A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP

2017 ◽  
Author(s):  
John P. McCloskey ◽  
Richard J. Smith
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Regulatory Commission ◽  
Computer Hardware ◽  
Data Sets ◽  
State University ◽  
Nuclear Operator ◽  
Training Simulator ◽  
Reactor Plant ◽  
Operator Training ◽  
Regulatory Commission

One of the requirements for validating nuclear reactor plant models is to benchmark the predicted results of selected transients against measured plant data or another qualified code. A major challenge with benchmarking is the criteria for validating a model are not always well defined and rely heavily on human judgment, thus introducing the possibility of human bias or inconsistent conclusions. The validation process can also be time consuming. A new method is presented to aid in the validation of nuclear reactor plant models, using the Automated Code Assessment Program (ACAP), which is a tool developed at Pennsylvania State University under contract by the U. S. Nuclear Regulatory Commission (NRC). The proposed method was developed specifically for real-time best-estimate nuclear operator training simulator transients. However, the tool can be easily customized for most applications (e.g., design models, steady state data). Four distinct statistical metrics and weightings were chosen, as deemed appropriate for transient nuclear operator training simulator applications. The metrics account for errors in magnitude and trend, and incorporate an experimental uncertainty. The four metrics are then combined into a single Figure of Merit (i.e., a statistical level of agreement between the predicted and benchmarking data sets). The use of ACAP for benchmarking is demonstrated by comparing experimental data from the Loss-of-Fluid-Test (LOFT) facility Large Break Loss-of-Coolant Experiment L2-5 with code predictions from a RELAP5-3D (Version 2.9.3+) model previously developed and published by Idaho National Laboratories. The proposed method is shown to have several advantages over conventional validation methods, in that it greatly reduces the possibility of human bias, generates reproducible results, can be easily automated to improve efficiency, and can be easily documented. After the initial validation, the tool can also be used to re-validate models after computer hardware changes, model changes, tool upgrades, and operating system upgrades.

Fracture Toughness Properties of SA540 Steels for Nuclear Bolting Applications

1977 ◽  
Vol 99 (3) ◽  
pp. 419-426
Author(s):  
R. R. Seeley ◽  
W. A. Van Der Sluys ◽  
A. L. Lowe
Keyword(s):  
Fracture Toughness ◽  
Impact Energy ◽  
Nuclear Reactor ◽  
Charpy Impact ◽  
Nuclear Regulatory Commission ◽  
Charpy Impact Energy ◽  
Energy Correlation ◽  
Impact Properties ◽  
Minimum Requirements ◽  
Regulatory Commission

Large bolts manufactured from SA540 Grades B23 and B24 are used on nuclear reactor vessels and require certain minimum mechanical properties. A minimum fracture toughness of 125 ksi in. (137 MPa m) at maximum operating stresses is required by the Nuclear Regulatory Commission for these bolts. This minimum toughness property was determined by a stress analysis of a bolt. Minimum required Charpy impact properties were calculated by a fracture toughness-Charpy impact energy correlation and the minimum calculated fracture toughness. The fracture toughness, yield strength and Charpy V notch impact properties were determined for five commercial heats of SA540 steels. Correlations between the fracture toughness and Charpy impact properties of these materials were evaluated. The toughness-impact energy correlation used to set the minimum required Charpy impact properties was found to be unduly conservative, and a different correlation of these properties is suggested. The SA540 steels investigated exhibited fracture toughness properties in excess of the NRC minimum requirements.

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.

scholarly journals Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmark Databases

2013 ◽  
Vol 2013 ◽  
pp. 1-12
Author(s):  
Maria Avramova ◽  
Diana Cuervo
Keyword(s):  
Nuclear Energy ◽  
Nuclear Regulatory Commission ◽  
Nucleate Boiling ◽  
State University ◽  
Energy Agency ◽  
Fine Mesh ◽  
The Us ◽  
Penn State ◽  
Regulatory Commission ◽  
Rod Array

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

Decommissioning the Georgia Tech Research Reactor

2001 ◽  
Author(s):  
Robert Eby ◽  
Lark Lundberg ◽  
Steve Marske ◽  
Nolan Hertel ◽  
Rod Ice
Keyword(s):  
Nuclear Reactor ◽  
Research Reactor ◽  
Cooling System ◽  
Cooling Water ◽  
Nuclear Regulatory Commission ◽  
Georgia Tech ◽  
Spent Fuel ◽  
Survey Report ◽  
Institute Of Technology ◽  
Regulatory Commission

Abstract The Georgia Tech Research Reactor (GTRR) is a 5-megawatt (MW) heavy-water-cooled nuclear reactor located on the Georgia Institute of Technology (Georgia Tech) campus in downtown Atlanta, Georgia. On July 1, 1997, Georgia Tech administration notified the U.S. Nuclear Regulatory Commission (NRC) of their intent to decommission the GTRR. In the summer of 1999, the NRC issued a license amendment to decommission the GTRR in accordance with NRC’s Regulatory Guide 1.86. In the spring of 1999, Georgia Tech and the State of Georgia contracted CH2M HILL to serve as the Executive Engineer to manage the decommissioning project. Later in the summer of 1999, the IT Corporation was selected as the Decommissioning Contractor. The Decommissioning Contractor began the dismantlement process at the Georgia Tech site in November, 1999. By February, 2000, reactor support systems such as the primary and secondary cooling water systems, and the bismuth cooling system were removed and packaged for off-site disposal. Reactor internals were removed in April, 2000. Removal of the bioshield occurred from May through November, 2000. Throughout January, 2001, various concrete structures, including the Spent Fuel Storage Hole, were decontaminated. Dismantlement and decontamination activities were completed by April, 2001. The Final Survey Report to the NRC is planned to be submitted to the NRC December, 2001, 2001. Final license termination by the NRC is anticipated in the spring of 2002.

scholarly journals Risk-Deformed Regulation: What Went Wrong with NFPA 805

2018 ◽  
Vol 4 (12) ◽  
pp. 2876 ◽  
Author(s):  
Raymond HV Gallucci
Keyword(s):  
Risk Reduction ◽  
Fire Protection ◽  
Nuclear Reactor ◽  
National Fire Protection Association ◽  
Nuclear Regulatory Commission ◽  
Official Position ◽  
Nuclear Plants ◽  
Potential Safety ◽  
Time Specific ◽  
Regulatory Commission

National Fire Protection Association (NFPA) Standard 805 was incorporated into Title 10 of the U.S. Code of Federal Regulations to allow commercial nuclear plants to transition their existing, deterministic fire protection licensing bases to ones that are “performance-based and risk-informed.”  Both the US Nuclear Regulatory Commission (NRC) and the commercial reactor industry championed this major leap forward in “risk-informed regulation.”  However, hidden behind all the “success” are compromises and manipulations that were necessary to make this “work,” as revealed in this article.  It is written by a former employee of the NRC (views do not nor ever did represent an official position), the first to receive a PhD on a thesis specifically related to fire probabilistic risk assessment (PRA) in nuclear plants, and later hired in 2003 as the expert in fire PRA for the Office of Nuclear Reactor Regulation (NRR).  He participated in the NFPA-805 program from the start in 2005 until mid-2014.  The perspectives here cover that time period, with some extended time specific to issues that the interested reader can find detailed in “Risk-Deformed Regulation:  What Went Wrong with NFPA 805” http://vixra.org/pdf/  (access latest version of entry 1805.0403).NFPA 805 will have been “successful” in that adopting plants are as safe as or safer than before, at a minimum having at least become more knowledgeable of potential safety weaknesses.  Plants that made effective changes will be safer than before, although “effective” conveys that some changes only may have “seemingly” reduced risk.  If such changes were prompted by questionable risk-reduction credits such as those cited later in this paper, then perhaps actual risk-reduction changes that could have been made were not.  At worst, the plant merely missed an opportunity to become “safer,” a consequence of the problems with “risk-deformed regulation.”

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.

An Innovative Approach to Nuclear Reactor Design Certification

1988 ◽  
Vol 110 (2) ◽  
pp. 161-165
Author(s):  
N. W. Brown ◽  
G. L. Gyorey ◽  
B. K. Genetti ◽  
M. A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Power Reactor ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
General Electric ◽  
Test Program ◽  
New Approach ◽  
The Us ◽  
Regulatory Commission ◽  
Made In

General Electric has proposed that the US Nuclear Regulatory Commission (NRC) consider adding an Appendix to 10CFR50 that would specifically address NRC Safety Review and Design Certification of advanced reactors through use of an experience building test program. The proposal was made in conjunction with the Department of Energy (DOE)-sponsored review of the General Electric advanced Liquid Metal Reactor (LMR) concept, Power Reactor Inherently Safe Module (PRISM). This paper provides a description of the proposed new 10CFR50 Appendix. It also provides the basis for the proposed new approach to Design Certification and outlines the plans that are in place for further review and consideration by the NRC.

scholarly journals Digitizing Native American Collections

2017 ◽  
Vol 45 (1) ◽  
pp. 11
Author(s):  
Nora Stewart
Keyword(s):  
Native American ◽  
Nuclear Regulatory Commission ◽  
State University ◽  
Technology Changes ◽  
Library School ◽  
Regulatory Commission ◽  
The Right

Note from the editor: DttP has been featuring student papers for a number of years at this point—I remember talking about it in my documents class in library school. (For the record, I did not get nominated, which was probably the right choice, even if I did find my paper about the Nuclear Regulatory Commission fascinating.) But as technology changes the work of libraries, so has it changed the products that our students are producing. What follows is an example of a libguide created by a Nora Stewart, student from Emporia State University. Although my presentation of this work is somewhat clumsy, I highly recommend that all of you visit the page (http://libguides.emporia.edu/c.php?g=531631&p=3637209) and look through the resources Nora has collected and placed into context.

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