Package Impact Models as a Precursor to Cladding Analysis

2010 ◽  
Author(s):  
Nicholas A. Klymyshyn ◽  
Harold E. Adkins ◽  
Christopher S. Bajwa ◽  
Jason M. Piotter
Keyword(s):  
Impact Analysis ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Fuel Cladding ◽  
Structural Dynamic ◽  
Fuel Pin ◽  
Regulatory Commission ◽  
Impact Models

The evaluation of spent nuclear fuel storage casks and transportation packages under impact loading is an important safety topic that is reviewed as part of cask and package certification by the United States Nuclear Regulatory Commission. Explicit dynamic finite element models of full systems are increasingly common in industry for determining structural integrity during hypothetical drop accidents. Full cask and package model results are also used as the loading basis for single fuel pin impact models, which evaluate the response of fuel cladding under drop conditions. In this paper, a simplified package system is evaluated to illustrate several important structural dynamic phenomena, including the effect of gaps between components, the difference in local response at various points on a package during impact, and the effect of modeling various simplified representations of the basket and fuel assemblies. This paper focuses on the package impact analysis, and how loading conditions for a subsequent fuel assembly or fuel cladding analysis can be extracted.

Package Impact Models as a Precursor to Cladding Analysis1

2012 ◽  
Vol 135 (1) ◽  
Author(s):  
Nicholas A. Klymyshyn ◽  
Harold E. Adkins ◽  
Christopher S. Bajwa ◽  
Jason M. Piotter
Keyword(s):  
Fuel Assembly ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Fuel Cladding ◽  
Fuel Pin ◽  
Regulatory Commission ◽  
Impact Models

The evaluation of spent nuclear fuel storage casks and transportation packages under impact loading is an important part of cask and package certification by the United States Nuclear Regulatory Commission. Finite element models are increasingly used for evaluating cask and package structural integrity during hypothetical drop accidents. Full cask and package model results are also used as the loading basis for single fuel pin impact models, which evaluate the response of fuel cladding under drop conditions. In this paper, a simplified package system is evaluated to illustrate the difference between local and bulk impact responses, the effect of simplified basket and fuel assembly representations, and the effect of gaps between components. This paper focuses on the package impact analysis and how loading conditions for a subsequent fuel assembly or fuel cladding analysis can be extracted. The results of this study suggest that detailed package system models are needed to determine cladding deceleration load histories.

The MacArthur Maze Fire and Roadway Collapse: A “Worst Case Scenario” for Spent Nuclear Fuel Transportation?

2012 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Harold Adkins ◽  
Judith Cuta ◽  
Nicholas Klymyshyn ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Steel Structure ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Severe Accident ◽  
Case Scenario ◽  
Worst Case ◽  
Work Related ◽  
Regulatory Commission

In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the “MacArthur Maze.” The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC’s analysis of the consequences of this type of severe accident.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

The Inclusion of Inner Surface Breaking Flaws in Probabilistic Fracture Mechanics Analyses of Reactor Vessels Subjected to Planned Normal Cool-Down Transients

2008 ◽  
Author(s):  
Terry Dickson ◽  
Mark EricksonKirk
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Regulatory Commission ◽  
Technical Basis

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses / publications, consistent with the assumptions utilized for this particular reactor in the PTS re-evaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify / generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.

Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Stephen A. Hambric ◽  
Samir Ziada ◽  
Richard J. Morante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Cycle Fatigue ◽  
Structural Integrity ◽  
Fatigue Cracking ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Boiling Water ◽  
Cycle Fatigue ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

Potential Effects of Recent Road Transportation Accidents on Radioactive Material Shipments

2008 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Darrell S. Dunn ◽  
Robert E. Shewmaker
Keyword(s):  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Radioactive Material ◽  
Road Transportation ◽  
Concrete Walls ◽  
The Us ◽  
A Chain ◽  
Safe Transport ◽  
Regulatory Commission

In 2007, two severe transportation accidents, involving primarily long-haul tractor trailers, occurred in the State of California. In the first, which occurred in Oakland in the “MacArthur Maze” section of Interstate 580, a tractor trailer carrying gasoline impacted an overpass support column and burst into flames. The subsequent fire, which burned for over 2 hours, led to the collapse of the overpass onto the remains of the tractor trailer, due to the loss of strength in the steel exposed to the fire. The second incident was a chain-reaction accident involving several tractor trailers in the I-5 “Newhall Pass” truck bypass tunnel in Santa Clarita. This accident also involved an intense fire that damaged the tunnel and required the closing of the tunnel for repairs to the concrete walls. The US Nuclear Regulatory Commission is studying both these accidents to examine any potential regulatory implications related to the safe transport of radioactive materials and spent nuclear fuel in the United States. This paper will provide a summary of that effort.

The Potential Impacts of Recent Transportation Accidents Involving Severe Fires on the Safe Shipment of Spent Nuclear Fuel

2010 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Concrete Walls ◽  
The Us ◽  
Food Commodities ◽  
A Chain ◽  
Safe Transport ◽  
Regulatory Commission

In 2007, two severe transportation accidents occurred in the state of California. The first occurred in Oakland on a section of Interstate 880 known as the “MacArthur Maze” and involved a tractor trailer carrying gasoline which impacted an overpass support column and burst into flames. The subsequent fire caused the collapse of a portion of the Interstate 580 overpass onto the remains of the tractor trailer in less than 20 minutes, due to a reduction of strength in the structural steel exposed to the fire. The second incident was a chain-reaction accident involving over thirty tractor trailers in the Interstate 5 “Newhall Pass” truck bypass tunnel in Santa Clarita. This accident also involved an intense fire, fueled mostly by produce and other food commodities, that damaged the concrete walls of the tunnel and required the tunnel to be closed for repairs. The US Nuclear Regulatory Commission (NRC) is in the process of studying both of these accidents to examine any potential regulatory implications related to the safe transport of spent nuclear fuel in the United States. This paper will summarize work recently completed on these severe transportation accidents by the NRC.

Risk-Based Fracture Evaluation of Reactor Vessels Subjected to Cool-Down Transients Associated With Shutdown: An Examination of the Effects of Different Modeling Approaches on Estimated Failure Probabilities

2006 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Asme Code ◽  
Computational Methodology ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Evaluation Project

The current regulations, as set forth by the United States Nuclear Regulatory Commission (USNRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. In 1999, the USNRC initiated the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project to determine if a technical basis could be established to support a relaxation in the current PTS regulations. The PTS re-evaluation project included the development and application of an updated risk-based computational methodology that incorporates several advancements applicable to modeling the physics of vessel fracture due to thermal hydraulic transients imposed on the RPV inner surface. The results of the PTS re-evaluation project demonstrated that there is a sound technical basis to support a relaxation of the current PTS regulations. The results of the PTS re-evaluation are currently under review by the USNRC. Based on the promising results of the PTS re-evaluation, the USNRC has recently applied the updated computational methodology to fracture evaluations of RPVs subjected to planned cool-down transients, associated with reactor shutdown, derived in accordance with ASME Section XI – Appendix G. The objective of these analyses is to determine if a sound technical basis can be established to provide a relaxation to the current regulations for the derivation of bounding cool-down transients as specified in Appendix G to Section XI of the ASME Code. This paper provides a brief overview of these analyses, results, and the implications of the results.

Historical Rail Accident Analyses Identifying Accident Parameters That Could Impact Transportation of Spent Nuclear Fuel

2010 ◽  
Author(s):  
Todd S. Mintz ◽  
George Adams ◽  
Marius Necsoiu ◽  
James Mancillas ◽  
Chris Bajwa ◽  
...  
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hazardous Materials ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Hazardous Material ◽  
Freight Train ◽  
Severe Fire ◽  
Regulatory Commission

As the regulatory authority for transportation of spent nuclear fuel (SNF) in the United States, the Nuclear Regulatory Commission (NRC) requires that SNF transportation packages be designed to endure a fully engulfing fire with an average temperature of 800 °C (1,475 °F) for 30 minutes, as prescribed in Title 10 of the Code of Federal Regulations (CFR) Part 71. The work described in this paper was performed to support NRC in determining the types of accident parameters that could produce a severe fire with the potential to fully engulf a SNF transportation package. This paper describes the process that was used to characterize the important features of rail accidents that would potentially lead to a spent nuclear fuel transport package being involved in a severe fire. Historical rail accidents involving hazardous material and long duration fires in the United States have been analyzed using data from the Federal Railroad Administration (FRA) and the Pipeline and Hazardous Materials Safety Administration (PHMSA). Parameters that were evaluated from this data include, but were not limited to, class of track where the accident occurred, class of hazardous material that was being transported, and number of railcars involved in the fire. The data analysis revealed that in the past 34 years of rail transport, roughly 1,800 accidents have led to the release of hazardous materials resulting in a frequency of roughly 1 accident per 10 million freight train miles. In the last 12 years, there have only been 20 accidents involving multiple car hazardous material releases that led to a fire. This results in an accident rate of 0.003 accidents per million freight train miles that involved multiple car releases and a fire. In all the accidents analyzed, only one involved a railcar carrying Class 7 (i.e., radioactive) hazardous material (HAZMAT).

Potential Effects of Historic Rail Accidents on the Integrity of Spent Nuclear Fuel Transportation Packages

2009 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hazardous Materials ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Radioactive Material ◽  
Material Transport ◽  
Long Duration ◽  
The Us ◽  
Regulatory Commission

The US Nuclear Regulatory Commission (NRC) completed an analysis of historical rail accidents (from 1975 to 2005) involving hazardous materials and long duration fires in the United States. The analysis was initiated to determine what types of accidents had occurred and what impact those types of accidents could have on the rail transport of spent nuclear fuel. The NRC found that almost 21 billion miles of freight rail shipments over a 30 year period had resulted in a small number of accidents involving the release of hazardous materials, eight of which involved long duration fires. All eight of the accidents analyzed resulted in fires that were less severe than the “fully engulfing fire” described as a hypothetical accident condition in the NRC regulations for radioactive material transport found in Title 10 of the Code of Federal Regulations, Part 71, Section 73. None of the eight accidents involved a release of radioactive material. This paper describes the eight accidents in detail and examines the potential effects on spent nuclear fuel transportation packages exposed to the fires that resulted from these accidents.

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