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Title: Thermal analysis on passive dry spent fuel storage for ABTR.


No abstract prepared.

; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE Office of Science (SC)
OSTI Identifier:
Report Number(s):
TRN: US1002751
DOE Contract Number:
Resource Type:
Resource Relation:
Conference: 2007 ANS Annual Meeting; Jun. 24, 2007 - Jun. 28, 2007; Boston, MA
Country of Publication:
United States

Citation Formats

Chikazawa, Y., Grandy, C., and Nuclear Engineering Division. Thermal analysis on passive dry spent fuel storage for ABTR.. United States: N. p., 2007. Web.
Chikazawa, Y., Grandy, C., & Nuclear Engineering Division. Thermal analysis on passive dry spent fuel storage for ABTR.. United States.
Chikazawa, Y., Grandy, C., and Nuclear Engineering Division. Mon . "Thermal analysis on passive dry spent fuel storage for ABTR.". United States. doi:.
title = {Thermal analysis on passive dry spent fuel storage for ABTR.},
author = {Chikazawa, Y. and Grandy, C. and Nuclear Engineering Division},
abstractNote = {No abstract prepared.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}

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  • Safe interim dry storage of spent nuclear fuel (SNF) must be maintained for a minimum of twenty years according to the Code of Federal Regulations. The most important variable that must be regulated by dry storage licensees in order to meet current safety standards is the temperature of the SNF. The two currently accepted models to define the maximum allowable initial storage temperature for SNF are based on the diffusion controlled cavity growth (DCCG) failure mechanism proposed by Raj and Ashby. These models may not give conservative temperature limits. Some have suggested using a strain-based failure model to predict themore » maximum allowable temperatures, but we have shown that this is not applicable to the SNF as long as DCCG is the assumed failure mechanism. Although the two accepted models are based on the same fundamental failure theory (DCCG), the researchers who developed the models made different assumptions, including selection of some of the most critical variables in the DCCG failure equation. These inconsistencies are discussed together with recommended modifications to the failure models based on more recent data.« less
  • As most of the nuclear power plants, on-site spent fuel pools (SFP) of Taiwan's plants were not originally designed with a storage capacity for all the spent fuel generated over the operating life by their reactors. For interim spent fuel storage, dry casks are one of the most reliable measures to on-site store over-filled assemblies from SFPs. The NUHOMS{sup R}-52B System consisting of a canister stored horizontally in a concrete module is selected for thermal evaluation in this paper. The performance of each cask in criticality, radioactive, material and thermal needs to be carefully addressed to ensure its enduring safety.more » Regarding the thermal features of dry storage casks, three different kinds of heat transfer mechanisms are involved, which include natural convection heat transfer outside and/or inside the canister, radiation heat transfer inside and outside the canister, and conduction heat transfer inside the canister. To analyze the thermal performance of dry storage casks, RELAP5-3D is adopted to calculate the natural air convection and radiation heat transfer outside the canister to the ambient environment, and ANSYS is applied to calculate the internal conduction and radiation heat transfer. During coupling iteration between codes, the heat energy across the canister wall needs to be conserved, and the inner wall temperature of the canister needs to be converged. By the coupling of RELAP5-3D and ANSYS, the temperature distribution within each fuel assembly inside canisters can be calculated and the peaking cladding temperature can be identified. (authors)« less
  • An important objective for a spent-fuel transportation and storage cask is to remove decay heat from the fuel array sufficiently to prevent clad degradation by maintaining it below a specified design limit. The purpose of this paper is to evaluate conduction and radiation models by the COBRA-SFS code through benchmarking by the Massachusetts Institute of Technology effective conductivity model. A bundle-lumping method is also presented that makes it feasible to numerically simulate large fuel bundles economically while achieving adequate accuracy.
  • A simplified two step thermal analysis method has been develop to evaluate (1) the mean temperature of the CANDU fuel bundles within a fuel basket in a given spent fuel dry storage canister by HEATING5 code with additional input data and heat transfer correlations in the step-1 analysis and (2) the maximum fuel rod temperature within a CANDU 37-element fuel bundle by MAXROT code developed here for step-2 analysis. In addition, the results of sample analysis is performed to examine the parametric effects of the site-specific ambient conditions on the maximum fuel temperature within a canister are presented. The comparisonmore » between the results of step-1 analysis and the mock-up test, in particular, is quite satisfactory. In essence, the two-step thermal analysis method proposed here is a code package that used the HEATING5 and MAXROT codes, respectively, for step-1 and step-2 calculations in series.« less