Adoption of TS-R-1 in the United States Nuclear Regulatory Commission Regulations for Type-B and Fissile Material

2003 ◽  
Vol 14 (1) ◽  
pp. 7-10
Author(s):  
D. Pstrak ◽  
P. Brochman ◽  
J. Cook ◽  
R. Lewis ◽  
R. Temps
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson

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.


Author(s):  
John Minichiello ◽  
Ernest B. Branch ◽  
Timothy M. Adams ◽  
Yasuhide Asada ◽  
Richard W. Barnes

The new rules for seismic piping design in Section III that were developed and included in the requirements in 1994 Addenda of the ASME Boiler and Pressure Vessel Code (B&PV Code) generated considerable discussion within the industry and from the United States Nuclear Regulatory Commission, (USNRC). The USNRC initiated a review of the results of the previous EPRI/NRC experimental program and the Japanese industry started its own experimental program. To accommodate and address developments resulting from these efforts, the ASME, B&PV Code established a Special Working Group (SWG) to continue the review and study of the questions and information generated. This paper reports on the efforts of this SWG which resulted in refinements of the revised rules. These refinements have been accepted for inclusion in Section III of the ASME, B&PV Code.


Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.


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):  
Ronald S. Hafner ◽  
Gerald C. Mok ◽  
Lisle G. Hagler

The U.S. Nuclear Regulatory Commission (USNRC) contracted with the Packaging Review Group (PRG) at Lawrence Livermore National Laboratory (LLNL) to conduct a single, 30-ft shallow-angle drop test on the Combustion Engineering ABB-2901 drum-type shipping package. The purpose of the test was to determine if bolted-ring drum closures could fail during shallow-angle drops. The PRG at LLNL planned the test, and Defense Technologies Engineering Division (DTED) personnel from LLNL’s Site-300 Test Group executed the plan. The test was conducted in November 2001 using the drop-tower facility at LLNL’s Site 300. Two representatives from Westinghouse Electric Company in Columbia, South Carolina (WEC-SC); two USNRC staff members; and three PRG members from LLNL witnessed the preliminary test runs and the final test. The single test clearly demonstrated the vulnerability of the bolted-ring drum closure to shallow-angle drops—the test package’s drum closure was easily and totally separated from the drum package. The results of the preliminary test runs and the 30-ft shallow-angle drop test offer valuable qualitative understandings of the shallow-angle impact. • A drum package with a bolted-ring closure may be vulnerable to closure failure by the shallow-angle drop, even if results of the steep-angle drop demonstrate that the package is resistant to similar damage. • Although there exist other mechanisms, the shallow-angle drop produces closure failure mainly by buckling the drum lid and separating the drum lid and body, which the bolted ring cannot prevent. • Since the closure failure by the shallow-angle drop is generated mainly by structural instabilities of a highly discontinuous joint, the phenomenon can be rather unpredictable. Thus, a larger-than-normal margin of safety is recommended for the design of such packages. • The structural integrity of the bolted-ring drum closure design depends on a number of factors. To ensure that the drum closure survives the shallow-angle drop, the following general qualitative rules should be observed: – The drum closure components should be quality products made of ductile materials, and the torque value for tightening the bolted ring should be included in the SAR and operating procedures to ensure quality. – The package should not be too heavy. – The package internal structure should be impact-absorbent and resistant to disintegration and collapse under high compressive load. However, a strong internal structure may defeat the purpose of protecting the containment vessel from damage during a free drop. • If not previously tested, drum packages with bolted-ring drum closures should be drop-tested at shallow angles. Due to the unpredictable nature of the behavior, the demonstration should be completed by test and on a case-by-case basis. The test plan should take into account the behavior’s sensitivity to the details of the package design and the impact condition. • Because the shallow-angle drop can open the drum closure, organizations using these types of drum packages should assess the consequences of exposing the radioactive contents in the containment vessel to unconsidered external elements or conditions. This work was supported by the United States Nuclear Regulatory Commission under a Memorandum of Understanding with the United States Department of Energy, and performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract W-7405-Eng-48.


Author(s):  
Barry J. Elliot ◽  
Jerry Dozier

Generic Aging Lessons Learned (GALL) report, License Renewal Standard Review Plan (SRP-LR), and regulatory guide were issued by the United States Nuclear Regulatory Commission (NRC) in June 2001. The intent of these documents was to provide the technical and process basis that will lead to a more effective, efficient and predictable license renewal process for industry and the NRC. The GALL report provides the aging effects on components and structures, identifies the relevant existing plant programs, and evaluates the program attributes to manage aging effects for License Renewal. The GALL report also identifies when existing plant programs would require further evaluation for License Renewal. The SRP-LR allows the applicant to reference the GALL report to demonstrate that the programs at the applicant’s facility correspond to those reviewed and approved in the GALL report. Programs that correspond to those in the GALL report will not need further detailed review by the staff. Implementation of the aging management program are verified as part of the license renewal inspection program. The GALL report identifies one acceptable way of demonstrating that components and structures have adequate aging management programs. However, applicants may propose alternatives to the programs identified in GALL. During the license renewal review, the NRC primarily focuses on areas where existing programs should be augmented or new programs developed for License Renewal. This paper will provide an overview of these documents and some of the lessons learned during a demonstration project in the application of the new guidance. This topic will be of interest to the U.S. participants considering License Renewal and desiring to know state-of-the-art information about License Renewal in the United States.


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