Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low-temperature heavy water moderator.

David William Hummel , David Raymond Novog

Department of Engineering Physics, McMaster University, Canada

h i g h l i g h t s

•AcoupledspatialkineticsandthermalhydraulicsmodelofthePT-SCWRwascreated

•Positivepowerexcursionsweredemonstratedduringaccident-liketransients

•Thereactorwillinherentlyself-shutdowninsuchtransientswithsomedelay

•Afast-actingshutdownsystemwouldlimittheconsequencesofthepowerpulse

a r t i c l e i n f o

Article history:

Received 3 June 2015

Received in revised form

21 November 2015

Accepted 7 December 2015

Available online 7 January 2016

a b s t r a c t

TheCanadianSupercriticalWater-cooledReactorconcept,asanevolutionoftheCANadaDeuterium Uranium(CANDU)reactor,includesbothpressuretubesandalowtemperatureheavywatermoderator ThecurrentPressureTubetypeSCWR(PT-SCWR)conceptfeatures64-elementfuelassembliesplaced withinHighEfficiencyRe-entrantChannels(HERCs)thatconnecttocoreinletandoutletplena.Among currentSCWRconceptsthePT-SCWRisuniqueinthattheHERCseparatesmultiplecoolantand mod-eratorregions,givingrisetocoupledneutronic-thermalhydraulicfeedbacksbeyondthosepresentin CANDUorcontemporaryLightWaterReactors.Theobjectiveofthisworkwasthustomodelthe cou-pledneutronic-thermalhydraulicpropertiesofthePT-SCWRtoestablishtheimpactofthesemultiple regionsonthecore’stransientbehavior.Tothatend,thefeaturesofthePT-SCWRwerefirstmodeled withtheneutrontransportcodeDRAGONtocreateadatabaseofhomogenizedandcondensed cross-sectionsandthermalhydraulicfeedbackcoefficients.Thesewereusedasinputtoacore-levelneutron diffusionmodelcreatedwiththecodeDONJON.Thebehavioroftheprimaryheattransportsystemwas modeledwiththethermalhydraulicsystemcodeCATHENA.Aprocedurewasdevelopedtocouplethe outputsofDONJONandCATHENA,facilitatingthree-dimensionalspatialneutronkineticsandcoupled thermalhydraulicanalysisofthePT-SCWRcore.Severalpostulatedtransientswereinitiatedwithinthe coupledmodelbychangingthecoreinletandoutletboundaryconditions.Decreasingcoolantdensity aroundthefuelwasdemonstratedtoproducepositivepowerexcursions(i.e.,thecoolantvoidreactivity aroundthefuelwaspositive),butsuchpowertransientswerefoundtobeinherentlyself-terminating

aslowdensitycoolantinevitablyreachesotherpartsoftheHERCgeometry(wherethevoidreactivity

ishighlynegative).Nevertheless,theobservedpowerexcursionspotentiallydemonstratetheneedfor fast-actingshutdownsystemintervention,similartoCANDUdesigns

©2015TheAuthors.PublishedbyElsevierB.V.ThisisanopenaccessarticleundertheCCBYlicense

(http://creativecommons.org/licenses/by/4.0/)

∗ Correspondence to: McMaster University, Dept of Engineering Physics, 1280

Main Street, West John Hodgins Engineering Building, Room A315, Hamilton,

Ontario, Canada L8S 4L7 Tel.: +1 9055259140x24924.

E-mail addresses: hummeld@mcmaster.ca, hummeld@gmail.com

(D.W Hummel).

1 Introduction

http://dx.doi.org/10.1016/j.nucengdes.2015.12.008

0029-5493/© 2015 The Authors Published by Elsevier B.V This is an open access article under the CC BY license (http://creativecommons.org/licenses/by/4.0/).

Fig 1.PT-SCWR HERC concept with 64-element fuel assembly.

Colton,2013)

2 Modeling methodology

(Marleauetal.,2008).InthisstudyDRAGONwasusedtogenerate

Fig 2.DRAGON spatial mesh for the infinite lattice (left) and reflector multicell (right).

PT-SCWR

(Hummeletal.,2013;HarrisonandMarleau,2013).Thisapproach

(Salaunetal.,2014)

(InternationalAtomicEnergyAgency,2012)

Fig 3. Evolution of the infinite lattice cell with burnup.

thecore(Penceretal.,2013).The84channelsweremodeledwith

Fig 5.PT-SCWR core modeled in DONJON.

section

(Hanna, 1998).It usesaone-dimensional, two-fluid

Wang,2013).Inlieuofthese,standardheattransfercorrelationsare

andWang,2013)

Novog,2014).Thisworkproceedsassumingthatpowershaping

Fig 6.CATHENA idealization of the PT-SCWR.

available

Fig 7.PT-SCWR channel power distributions used as initial conditions.

Fig 8.Transient coupling procedure used for DONJON and CATHENA.

(Salaunetal.,2015)

3 Transient simulation results

Figs.9–11showthetransientresultsfroma2.5◦Cstepreduction

Fig 10.Axial profiles during inlet temperature step down at BOC.

Figs.10and11showaxialdistributionsinthecoreaveraged

Fig 12.Integral core parameters during inlet temperature step up.

before

Fig 14.Axial profiles during outlet pressure transient at MOC.

Fig 15.Integral core parameters during inlet pressure transient.

Figs.15and16showatransientinitiatedbyasimilarreduction

Fig 17.Integral core parameters during flow reversal transient at MOC.

Figs.17and18showa“flowreversal”transientwheretheinlet

Fig 19.Integral core parameters during LOCA-like transient at MOC.

seconds

Figs.21and 22showa “flowrundown”transientwherethe

Fig 21.Integral core parameters during flow rundown transient at MOC.

Fig 22.Axial profiles during flow rundown transient at MOC.

4 Conclusions

postulatedtransientsinitiated bychangestothecoreboundary

design:

self-terminating

CANDU

Acknowledgments

References

Chang, J.J., Bo, W.R., Jong, Y.J., Ho, C.S LOCA power pulse characteristics in a

CANDU-6 with CANFLEX-RU fuel Proceedings of the Korean Nuclear Society Autumn Meeting, Yongpyong, South Korea, 2003.

Hanna, B.N., 1998 CATHENA: A thermalhydraulic code for CANDU analysis Nuclear Eng Des 180, 113–131.

Harrison, G., Marleau, G., 2013 Simulation strategy for the evaluation of neutronic properties of a Canadian SCWR fuel channel Sci Technol Nuclear Install 2013, http://www.hindawi.com/journals/stni/2013/352757/cta/.

Hummel, D.W., Novog, D.R Optimized channel inlet orifice sizing for the pressure tube type supercritical water cooled reactor The 19th Pacific Basin Nuclear Conference (PBNC 2014), Vancouver, Canada, 2014.

Hummel, D.W., Langton, S.E., Ball, M.R., Novog, D.R., Buijs, A Description and pre-liminary results of a two-dimensional lattice physics code benchmark for the Canadian pressure tube supercritical water-cooled reactor (PT-SCWR), The 6th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-6), Shenzhen, China, 2013.

International Atomic Energy Agency, “WIMS Library Update Project,” 28 January

2008 [Online] Available from http://www-nds.iaea.org/wimsd (accessed 29.10.12).

Kozlowski, T., Downar, T.J., 2006 Pressurized water reactor MOX/UO2 core transient benchmark—final report Nuclear Energy Agency—Organization for Economic Co-operation and Development, Paris, France (NEA/NSC/DOC(2006)20) Leung, L.K.H., Yetisir, M., Diamond, W., Martin, D., Pencer, J., Hyland, B., Hamil-ton, H., Guzonas, D., Duffey, R, A next generation heavy water nuclear reactor with supercritical water as coolant, International Conference on Future of Heavy Water Reactors, Ottawa, Canada, 2011.

Marleau, G., Hébert, A., Roy, R., 2008 A User Guide for DRAGON 3.06 Institute of Nuclear Engineering, École Polytechnique de Montréal, Montréal, Canada (Tech.

Re IGE-174 Rev 8).

Pencer, J., Colton, A Progression of the lattice physics concept for the Canadian super-critical water reactor 34th Annual Conference of the Canadian Nuclear Society, Toronto, Canada, 2013.

Pencer, J., Watts, D., Colton, A., Wang, X., Blomely, L., Anghel, V., Yue, S Core neutron-ics for the Canadian SCWR conceptual design The 6th International Symposium

on Supercritical Water-Cooled Reactors (ISSCWR-6), Shenzhen, China, 2013 Pencer, J., McDonald, M., Anghel, V Parameters for Transient response modelling for the Canadian SCWR The 19th Pacific Basin Nuclear Conference (PBNC 2014), Vancouver, Canada, 2014.

Salaun, F., Hummel, D.W., Novog, D.R., The impact of the radial reflector on the 8-group cell-averaged cross-sections for the SCWR 62-element lattice cell 2014 Canada-China Conference on Advanced Reactor Development (CCCARD-2014), Niagara Falls, Canada, 2014.

Salaun, F., Sharpe, J.R., Hummel, D.W., Buijs, A., Novog, D.R Optimization of the PTSCWR control blade sequence using PARCS and DAKOTA The 7th Interna-tional Symposium on Supercritical Water-Cooled Reactors (ISSCWR-7), Helsinki, Finland, 2015.

Shan, J., Chen, W., Rhee, B.W., Leung, L.K.H., 2010 Coupled neutronics/thermal-hydraulics analysis of CANDU-SCWR fuel channel Ann Nuclear Energy 37, 58–65.

Varin, E., Hébert, A., Roy, R., Koclas, J., 2005 A User Guide for DONJON Version 3.01 Institute of Nuclear Engineering, École Polytechnique de Montréal, Montréal, Canada (Tech Re IGE-208 Rev 4).

Wang, D.F., Wang, S A CATHENA model of the Canadian SCWR concept for safety Analysis The 6th International Symposium on Supercritical Water-Cooled Reac-tors (ISSCWR-6), Shenzhen, China, 2013.

Wu, P., Shan, J., Gou, J., Zhang, B., Zhang, B., Wang, H LOCA/LOECC analysis for Canadian-SCWR 7th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-7), Helsinki, Finland, 2015.

Yang, P., Cao, L., Wu, H., Wang, C., 2011 Core design study on CANDU-SCWR with 3D neutronics/thermal-hydraulics coupling Nuclear Eng Des 241, 4714–4719 Yetisir, M., Gaudet, M., Rhodes, D Development and integration of Canadian scwr concept with counter-flow fuel assembly, The 6th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-6), Shenzhen, China, 2013.