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Progress in Nuclear Energy 50 (2008) 7e14www.elsevier.com/locate/pnucene

Review

Overview of accident analysis in nuclear research reactors

Tewfik Hamidouche a,*, Anis Bousbia-Salah b,*, El Khider Si-Ahmed c, Francesco D’Auria b

a Division de l’Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d’Alger (CRNA),02 Boulevard Frantz Fanon, B.P. 399, 16000 Alger, Algerie

b Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi 2, 56126 Pisa, Italyc Laboratoire de Mecanique des Fluide Theorique et Appliquee, Faculte de Physique, Universite des Sciences et de la Technologie,

Houari Boumedienne, Alger, Algerie

Abstract

Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. Nowadays, it becomespossible to switch to new generation of computational tools in order to get better realistic simulations of complex phenomena and transients. Thechallenge today is to revisit safety features of the existing research reactors in order to verify that the safety requirements still met and whennecessary to introduce some amendments, coming from not only the new requirements but also, in order to introduce new equipments fromrecent advancement of new technologies. The objective of this work is to give an overview of the state of the art in performing safety analysisof research reactors and to emphasize the need and the provision to achieve such goals.� 2007 Elsevier Ltd. All rights reserved.

Keywords: Raserch reactors safety; Accident analysis; Reactivity accidents; LOCA; Loss of heat removal; Loss of flow accidents; Validation; Channel codes; Best

estimate computational tools

1. Foreword

Nuclear research reactors have been of great support in thedevelopment of nuclear science and technology. According tothe IAEA inventory (IAEA-TECDOC-1387, 2004), 651 nu-clear Research Reactors (RRs) have been built in 56 countriesaround the world. To date almost 284 research reactors arecurrently in operation, 258 are shut down and 109 have beendecommissioned. More than half of all operating research reac-tors reached their first criticality more than thirty years ago.Therefore, some face concerns of obsolescence of equipment,lack of experiment programmes, lack of funding for operationand maintenance, loss of expertise, equipment ageing andretirement of staff. The 258 research reactors that no longer op-erate are in some form of shutdown. Some of these are expectedto restart sometime in the future, some are waiting

* Corresponding authors.

E-mail addresses: thamidouche@comena-dz.org (T. Hamidouche), b.salah@

ing.unipi.it (A. Bousbia-Salah).

0149-1970/$ - see front matter � 2007 Elsevier Ltd. All rights reserved.

doi:10.1016/j.pnucene.2007.11.089

decommissioning, with or without nuclear fuel in the facility,and the remaining reactors have no clear definition about theirfuture. The undefined status of these remaining research reac-tors gives rise to safety concerns, in particular in the areas ofloss of corporate memory, personnel qualification, mainte-nance of components and systems, and preparation and main-tenance of documentation.

In general, the purpose of nuclear research reactors is notfor energy generation; the maximum power generated withindoes not exceed 100 MW. They are commonly devoted forgeneration of neutrons for different scientific and social pur-poses. However, high power densities are involved in thecore and specific features are necessary to ensure safe utiliza-tion of these installations. In addition to their particular char-acteristics, including large variety of designs, wide range ofpowers, different modes of operation and purposes of utiliza-tion, special attention should be focused for their safety as-pects. Thus accurate safety evaluations, for instance in caseof core reloading, planned power up-rating, or as part of re-quired analyses of occurred events, should be considered.Generally speaking, the content of the safety report has to

8 T. Hamidouche et al. / Progress in Nuclear Energy 50 (2008) 7e14

be modified whenever a new type or design of fuel is to beused in the reactor core. As the existing plants have well estab-lished licensing procedures, including well founded analysismethods, the application of new innovative analysis methodshas to be thoroughly evaluated, with specific emphasis to thecapabilities of producing results that in general terms mightbe beneficial related to the RR operations.

An attempt to perform standardized safety analyses for RRwas proposed by the International Atomic Energy Agency eIAEA (IAEA-TECDOC-233, 1980) in the framework of coreconversion from the use of highly enriched uranium fuel tothe use of low enriched uranium fuel. In this regard, the facil-ity operator would be required to submit an amendment to, ora revision of, the Safety Report. For this purpose, a safety-related benchmark problem for an idealized generic 10 MWMTR light-water pool-type reactor was specified in order tocompare computational methods used in various research cen-ters and institutions. The related benchmark problem coverslarge steady state kinetic and thermalehydraulic calculationsand wide range of hypothetical dynamic transient conditions.However, almost all of the safety analyses have so far beenperformed using conservative computational tools (IAEA-TECDOC-643, 1990; Woodruff, 1984; Baggoura et al., 1994).

Nowadays, an established international expertise in relationto computational tools, procedures for their application, includ-ing best estimate methods supported by uncertainty evaluation,and comprehensive experimental database exists within thesafety technology of Nuclear Power Plants (NPP). The

Fig. 1. UMLRR noda

importance of transferring NPP safety technology tools andmethods to RR safety technology has been noted in recentIAEA activities. However, the ranges of parameters of interestto RR are different from those for NPP. This is namely true forfuel composition, system pressure, adopted materials and over-all system geometric configuration. The large variety of researchreactors prevented so far the achievement of systematic and de-tailed lists of initiating events based upon qualified PSA (Prob-abilistic Safety Assessment) studies with results endorsed by theinternational community. However, bounding and generalizedlists of events are available from IAEA documents and can beconsidered for deeper studies in the area.

In the area of acceptance criteria, established standards ac-cepted by the international community are available. Thereforeno major effort is needed, but an effort appears worthwhile tocheck that those standards are adopted and that the relatedthresholds are fulfilled.

The importance of suitable experimental validation is rec-ognized. A large amount of data exists as the kinetic dynamiccore behavior form SPERT reactors tests (Forbes and Nyer,1961). However, not all data are accessible to all institutionsand the relationship between the range of parameters of exper-iments and the range of parameters relevant to RR technologyis not always established. However, code-assessment throughrelevant set of experimental data is recorded and properlystored.

An established technology exists for development, qualifi-cation and application of system thermalehydraulics codes

lization scheme.

Fig. 2. Pool heat-up during the heat-up experiment.

Fig. 3. Sate of the liquid level in the pool and the core after the postulated

LOCA.

9T. Hamidouche et al. / Progress in Nuclear Energy 50 (2008) 7e14

suitable to be adopted for accident analysis in research reactors.This derives from NPP technology. The applicability of systemcodes like RELAP5, COBRA and MARS to the research reac-tor needs has been confirmed from recent IAEA activities. Def-initely, system codes are mature for application to transientanalysis in research reactors. However, code limitations havebeen found in predicting pressure drops as a function of massflux at low values of mass flux when nucleate boiling occurs.The importance of the Whittle and Forgan experiments shall

Fig. 4. Status of a partially submerged fuel element.

be mentioned, as well as the dependence of results from thenoding (cell subdivision) adopted by the code users.

Several code user choices, including time step may havea significant effect upon prediction, thus confirming the needfor detailed code user guidelines. Furthermore, code validationmust be demonstrated for the range of parameters of interest toresearch reactors. The crucial role of uncertainty in researchreactor technology has been emphasized,

(a) for the design, with main reference to the prediction of thenominal steady state conditions and,

(b) for the safety issues, with main reference to the predictionof the time evolution of significant safety parameters.

Fig. 5. Partially submerged fuel element model.

Fig. 6. Evolution of the maximum fuel plate temperature for various core immersion heights.

10 T. Hamidouche et al. / Progress in Nuclear Energy 50 (2008) 7e14

It has been observed that suitable-mature methods exist, butthe spread of these methods and procedures within the commu-nity of scientists working in research reactor technology islimited (IAEA-RCP J7-10.10., 2002). Therefore, the purposeof the present paper is to provide an overview of the accidentanalysis technology applied to the research reactor, with empha-sis given to the capabilities and limits of the used computationaltools.

2. Examples from code qualification process

Currently, with widespread use of research reactor, there isa real need to get more realistic simulations of the phenomenainvolved during steady state and transient conditions, and even-tually the identification of design/safety requirements that canbe relaxed or enhanced (Hamidouche et al., 2004). Several

Fig. 7. MTR benchmark transversal core cross-section.

attempts were performed to assess the applicability of Best Es-timate codes to RR operating conditions (Woodruff et al., 1996;Hamidouche et al., 2004). Relevant assessments were appliedagainst the following cases.

2.1. UMLRR research reactor

The UMLRR is a 1 MW MTR pool-type nuclear researchreactor. The main features connected with this reactor arethe fact that operational experimental data are available onlineand constitutes a real source for detailed measurement data forvarious code validations and system analyses (Bousbia-Salahet al., 2006a). Fig. 1 shows the REALP5 nodalization of thereactor, while Fig. 2 shows typical results obtained for the sim-ulation of a pool heat-up experiment.

Fig. 8. Radial fast flux distribution of 93% e fresh core.

Fig. 9. Radial thermal flux distribution of 93% e fresh core.

0.0 0.2 0.4 0.6 0.8 1.0Time (sec)

20.0

40.0

60.0

80.0

100.0

Tem

peratu

re ( C

)

HEU $1.5/0.5 sec

RELAP5PARETRETRAC

Fig. 11. Outlet fluid temperature during RIA transient.

11T. Hamidouche et al. / Progress in Nuclear Energy 50 (2008) 7e14

2.2. NUR research reactor

120.00HEU SLOFA

In this case, a general approach used at the NUR MTR pool-type research reactor for neutron flux optimization in irradiationchannels is considered. In order to allow the implementation ofthe new core configuration in the operation scheme of thereactor, detailed neutronic and thermal hydraulic studies areperformed. The safety steady state analysis were carried outusing combined lattice code WIMS-D4 (Askew et al., 1966)with CITATION diffusion code (Fowler et al., 1971) whereastransients calculations were performed using channel codessuch as PARET (Obenchain, 1969) and the in-house LODEHR(Bousbia-Salah et al., 2006b,c) codes. The objective was to in-sure that the allowed operational safety limits of the reactor arenot exceeded (Meftah et al., 2006).

The schematic representation of core uncovering duringa loss of coolant transient is shown in Figs. 3 and 4. Fig. 5shows the adopted nodalization for the LODEHR code, which

Fig. 10. Typical representation of an MTR research reactor applied for IAEA

Benchmark reactors.

take into account the level tracking approach. While Fig. 6shows the transient evolution of the core temperature duringthe postulated core uncovery accident.

2.3. The IAEA MTR fuel type research reactor

The IAEA Benchmark is based upon one of the SPERTseries test reactors (Forbes and Nyer, 1961). The reactor isan open pool 10 MW MTR fuelled core type. The core config-uration is shown in Fig. 7. The grid contains 21 Standard MTRFuel Elements (SFEs) and 4 Control Fuel Elements (CFEs).The SFEs contain 23 standard plates whereas the CFEs contain17 standard plates with special region to receive the 4 forktype absorber blades. The core is cooled and moderated bydownward forced circulation of light water. The benchmarkconsists in performing static and dynamic calculations. Thestatic computations were performed (Bousbia-Salah et al., in

0.00 20.00 40.00 60.00 80.00 100.00Time (Sec)

0.00

20.00

40.00

60.00

80.00

100.00

Tem

peratu

re ( C

)

RELAP5PARETRETRAC

-1.0

0.0

1.0

2.0

Relative In

let M

ass F

lo

w

Flow Inversion

Fig. 12. Fuel temperature and relative mass flow rate during slow LOFA.

4 8 12 16 20 24Inlet Mass Flux [Kg/m2/s] x 1000

0.0

0.2

0.4

0.6

0.8

1.0

Pressu

re D

ro

p [M

Pa]

THTL hot cases

Pe=1.7MPa

RELAP5 - FE714BExp - FE714B RELAP5 - FE714CExp - FE714CRELAP5 - FE105BExp - FE105B

Fig. 13. Channel pressure drop for THTL hot cases.

Fig. 15. Comparisons of SPERT III-C 19/52 data with PARET calculations.

12 T. Hamidouche et al. / Progress in Nuclear Energy 50 (2008) 7e14

press) using the Monte Carlo MCNP5 code (Jeremy, 2003).They concern criticality estimations and derivation of keykinetic parameters, as the spatial neutron flux distributionand the multiplication factor.

Figs. 8 and 9 show the MCNP5 results for the thermaland fast neutron flux, respectively. On the other hand, the dy-namic calculations were performed (Hamidouche et al., 2004)using the Best Estimate thermalehydraulic system code RE-LAP5/3.3. A standard representation of an MTR RR as shownin Fig. 10 has been adopted and the nodalization has been de-veloped and qualified according to the qualification procedureused at the University of Pisa (Hamidouche et al., 2004). Thebenchmark problem consists in analyzing protected transientsin MTR Highly Enriched Uranium (HEU) and Low EnrichedUranium (LEU) cores. The analysis concerns mainly rapidkinetic transient initiated by positive Reactivity Induced Acci-dents (RIA), and relatively slow transient (thermalehydraulicdriven) cases related to complete LOss of core coolant Flow

2000 4000 6000 8000 10000Mass Flux [kg/m2s]

0

20

40

60

80

100

Pre

ss

ure

D

ro

p [K

Pa

]

W&F TS2

1.23MW/m2 - RELAP51.23MW/m2 - Exp1.77MW/m2 - RELAP51.77MW/m2 - Exp2.18MW/m2 - RELAP52.18MW/m2 - Exp

Fig. 14. Pressure drop characteristic curve for W&F TS2.

Accidents (LOFA). Representative results are given in Figs.11 and 12, where comparison is made, when applicable,with reference results obtained by the RR devoted codesPARET and RETRAC-PC (Bousbia-Salah and Hamidouche,2005).

2.4. The Whittle and Forgan and THTL tests facilities

Relevant experimental data related the separate effect Onsetof Flow Instabilities (OFI) in the Oak Ridge National Labora-tory Thermal Hydraulic Test Loop (THTL) (Siman-Tov et al.,1994; Siman-Tov et al., 1995), as well as some representativeWhittle and Forgan (W&F) experiments (Forgan and Whittle,1966; Whittle and Forgan, 1967), and RELAP5 calculation re-sults are compared in Fig. 13 for the THTL case and in Fig. 14for the W&F case (Hamidouche and Bousbia-Salah, 2006).The objective of the work was to investigate the RELAP5/3.2 system code capabilities in predicting phenomena thatcould be encountered under abnormal research reactor’soperating conditions. Similar work was performed for thevalidation of ATHLET code (Lerchl and Austregesilo Filho,1995) under research reactor operating conditions (Hainounand Schaffrath, 2001).

2.5. SPERT tests

Validation process was also performed against experimentaltests as (Bousbia-Salah and Berkani, 2001) and those issuedfrom the full scale SPERT (Special Power Excursion ReactorTest) cases (Montgomery et al., 1957). Through these testsa check of the models of codes such PARET, RETRAC-PC,and RELAP5 is carried out. The SPERT III system was designedto investigate the effects of the initial system conditions on thecore behaviour under step reactivity insertions’ events. TheSPERT III reactor is light-water-cooled and moderated MTRcore with operating power of 60 MW. The power self-limiting tests were performed by considering step positive reac-tivity insertions ranging from 0.2 $ to 1.25 $ (Toole, 1964).

Results of the validation process are shown in Fig. 15, whichsummarises the calculation obtained for a large spectrum of

13T. Hamidouche et al. / Progress in Nuclear Energy 50 (2008) 7e14

self-limited transients in the SPERT III-C 19/52 core. Accept-able agreement is obtained when some adjustment of thefeedback parameters are performed (Bouaouina et al., 2007).

3. Topics of interest for accident analysis in RR andcurrent status

A list of topics relevant to the deterministic accident anal-ysis in research reactors is provided below.

� Postulated Initiating Events. The identification of accidentscenarios, typically derived by considering probability ofoccurrences and severity of consequences, constitutes thefirst step needed for performing deterministic safetyanalyses.� Acceptance criteria. The availability of ‘thresholds of

acceptability’ for consequences of accident, as a functionof the probability of occurrence of the event, constitutethe second requirement for deterministic safety analyses:namely results of the analyses shall be compared with‘limiting values’. Acceptance criteria are imposed bynational authorities and are not connected with the deter-ministic safety analysis.� Experimental database. Computational tools are used to

perform deterministic analyses and experimental data areneeded to demonstrate the quality of those tools. Experi-mental data can also be used directly to improve the designand the performance of research reactors.� Qualification of system codes. System codes, such as RE-

LAP5 (Fletcher and Schultz, 1995), CATHARE (Barre andBernard, 1990), ATHLET, CATHENA (Hanna, 1998),widely used within the safety technology of NPP (IAEATECDOC-1351, 2003; NEA/CSNI/R(99)10, 1998), couldbe used for the deterministic analysis of accidents inRR. The application must be based upon the evaluationof the complexity of the transient: in a number of casesowing to the ‘simple’ configuration of RR comparedwith NPP, simpler tools including analytical-hand calcula-tions should be used. However, for the cases when systemcodes are needed, proper demonstrations of qualificationmust be provided.� Uncertainty in research reactor technology. The role of

uncertainty has been considered at 2 levels: (a) design ofresearch reactors, mostly addressed to the calculation ofnominal steady state operating conditions, and (b) evaluat-ing the results from best estimated predictions performedby thermalehydraulic system codes, mostly addressed tothe calculation of transient scenarios. Origins and impactsof uncertainty within both the frameworks are discussed in(D’Auria et al., 2006; Bousbia-Salah et al., 2006c).Methods and procedures to deal with uncertainty are pre-sented in the same reference.

4. Conclusions and recommendations

The demonstration of applicability of qualified bestestimate system codes to RR accident analysis constitutes

the key message from this paper: a proper accident analysistechnology should be developed for RR that could benefit ofthe experience available from NPP, considering that the risklevel and the cost associated with RR are orders of lower mag-nitude. Recommendations are provided hereafter distinguish-ing between potential RR system thermalehydraulic codeusers and decision makers in the area.

Recommendations to the users of computational tools areas follows:

� To consider experimental data and to perform code-to-experiment comparison before any code application toprediction relevant to the RR design or safety analysis.� To demonstrate that any code adopted for design and

safety is qualified.� To consider that any best estimate code, even though sup-

ported by the use of the optimized procedures, producesresults affected by an unknown error, i.e. uncertainty.

Recommendations to decision makers focus on establishingan international understanding in the area:

B To plan ‘‘benchmark’’ exercises in conditions whereneutron kinetics and natural circulation are relevant.

B To promote the use of PSA techniques, establishingdetailed PIE (Postulated Initiating Events) lists.

B To make an effort to establish ‘validation-matrices’ forcomputational tools.

B To plan suitable training in the area of RR accidentanalysis.

B To consider innovative techniques including of CFD(Computational Fluid Dynamics) and coupled three-dimensional neutron kinetics codes and thermalehydraulicsystem codes.

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