The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis

In September 1988, the United States Nuclear Regulatory Commission issued a revised emergency core cooling system rule for light water reactors that allows, as an option, the use of best estimate plus uncertainty methods in safety analysis. To support the 1988 licensing revision, the United States Nuclear Regulatory Commission and its contractors developed the code scaling, applicability and uncertainty evaluation methodology to demonstrate the feasibility of the best estimate plus uncertainty approach. The phenomena identification and ranking table (PIRT) process, Step 3 in the code scaling, applicability and uncertainty methodology, was originally formulated to support the best estimate plus uncertainty licensing option. Through further development and application, the PIRT process has shown additional utility as a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. The generic PIRT process, including typical and common illustrations from prior applications that promoted further development of the process, are described. Analysis of the results of the prior applications is also described. The analysis results provide information that can help guide future applications of the process in a graded approach based on phenomena relative importance.     

AREVA's realistic large break LOCA analysis methodology

The AREVA realistic large break loss-of-coolant-accident (LOCA) analysis methodology received approval by the United States Nuclear Regulatory Commission (USNRC) in April 2003. The methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. In addition, it conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the USNRC in the late 1980s; however, it also incorporates the nonparametric statistical approach originally incorporated in the Gesellschaft fur Anlagen und Reaktorsicherheit (GRS) methodology for LOCA analysis. This paper provides a methodology summary, lessons learned and insight about current LOCA issues.     

Reduced seismic loads for temporary conditions

A new methodology, developed under an EPRI Tailored Collaboration Project, to calculate and apply reduced seismic loads (RLSs) for evaluation of temporary conditions (TCs) in nuclear power plants using design-basis (DB) allowables is described. The methodology, which was submitted to Nuclear Regulatory Commission (NRC) through the Nuclear Energy Institute (NEI), calculates load reduction factors based on an allowed limit for time-averaged increase in seismic core damage frequency within the duration of a refueling cycle. For this allowable in the range 5×10⁻⁶ to 1×10⁻⁵ per reactor year, substantial reduction relative to DB seismic load is possible. The methodology is equally applicable to plants with and without seismic probabilistic risk analysis model.     

剂量计算工作组的生物动力学和剂量学模型

剂量计算工作组负责国际放射防护委员会第２委员会的剂量系数的计算。该工作组最近已系统地描述了提供了一种年龄依赖的剂量学方法。本文对该方法学作一评述，并对系统描述中的某些具体细节作讨论，在最后结论中提到了工作组今后可能开展的一些活动。     

Risk-informed extension of the inservice inspection interval for pressurized water reactor vessel from 10 to 20 years

The Pressurized Water Reactor Owners Group (PWROG) methodology for extending the inservice inspection interval for reactor vessel welds in pressurized water reactors is described in technical report, WCAP-16168-NP, Revision 1. This report presents a risk-informed basis for extending the interval between inspections from the current interval of 10-20 years. This paper discusses the background of the PWROG inservice inspection interval extension methodology and the results of pilot plant studies on typical Westinghouse, Combustion Engineering, and Babcock and Wilcox plant designs. These pilot plant studies showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. This paper also discusses the application of the extended interval to other pressurized water reactors and provides examples of the application.     

Study of a methodology of identifying important research problems by the PIRT process

In this paper, we propose a new methodology of identifying important research problems to be solved to improve the performance of some specific scientific technologies by the phenomena identification and ranking table (PIRT) process which has been used as a methodology for demonstrating the validity of the best estimate simulation codes in US Nuclear Regulatory Commission (USNRC) licensing of nuclear power plants. The new methodology makes it possible to identify important factors affecting the performance of the technologies from the viewpoint of the figure of merit and problems associated with them while it keeps the fundamental concepts of the original PIRT process. Also in this paper, we demonstrate the effectiveness of the new methodology by applying it to a task of extracting research problems for improving an inspection accuracy of ultrasonic testing or eddy current testing in the inspection of objects having cracks due to fatigue or stress corrosion cracking.     

Intersystem LOCA risk assessment: methodology and results

The United States Nuclear Regulatory Commission is sponsoring a research program to develop an improved understanding of the human factors, hardware and accident consequence issues that dominate the risk from an intersystem loss-of-coolant accident (ISLOCA) at a nuclear power plant. To accomplish the goals of this program, a mehtodology has been developed for estimating ISLOCA core damage frequency and risk. The steps in this methodology are briefly described, along with the results obtained from an application of the methodology at three pressurized water reactors. Also included are the results of a screening study of boiling water reactors.     

Methods for assessing NPP containment pressure boundary integrity

Research is being conducted to address aging of the containment pressure boundary in light-water reactor plants. Objectives of this research are to (1) understand the significant factors relating to corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and of liners of concrete containments; (2) provide the U.S. Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation. Activities include development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of candidate techniques for inspection of inaccessible regions of containment metallic pressure boundaries; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion.     

Application of NASA Kennedy Space Center system assurance analysis methodology to nuclear power plant systems designs

In May of 1982, the Kennedy Space Center (KSC) entered into an agreement with the Nuclear Regulatory Commission (NRC) to conduct a study to demonstrate the feasibility and practicality of applying the KSC System Assurance Analysis (SAA) methodology to nuclear power plant systems designs. North Carolina's Duke Power Company expressed an interest in the study and proposed the nuclear power facility at CATAWBA for the basis of the study. In joint meetings of KSC and Duke Power personnel, an agreement was made to select two CATAWBA systems, the Containment Spray System and the Residual Heat Removal System, for the analyses. Duke Power provided KSC with a full set of Final Safety Analysis Reports (FSAR) as well as schematics for the two systems. During Phase I of the study the reliability analyses of the SAA were performed. During Phase II the hazard analyses were performed. The final product of Phase II is a handbook for implementing the SAA methodology into nuclear power plant systems designs. The purpose of this paper is to describe the SAA methodology as it applies to nuclear power plant systems designs and to discuss the feasibility of its application.     

Probability-based load combinations for design of Category I structures — Overview of research program and recent results

This paper discusses the probability-based load combinations for the program dealing with the design of Category I structures, currently being worked on at Brookhaven National Laboratory (BNL) for the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission (NRC). The objective of this program is to develop a probabilistic approach for the safety evaluations of reactor containments and other seismic Category I structures subjected to multiple static and dynamic loadings. Furthermore, on the basis of the developed probabilistic approach, a load combination methodology for the design of seismic Category I structures will also be established.The major tasks of this program are: (1) establish probabilistic representations for various loads and structural resistance, (2) select appropriate structural analysis methods and identify limit states of structures, (3) develop a reliability analysis method applicable to nuclear structures, (4) apply the developed methodology to existing Category I structures in order to evaluate the reliability levels implied in the current design criteria, and (5) recommend load combination design criteria for Category I structures. When the program is completed, it will be possible to (1) provide a method that can evaluate the safety margins of existing containment and other Category I structures and (2) recommend probability-based load combinations and load factors for the design of Category I structures.At the present time, a reliability analysis method for seismic Category I concrete structures has been completed. By utilizing this method, it is possible to evaluate the safety of structures under various static and dynamic loads. In this paper, results of a reliability analysis of a realistic reinforced concrete containment structure under dead load, accidental pressure, and earthquake ground acceleration are presented to demonstrate the feasibility of the methodology.     

Estimating internal dose coefficients of short-lived radionuclides in accordance with ICRP 2007 recommendations

For radiation protection in high-energy accelerator facilities, internal dose coefficients of short-lived radionuclides were estimated using the dosimetric methodology in accordance with the International Commission on Radiological Protection (ICRP) 2007 Recommendations. A computational program was developed for estimating the dose coefficients. The program was verified by confirming whether it could reproduce the dose coefficients provided by ICRP for intakes of representative radionuclides. In addition, the estimated dose coefficients of short-lived radionuclides were compared to the values generated by Dose and risk CALculation software (DCAL), which is based on a dosimetric methodology that is in accordance with the ICRP 1990 Recommendations, to discuss the reasons why the dose coefficients were changed by the revision of the dosimetric methodologies. The comparison revealed a decreasing trend of dose coefficients in the case of inhalation upon revision of the dosimetric methodologies. By contrast, in the case of ingestion, the dose coefficients tended to increase.     

Quantifying reactor safety margins part 4: Uncertainty evaluation of lbloca analysis based on trac-pf1/mod 1

The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. The rule also requires an estimation of the uncertainty in the calculated system response and a comparison of the resulting bound with the acceptance limits of 10CFR Part 50. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling, Applicability and Uncertainty Methodology (CSAU). The last of the three elements of the methodology - Uncertainty Evaluation - is described in this paper.     

介绍德国辐射防护委员会对R.Clarke先生辐射防护新建议的立场

本文介绍现任国际放射防护委员会（ICRP）主席R.Clarke博士提出的辐射防护新建议的重要背景和主要内容，以及德国辐射防护委员会（SSK）对这个建议所持的立场。     

Letter to the editor on “Evaluation of subcooled critical heat flux correlations for tubes with and without internal twisted tapes” by F. Inasaka and H. Nariai and “Consideration of CHF margin prediction by subcooled or low quality CHF correlations” by P. Hejzlar and N.E. Todreas

Some comments are presented concerning CHF correlations which have the property MCHFR = MCPR and on the limitations of our knowledge of how to apply such models to three dimensional situations and rapid transients.     

Probabilistic assessment of the seismic hazard for Eastern United States nuclear power plants

The purpose of the seismic hazard characterization of the Eastern United States project, for the Nuclear Regulatory Commission, was to develop a methodology and data bases to estimate the seismic hazard at all the plant sites east of the Rocky Mountains. A summary of important conclusions reached in this multi year study is presented. The magnitude and role of the uncertainty in the hazard estimates is emphasized in regard to the intended final use of the results.     

核设施退役中几个值得重视的问题

本文阐述了核设施退役中一些重视的问题，包括：退役的级别问题；退役必须有周密的计划；搞好源项调查；优选去污工艺；采用安全可靠的切割解体技术，必须高度重视废物最少化；高度重视安全问题；重视人员培训和安全文化素养的提高；安排好废物的最终出路；相关标准和导则的编制。     

Assessing dose of the representative person for the purpose of radiation protection of the public. ICRP publication 101. Approved by the Commission in September 2005

The Commission intended that its revised recommendations should be based on a simple, but widely applicable, system of protection that would clarify its objectives and provide a basis for the more formal systems needed by operating managers and regulators. The recommendations would establish quantified constraints, or limits, on individual dose from specified sources. These dose constraints apply to actual or representative people who encounter occupational, medical, and public exposures. This report updates the previous guidance for estimating dose to the public. Dose to the public cannot be measured directly and, in some cases, it cannot be measured at all. Therefore, for the purpose of protection of the public, it is necessary to characterise an individual, either hypothetical or specific, whose dose can be used for determining compliance with the relevant dose constraint. This individual is defined as the 'representative person'. The Commission's goal of protection of the public is achieved if the relevant dose constraint for this individual for a single source is met and radiological protection is optimised. This report explains the process of estimating annual dose and recognises that a number of different methods are available for this purpose. These methods range from deterministic calculations to more complex probabilistic techniques. In addition, a mixture of these techniques may be applied. In selecting characteristics of the representative person, three important concepts should be borne in mind: reasonableness, sustainability, and homogeneity. Each concept is explained and examples are provided to illustrate their roles. Doses to the public are prospective (may occur in the future) or retrospective (occurred in the past). Prospective doses are for hypothetical individuals who may or may not exist in the future, while retrospective doses are generally calculated for specific individuals. The Commission recognises that the level of detail afforded by its provision of dose coefficients for six age categories is not necessary in making prospective assessments of dose, given the inherent uncertainties usually associated with estimating dose to the public and with identification of the representative person. It now recommends the use of three age categories for estimating annual dose to the representative person for prospective assessments. These categories are 0-5 years (infant), 6-15 years (child), and 16-70 years (adult). For practical implementation of this recommendation, dose coefficients and habit data for a 1-year-old infant, a 10-year-old child, and an adult should be used to represent the three age categories. In a probabilistic assessment of dose, whether from a planned facility or an existing situation, the Commission recommends that the representative person should be defined such that the probability is less than about 5% that a person drawn at random from the population will receive a greater dose. If such an assessment indicates that a few tens of people or more could receive doses above the relevant constraint, the characteristics of these people need to be explored. If, following further analysis, it is shown that doses to a few tens of people are indeed likely to exceed the relevant dose constraint, actions to modify the exposure should be considered. The Commission recognises the role that stakeholders can play in identifying characteristics of the representative person. Involvement of stakeholders can significantly improve the quality, understanding, and acceptability of the characteristics of the representative person and the resulting estimated dose.     

Testing to determine relay seismic ruggedness

The seismic qualification of equipment in operating nuclear plants has been identified as a potential safety concern in U.S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment in Operating Nuclear Power Plants”. In response to this concern, the Seismic Qualification Utility Group (SQUG), with support from the Electric Power Research Institute (EPRI), has undertaken a program to demonstrate the seismic adequacy of essential equipment by the use of actual experience with such equipment in plants which have undergone significant earthquakes and by the use of available test data for similar equipment. An important part of this program is the development of the methodology and test data for verifying the functionality of electrical relays used in essential circuits needed for plant shutdown during a seismic event. This paper describes the EPRI supported relay testing program to supplement existing relay test data. Many old relays which are used in safe shutdown systems of SQUG plants and for which seismic test data do not exist have been shake-table tested. The testing performed on these relays and the test results for two groups of relays are summarized in this paper.