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Title: Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

Abstract

The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

Authors:
Publication Date:
Research Org.:
Yucca Mountain Project, Las Vegas, Nevada (US)
Sponsoring Org.:
US Department of Energy (US)
OSTI Identifier:
790801
Report Number(s):
CAL-WIS-TH-000011, Rev. 00
MOL.20011101.0005, DC#28027; TRN: US0301700
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 17 Oct 2001
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 12 MANAGEMENT OF RADIOACTIVE WASTES, AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; BOILING; DESIGN; FUEL ASSEMBLIES; NUCLEAR FUELS; SPENT FUELS; STORAGE; THERMAL ANALYSIS; THERMAL CONDUCTIVITY; WASTES; WATER; YUCCA MOUNTAIN

Citation Formats

Matthew D. Hinds. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity. United States: N. p., 2001. Web. doi:10.2172/790801.
Matthew D. Hinds. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity. United States. doi:10.2172/790801.
Matthew D. Hinds. 2001. "Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity". United States. doi:10.2172/790801. https://www.osti.gov/servlets/purl/790801.
@article{osti_790801,
title = {Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity},
author = {Matthew D. Hinds},
abstractNote = {The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.},
doi = {10.2172/790801},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2001,
month =
}

Technical Report:

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  • This paper summarizes the development of a reliable methodology for the prediction of peak spent nuclear fuel cladding temperature within the waste disposal package. The effective thermal conductivity method replaces other older methodologies.
  • This report describes a modeling technique to determine effective conductivities and comparison with experimental data that were obtained on a full-scale mock-up of a breeder reactor fuel assembly subjected to conditions that occur during normal shipping. The analysis involved the determination of nonlinear effective conductivities that can be used in pure thermal conduction heat transfer codes to simulate the coupled phenomena of conduction, convection and thermal radiative heat transfer in the gas regions between individual elements. The results obtained in this study indicate that the temperature of the fuel pins in a Liquid Metal Fast Breeder Reactor (LMFBR) assembly, undermore » normal conditions of transport within a shipping cask, can be estimated to within +-35/sup 0/F over the temperature range (400 less than or equal to T less than or equal to 650/sup 0/F) with the use of an effective conductivity within a pure conduction heat transfer code. Results also indicated convection is a dominate heat transfer agent over the experimental temperature range, this finding was not shown in previous reports that analyzed experimental data from heat transfer within simulated breeder reactor fuel pin assemblies since: (1) the previous experiments were conducted at higher temperatures, at which radiation heat transfer dominates; (2) experiments were conducted only at atmospheric pressure; and/or (3) detailed analysis was not performed.« less
  • A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded tomore » seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.« less
  • A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing a pressurized water reactor spent fuel assembly having a decay heat level of approximately 1.25 kW. The fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell is located 50 feet from an adjacent drywell. Instrumentation provided to measure canister, liner and soil temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and drywell liner and thermocouples which were attached to plastic pipe and grouted intomore » holes in the soil. Temperatures from the isolated drywell and the adjacent soil were recorded throughout the nine month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (323{sup 0}F and 262{sup 0}F, respectively) during October, 1980. Thereafter, all temperatures began to decrease in response to the decay heat and seasonal atmospheric temperature changes. This thermal response is comparable to that of the approximately 1.0 kW spent fuel assemblies previously tested at E-MAD where peak canister and liner temperatures of 254{sup 0}F and 203{sup 0}F were recorded. A previously developed computer model was utilized to predict the thermal response of the surrounding soil are presented and are compared with the test data.« less