Piping research overview

The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses.     

Leak-before-break behaviour of austenitic and ferritic pipes containing circumferential defects

In recent years several research projects have been carried out at MPA Stuttgart to investigate the leak-before-break (LBB) behaviour of pressure-bearing components which are relevant to plant safety. In these investigations the test pipes have for the most part been made of ferritic material. International research programmes such as, for example, the Degraded Piping Programme (Wilkowski et al., 1986 and Wilkowski et al., 1989. Degraded Piping Program, Phase II. Report NUREG/CR-4082, vol. 4, Sept. 1986, and vol. 8, March 1989, Battelle, Columbus, Ohio, USA) or the IPIRG-Program (Schmidt et al., 1991. The International Piping Integrity Research Group (IPIRG), Program—An Overview. SMiRT 11 Proceedings, Paper G23/1, Tokyo, Japan, August 1991) have also dealt with pipes made of austenitic materials. However, they were fabricated of not stabilized quality. To take into account the material of comparable components of German nuclear power plants, the experiments reported in the following are focussed on pipes made of Ti- and Nb-stabilized austenitic material. The results presented below relate to pipes containing circumferential defects subjected to internal pressure and external bending loading. As regards the ferritic components an overview of the experimentally determined results is presented. The predictive capability of engineering calculational methods are presented by way of example. The current programme of investigations is presented together with the testing techniques and the initial results.     

Engineering initiatives in nuclear reactor regulation

In this paper relevant engineering initiatives that are currently being developed by the United States Nuclear Regulatory Commission (NRC) to enhance regulatory effectiveness are described. The broad issues addressed are: Piping Design and Non-Destructive Examination, Pressurized Thermal Shock, Containment Integrity During Severe Accidents, and Equipment Qualification.     

Piping dynamic reliability and code rule change recommendations

The conservative nuclear piping design criteria for seismic and dynamic loads have led to piping systems with excessive numbers of snubbers. To improve this undesirable situation, a Piping and Fitting Dynamic Reliability Program was initiated by the Electric Power Research Institute (EPRI) in 1985 with cooperation from the U.S. Nuclear Regulatory Commission (NRC). The objective of the program is to develop improved, realistic, and defensible ASME design rules by taking advantage of the inherent dynamic margins in the nuclear piping system. The research results have demonstrated that piping systems have large reserve dynamic capacity and the dynamic failure mode is due to fatigue or fatigue-ratcheting rather than plastic collapse. Based on such physical evidence, a set of code rule change recommendations is suggested in its preliminary form.     

The IPIRG dynamic pipe loop test facility

The International Piping Integrity Research Group (IPIRG) Program was a group program conducted at Battelle, managed by the U.S. Nuclear Regulatory Commission, and funded by a consortium of organizations from nine nations. A unique pipe loop test facility was designed and constructed for the program to evaluate the behavior of nuclear piping containing flaws and subjected to high rate loading typical of high amplitude seismic events. The facility was carefully designed with rigid anchors and special support bearings to provide well-controlled boundary conditions that can be accurately modelled in numerical analyses. Extensive instrumentation provided pipe system response data and pipe fracture data that serve as a test bed to evaluate various structural and fracture analyses.     

Evaluation of nuclear facility decommissioning projects program-status

Major studies have been undertaken in recent years by the US Nuclear Regulatory Commission (NRC) and others on the technology, safety and costs associated with decommissioning nuclear facilities. The program described in this presentation is being undertaken by the NRC to compile and evaluate the activities of ongoing decommissioning projects. Assessment and evaluation of the methods, impacts, and costs will provide a basis for evaluating licensee's decommissioning proposals and for future decommissioning direction and regulation.Program participants include the US Nuclear Regulatory Commission (NRC) through the Office of Regulatory Research, UNC Nuclear Industries (UNC) through the Decommissioning Programs Department, and nuclear facility licensees.     

Managing aging in nuclear power plants

Aging degradation in nuclear power plants must be controlled to prevent safety margins from declining below limits provided in plant design bases. The NPAR Program and other aging-related programs conducted under the auspices of the NRC Office of Research are developing needed technical guidance for control of aging. Results from these programs, together with relevant information developed by industry and elsewhere, are implemented through various ongoing NRC and industry programs and initiatives as well as by means of conventional regulatory instruments. The aging control process central to these efforts consists of three key elements: (1) selection of components, systems, and structures (CSS) in which aging must be controlled, (2) understanding of the mechanisms and rates of degradation in these CSS, and (3) managing degradation through effective surveillance and maintenance. These elements are addressed in Recommended Practices Guidance that integrates information developed under NPAR and other studies of aging into a systems-oriented format that tracks directly with the Safety Analysis Reports and with the NRC Standard Review Plan (NUREG-0800).     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

Repair materials and techniques for concrete structures in nuclear power plants

The paper summaries portions of work of the Structural Aging Program, sponsored by the Nuclear Regulatory Commission (NRC). The paper addresses the assessment and repair of concrete structures in nuclear power plants. It presents the results of a survey of the the nuclear power plants in the United States to identify susceptible concrete components, rates of occurrence of deterioration, and to determine the durability of repairs. The paper describes deterioration mechanisms and discusses their effect. Repair techniques are described. Evaluation techniques and nondestructive test techniques are also discussed.     

压水堆-回路管道的铸造工艺及其国产化

工艺评定表明,1 000 Mw压水堆核电厂(CPR1000)原选用的主管道铸件Z3CN20-09M(法国牌号)不锈钢的化学成分符合RCC-M采购技术规范,但力学性能并不能完全满足压水堆核岛机械设备设计和建造规范(RCC-M)的要求.本文从金属学角度分析了Z3CN20-09M不锈钢抗蚀性特点和力学性能强化机理,确立了主管道铸件冶炼化学成份的内控标准,使CPR1000核电厂核岛主管道铸件(以下简称主管道铸件)的工艺评定在保持抗蚀性和可焊性特点前提下,各项力学性能指标均满足RCC-M标准,且有较大的裕度,离散度小,质量稳定,综合性能达到领先水平.     

Loose-parts monitoring: Present status of the technology, its implementation in U.S. reactors, and some recommendations for achieving improved performance

At the request of the U.S. Nuclear Regulatory Commission (NRC), an assessment of the technical development status of loose-parts monitoring systems (LPMS) and their performance record to date in commercial light-water-cooled nuclear reactor plants was made during the spring of 1977, using an on-site personal interview and equipment demonstration approach. Our study revealed that while presently demonstrated LPMS technology does indeed provide a capability for detecting the presence of those relatively massive loose parts that would likely constitute a serious operational or safety hazard to the plant, it unfortunately affords little information useful to the determination of the parts' safety significance and has not yet attained the levels of sophistication and reliability ordinarily associated with safety systems. We also found a definite need for specification of the functional requirements for LPMS, in the form of a clear and comprehensive statement of NRC policy regarding the formulation and implementation of safety-oriented, yet operationally practicable, loose-parts monitoring programs for both existing and future nuclear generating stations so that overall objectives of both the utilities and the regulatory agency might be satisfied simultaneously.

While it is our best technical judgment that loose-parts monitoring programs providing reliable detection (but not characterization) capabilities could be implemented with today's technology, the path on which the nuclear utility industry should proceed in order to meet NRC expectations is not completely clear. A Regulatory Guide entitled “Loose Part Detection Program for the Primary System of Light-Water-Cooled Reactors,” soon to be issued for public comment, constitutes a first step towards satisfying this need for guidance and goal establishment.     

A reliability program approach to operational safety

A Reliability Program (RP) model based on proven reliability techniques is being formulated for potential application in the nuclear power industry. Methods employed under NASA and military direction, commercial airline and related FAA programs were surveyed and a review of current nuclear risk-dominant issues conducted. The need for a reliability approach to address dependent system failures, operating and emergency procedures and human performance, and develop a plant-specific performance data base for safety decision making is demonstrated.Current research has concentrated on developing a Reliability Program approach for the operating phase of a nuclear plant's lifecycle. The approach incorporates performance monitoring and evaluation activities with dedicated tasks that integrate these activities with operation, surveillance, and maintenance of the plant. The detection, root-cause evaluation and before-the-fact correction of incipient or actual systems failures as a mechanism for maintaining plant safety is a major objective of the Reliability Program.Embodied within the approach are (1) determination of acceptable safety system performance criteria and associated alert levels: (2) tracing and/or trending of in-plant and industry systems performance and management of the associated surveillance, maintenance, and reportable event data base; (3) determination of risk-importance prioritized systems, components, and root-causes and ad hoc response to inplant safety problems or potentially applicable industry problems identified by NRC or INPO; and (4) a closed-loop failure reporting and corrective action program for correcting performance criteria violations or identified problems either through changes in operation or maintenance or through changes in utility management practices.     

核安全法规和标准中核电厂工况分类的相互对照

IAEA文件No．NS-R-1-2000和我国的HAF102将核电厂工况（状态）划分为正常运行、预计运行事件、设计基准事故、严重事故。美国的RG1．70和我国的EJ／T312将核电厂工况划分为正常运行、中等频率事故、稀有事故、极限事故。本文引述了相关文件给出的各工况（状态）的发生频率，分析并提供了这两种工况（状态）划分方法之间的对照关系。     

Use of PRA and safety goals in nuclear power plant regulation

Probabilistic risk assessments (PRAs) have been performed on a number of nuclear power plants, both by the NRC and industry. The NRC has used risk perspectives gained from PRAs, both in an absolute as well as a relative sense, as an aid in making decisions on plant-specific as well as generic safety issues. However, substantial uncertainties pervade present-day risk assessments, which makes the application of the results of such analyses difficult at best in the regulation of nuclear power. Nonetheless, the Commission approved in January 1983 a policy statement on safety goals for public comment and a two year evaluation period. These safety goals include quantitative design objectives which could serve in the future as risk benchmarks for use by the NRC as part of the decision making process on matters relating to nuclear safety. While the Commission's policy statement explicitly excludes the safety goals from use both in licensing cases and in regulation for the two year evaluation period, PRA will be used generically and on a plant-specific basis more and more to assess the importance of new safety issues, prioritize resources within the agency, and test the adequacy of (or in some instances the need for) NRC's regulations.     

U.S. NRC accident management program

The relation between the NRC and industry accident management programs is outlined. Three research goals are defined, separating the NRC research program into short-term and long-term activities. The process for identification and initial assessment of candidate accident management strategies is presented and strategies selected for detailed evaluation are identified.     

对美国核管会运行经验反馈的研究

研究了美国核管会的运行经验反馈体系和流程,以及与经验反馈工作相关的活动,其中一些成熟做法和经验对我国运行经验反馈工作将起到很好的借鉴作用.     

PRAAGE88: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

The effects of time dependent failure rates caused by the aging of components are becoming increasingly important in probabilistic risk assessment (PRA) and reliability analyses of nuclear power plant systems. In the NRC Nuclear Plant Aging Research (NPAR) program, the effects of aging in nuclear systems are being evaluated through the use of time varying failure rates that are determined as a function of the age of the system. These analyses involve complex systems and include various sensitivity studies; thus, the PRAAGE88 computer code was developed to facilitate these calculations. PRAAGE88 is an IBM PC based code that computes system unavailability, component unavailability, and various importance measures for use in evaluating the effect of aging on reactor systems. This paper describes the methodology utilized in the code, its capabilities and areas of application.     

The three-level scaling approach with application to the Purdue University Multi-Dimensional Integral Test Assembly (PUMA)

The three-level scaling approach was developed for the scientific design of an integral test facility and then it was applied to the design of the scaled facility known as the Purdue University Multi-Dimensional Integral Test Assembly (PUMA). The NRC Technical Program Group for severe accident scaling developed the conceptual framework for this scaling methodology. The present scaling method consists of the integral system scaling, whose components comprise the first two levels, and the phenomenological scaling constitutes the third level of scaling. More specifically, the scaling is considered as follows: (1) the integral response function scaling, (2) control volume and boundary flow scaling, and (3) local phenomena scaling. The first two levels are termed the top-down approach while the third level is the bottom-up approach. This scheme provides a scaling methodology that is practical and yields technically justifiable results. It ensures that both the steady state and dynamic conditions are simulated within each component, and also scales the inter-component mass and energy flows as well as the mass and energy inventories within each component.