Pressurized Thermal Shock Rule Research Articles

Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

The radiation embrittlement behavior of reactor pressure vessel shell is influenced by the chemistry concentration of metal materials. This paper aims to study the effects of copper and nickel content variations on the fracture risk of pressurized water reactor (PWR) pressure vessel subjected to pressurized thermal shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel assumed with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the U.S. NRC's new PTS rule are applied as the loading condition. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

DEVELOPMENT OF THE ALTERNATE PRESSURIZED THERMAL SHOCK RULE (10 CFR 50.61a) IN THE UNITED STATES

In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not common, did occur, and because the results of in-reactor materials surveillance programs showed that RPV steels and welds, particularly those having high copper content, experience a loss of toughness with time due to neutron irradiation embrittlement. These recognitions motivated analysis of PTS and the development of toughness limits for safe operation. It is now widely recognized that state of knowledge and data limitations from this time necessitated conservative treatment of several key parameters and models used in the probabilistic calculations that provided the technical of the PTS Rule, 10 CFR 50.61. To remove the unnecessary burden imposed by these conservatisms, and to improve the NRC's efficiency in processing exemption and license exemption requests, the NRC undertook the PTS re-evaluation project. This paper provides a synopsis of the results of that project, and the resulting Alternate PTS rule, 10 CFR 50.61a.

Role of probabilistic analysis in integrity assessments of reactor pressure vessels exposed to pressurized thermal-shock conditions

In recent years, probabilistic analyses have been playing an increasing role in safety assessments for nuclear power plants (NPPs) in the USA. An important example is the treatment of integrity analyses for the primary reactor pressure vessel (RPV) when it is exposed to pressurized thermal-shock (PTS) events. The U.S. Nuclear Regulatory Commission’s (NRC) approach to ensuring PTS integrity over the RPV lifetime has been to establish limits that ensure that the likelihood of through-wall-crack (TWC) development is kept below a prescribed very low level. The key elements for the probabilistic fracture mechanics (PFM) analysis package include thermal-hydraulic analysis input, neutron exposure models, embrittlement models for RPV steels, representations of fracture properties data, flaw characterization models, PFM models, and computational computer codes. The state of knowledge that existed when the NRC initially formulated the PTS rule in the 1980s necessitated that the rule and its technical bases use conservative treatment of parameters and models. Advancements made in the ensuing 20 years have motivated the NRC to undertake a comprehensive update of these technical bases to support potential revisions to the rule. The current status of the key technical elements has now been established, and an updated PFM methodology is being drafted and used for preliminary analyses of three U.S. PWRs. The frequencies of vessel failure predicted by these analyses have been compared with those calculated under the early PTS methodology and are encouraging that a revision to the PTS rule may be forthcoming. Because the PFM calculations utilize a series of deterministic LEFM fracture analyses, it is important to know the degree to which the LEFM methods are applicable to the analysis of thick-wall vessels. U.S. efforts to establish this applicability have been extensive and included a significant number of fracture experiments using large-scale vessels to validate analysis methods and demonstrate margins within design codes. This paper includes a summary of results from several of those key experiments.

Developing a generalized flaw distribution for reactor pressure vessels

The US Nuclear Regulatory Commission (NRC) is re-evaluating the guidance and criteria in the code of federal regulations as it relates to reactor vessel integrity, specifically pressurized thermal shock (PTS). Recent ultrasonic examination of considerable vessel material at Pacific Northwest National Laboratory (PNNL) and industry experiences with Yankee Rowe have provided the NRC with a better understanding of PTS issues. The re-evaluation of PTS will consider a risk-informed approach to the PTS rule and also provide important benefits for licensees considering license renewal. Pressurized thermal shock transients can lead to reactor vessel failure. These transients have occurred at operating reactors but, to date, they have not resulted in vessel failure. To properly determine the potential or probability for vessel failure from a PTS event, an accurate estimate of fabrication flaws is necessary. The characteristics of the fabrication flaw are inputs to fracture mechanics structural calculations that will determine the probability of vessel failure during a PTS event. Also, the results will indicate the sizes and locations of flaws that are most likely to cause failures. This information is also an integral input to the overall pressure vessel safety program. In order to obtain an accurate estimate of fabrication flaws to address PTS events for all classes of reactors, a generic flaw distribution must be developed. An expert judgment process will be used in conjunction with empirical data from PNNL, reactor pressure vessel studies and modeling (RR- PRODIGAL Code) in developing generalized flaw distributions. This paper will demonstrate the important relationship between reactor vessel integrity and flaw distributions in reactor pressure vessel material, discuss the PNNL work to date on developing flaw density and distributions for domestic RPVs, and describe the expert judgment process that was used to verify that a generalized flaw distribution can be properly developed and then assist in developing a generalized flaw distribution.

The effect of flaw orientation on the integrity of PWR pressure vessel at the events of pressurized thermal shock

The reactor pressure vessel (RPV) is the most critical component in nuclear power plants, housing the reactor core and serving as a part of the primary system pressure boundary. Because of its proximity to the reactor core, the RPV is subjected to high fast neutron flux, losing ductility and fracture toughness. At the events of pressurized thermal shock (PTS), highly embrittled RPV may not have a sufficient safety margin for fast fracture. The US NRC PTS rule requires that the reference temperature ( RT PTS) should be limited to ensure sufficient safety margins against PTS. RT PTS=270°F was defined as the screening criterion for axial welds based on extensive quantitative evaluation of associated risks. For circumferential welds, a technical margin of 30°F was added to account for the effects of flaw orientation without same level of quantitative analysis. In this paper, the validity of the technical margin for circumferential welds is examined by comparing the quantitative risks depending on the flaw orientation. First, the result of the original work on axial welds was reproduced. Then, the risk associated with circumferential flaws was evaluated at the identical condition except for flaw orientation. The difference in screening criteria due to flaw orientation was at least 55°F, suggesting that current PTS screening criteria for circumferential flaws do not represent the same level of associated risks as that for axial flaws.

Predicting crack arrest in reactor pressure vessels

Abstract The pressurized thermal shock (PTS) issue has provided increased motivation for the search for a reasonably accurate crack arrest prediction methodology. This issue has assumed greater significance recently as a consequence of the imposition of Regulatory Guide 1.99 Revision 2 procedures for determining the effects of radiation embrittlement in the context of the screening criteria in the PTS rule that is used by the United States Nuclear Regulatory Commission to assess the integrity of reactor pressure vessels. The currently accepted procedure for predicting crack arrest is the so-called K Ia procedure, which is based on static linear elastic fracture mechanics principles, with a crack being presumed to arrest when the crack tip stress intensity factor K I ST falls below a value K Ia . The present paper reviews recent EPRI sponsored research, which shows that the static procedure is overly conservative when it is applied to the first arrest of a deep crack in the thickness of a reactor vessel. This conclusion is clearly important when assessing the consequences of the imposition of the procedures of Regulatory Guide 1.99 Revision 2. A more accurate crack arrest prediction procedure, i.e. the Combustion Engineering constrained static procedure or the reflectionless stress intensity factor procedure which are very similar in concept and their arrest prediction, should be considered to assess the impact of its use in the context of the screening criteria limits in the PTS rule.