Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

Williams, Brian G.¹  Schultz, Richard R.²  McEligot, Don M.³  McCreery, Glenn⁴ 

[1] Idaho State Univ., Pocatello, ID (United States);Idaho National Lab. (INL), Idaho Falls, ID (United States);Univ. of Idaho, Moscow, ID (United States);Battelle Energy Alliance, LLC, Idaho Falls, ID (United States)

关键词: HEAT TRANSFER;    HTGR TYPE REACTORS;    REACTOR CORES;    PHYSICAL RADIATION EFFECTS;    DEFORMATION;    HELIUM;    COOLANTS;    RHR SYSTEMS;    BYPASSES;   

DOI  :  10.2172/1214656
RP-ID  :  DOE/NEUP--10--876-Final-Report-Appendix-F
PID  :  OSTI ID: 1214656
Others  :  Other: 10-876 Final Report and Appendix F
Others  :  TRN: US1600277

美国|英语

来源: SciTech Connect

【 摘 要 】

A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.