Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

Williford, R.E.; Mohr, C.L.; Lanning, D.D.; Cunningham, M.E.; Rausch, W.; Bradley, E.R.

doi:10.2172/5456977

Cited by 10 publications

(8 citation statements)

References 8 publications

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“…A rod displaying soft PCMI in operation also displays pellet deformability in rod radial compression tests. The occurrence of "soft" PCMI is consistent with the results of several experiments in the Halden Reactor, including NRC-sponsored tests, in which fuel temperature and rod elongation were simultaneously measured as a function of burnup and LHGR (Williford, 1980 …”

Section: 13supporting

confidence: 86%

Predictive Bias and Sensitivity in NRC Fuel Performance Codes

Geelhood¹,

Luscher²,

Senor³

et al. 2009

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Section: 13supporting

confidence: 86%

Predictive Bias and Sensitivity in NRC Fuel Performance Codes

Geelhood¹,

Luscher²,

Senor³

et al. 2009

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“…A complete discussion of the basic concepts used in fuel relocation and mechanical response modeling have appeared in previous NRC-RSR quarterly reports. In addition, a formal report is now being published (Williford et al 1980). …”

Section: Future Plansmentioning

confidence: 99%

Reactor Safety Research Programs

Dotson¹

1980

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“…A formal report on the basic concepts used in fuel relocation and mechanical response modeling is now being prepared for publication (Williford et al 1980). …”

Section: Fuel Relocation Data Analysismentioning

confidence: 99%

Reactor Safety Research Programs Quarterly Report January - March 1980

Hagen¹

1980

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This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.; i i CONTENTS

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Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

Cited by 10 publications

References 8 publications

Predictive Bias and Sensitivity in NRC Fuel Performance Codes

Predictive Bias and Sensitivity in NRC Fuel Performance Codes

Reactor Safety Research Programs

Reactor Safety Research Programs Quarterly Report January - March 1980

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