University of Illinois Urbana-Champaign

Risk of misloading spent nuclear fuel cask for light water reactors

Jarrah, Ibrahim

Content Files
JARRAH-THESIS-2018.pdf

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https://hdl.handle.net/2142/102866

Description

Title
Risk of misloading spent nuclear fuel cask for light water reactors
Author(s)
Jarrah, Ibrahim
Issue Date
2018-12-14
Director of Research (if dissertation) or Advisor (if thesis)
Uddin, Rizwan
Committee Member(s)
Mohaghegh, Zahra
Department of Study
Nuclear, Plasma, & Rad Engr
Discipline
Nuclear, Plasma, Radiolgc Engr
Degree Granting Institution
University of Illinois at Urbana-Champaign
Degree Name
M.S.
Degree Level
Thesis
Date of Ingest
2019-02-07T20:44:31Z
Keyword(s)
  • Spent fuel cask
  • misload
  • PRA
  • Risk
Abstract
The spent fuel dry cask should remain subcritical under normal, abnormal, and accident conditions. The cask may become susceptible to criticality if it is misloaded with assemblies that do not conform with the Certificate of Compliance (CoC). Assessment of probability of criticality for a misloaded cask that subsequently experiences an accident during the transportation process is also of interest. To avoid misloading, the cask loading process involves several verification steps to make sure that all of the loaded assemblies satisfy the CoC requirements. However, most of the loading and verification steps are carried out by humans with finite probabilities for errors, which need to be quantified. Quantification of the risk of having a misloaded cask may reduce the conservatism in the cask designs and eliminate unnecessary steps in the spent fuel handling and loading procedure. In the first part of this study, the probability of misloading a cask with at least one light water reactor, pressurized water reactor (PWR) and boiling water reactor (BWR), fuel assembly is quantified first using the event tree method. An event tree and associated fault trees are developed for the cask loading procedures. Probability distribution functions (PDFs) for all of the human errors are obtained using the Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) human reliability analysis method. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software is used to quantify the event tree and to calculate the probability of misload. The probability of misloading a cask with at least one fuel assembly is first determined for PWR fuel and it was found to be 5.56E-06, which agrees well with that reported in the literature. The probability of misloading a cask with at least one fuel assembly for the BWR fuel is found to be 2.95E-05. The impact of the cask capacity on the probability of misload is quantified and discussed. The Fussell-Vesely (FV) importance measure is performed to determine the tasks that contribute the most to having a misloaded cask. The effects of the available time to perform a task and the stress level of the operator on the final probability of misload are studied. The available time and stress are found to have a significant impact on the final misload probability. Based on the neutronic calculations, the cask needs to be misloaded with more than one fuel assemblies in order to become susceptible to criticality. In the second part of this study, an event tree is built to predict the multiple misloads scenarios. Six multiple misloads scenarios are identified from the tree. The probabilities of the six scenarios and the total probability are calculated for casks for both reactor types. The probabilities calculated using this method are found to be 6.73E-07 and 7.55E-06 for PWR and BWR fuels, respectively. In addition, the probability of multiple misloads is calculated as a function of the cask capacity.
Graduation Semester
2018-12
Type of Resource
text
Permalink
http://hdl.handle.net/2142/102866
Copyright and License Information
Copyright 2018 Ibrahim Jarrah

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Dissertations and Theses - Nuclear, Plasma, and Radiological Engineering