Potential Ways to Modify the NRC Safety Goal Policy

Author(s):  
Steven A. Arndt ◽  
Richard Denning

There have been significant discussions over the past few years by the U.S. Nuclear Regulatory Commission (NRC) staff and the Advisory Committee on Reactor Safeguards (ACRS), as to the adequacy of the safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the U.S. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have also been used as surrogates for the QHOs. This study will examine a potential approach to update the safety goals that includes the establishment of new qualitative goals associated with the comparative risk of generating electricity by viable competing technologies, and the development of preliminary tests in support of a new qualitative goal.

Author(s):  
Paul H. Genoa

Over the past few years, the U.S. nuclear power industry has gained substantial experience and appreciation of the technical complexity and rigor required to meet a performance-based site clean-up standard. Five large power reactors and several smaller ones are now well along the path to license termination. They have not been on this journey alone. There has been a steep learning curve for all stakeholders involved in the process including state and federal radiation regulators, legislators, and the public. We have all learned that the translation of results from a post remediation survey interpreted through pathway modeling for comparison with a dose-based clean-up standard is for many a leap of faith. Our regulator has an understandable desire to address this uncertainty by demanding conservative analysis at each turn. As a result, it is extremely demanding to demonstrate that a clean-up standard in the 0.15–0.25 mSv/a range has been met. It is not likely that a standard in the 10 μSv/a level, typically associated with radiological clearance standards, can be practically demonstrated while still meeting the current expectations of U.S. Nuclear Regulatory Commission for technical rigor.


Author(s):  
Ronald C. Lippy

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.


1980 ◽  
Vol 24 (1) ◽  
pp. 123-123
Author(s):  
Linda O. Hecht

Due to the concern for safety the nuclear power industry in the United States has fostered the use of reliability analysis to assess system performance and the impact of system failure on overall plant safety. The need for system and component failure rate data has been recognized and has spurred such efforts as NPRDS (Nuclear Power Research Data System) and IEEE's Std 500 (The Reliability Data Manual). Reliability modeling techniques have been developed for application to nuclear systems and are presently being considered by the Nuclear Regulatory Commission for licensing purposes.


Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.


Author(s):  
David Alley

This paper provides a historical perspective on the need for, and development of, buried and underground piping tanks programs at nuclear power plants. Nuclear power plant license renewal activities, Nuclear Regulatory Commission Buried Piping Action Plan, and the rationale for addressing the issue of buried pipe through an industry initiative as opposed to regulation are discussed. The paper also addresses current NRC activities including the results of Nuclear Regulatory Commission inspections of buried piping programs at nuclear power plants as well as Nuclear Regulatory Commission involvement in industry and standards development organizations. Finally, the paper outlines the Nuclear Regulatory Commission’s future plans concerning the issue of buried piping at US nuclear power plants.


2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Stephen A. Hambric ◽  
Samir Ziada ◽  
Richard J. Morante

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).


Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.


Author(s):  
Jerry McNeish ◽  
Peter Swift ◽  
Rob Howard ◽  
David Sevougian ◽  
Donald Kalinich ◽  
...  

The development of a deep geologic repository system in the United States has progressed to the preparation of an application for a license from the U.S. Nuclear Regulatory Commission. The project received site recommendation approval from the U.S. President in early 2002. The next phase of the project involves development of the license application (LA) utilizing the vast body of information accumulated in study of the site at Yucca Mountain, Nevada. Development of the license application involves analyses of the total system performance assessment (TSPA) of the repository, the TSPA-LA. The TSPA includes the available relevant information and model analyses from the various components of the system (e.g., unsaturated geologic zone, engineered system (waste packaging and drift design), and saturated geologic zone) (see Fig. 1 for nominal condition components), and unites that information into a single computer model used for evaluating the potential future performance or degradation of the repository system. The primary regulatory guidance for the repository system is found in 10 CFR 63, which indicates the acceptable risk to future populations from the repository system. The performance analysis must be traceable and transparent, with a defensible basis. The TSPA-LA is being developed utilizing state-of-the-art modeling software and visualization techniques, building on a decade of experience with such analyses. The documentation of the model and the analyses will be developed with transparency and traceability concepts to provide an integrated package for reviewers. The analysis relies on 1000’s of pages of supporting information, and multiple software and process model analyses. The computational environment represents the significant advances in the last 10 years in computer workstations. The overall approach will provide a thorough, transparent compliance analysis for consideration by the U.S. Nuclear Regulatory Commission in evaluating the Yucca Mountain repository.


Author(s):  
William Greenman ◽  
Kimberly Cole

Abstract In the United States, mixed-waste is typically defined as waste that contains both radioactive constituents and non-radioactive constituents that pose a threat to human health or the environment (hazardous waste). Prior to 1986 the U.S. Nuclear Regulatory Commission (NRC) had sole regulatory authority over mixed-waste because of its radioactive constituents. In 1986, however, the U.S. Environmental Protections Agency (EPA) was granted regulatory authority over the hazardous constituents in mixed-waste; and, a system of dual regulation was created. Dual regulation of mixed-waste by the EPA and the NRC has caused significant problems for the regulated community. The burden of dual regulation has contributed to the slow development of treatment technologies, and to the overall lack of treatment capacity available to U.S generators of mixed-waste. This paper reviews the requirements that the EPA and the NRC mandate with regard to mixed-waste generation, treatment and disposal; and it explores technical impacts of those requirements as they relate to generators, treatment facilities and the public.


Author(s):  
Joseph S. Miller

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.


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