Abstract
During the in-vessel stage of a severe accident in a CANDU 6 reactor, decay heat from a collapsed core would be rejected through the calandria walls into the surrounding water. At the step in the calandria wall, the subshell and annular plate meet at a right angle pointing into the calandria. The geometry at this joint could concentrate the exiting heat flux, potentially leading to calandria failure. Finite element analysis is used to study the heat transfer near the welded joint. Different weld profiles, boundary conditions, and decay heat characteristics are considered, and the local concentration of exiting heat flux is calculated.
Abstract
Während eines schweren Unfalls in einem CANDU 6-Reaktor wird eine Phase erreicht, während der im Reaktorbehälter die Nachzerfallswärme eines kollabierten Kerns durch die Wände der Calandria in das umgebende Wasser abgeleitet wird. An der Stufe in der Calandriawand treffen die Unterschale und die ringförmige Platte in einem rechten Winkel aufeinander, der in die Calandria zeigt. Die Geometrie an dieser Verbindungsstelle könnte den austretenden Wärmestrom verstärken, was zu einem Versagen der Calandria führen könnte. Mit Hilfe der Finite-Elemente-Analyse wird der Wärmeübergang in der Nähe der Schweißnaht untersucht. Es werden verschiedene Schweißprofile, Randbedingungen und Zerfallseigenschaften berücksichtigt und die lokale Konzentration des austretenden Wärmestroms berechnet.
Acknowledgements
I thank J.H. Spencer and J.R. Buell for critical review of the manuscript. This work was funded by Atomic Energy of Canada Ltd. under the Federal Nuclear Science and Technology program.
References
1 Mathew, P. M.; Nitheanandan, T.; Bushby, S. J.: Severe core damage accident progression within a CANDU 6 calandria vessel. European Review Meeting on Severe Accident Research, Nesseber, Bulgaria, September 23–25, 2008Search in Google Scholar
2 Park, S.-Y.; Jin, Y.-H.; Song, Y.-M.: An investigation of an in-vessel corium retention strategy for the Wolsong pressurized heavy water reactor plants. Nuclear Technology 158 (2007) 109–115, DOI:10.13182/NT07-A382910.13182/NT07-A3829Search in Google Scholar
3 Ma, W.; Yuan, Y.; Sehgal, B. R.: In-vessel melt retention of pressurized water reactors: historical review and future research needs. Engineering 2 (2016) 103–111, DOI:10.1016/J.ENG.2016.01.01910.1016/J.ENG.2016.01.019Search in Google Scholar
4 Dupleac, D.: Analysis of late phase severe accident phenomena in CANDU plant. Journal of Nuclear Engineering and Radiation Science 3 (2017) 020904, DOI:10.1115/1.403541610.1115/1.4035416Search in Google Scholar
5 Mohta, K.; Gokhale, O. S.; Gupta, S. K.; Mukhopadhyay, D.; Chattopadhyay, J.: Structural integrity assessment of calandria of 540 MWe PHWR for in-vessel corium retention. Nuclear Engineering and Design 367 (2020) 110791, DOI:10.1016/j.nucengdes.2020.11079110.1016/j.nucengdes.2020.110791Search in Google Scholar
6 Spencer, J.: Measurement of critical heat flux in a CANDU end shield consisting of a vertical surface abutting a packed bed of steel shielding balls. CNL Nuclear Review 6 (2017) 79–90, DOI:10.12943/CNR.2016.0000210.12943/CNR.2016.00002Search in Google Scholar
7 Nicolici, S.; Dupleac, D.; Prisecaru, I.: Numerical analysis of debris melting phenomena during late phase CANDU 6 severe accident. Nuclear Engineering and Design 254 (2013) 272–279, DOI:10.1016/j.nucengdes.2012.09.02310.1016/j.nucengdes.2012.09.023Search in Google Scholar
8 David, R.: Calculation of heat fluxes exiting a CANDU calandria during in-vessel retention. Nuclear Technology 205 (2019) 1488–1494, DOI:10.1080/00295450.2019.159758110.1080/00295450.2019.1597581Search in Google Scholar
9 Ansys MechanicalAPDL, Release 17.0, Help System, Ansys, Inc., 2015Search in Google Scholar
10 Welded Steel Construction (Metal Arc Welding). Canadian Standards Association, W59–1989, Toronto, 1989Search in Google Scholar
11 Rempe, J. L.; Chávez, S. A.; Thinnes,G. L.; Allison, C. M.; Korth,G. E.; Witt, R. J.; Sienicki, J. J.;Wang, S. K.; Stickler, L. A.; Heath, C. H.; Snow, S. D.: Light WaterReactor Lower Head Failure Analysis. U.S. Nuclear Regulatory Commission, NUREG/CR-5642,Washington, 1993, PMid:8431759; DOI:10.2172/1019157010.2172/10191570Search in Google Scholar
12 Verma, P. K.; Kulkarni, P. P.; Pandey, P.; Prasad, S. V.; Nayak, A. K.: Critical heat flux on curved calandria vessel of Indian PHWRs during severe accident condition. Journal of Heat Transfer 143 (2021) 022101, DOI:10.1115/1.404882310.1115/1.4048823Search in Google Scholar
13 Nishikawa, K.; Fujita, Y.; Uchida, S.; Ohta, H.: Effect of surface configuration on nucleate boiling heat transfer. International Journal of Heat and Mass Transfer 27 (1984) 1559–1571, DOI:10.1016/0017-9310(84)90268-010.1016/0017-9310(84)90268-0Search in Google Scholar
14 Incropera, F. P.; DeWitt, D. P.; Bergman, T. L.; Lavine, A. S.: Fundamentals of Heat and Mass Transfer, 6th ed. John Wiley& Sons, Hoboken, USA, 2007, p. 624Search in Google Scholar
15 Zhang, L.; Zhou, Y.; Zhang, Y.; Tian, W.; Qiu, S.; Su, G.: Natural convection heat transfer in corium pools: a review work of experimental studies. Progress in Nuclear Energy 79 (2015) 167–181, DOI:10.1016/j.pnucene.2014.11.02110.1016/j.pnucene.2014.11.021Search in Google Scholar
16 Huang, Y.; Han, Z.; Pan,D.; Cheng, C.; Zhou, H.: Analyze the singularity of the heat flux with a singular boundary element. Engineering Analysis with Boundary Elements 120 (2020) 59 –66, DOI:10.1016/j.enganabound.2020.08.00410.1016/j.enganabound.2020.08.004Search in Google Scholar
17 He, Y.; Yang, H.; Deeks, A. J.: An element-free Galerkin scaled boundary method for steady-state heat transfer problems. Numerical Heat Transfer, Part B: Fundamentals 64 (2013) 199–217, DOI:10.1080/10407790.2013.79177710.1080/10407790.2013.791777Search in Google Scholar
18 Cheung, F. B.; Haddad, K. H.; Liu, Y. C.: Critical Heat Flux (CHF) Phenomenon on a Downward Facing Curved Surface. U.S. Nuclear Regulatory Commission, NUREG/CR-6507,Washington, 1997, DOI:10.2172/49156010.2172/491560Search in Google Scholar
19 Park, R.-J.; Kang, K.-H.; Kim, J.-T.; Lee, K.-Y.; Kim, S.-B.: Experimental and analytical studies on the penetration integrity of the reactor vessel under external vessel cooling. Nuclear Technology 145 (2004) 102–114, DOI:10.13182/NT04-A346310.13182/NT04-A3463Search in Google Scholar
© 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany
