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.

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.

Benchmarking of a Thermal Finite Element Approximation Scheme for Externally Cooled Spent Fuel Storage Casks

2007 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Ian F. Spivack
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Finite Element Approximation ◽  
Nuclear Regulatory Commission ◽  
Element Approximation ◽  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Regulatory Commission

The US Nuclear Regulatory Commission (NRC) is responsible for licensing spent fuel storage casks under Title 10 of the Code of Federal Regulations Part 72 (10 CFR Part 72). Under these regulations, storage casks must be evaluated to verify that they meet various criteria, including acceptable thermal performance requirements. The purpose of the evaluation described in this paper is to establish the effectiveness of a medium-effort modeling approach and associated simplifying assumptions in closely approximating spent fuel cask component temperature distributions. This predictive evaluation is performed with the ANSYS® code, and is applicable to externally cooled cask designs. The results are compared against experimental measurements and predictions of the COBRA-SFS finite-difference code developed at Pacific Northwest National Laboratory.

Preparing for the Extended Loss of AC Power (ELAP) Event in the USA

2014 ◽  
Author(s):  
Joseph S. Miller
Keyword(s):  
Nuclear Power ◽  
Task Force ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Spent Fuel ◽  
Ac Power ◽  
The Us ◽  
External Events ◽  
Near Term ◽  
Regulatory Commission

The United States utilities started preparing for external events that could lead to a loss of all ac power in the 1980’s, when the Station Blackout (SBO) rulemaking was first introduced by the United States Nuclear Regulatory Commission (USNRC). Following the events at the Fukushima Dai-ichi nuclear power plant on March 11, 2011, the USNRC established a senior-level agency task force referred to as the Near-Term Task Force (NTTF). The NTTF was tasked with conducting a systematic, methodical review of Nuclear Regulatory Commission (NRC) regulations and processes to determine if the agency should make additional improvements to these programs in light of the events at Fukushima Dai-ichi. As a result of this review, the NTTF developed a comprehensive set of recommendations, documented in SECY-11-0093, “Near-Term Report and Recommendations for Agency Actions Following the Events in Japan,” dated July 12, 2011. Documentation of the staff’s efforts is contained in SECY-11-0124, “Recommended Actions to be Taken without Delay from the Near-Term Task Force Report,” dated September 9, 2011, and SECY-11-0137, “Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated October 3, 2011. To satisfy some of the NRC’s recommendations, the industry described its proposal for a Diverse and Flexible Mitigation Capability (FLEX), as documented in Nuclear Energy Institute’s (NEI) letter, dated December 16, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11353A008). FLEX was proposed as a strategy to fulfill the key safety functions of core cooling, containment integrity, and spent fuel cooling. The events at Fukushima Dai-ichi highlight the possibility that extreme natural phenomena could challenge the prevention, mitigation and emergency preparedness defense-in-depth layers. At Fukushima, limitations in time and unpredictable conditions associated with the accident significantly challenged attempts by the responders to preclude core damage and containment failure. During the events in Fukushima, the challenges faced by the operators were beyond any faced previously at a commercial nuclear reactor. NRC Order 12-049 (Ref. 1) and NRC Interim Staff Guidance JLD-ISG-2012-01 (Ref. 6) provided additional requirements to mitigate beyond-design-basis external events. These additional requirements impose guidance and strategies to be available if the loss of power, motive force and normal access to the ultimate heat sink to prevent fuel damage in the reactor and spent fuel pool affected all units at a site simultaneously. The NEI submitted document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation Guide” in August 2012 (ADAMS Accession No. ML12242A378) to provide specifications for the nuclear power industry in the development, implementation, and maintenance of guidance and strategies in response to NRC Order EA-12-049. The US utilities are currently proposing modifications to their plants that will follow specifications provided in NEI 12-06. This paper presents some of the NEI 12-06 requirements and some of the proposed modifications proposed by the US utilities.

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.

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.”

scholarly journals Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology

2008 ◽  
Vol 2008 ◽  
pp. 1-13 ◽  
Author(s):  
Robert P. Martin ◽  
Larry D. O'Dell
Keyword(s):  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Lessons Learned ◽  
Technical Program ◽  
Methodology Development ◽  
Analysis Methodology ◽  
Loss Of Coolant Accident ◽  
Core Cooling ◽  
Regulatory Commission

The AREVA NP Inc. realistic large-break loss-of-coolant-accident (LOCA) analysis methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. This methodology conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the United States Nuclear Regulatory Commission in the late 1980s. In addition, several practical considerations were revealed with the move to a production application. This paper describes the methodology development within the CSAU framework and utility objectives, lessons learned, and insight about current LOCA issues.

Seismic Considerations for the Transition Break Size: An Overview of the Methodology

2006 ◽  
Author(s):  
G. Wilkowski ◽  
H. Xu ◽  
P. Krishnaswamy ◽  
N. Chokshi ◽  
S. Shaukat ◽  
...  
Keyword(s):  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Federal Regulations ◽  
Regulatory Requirements ◽  
Technical Requirements ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Primary Focus ◽  
Core Cooling ◽  
Regulatory Commission

This paper describes the results of a study, sponsored by the U.S. Nuclear Regulatory Commission (NRC), to assess potential seismic effects on the postulated transition break size (TBS) in the proposed risk-informed revision of the regulatory requirements for the emergency core cooling system (ECCS) contained in Title 10, Section 50.46, of the Code of Federal Regulations (10 CFR 50.46). The full report was provided on the NRC web site in mid-December 2005. The primary focus of this paper is to provide a summary of the study’s approach and results which was conducted to facilitate review and comment concerning the proposed rule and statement of considerations (SOC), entitled “Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements; Proposed Rule,” which the NRC published in the Federal Register (70 FR 67598) on November 7, 2005.

Overview of Current Practice for Analysis and Design of Independent Spent Fuel Storage Installation Pads

2019 ◽  
Author(s):  
Gunup Kwon ◽  
Khaled Ata
Keyword(s):  
Nuclear Power ◽  
Seismic Analysis ◽  
Storage System ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Loading Conditions ◽  
Spent Fuel ◽  
Analysis And Design ◽  
License Application ◽  
Regulatory Commission

Abstract Nuclear power plant spent fuels are initially stored in the spent fuel pool. Then, the water cooled fuels are transferred in a concrete or steel cask and transported outside of the Fuel Handling Building (FHB) or the Reactor Building (RB) for long term on site storage. The spent fuel casks are typically stored on a slab-on-grade pad. The slab-on-grade pad is designed according to the U.S. Nuclear Regulatory Commission NUREG-1536 and NUREG-1567. The two Standard Review Plans provide guidance to the regulators for the review of cask storage system license application. The ISFSI pad analysis and design have to consider various loading conditions, such as earthquake and tornado loadings as well as normal operating loading conditions. Seismic analysis of the ISFSI pad requires considering interaction between the pad and the supporting soil. Various cask loading configurations on the pad also have to be considered. Due to the lack of specific guidelines, many ISFSI pad designs show overly conservative reinforcement. This study provides guidelines and procedure for the design of the ISFSI pad that are typically used in the nuclear industry. It is considered that the guidelines and practices described in this study help design engineers understand general guidance provided in the NRC Standard Review Plans.

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