paper relap5

Annals of Nuclear Energy 37 (2010) 1341–1350

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Annals of Nuclear Energy

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Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

Patrícia A.L. Reis a,b, Antonella L. Costa a,b,*, Cláubia Pereira a,b, Maria A.F. Veloso a,b, Amir Z. Mesquita c,Humberto V. Soares a,b, Graiciany de P. Barros a,b

a Departamento de Engenharia Nuclear – Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos No. 6627, Campus Pampulha, PCA 1, CEP 31270-901,Belo Horizonte, MG, Brazilb Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brazil1c Centro de Desenvolvimento da Tecnologia Nuclear – CDTN/CNEN, Av. Antônio Carlos, 6627, Campus UFMG, Belo Horizonte, Brazil

a r t i c l e i n f o

Article history:Received 18 February 2010Received in revised form 14 May 2010Accepted 18 May 2010Available online 15 June 2010

Keywords:RELAP5IPR-R1TRIGAThermal hydraulic

0306-4549/$ - see front matter � 2010 Elsevier Ltd. Adoi:10.1016/j.anucene.2010.05.013

* Corresponding author at: Departamento de EngEngenharia da Universidade Federal de Minas GeraisCampus Pampulha, PCA 1, CEP 31270-901, Belo Horiz34096688.

E-mail addresses: [email protected] (P.A.Lcom (A.L. Costa), [email protected] (C. Per(M.A.F. Veloso), [email protected] (A.Z. Mesquita), [email protected] (Graiciany de P. Barros).

1 http://www.cnpq.br/programas/inct/_apresentahtml.

a b s t r a c t

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it iswidely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has beenalso applied for thermal hydraulic analysis of nuclear research systems with good predictions. This workis a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGAresearch reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear TechnologyDevelopment Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation.The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reac-tors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improveresults in the transient condition respect to preceding investigations.

� 2010 Elsevier Ltd. All rights reserved.

1. Introduction

Interest in safety issues of nuclear research reactors is nowa-days increasing due their enlarged commercial exploitation com-monly devoted for generation of neutrons for different scientificand social purposes (Adorni, 2007; Bokhari et al., 2002; Khateret al., 2007; Khedr et al., 2005). The main activity of a nuclear re-search reactor is not for energy generation reaching maximumpower operation of about 100 MW. In spite of this, specific featuresare necessary to ensure safe utilization of such installations. There-fore, several codes have been used focusing special attention for re-search reactors safety analysis and valuation of specificperturbation plant processes. The RELAP5 system code was devel-oped to simulate transient scenarios in power reactors such asPWR and BWR but recent works as, for example (Antariksawanet al., 2005; Khedr et al., 2005; Marcum et al., 2010), have been

ll rights reserved.

enharia Nuclear – Escola de, Av. Antonio Carlos n 6627,onte, MG, Brazil Tel.: +55 31

. Reis), [email protected]), [email protected]@ig.com.br (H.V. Soares),

cao/inct_reatores_nucleares.

performed to investigate the applicability of the code to researchreactors operating conditions (TRIGA 2000, MTR, Oregon StateTRIGA), respectively.

Specifically, the Training, Research, Isotope, General Atomic(TRIGA) reactors are constructed in a variety of configurationsand capabilities, with steady-state power levels ranging from20 kW to 16 MW offering true ‘‘inherent safety”. TRIGA is a pooltype reactor that can be installed without a containment buildingbeing designed for use by scientific institutions and universitiesfor purposes such as graduate education, private commercial re-search, non-destructive testing and isotope production.

In the present work, the IPR-R1 TRIGA reactor, Mark-I model, in-stalled in Brazil, in operation since 1960, has been modeled for RE-LAP5 code with the aim of to reproduce the measured steady-stateas well as transient conditions. The development and the calcula-tion for the validation of a RELAP5 model for the IPR-R1 TRIGA re-search reactor have been presented. The version MOD3.3 was usedto perform the simulations. The model validation at 50 kW waspresented in a preceding work (Costa et al., 2010) using experi-mental data and also comparisons with data calculated using theSTHIRP-1 code (Veloso, 2004). However, the modeling was not ableto predict transient conditions adequately. Modifications werethen performed in the model mainly respect to the pool nodaliza-tion adding a cross flow model to permit a better heat removalfrom the core in natural circulation conditions. The current resultsobtained with this new nodalization demonstrate that the IPR-R1

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1342 P.A.L. Reis et al. / Annals of Nuclear Energy 37 (2010) 1341–1350

TRIGA model is representative of the reactor behavior consideringsteady-state and transient operation conditions as it is being de-scribed in the next sections.

1.1. RELAP5 and STHIP-1 codes

RELAP5 computer code is a LWR transient analysis code devel-oped mainly by the Idaho National Engineering Laboratory (INEL)for the U.S. Nuclear Regulatory Commission (NRC) for use in rule-making, licensing audit calculations, evaluation of operator guide-lines, and as a basis for a nuclear plant analyzer (US NRC, 2001).Specific applications have included simulations of transients inLWR systems such as loss of coolant, anticipated transients with-out scram (ATWS), and operational transients such as loss of feed-water, loss of offsite power, station blackout and turbine trip.However, some recent works have been performed to access theapplicability of the code to research reactors operating conditions.

The STHIRP-1 computer program was developed at the NuclearEngineering Department in the Federal University of Minas Gerais,Brazil, to simulate the IPR-R1 reactor. It uses the sub-channelsanalysis methodology to simulate, under steady-state and tran-sient conditions, the thermal and hydraulic phenomena occurringinside the core of water cooled research reactor under a naturalconvection regime (Veloso, 2004). The models and empirical corre-lations necessary to describe the flow phenomena which cannot bedescribed by theoretical relations were selected according to thecharacteristics of the IPR-R1 TRIGA research reactor operation.The program was validated against the IPR1 TRIGA model. The cal-culation results in comparison with the experimental data indicatethat the program reproduces the experimental data with goodagreements.

1.2. IPR-R1 TRIGA research reactor

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufacturedby the General Atomic Company and installed at Nuclear EnergyDevelopment Centre (CDTN) of Brazilian Nuclear Energy Commis-sion (CNEN), in Belo Horizonte, Brazil. It is a light water moderatedand cooled, graphite-reflected, open-pool type research reactor.

Fig. 1. Schematic representation of the IP

IPR-R1 works at 100 kW but it will be briefly licensed to operateat 250 kW. It presents low power, low pressure, for application inresearch, training and radioisotopes production. The reactor ishoused in a 6.625 m deep pool with 1.92 m of internal diameterand filled with demineralized light water. A schematic reactor dia-gram is illustrated in Fig. 1.

The main aim of the water in the pool is for cooling, as well asmoderator, neutron reflector and it is able to assure an adequateradioactive shielding. The reactor cooling occurs predominantlyby natural convection, with the circulation forces governed bythe water density differences. The heat removal generated fromthe nuclear fissions is performed pumping the pool water througha heat exchanger. The core has a radial cylindrical configurationwith six concentric rings (A, B, C, D, E, F) with 91 channels ableto host either fuel rods or other components like control rods,reflectors and irradiator channels. There are in the core 63 fuel ele-ments constituted by a cylindrical metal cladding filled with ahomogeneous mixture of zirconium hydride and Uranium 20% en-riched in 235U isotope. There are 59 fuel elements covered with alu-minum and four fuel elements with stainless steel. Details of IPR-R1 fuel elements are presented in Fig. 2. They have three axial sec-tions, upper and lower reflector (graphite), and the central portionfilled with fuel (U-ZrHx) which is 0.356 and 0.381 m in length forelements coated with aluminum and stainless steel, respectively.

The radial power distribution was calculated in precedingworks using the WIMSD4C and CITATION codes (Dalle, 2003; Dalleet al., 2002) and also experimental data (Souza, 1999). The radialfactor is defined as the ratio of the average linear power densityin the element to the average linear power density in the core.Fig. 3 shows the radial relative power distribution. The main fea-tures of the TRIGA IPR-R1 are presented in Table 1 (CDTN/CNEN,2000).

General Atomics developed the manufacturing process of fuelcontaining high concentrations of hydrogen using a metal alloyof uranium and zirconium, to fulfill a core requirement of intrinsicsafety (General Atomics, 1959). The characteristic of this fuel is theprompt negative temperature coefficient (�1.1 ± 0.2¢/�C) that pro-vides safety and automatically limiting the power when excessreactivity is suddenly inserted (Mesquita e Souza, 2010). The aver-

R-R1 (out of scale, measure in mm).

Fig. 2. IPR-R1 TRIGA fuel element design (measure in mm).

P.A.L. Reis et al. / Annals of Nuclear Energy 37 (2010) 1341–1350 1343

age energy of thermal neutrons follows the temperature in the fuelelements. Therefore, an increase in power causes an increase in thefuel temperature and, consequently, an increase in neutron energyand a decrease in reactivity due to:

– reduction of the cross sections of 235U due to increased energyof the neutrons.

– increase of parasitic absorption of neutrons due to Dopplerbroadening in the resonances of 238U, present in large quantitiesin the nucleus.

– increased leakage of neutrons.

Therefore, an increase of fuel temperature causes immediatereduction in the neutron population, reducing the reactor powerfaster than any device engineering. The fuel acts as an automaticpower regulator. This propriety of the fuel elements characterizinga high negative prompt temperature coefficient is the main reasonof the inherent safety behavior of the TRIGA reactors.

2. Nodalization description

Each of the 63 fuel elements was modeled separately and 63heat structure (HS) components were associated with 13 corre-sponding hydrodynamic pipe components constituting 13 hydro-dynamic channels (201–213), as can be verified in Fig. 4. Table 2presents some characteristics of the 13 regions.

Fig. 5 shows the RELAP5 general nodalization developed to sim-ulate the IPR-R1. The reactor pool was modeled using two pipecomponents, each one composed by ten volumes. As it can be ver-ified in the Fig. 5, both components (0 2 0 and 0 5 0) have their vol-umes connected by single junctions to characterize a cross flowmodel. The pool nodalization has been modified respect to a pre-ceding work (Costa et al., 2010) to improve transient predictionsas it is being clearly demonstrated in Section 4. A time dependentvolume was used to simulate the atmospheric pressure on the poolsurface. The natural convection system and the primary loop circu-lation have been modeled. The secondary loop, composed mainly

Fig. 3. Radial relative power distribution.


by the external cooling tower was not modeled in the presentnodalization because the primary circuit was sufficient to guarantythe heat removal of the coolant.

The point kinetics model was used in the current simulations. Adetailed representation of each element is, however, essential toproperly take into account the radial power distribution associatedwith the position of the fuel elements. The axial power distributionwas calculated considering a cosine profile and taking into accountalso that the power is cut off in the extremes of the element duethe presence of the graphite as it is sketched in Fig. 6. Althoughthe above modeling procedure is approximated, it is used here toconserve the actual axial and radial power distribution fixed.

3. Steady-state predictions

The validation of a RELAP5 nodalization implicates that the mod-el reproduces the measured steady-state conditions of the systemwith acceptable margins. The nodalization may be considered qual-ified when it has a geometric fidelity with the system, it reproducesthe measured steady-state condition of the system, and it demon-strates satisfactory time evolution conditions (D’Auria et al.,1999). However, sometimes a nodalization qualified to simulatedetermined condition may not be suitable to simulate other typeof situation being necessary modifications and re-qualification.

The RELAP5 steady-state calculation has been performed at 50and 100 kW. The temperature values at the inlet and outlet ofthe thermal hydraulic channels three, eight and 13 calculated usingRELAP5 can be verified in Tables 3 and 4, for 50 and 100 kW,respectively. The calculated values were compared with the avail-able experimental data (inlet and outlet channel temperature) andwith STHIRP-1 calculation data (outlet channel temperature).Chromel–alumel calibrated thermocouples were used to collectthe coolant temperature data and the measured values have amaximum error of ±1 �C (Veloso, 2004).

As it can be verified in the Table 3, considering operation at50 kW, the results for both codes are in good agreement in compar-ison with the experimental data. The error obtained using the RE-LAP5 calculation is into the range of the maximum acceptable errorsuggested for coolant temperature (0.5%) by the RELAP5 users(D’Auria and Galassi, 1998).

STHIRP-1 code reached values of outlet temperature withminimum error demonstrating an excellent reproduction of thereactor behavior. STHIRP-1 was developed specially accordingwith the IPR-R1 TRIGA characteristics and the core region wasmodeled with 104 TH channels against 13 TH channels in thepresent RELAP5 model. STHIRP-1 inlet temperature values arenot available. The inlet temperature values calculated using theRELAP5 code presented the same values for the three consideredchannels.

Fig. 4. Representation of the 13 TH channels in RELAP5 model.

Table 1Main thermal hydraulic and kinetic IPR-R1 core data (CDTN/CNEN, 2000).

Core power (kW) 250 0 (in process oflicensing)

Delayed neutron fraction 0.0079Prompt neutron generation time (s) 1004

Isothermal coefficient (cents/kW) 0.44Temperature reactivity feedback

(cents/�C)�11

Pressure of operation (kPa) 158 7Main moderator Zirconium hydride

(hydrogen)Cladding Aluminum or stainless

steel (SS)

Coolant Light water

Reflector Graphite

Al cladding SScladding

Number of fuel elements 59 4Fuel U–ZrHj o U–ZrH16

Zr concentration (% weight) 9.10 899U concentration (% weight) 8.0 8.5H concentration (%w eight) 1.0 1.6235U enrichment 20% 20%Cladding material A11100-F SS AISI-

304Fuel diameter (mm) 35.6 36.3Total height (mm) 722.4 720.6Gap width (mm) 0.09 0.14Total diameter (mm) 37.3 37.6Active height (mm) 355 6 381.0Graphite reflector height (mm) 101 6 88.1Gap material Helium Helium

Table 2Main features of the 13 thermal hydraulic (TH) regions.

THchannel

Number offuelelements

THchannelidentifier

Mass flowarea (nf)

HSidentifier

HS position(see Fig. 3)

1 6 201 0.003595 201–206 B1, B2, B3,B4, B5, B6

2 5 202 0 002779 207–211 C2, D2, C3,D3, D4

3 5 203 0.002779 212–216 C4, C5, D5,D6, D7

4 4 204 0002759 217–220 C6, D8, D9,D10

5 5 205 0.002779 221–225 C8, C9, D11,D12, D13

6 5 206 0.002779 226–230 C1O, C11,D14, 15,D16

7 4 207 0.002756 231–234 C12, D17,D18, D1

8 5 208 0005477 235–239 E2, E3, E4,E5, F6

9 5 209 0.005495 240–244 E6, E7, E8,E9, F11

10 4 210 0.005436 245–248 E1O, E11,E12, E13

11 5 211 0.005477 249–253 E14, E15,E16, E17,F21

12 5 212 0.005477 254–258 E18, E19,E20, E21,F26

13 213 0.005477 259–263 E22, E23,E24, E1, F1

Total = 63 Total = 63


Fig. 5. IPR-R1 TRIGA nodalization in the RELAP5 model.

Fig. 6. Prediction of the axial power distribution function in a TRIGA fuel element.

Table 3Experimental and calculated results at 50 kW of power operation.

THchannel

Outlet channel temperature (K) Inlet temperature (K)

Exp. RELAP5 Error(%)a

STHIRP-1

Error(%)a


3 300.0 298.4 0.5 299.0 0.4 294.1 294.7 0.18 298.0 296.4 0.5 297.6 0.1 296.1 294.7 0.513 298.0 296.4 0.5 298.5 0.1 296.1 294.7 0.5

a Error = 100 � (Calculation � Experimental)/Experimental.

Table 4Experimental and calculated results at 100 kW of power operation.

THchannel

Outlet channel temperature (K) Inlet temperature (K)


STHIRP-1

Error(%)a


3 304.0 301.3 0.9 305.3 0.4 294.0 295.7 0.68 300.5 298.8 0.8 300.1 0.2 295.5 295.7 0.113 301.5 298.8 1.1 303.6 0.6 296.5 295.7 0.3

a Error = 100 � (Calculation � Experimental)/Experimental.


Results performed at 100 kW of power operation are shown inTable 4. The error found for RELAP5 calculation is a few overesti-mated in comparison with the error suggested for coolant temper-ature (0.5%) by the RELAP5 users. However, considering the errorfrom the experimental data (±1 �C) the values predicted using RE-

LAP5 are perfectly acceptable for the present model validation pro-cess for operation power up to 100 kW.

Figs. 7 and 8 show the RELAP5 calculation for the inlet and out-let temperature for the TH channel 1, at 50 and 100 kW of power,respectively. Such channel was chosen because it concentrates theHS with higher values of radial power. As it can be verified, afterabout 2500 s of calculation, the temperatures reach steady-statecondition. The temperature stable values are in good agreementwith the experimental available data.

In addition, Fig. 9 shows the fuel and cladding temperature evo-lution for the heat structure 1 (HS-B1) in the channel one at themid high predicted using RELAP5. As it can be observed, theseparameters are completely stable and the difference of tempera-ture between the two curves is approximately 40 K. The HS-B1 cor-responds to the fuel element in the position B1 according to Fig. 3.Fig. 10 illustrates the time evolution for the HS-B1 fuel tempera-ture at four different axial levels.

The axial fuel temperature distribution follows the cosine pro-file function, reaching higher temperatures in the central parts ofthe element following the same power profile, as it is demon-strated in Fig. 11 for HS-B1.

Fig. 7. Inlet and outlet coolant temperature for the channel 1 at 50 kW predicted by the RELAP5.

Fig. 8. Inlet and outlet coolant temperature for the channel 1 at 100 kW predicted by the RELAP5.


4. Transient simulation

Several transient events are possible to occur in research reac-tors, as for example, loss of electrical power supplies, insertion ofexcess reactivity, loss of flow, loss of coolant, erroneous handlingor failure of equipment or components, special internal events,external events and human errors (IAEA, 2005). In spite of theIPR-R1 to be inherently safe, situations that can disturb the normalreactor operation are possible to occur. The RELAP5 model pre-sented in this work demonstrated to reproduce very well the stea-dy-state conditions. Therefore, in addition to the validation processof the modeling, a transient event was investigated using the codeand the results has been compared with available experimental

data. The investigated event is the forced recirculation off andmay be caused by the recirculation pump failure, bringing the reac-tor to operate in natural circulation conditions.

In the experiment, the reactor operated during about 2.5 h withthe forced cooling system switched off and with an indication of100 kW at the linear neutronic channel (Mesquita et al., 2009).The measurements have demonstrated an average temperature-rise rate of about 4.8 �C/h. At inlet and outlet of a thermal hydraulicchannel the temperature values were verified to increase about5.3 �C/h in both cases (Mesquita et al., 2009).

To perform the simulation using the RELAP5, the valve in theprimary system (number 600 in the nodalization) has been closedat 3000 s of calculation after the system to reach steady-state

Fig. 9. Cladding and fuel temperatures time evolution for an element in the TH channel 1 predicted using RELAP5.

Fig. 10. RELAP5 axial fuel temperature prediction at four levels of HS-B1.


condition. After the beginning of the transient at 3000 s, the tem-peratures increase as consequence of no energy removal from thepool since the primary was off. The coolant temperature at inletand outlet of the TH channel one then increases gradually withrates of about 4.9 �C/h and 4.6 �C/h, respectively, demonstratingvery good agreement with the experimental available data. Bothcurves are illustrated in the Fig. 12.

The modification of the pool nodalization represented an essen-tial process to improve the transient predictions using the RELAP5.The insertion of the cross flow model makes possible better removal

of heat from the core during natural circulation condition dueimprovement on the coolant flow between the pool pipe volumes.Fig. 13 illustrates the coolant temperature predicted using the codeconsidering the nodalization presented in this paper with cross flowmodel and that in the preceding work without cross flow model(Costa et al., 2010), both at 100 kW of power operation. The curvesshow clearly that the model using cross flow presents a tempera-ture-rise rate (4.9 �C/h) much more approximated to the experimen-tal rate (4.8 �C/h) than the model without cross flow model (30.0 �C/h), demonstrating the efficiency of the changes in the nodalization.

Fig. 11. Axial fuel, cladding and coolant temperature distributions predicted using RELAP5.

Fig. 12. Inlet and outlet coolant temperature for the channel 1 at 100 kW predicted by the RELAP5 after forced recirculation off at 3000 s.


5. Conclusions

In this work, a nodalization for the IPR-R1 TRIGA research reac-tor performed using the RELAP5/MOD3.3 code was presented as acontribution to the assessment of such code for research reactorsafety analysis. The nodalization was validated against experimen-tal data from steady-state conditions at 50 and 100 kW of poweroperation. RELAP5 results have been also compared with dataobtained using the STHIRP-1 thermal hydraulic code. The resultsdemonstrate good agreement between the codes with little dis-crepancies which could be explained by the different empirical cor-relations embedded within each code. The little discrepancies canbe also related to the differences on the nodalization methodologiesadopted for each code. Future investigations will be performed to

verify the effect of the number of TH channels in the calculation re-sults as a sensitivity investigation. Moreover, other transient calcu-lations must be performed and compared with experimental dataaiming to complete the code validation process. Considering thethree basic aspects necessary to qualify a nodalization for a system(geometric fidelity, reproduction of the measured steady-state con-ditions and satisfactory time evolution conditions), it is possible toconclude that the RELAP5 model presented in this work was quali-fied to represent adequately the IPR-R1 TRIGA research reactor insteady-state as well as in transient situations.

Future works about the present model include sensitivity inves-tigations including core cross flow calculation and also analysisadding the coupled neutron kinetic calculation using, for example,the RELAP5–3D code.

Fig. 13. Forced recirculation off transient–prediction using two different types of pool nodalization.


Acknowledgements

The authors are grateful to CAPES, CDTN, CNPq and FAPEMIG forthe support.

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