Preliminary Results of the United States Nuclear Regulatory Commission's Pressurized Thermal Shock Rule Reevaluation Project

2008 ◽  
pp. 167-167-16
Author(s):  
TL Dickson ◽  
PT Williams ◽  
BR Bass ◽  
MT Kirk
Keyword(s):  
United States ◽  
Thermal Shock ◽  
The United States ◽  
Preliminary Results ◽  
Pressurized Thermal Shock

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.

Derivation of the New Pressurized Thermal Shock Screening Criteria

2011 ◽  
Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou
Keyword(s):  
United States ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Technical Basis

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.

The Impact of an Improved Flaw Model on a Pressurized Thermal Shock Evaluation

2002 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Computational Models ◽  
The United States ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Evaluation Program ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
The Impact ◽  
Flaw Characterization

The current regulations for pressurized thermal shock (PTS) were derived from computational models that were developed in the early-mid 1980s. The computational models utilized in the 1980s conservatively postulated that all fabrication flaws in reactor pressure vessels (RPVs) were inner-surface breaking flaws. It was recognized at that time that flaw-related data had the greatest level of uncertainty of the inputs required for the probabilistic-based PTS evaluations. To reduce this uncertainty, the United States Nuclear Regulatory Commission (USNRC) has in the past few years supported research at Pacific Northwest National Laboratory (PNNL) to perform extensive nondestructive and destructive examination of actual RPV materials. Such measurements have been used to characterize the number, size, and location of flaws in various types of welds and the base metal used to fabricate RPVs. The USNRC initiated a comprehensive project in 1999 to re-evaluate the current PTS regulations. The objective of the PTS Re-evaluation program has been to incorporate advancements and refinements in relevant technologies (associated with the physics of PTS events) that have been developed since the current regulations were derived. There have been significant improvements in the computational models for thermal hydraulics, probabilistic risk assessment (PRA), human reliability analysis (HRA), materials embrittlement effects on fracture toughness, and fracture mechanics methodology. However, the single largest advancement has been the development of a technical basis for the characterization of fabrication-induced flaws. The USNRC PTS-Revaluation program is ongoing and is expected to be completed in 2002. As part of the PTS Re-evaluation program, the updated risk-informed computational methodology as implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, including the improved PNNL flaw characterization, was recently applied to a domestic commercial pressurized water reactor (PWR). The objective of this paper is to apply the same updated computational methodology to the same PWR, except utilizing the 1980s flaw model, to isolate the impact of the improved PNNL flaw characterization on the PTS analysis results. For this particular PWR, the improved PNNL flaw characterization significantly reduced the frequency of RPV failure, i.e., by between one and two orders of magnitude.

Preliminary Results from the Air Mass Transformation Experiment (AMTEX)

1976 ◽  
Vol 57 (11) ◽  
pp. 1346-1355 ◽  
Author(s):  
D. H. Lenschow ◽  
E. M. Agee
Keyword(s):  
United States ◽  
Experimental Period ◽  
The United States ◽  
Published Data ◽  
Data Set ◽  
Air Mass ◽  
Preliminary Results ◽  
Cold Air Outbreak ◽  
Transformation Experiment ◽  
Transformation Processes

The field phases of AMTEX, a GARP subprogram on air-sea interaction implemented by Japan, were conducted over the East China Sea in the environs of Okinawa, Japan, during the last two weeks of February in 1974 and 1975. Investigators from Australia, Canada, and the United States also participated in this experiment. The weather was generally very favorable for this study of air mass transformation processes in 1975 because of an extensive cold air outbreak during most of the experimental period. A basic synoptic data set was obtained from 6 h soundings from an array of aerological stations centered at Okinawa. In addition, satellite, hourly surface and surface marine, oceanographic, boundary layer, radiation, radar, cloud physics, and aircraft data were obtained and have been or will be available in published data reports or on magnetic tape. Preliminary results from 1974 and 1975 reported at the Fourth AMTEX Study Conference and joint United States–Japan Cooperative Science Program Seminar, “Air Mass Transformation Processes over the Kuroshio in Winter,” held in Tokyo, 26–30 September 1975, are presented and discussed.

The Sensitivity of Pressurized Thermal Shock Analysis Results to Alternative Models for Weld Flaw Distributions

2003 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
Evaluation Study ◽  
The United States ◽  
Metal Plate ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Pressurized Thermal Shock ◽  
Technical Basis ◽  
The Impact

The United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project in 1999 to determine if improved technologies can provide a technical basis to reduce the conservatism in the current regulations for pressurized thermal shock (PTS) 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 renewal considerations. During the PTS re-evaluation study, an improved risk-informed computational methodology was developed that provides a more realistic characterization of PTS risk. This updated methodology was recently applied to three commercial PWRs. The results of this study provide encouragement that a technical basis can be established to support a relaxation of current PTS regulations. One significant model improvement applied in the PTS re-evaluation study was the development of flaw databases derived from the non-destructive and destructive examinations of material from cancelled reactor pressure vessels (RPV). Empirically-based statistical distributions derived from these databases and expert illicitation were used to postulate the number, size, and location of flaws in welded and base metal (plate and forging) regions of an RPV during probabilistic fracture mechanics (PFM) analyses of RPVs subjected to transient loading conditions such as PTS. However, limitations in the available flaw data have required assumptions to be made to complete the risk-based flaw models. Sensitivity analyses were performed to evaluate the impact of four flaw-related assumptions. Analyses addressed: 1) truncations of distributions to exclude flaws of extreme depth dimensions, 2) vessel-to-vessel differences in flaw data, 3) large flaws observed in weld repair regions, and 4) the basis for estimating the number of surface breaking flaws. None of the four alternate weld flaw models significantly impacted calculated vessel failure frequencies or invalidated the tentative conclusions derived from the USNRC PTS re-evaluation study.

Nuclear Medicine Exposure in the United States, 2005-2007: Preliminary Results

2008 ◽  
Vol 38 (5) ◽  
pp. 384-391 ◽  
Author(s):  
Fred A. Mettler ◽  
Mythreyi Bhargavan ◽  
Bruce R. Thomadsen ◽  
Debbie B. Gilley ◽  
Jill A. Lipoti ◽  
...  
Keyword(s):  
United States ◽  
Nuclear Medicine ◽  
The United States ◽  
Preliminary Results

Intracrustal complexity in the United States midcontinent: Preliminary results from COCORP surveys in northeastern Kansas

Geology ◽  
1983 ◽  
Vol 11 (1) ◽  
pp. 25 ◽  
Author(s):  
L. Brown ◽  
L. Serpa ◽  
T. Setzer ◽  
J. Oliver ◽  
S. Kaufman ◽  
...  
Keyword(s):  
United States ◽  
The United States ◽  
Preliminary Results

Problems involved in the cultivation of maize for fodder and ensilage. I. The choice of variety

1959 ◽  
Vol 52 (1) ◽  
pp. 95-105 ◽  
Author(s):  
E. S. Bunting ◽  
L. A. Willey
Keyword(s):  
United States ◽  
Western Europe ◽  
The United States ◽  
Preliminary Results ◽  
The Past ◽  
Maize Cultivation ◽  
Green Fodder ◽  
Western Germany

Since the introduction of silos and associated techniques of ensilage production in the late 1870's, maize has been the principal silage crop in the United States. Many British authorities agree that maize is a ‘splendid silage crop, highly nutritious, heavy yielding and easy to cut and handle’ (Bond, 1948, see also Watson & Smith, 1956; Woodman & Amos, 1944) but, nevertheless, it is very rarely grown in Britain for silage. In this country, as in most countries of Western Europe, the predominant aim of maize cultivation has been to produce succulent green fodder for direct feeding to animals during time of drought and consequent grass shortage. The acreage grown in England is small, but in Western Germany in 1955 there were about 100,000 acres of fodder maize (Becker, 1956), while in France, in 1954, the acreage exceeded 500,000 (Desroches, 1955). Recently in Western Europe considerable interest has been shown in the possibilities of maize as a silage crop, and within the past few years preliminary results have been reported from Holland (Becker, 1956; Anon. 1954, 1955); Denmark (Bagge & Hansen, 1956); Belgium (Lacroix, 1955; Ledent, 1955); Germany (Jungehulsing, 1955; Schell, 1954); France (Desroches, 1955), and Switzerland (Bachmann, 1952).

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