ASSESSMENT OF ASSERT SUBCHANNEL CODE FOR THE CANADIAN SCWR BUNDLE AT SUPERCRITICAL CONDITIONS

摘要：

The ASSERT-PV subchannel code has been applied to predict maximum cladding temperatures inthe Canadian Super-Critical Water-Cooled Reactor (SCWR) fuel bundle. ASSERT-PV subchannelcode was developed and qualified for predictions of flow and enthalpy distributions, critical heatflux, and post-dryout cladding temperatures for CANDU1 bundles. Recently, the code has beenupdated with supercritical fluid properties and supercritical heat transfer correlations, and verifiedagainst predictions of another subchannel code. Maximum cladding temperatures are determinedfor the reference bundle design with axial and radial power profiles corresponding to beginning andend of cycles. The power profiles are also varied to assess their effects on maximum claddingtemperatures.

作者: E.N.Onder Y.F.Rao J.Pencer

作者单位: Atomic Energy of Canada Limited,Chalk River Laboratories,Chalk River,Ontario,Canada

会议类型: 国际会议

会议名称: 3rd China-Canada Joint Workshop on Supercritical Water-Cooled Reactors(第三届中国-加拿大超临界水堆研讨会 (CCSC-2012)

会议地点: 西安

会议语种:英文

页码: 1-11

在线出版日期: 2012-04-18（万方平台首次上网日期，不代表论文的发表时间）