compared against previous

data obtained by various conservative channel codes. Differences between the two modeling

approaches are afterwards emphasized and discussed.

� 2004 Elsevier Ltd. All rights reserved.

*Corresponding author.

E-mail addresses: [email protected] (T. Hamidouche), [email protected] (A. Bousbia-

Salah).

0306-4549/$ - see front matter � 2004 Elsevier Ltd. All rights reserved.

doi:10.1016/j.anucene.2004.03.008

1386 T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402

1. Introduction

Enlarged commercial exploitation of nuclear Research Reactors (RR) has in-

creased the consideration toward their safety issues. RR safety analyses have been,

so far, performed using conservative computational approaches (Woodruff, 1984;

IAEA, 1990; Housiadas, 2000; Hamidouche et al., 2002). However, recent avail-ability of powerful computers and computational techniques together with the

continuing increase in operational experience imposes the revisiting of those areas.

The application of Best-Estimate (BE) method constitutes a real necessity in order to

get more realistically simulations of the phenomena involved during steady state and

transient conditions and eventually the identification of design/safety requirements

that can be relaxed (Bousbia Salah et al., 2003).

The global aim of the current work constitutes an attempt to apply the BE

coupled code approach under the operating conditions of RR. For this purpose, theIAEA 10 MW MTR problem (IAEA, 1980) is considered since it covers a wide

range of hypothetical typical RR transient conditions. The calculations concern

particularly:

• rapid transient initiated by positive Reactivity Induced Accidents (RIA) during an

hypothetic control rod withdrawal, and

• relatively slow transient cases related to Loss Of core coolant Flow Accidents

(LOFA) as a consequence of main cooling pump failure.

In the current framework, only the first step of the coupled code calculationprocedure is considered. In other words, the Thermal-Hydraulic System (THS) code

RELAP5 is used alone taking into account its point Neutron Kinetic (NK) model.

The second step, under progress, will concern three dimensional (3D) calculations

using a 3D NK code.

Unlike the precedent IAEA Benchmark problem calculations based upon channel

code models (IAEA-TECDOC-643, 1990), conservative hypothesis related to the

consideration of imposed core boundary conditions are herein overcome. In fact, a

more realistic simulation is obtained through the consideration of a whole nodal-ization of a typical pool research reactor cooling loop. This will allow, among all, an

adequate simulation of:

• the dynamic interaction between the reactor cooling loop and the core kinetics,

• the flow reversal phenomenon leading to natural circulation regime.

Nevertheless, conservatism related to the consideration of representative core

channels and fixed hot channels peaking factors is kept.

2. IAEA research reactor Benchmark problem

The Benchmark problem consists in some protected transients in MTR Highly

Enriched Uranium (HEU) and Low Enriched Uranium (LEU) cores. These generic

reactors are representative of medium power research reactors with high fissile

loading and more demanding thermal-hydraulic requirements. They are used only to

give an indication of the reliability of the methods adopted. The main Benchmark

Table 1

Main Benchmark problem operating conditions

Core material

Nuclear fuel MTR

Fuel element Plate-type clad in Al

Coolant Light water (downward forced flow)

Moderator Light water

Reflector Graphite-light water

Core thermalhydraulics HEU LEU

Fuel thermal conductivity (W/cmK) 1.58 0.5

Cladding thermal conductivity (W/cmK) 1.80

Radial peaking factor 1.4

Axial peaking factor 1.5

Engineering factor 1.2

Inlet coolant temperature (�C) 38.0

Operating pressure (bar) 1.7

Fuel element dimensions

Length (cm) 8.00

Width (cm) 7.60

Height (cm) 60.0

Number of plates/fuel element ECS/ECN 23 / 17

Plate meat (mm) 0.51

Width (cm) active/ total 6.30/6.65

Height (cm) 60.0

Water channel between plates, mm 2.23

Plate clad thickness (mm) 0.38

Core kinetics HEU LEU

Effective delayed neutron Fraction 7.607E) 3 7.275E) 3

Prompt neutron generation time (ls) 55.96 43.74

Void feedback coefficient ($/% void) 0.3257 0.4047

Doppler feedback coefficient ($/�C) 3.6E) 5 3.31E) 3

Coolant temperature feedback ($/�C) 1.537E) 3 1.082E) 2

T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402 1387

specifications are outlined in Table 1, and detailed specifications data related to the

kinetics and thermal-hydraulic parameters are given in IAEA (1990). The Bench-

mark problem considers two categories of transients. The rapid one is governed bykinetic processes, whereas the slow category is associated mainly with thermal-

hydraulic phenomena.

2.1. Kinetic (or overpower) transients

This category of transients is characterized by positive reactivity addition into the

core. This leads to core overpower situations where the cooling system, even though

it works under nominal conditions, is not able to evacuate all the heat released intothe core. The leading RIA events are summarized in Table 2. As commonly known,

RIA events are characterized by prompt core power excursion course followed by

strong coupled feedback effects related to Doppler, coolant temperature, and void (if

present) effects. The Fast RIA (FRIA) transients, considered herein, are initiated by

a super prompt ramp positive reactivity addition of $1.5/0.5 s in both HEU and LEU

Table 2

Main Benchmark initial and boundary conditions

Transient key parameters RIA HEU, LEU LOFT HEU, LEU

Initial power 1.0 W 12.0 MW

Steady state duration time before

transient

50 s 50 s

Coolant flow direction Upward (downward in the current

framework)

Downward

Rate of external positive reactivity

addition

($) 1.5/0.5 (Fast RIA) 0.0

($) 1.3/0.5 (Slow RIA-LEU only)

($) 0.09/s (HEU only)

($) 0.10/s (LEU only)

Loss of flow decay period (s) – 1.0 (Fast LOFA)

– 25.0 (Slow LOFA)

Scram setting point 12 MW (120% of nominal power) 85% of nominal core

coolant flow rate

Delay time before Scram 0.025 s 0.2 s

Shut down reactivity )$10.0 in 0.5 s )$10.0 in 0.5 s

1388 T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402

cores. For the Slow RIA (SRIA), it is considered a positive reactivity insertion of 9

!/s in the HEU core and 10 !/s in the LEU core. The initial conditions are assumed to

be at an operating power of 1 W and full downward cooling flow (not as the

Benchmark specifications which consider initial upward flow). The safety system trip

point is set at 12 MW, it induces the shutdown negative reactivity of )$10 in 0.5 s

with a response delay time of 0.025 s.

2.2. Thermal-hydraulic transients

This category is characterized by a core heat up due to malfunction of the cooling

system even if the reactor power is operating at its nominal value. The transients

related to the LOFA cases involve strong thermal-hydraulic interactions between the

core and the coolant loop. The flow decay is modeled as an exponential ðexpð�t=T ÞÞdecrease with a period T equal to 1 and 25 s for the Fast LOFA (FLOFA) and the

Slow LOFA (SLOFA) cases, respectively. The LOFA transients are initiated at a

nominal core power of 12 MW and full core downward cooling flow conditions. Thereactor scrams when the flow decay is reduced by 15%, with a response delay time of

0.2 s.

Furthermore, when the flow decay reaches 15% of its initial value, the Natural

Convection Valve (NCV) opens to allow flow reversal and the establishment of

passive decay heat removal process by natural circulation flow.

3. Plant nodalization

The IAEA-10 MW HEU and LEU cores consist of 5� 6 grid core containing 21

MTR fuel elements and 4 control elements (see Fig. 1). The core is reflected by

W G G G G G

W ECS5%

ECS25%

ECS25%

ECS5%

W

ECS5%

ECC25%

ECS45%

ECS45%

ECC25%

ECS5%

ECS25%

ECS45%

ECS45%

H2O+

ALECS45%

ECS45%

ECS25%

ECS5%

ECC25%

ECS45%

ECS45%

ECC25%

ECS5%

W ECS5%

ECS25%

ECS25%

ECS25%

W

W G G G G W

Standard Fuel elementECSWater Element W

G ECCGraphite Control fuel element

% % Consumed Uranium

Fig. 1. Benchmark core configuration.

T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402 1389

graphite on two opposite sides and surrounded by light water. The standard fuel

elements contain 23 plates whereas the control fuel elements contain 17 standard

plates with special region to receive the 4 fork type absorber blades.No information was made available for the reactor cooling loop in the Bench-

mark specification volumes (IAEA, 1980). However, in the current framework, a

standard nodalization performed for a typical MTR research reactor is considered

(Pierro et al., 2003). In fact, an adequate simulation of the interaction between the

reactor cooling loop and the core (dynamic) kinetics as well as flow reversal and

natural circulation phenomena during the LOFA flow decay, needs to take into

account the whole coolant loop components. The adopted plant nodalization for the

Benchmark problem is shown in Fig. 2. It includes the main standard components ofa RR such as the core zone, the reactor pool, the holdup tank, the main coolant

pump, the heat exchanger, and a representative NCV (see also Table 3). The reactor

Fig. 2. Reactor nodalization.

Table 3

Main nodalization components and their relative reference code

Component Reference code

Hot channel 100

Average channel 105

Main coolant pump 310

Natural Convection Valve (NCV) 245

Heat exchanger 342

Holdup tank 280

Reactor pool 110 and 120

Pool atmosphere simulator 225

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T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402 1391

pool, upper the core zone, was modeled to get an operating pressure of 1.7 bar, as

well as to involve all the pool water during a natural circulation flow. For this

purpose, the pool was divided into two big volumes (element number 110 and 120)

connecting the core upper plenum with the NCV (element number 245). The position

of NCV valve was chosen to allow the buoyancy forces to overcome the frictional

pressure loses and allows the establishment of natural circulation flow between thereactor pool and the core zone.

In this framework, since the point kinetic model is used, conservative hypothesis

are assumed. These topics concern the assumption of two representative core re-

gions: the hot and the average channels, as well as fixed axial profile of the core

power. Additional core kinetics and thermal-hydraulics properties are summarized

in Table 1.

4. Results and discussion

To emphasize the differences between the conservative and BE codes, the RE-

LAP5 results are compared with two representative channel codes usually used for

research reactors safety analysis, namely PARET (Obenchain, 1969) and RETRAC-

PC (an enhanced version of the RETRAC code, Baggoura et al., 1994). All the re-

sults given in the tables (Tables 4–7) are issued from (IAEA, 1990) except those

obtained by PARET and RETRAC-PC which are gathered from in-house modelingand calculations using updated version of these codes.

4.1. RIA transients

Typical power behavior following a positive reactivity insertion in the HEU and

LEU cores is well predicted by RELAP5/Mod3.2 calculations (see Figs. 3 and 4, for

FRIA and SRIA, respectively). Differences between the power responses, of HEU

and LEU cores, are due essentially to the rate of inserted reactivity, which governsthe excursion period, and more particularly to their relative prompt neutron life

time. As can be seen in Fig. 3, the power response exhibits exponential rise during

the earlier transient instant, it is slightly slowed by prompt Doppler feedback effect.

The energy released during the transient did not alter significantly the power course

since the coolant temperature rise occurs during the control rods insertion period.

However, for the SRIA, a beginning of self-limiting power behavior (more visible

for the LEU curve, Fig. 4) is observed when the power trend begins to quench under

the delayed feedback effect of the coolant temperature rise. On the other hand, toillustrate the differences between the two code categories, namely the channel and

system codes, only results related to the HEU hot channel calculations are presented

for consistency. The power trends shown in Fig. 5 seem to be identical; however,

Table 4 shows differences in power peak values. The differences are more empha-

sized for the coolant temperature response (see Fig. 6). The coolant outlet tem-

perature calculated by RELAP5 exhibits a small delayed excursion response due to

the inertial effect of the whole cooling loop. The magnitude of this time delay is

Table 4

HEU and LEU – FRIA results

Program name RELAP5/3.2 PARET RETRAC-PC COSTAX-BOIL EUREKA-PT COBRA III-C

Laboratory UPISA ANL LAS JEN JAERI Interatom

Fast reactivity insertion transient $1.5/0.5 s

Minimum period (ms) LEU 12.0 (0.572) 12.13 (0.573) 12.24 (0.572) 13.5 (0.500) 12.2 (0.576) 12 (NA)

HEU 14.70 (0.609) 14.66 (0.609) 14.67 (0.608) 14.5 (0.500) 15.2 (0.619) 14 (0.604)

Peak power (MW) LEU 150.37

(0.612)

148.29

(0.613)

141.14 (0.612) 116.1 (0.638) 143.8 (0.616) 143.9 (0.608)

HEU 131.17

(0.655)

129.01

(0.655)

128.44 (0.655) 132.7 (0.659) 114.8 (0.664) 135.1 (0.650)

Peak clad temperature (�C) LEU 166.55

(0.629)

155.76

(0.629)

155.94 (0.626) 156.6 (0.654) 149.2 (0.627) 168.2 (0.625)

HEU 163.31

(0.673)

155.25

(0.672)

162.04 (0.668) 162.3 (0.675) 147.3 (0.678) 160.0 (0.665)

Onset of nucleate boiling (s) LEU None 0.615–0.668 0.614–0.664 NA NA NA

HEU None 0.656–0.715 0.653–0.719 NA NA NA

Peak coolant temperature (�C) LEU 78.01 (0.728) 82.0 (0.706) 79.42 (0.706) 80.4 (0.711) 62.7 (0.762) 63.2 (0.740)

HEU 78.90 (0.770) 84.32 (0.76) 82.97 (0.745) 108.7 (0.765) 62.3 (0.820) 70.7 (0.783)

Fast reactivity insertion transient $1.35/0.5 s

Minimum period (ms) LEU 17.66 (0.656) 17.52 (0.656) 17.60 (0.655) 19.2 (0.500) 17.1 (0.660) 17 (NA)

Peak power (MW) LEU 64.36 (0.693) 63.16 (0.694) 61.62 (0.693) 51.8 (0.729) 61.5 (0.697) 62.9 (0.688)

Peak clad temperature (�C) LEU 113.23

(0.718)

105.63

(0.718)

107.60 (0.716) 102.1 (0.756) 107.2 (0.722) 105.1 (0.710)

Peak coolant temperature LEU 56.18 (0.826) 57.86 (0.815) 56.84 (0.804) 54.9 (0.840) 55.2 (0.827) 52.0 (0.840)

Quantities between parentheses indicate time in seconds at which values occur.

NA: data not available in TECDOC-643.

UPISA: University of PISA (Italy); ANL: Argonne National Laboratory (USA); LAS: Laboratoire d’Analyse de Suret�e (Algeria); JEN: Junta de Energia

Nuclear (Spain); JAERI: Japan Atomic Energy Research Institute (Japan); INTERATOM: (Germany).

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Table 5

HEU and LEU – SRIA results

Program name RELAP5/3.2 PARET RETRAC-PC COSTAX-BOIL EUREKA-PT COBRA III-C

Laboratory UPISA ANL LAS JEN JAERI Interatom

Slow reactivity insertion 0.10$/s – HEU core

Minimum period (s) HEU 0.099 (10.62) 0.152 (10.62) 0.148 (10.62) 0.145 (10.61) NA 0.10

Peak power (MW) HEU 13.69 (10.65) 13.74 (10.65) 14.05 (10.64) 14.05 (10.64) 13.75 (10.668) 14.36 (10.59)

Peak clad temperature (�C) HEU 71.70 (10.66) 69.30 (10.67) 75.01 (10.66) 75.01 (10.66) 69.2 (10.693) 69.2 (10.62)

Peak coolant temperature (�C) HEU 47.98 (10.74) 48.21 (10.73) 48.05 (10.37) 48.05 (10.73) 47.7 (10.773) 45.2 (10.70)

Power (MW) at 20 s HEU 0.0078 0.0055 0.0049 0.0049 0.006 NA

Clad temperature (�C) at 20 s HEU 38.0 38.02 38.01 38.01 NA NA

Coolant temperature (�C) at 20 s HEU 38.0 38.00 38.00 38.00 NA NA

Slow reactivity insertion 0.09$/s – LEU core

Minimum period (s) LEU 0.088 (11.92) 0.85 (11.87) 1.10 (11.80) 1.11 (11.80) NA 0.11

Peak power (MW) LEU 12.34 (11.94) 12.36 (11.9) 12.29 (11.82) 12.29 (11.82) 12.35 (11.923) 12.18 (12.053)

Peak clad temperature (�C) LEU 81.12 (11.95) 77.59 (11.91) 78.52 (11.83) 78.52 (11.83) 78.5 (11.933) 78.1 (12.06)

Peak coolant temperature (�C) LEU 53.15 (11.99) 53.73 (11.94) 53.52 (11.83) 53.52 (11.88) 52.8 (11.978) 51.1 (12.10)

Power (MW) at 20 s LEU 0.022 0.0147 0.128 0.0128 0.015 NA

Clad temperature (�C) at 20 s LEU 38.0 38.06 38.04 38.04 NA NA

Coolant temperature (�C) at 20 s LEU 38.0 38.02 38.00 38.00 NA NA

Quantities between parentheses indicate time in seconds at which values occur.

NA: data not available in TECDOC-643.

UPISA: University of PISA (Italy); ANL: Argonne National Laboratory (USA); LAS: Laboratoire d’Analyse de Suret�e (Algeria); JEN: Junta de Energia

Nuclear (Spain); JAERI: Japan Atomic Energy Research Institute (Japan); INTERATOM: (Germany).

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1393

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2 .0

Time (sec)

1.0E+0

1.0E+1

1.0E+2

1.0E+3

1.0E+4

1.0E+5

1.0E+6

1.0E+7

1.0E+8

1.0E+9

Po

wer

(W

atts

)

RIA TRANSIENTS

LEU - $1.35 / 0.5sec

HEU - $ 1.5 /0.5 sec

LEU - $ 1.5 / 0.5 sec

Fig. 3. Core power evolutions during RIA transients.

0.0 4.0 8.0 12.0 16.0 20.0

Time (sec)

1.0E+0

1.0E+1

1.0E+2

1.0E+3

1.0E+4

1.0E+5

1.0E+6

1.0E+7

1.0E+8

Pow

er (W

atts

)

SRIA TRANSIENTS

HEU - $ 0.1 /sec

LEU - $ 0.09 / sec

Fig. 4. Core power evolutions during SRIA transients.

1394 T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402

proportional to the ratio of the length of the pipe versus its flow area (i.e. L=A)(Lewis, 1979). This effect is not observed in the channel codes calculations and, as

shown in Fig. 6 a prompt coolant temperature response is predicted. Furthermore,

0.0 0.3 0.5 0.8 1.0

Time (sec)

1.0E-1

1.0E+0

1.0E+1

1.0E+2

1.0E+3

1.0E+4

1.0E+5

1.0E+6

1.0E+7

1.0E+8

1.0E+9

Po

wer

(W

atts

)

-6.0

-4.5

-3.0

-1.5

0.0

1.5

3.0

Rea

ctiv

ity (D

olla

rs)

RELAP5

PARET

RETRAC

Scram

Fig. 5. Power and compensated reactivity during HEU-FRIA transient.

0.0 0.2 0.4 0.6 0.8 1.0

Time (sec)

20.0

40.0

60.0

80.0

100.0

Tem

pera

ture

( C

)

HEU $1.5/0.5 sec

RELAP5

PARET

RETRAC

Time of Peak power

Fig. 6. Coolant outlet temperature during HEU-FRIA transient.

T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402 1395

while channel codes assume fixed core inlet flow, the RELAP5 calculation predicts a

dynamic variation of this parameter (see Fig. 7). The core flow rate exhibits a pulse

reduction during the power excursion phase due to interactions between the core

and the coolant loop.

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0

Time (sec)

-138.0

-137.0

-136.0

-135.0

-134.0

-133.0

-132.0

-131.0

-130.0

Mas

s Fl

ow (k

g/se

c/m

2)

HEU $1.5/0.5 sec

Core outlet Flow

core inlet Flow

Time of Trip

Fig. 7. Inlet and outlet core flow during HEU-FRIA transient.

0.0 0.2 0.4 0.6 0.8 1.0

Time (sec)

0.0

40.0

80.0

120.0

160.0

200.0

Tem

pera

ture

( C

)

HEU $1.5/0.5 sec

RELAP5

PARET

RETRAC

Fig. 8. Clad surface temperature during HEU-FRIA transient.

1396 T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402

T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402 1397

In fact, as can be seen in Fig. 2, some part of the recirculation loop are above the

core level (element number 260), and if an overheat of the core takes place, the fluid

circulation tends to turn back under the effect buoyancy forces (differences in static

head between the cold water in the piping and the hot water in the core). The role of

these buoyancy forces is more pronounced in the case of LOFA transients.

Conversely, it should be noted that PARET and RETRAC-PC predict the Onsetof subcooled Nucleate Boiling (ONB) regime for a short time just before the peak

power time occurrence (see Table 4). This two-phase flow regime takes place when

the cladding temperature is higher than the ONB temperature ðTONBclad � 126 �C).

However, no boiling is predicted by the RELAP5/Mod 3.2 code even though higher

cladding temperature is obtained (up to 160 �C) (see Fig. 8 and Table 4). Similar

result was obtained by another RELAP5/Mod3 version (Woodruff et al., 1996), even

though no nodalization for an external loop was made. This may be due, as outlined

by (Kon�ear and Mavko, 2003), to inadequate RELAP5 Lahey’s model in predictingthe ONB under low pressure operating conditions. The higher predicted peak power

by RELAP5 (see Table 4) could be explained by the absence of void feedback

contribution during the power excursion course.

For the SRIA, as outlined in Table 5, differences between REALP5 and channel

codes results are practically insignificant for both power and temperature responses.

In fact, the core and the coolant loop interactions are weak since the dynamic of the

transient is very slow in this case.

4.2. LOFA transients

At a certain moment of the LOFA course, the buoyancy forces, due to the coolant

heat up by the decay heat generation, become important in comparison with the

decaying pump active forces. A mixed convection flow establishes followed by a flow

reversal and natural circulation regime. As pointed out by the TECDOC-643

Benchmark specifications (volume 1), the LOFA calculations are ended when the flow

decay reaches 15% of its nominal value. This limitation is mainly due to the inade-quacy of almost of the used channel computational tools, reported in Tables 6 and 7,

in performing further calculations involving thermal-hydraulic interactions between

the core and its cooling loop. However, the use of RELAP5 code allows the simu-

lation of all the leading processes to the natural convection regime. In the current

framework, only some representative LOFA results for HEU core are displayed.

The main output parameters of the transients, calculated by RELAP5/Mod3.2

code, are summarized in Table 6 for FLOFA and Table 7 for SLOFA in comparison

with the results of some channel codes as reported in (IAEA-TECDOC 643).Practically, no differences are observed between the HEU and LEU power responses.

This expected behavior is due to weak kinetic feedback effects involved during such

transients. On the other hand, the temperature responses are reported in Figs. 9 and

10. The fuel and the coolant temperatures exhibit a steady rise due to the degra-

dation of the core cooling process during the flow decay. After the scram of the

reactor power, a sharp decrease of core temperature is observed. However, due to the

combined effect of constant decay heat and continuous reduction of the core flow

Table 6

HEU and LEU – FLOFA results

Program name RELAP5/3.2 PARET RETRAC-PC COSTAX-BOIL EUREKA-PT COBRA III-C

Laboratory UPISA ANL LAS JEN JAERI INTERATOM

Power core at scram

(85% of nominal flow)

(MW)

LEU 11.83 (0.190) 11.86 (0.295) 11.72 (0.185) 11.67 NA 11.4 (0.363)

HEU 11.87 (0.200) 11.85 (0.295) 11.75 (0.192) 1167 NA 11.5 (0.363)

1st Peak cladding

temperature (�C)LEU 92.58 (0.400) 89.46 (0.505) 87.92 (0.400) 93.9 (0.37) 97.1 (0.40) 89.3 (0.363)

HEU 91.28 (0.408) 89.43 (0.505) 91.74 (0.385) 94.0 (0.37) 98.4 (0.40) 89.5 (0.380)

1st Peak coolant

temperature (�C)LEU 59.50 (0.504) 60.84 (0.601) 59.92 (0.465) 59.3 (0.43) 58.1 (0.48) 56.4 (0.460)

HEU 59.53 (0.503) 60.94(0.602) 60.04 (0.481) 59.4 (0.43) 58.4(0.48) 56.5 (0.460)

Flow inversion time (s) LEU 7.40 4.415 7.36a NA NA NA

HEU 7.40 4.415 7.66a NA NA NA

Coolant temperature

at 15% of nominal flow

LEU 46.70 47.15 45.63 NA NA NA

HEU 46.79 47.16 45.21 NA NA NA

2nd Peak cladding

temperature (�C)LEU 120.73 (10.00) 105.90 (8.51) 128.25 (7.36)a NA 95.2 (10.0) NA

HEU 121.00 (10.20) 105.76 (8.91) 112.12 (7.36)a NA 106.0 (10.0) NA

2nd Peak cooling

temperature (�C)LEU 105.3 (11.90) 101.67 (9.14) 69.76a NA 49.3 (10.0) NA

HEU 107.06 (11.10) 101.68 (9.13) 54.37a NA 48.3 (10.0) NA

Quantities between parentheses indicate time in seconds at which values occur.

NA: data not available in TECDOC-643.

UPISA: University of PISA (Italy); ANL: Argonne National Laboratory (USA); LAS: Laboratoire d’Analyse de Suret�e (Algeria); JEN: Junta de Energia

Nuclear (Spain); JAERI: Japan Atomic Energy Research Institute (Japan); INTERATOM: (Germany).aData given when RETRAC-PC calculations stopped (no flow inversion).

1398

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y31(2004)1385–1402

Table 7

HEU and LEU – LOFA results

Program name RELAP5/3.2 PARET RETRAC-PC COSTAX-BOIL EUREKA-PT COBRA III-C

Laboratory UPISA ANL LAS JEN JAERI Interatom

Power core at scram (MW) LEU 11.56 (4.102) 11.64 (3.87) 11.56 (4.050) 11.7 (4.06) NA 11.46 (4.263)

HEU 11.62 (4.098) 11.62 (3.915) 11.61 (4.047) 11.8 (4.06) NA 11.55 (4.263)

1st Peak cladding

temperature (�C)LEU 88.41 (4.299) 84.56 (4.07) 84.63 4.240) 90.3 (4.27) 96.1 (4.20) 85.5 (4.263)

HEU 88.67 (4.305) 84.51 (4.07) 84.69 (4.160) 90.7 (4.27) 96.4 (4.20) 85.8 (4.263)

1st Coolant peak

temperature (�C)LEU 57.97 (4.300) 58.83 (4.075) 58.82 (4.272) 58.1 (4.27) 57.5 (4.30) 55.4 (4.263)

HEU 58.78 (4.305) 58.81 (4.075) 58.83 (4.16) 58.3 (4.27) 57.7 (4.30) 55.6 (4.263)

Clad temperature at

15% of flow (�C)LEU 49.38 (47.50) 48.61 (43.58) 49.21 (47.19) NA 41.1 (10.0) NA

HEU 49.34 (47.39) 48.61 (43.57) 49.56 (43.83) NA 41.1 (10.0) NA

Cool. temperature at

15 % of flow (�C)LEU 43.50 43.43 42.42 NA 39.0 (10.0) NA

HEU 43.47 43.42 42.48 NA 39.0 (10.0) NA

Flow inversion time (s) LEU 57.40 62.84 57.26a NA NA

HEU 57.40 61.82 57.26a NA NA

2nd Clad peak

temperature (�C)LEU 98.48 (61.58) 98.23 (70.74) 79.29 (57.26)a 43.9 (45.0) NA NA

HEU 98.48 (61.58) 98.23 (70.74) 77.87 (57.26)a 43.9 (45.0) NA NA

2nd Coolant peak

temperature (�C)LEU 87.36 (64.20) 94.21 (70.74) 58.77 (57.26)a 44.0 (45.0) NA NA

HEU 87.11 (64.19) 98.23 (70.74) 65.20 (57.26)a 43.9 (45.0) NA NA

Quantities between parentheses indicate time in seconds at which values occur.

NA: data not available in TECDOC-643.

UPISA: University of PISA (Italy); ANL: Argonne National Laboratory (USA); LAS: Laboratoire d’Analyse de Suret�e (Algeria); JEN: Junta de Energia

Nuclear (Spain); JAERI: Japan Atomic Energy Research Institute (Japan); INTERATOM: (Germany).aData given when RETRAC-PC calculations stopped (no flow inversion).

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y31(2004)1385–1402

1399

0.00 4.00 8.00 12.00 16.00 20.00

Time (Sec)

20.00

40.00

60.00

80.00

100.00

120.00

140.00

Tem

per

atu

re (

C)

HEU FLOFA

RELAP

PARET

RETRAC

-1.0

0.0

1.0

2.0

3.0

Rel

ativ

e In

let M

ass

Flow

Flow Inversion

Fig. 9. Clad surface temperature and relative inlet mass flow rate during HEU-FLOFA transient.

0.00 20.00 40.00 60.00 80.00 100.00

Time (Sec)

0.00

20.00

40.00

60.00

80.00

100.00

120.00

Tem

per

atu

re (

C)

HEU SLOFA

RELAP5

PARET

RETRAC

-1.0

0.0

1.0

2.0

Rel

ativ

e In

let M

ass

Flow

Flow Inversion

Fig. 10. Clad surface temperature and relative inlet mass flow rate during HEU-SLOFA transient.

1400 T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402

rate, the core temperatures exhibit a second rise. The increase is further sustained as

the flow regime passes to laminar regime and to mixed flow when the NCV opens.

The core temperatures begin to decrease only when the natural circulation flow is

fully established. In Figs. 9 and 10 are also displayed the temperature responses and

the flow evolutions as calculated by PARET and RETRAC-PC codes. As it is ex-

pected, differences appear during the flow inversion phase. Indeed, this phenomenon

is governed by the balance between the system inertia ðL=AÞ and the buoyancy forces

T. Hamidouche et al. / Annals of Nuclear Energy 31 (2004) 1385–1402 1401

involved during the flow decay process. In comparison with PARET results, RE-

LAP5 predicts delayed flow inversion for FLOFA while in SLOFA cases the flow

inversion occurs earlier. This could be explained by the fact that unlike PARET

calculations, RELAP5 model take into account two key transient parameters; the

coolant loop inertia and more realistic buoyancy forces effects. For FLOFA, the

delayed response predicted by RELAP5 is due to the system inertia effect which isnot considered by PARET calculations. Whereas for slow transients, the flow in-

version is mainly governed by differences in modeling the buoyancy forces effect.

In general, differences between the RELAP5 and channel codes predictions for the

LOFA transients are mainly related to the thermal-hydraulic interactions between

the core and the coolant loop. These mechanisms could not, in any case, be ade-

quately simulated without considering the effect of different components of the real

plant configuration.

5. Conclusions

Increased consideration to safety issues for research reactors exploitation has

emerged as a consequence of their enlarged commercial exploitation. So far,

conservative computational tools were used to perform safety analyses for the design

and exploitation of such reactors. Nowadays it becomes necessary to review such

limiting tools by using Best-Estimate calculation methods. The current work consti-tutes a first attempt to apply this technique to the Research Reactors operating con-

ditions. For these purpose the well-known IAEA Research Reactors Benchmark

problem is considered. In general, for all the considered transients, the obtained results

show similar trendswith some specified channel codes results.However, as emphasized

by the comparative study, the RELAP5 simulation seems to be more realistic since it

take into account the interaction between the coolant loop and the core dynamic,

especially, during fast power excursion and loss of flow transients. Nevertheless, it

seems that low pressure models embedded into the current RELAP5/3.2 version arenot capable to predict correctly the onset of subcooled nucleate boiling flow regime.

A further investigation of the current Benchmark problem will be performed by

taking into account a more realistic neutron kinetic response using coupled THS-3D-

NK code simulations.

Acknowledgements

This work was performed under the auspices of Nuclear Research Center of

Algiers, in cooperation with the International Atomic Energy Agency (IAEA)through a contract Research Project (RP-ALG-12299).

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