Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences.The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA.A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach.This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR,which is the representative case for high pressure sequences.MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident.In addition,a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components.This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.     

基于分离变量的地震PSA相关性分析方法研究

为解决现有地震概率安全评价（PSA）相关性分析简化假设存在的问题,建立更准确反映核电厂构筑物、系统和部件（SSC）地震相关性的分析方法,对基于分离变量的易损度相关性分析开展了研究。结合易损度模型对分析方法进行了理论推导,并对方法的实施过程进行了介绍。利用该方法对不同条件下SSC的联合失效开展案例分析,得到了联合失效的易损度曲线和失效频率分析结果,并与现有相关性简化假设得到的结果进行了对比。研究结果表明,基于分离变量的地震PSA相关性分析方法能弥补现有方法的不足,支持核电厂地震PSA开发和应用。     

地震概率安全分析中地震易损度不确定性分布的研究

核电厂地震概率安全分析(PSA)中,构筑物和设备的地震易损度是在给定地面运动强度条件下的条件失效概率。地震易损度的不确定性分布较为复杂,在地震PSA定量化过程中难于处理。本文针对地震易损度的数学模型进行研究,采用数值方法求解地震易损度的均值和方差。在均值和方差相等的条件下,以几种常见的不确定性分布类型近似地震易损度的不确定性分布。通过比较可以看出,Beta分布可以较为准确地描述地震易损度的不确定性分布。     

核电厂地震PSA中的风险定量化研究

论述了核电厂地震概率安全评价（PSA）定量化方法和工具的现状,指出了定量化工具面临的挑战和存在的问题。根据定量化的概率论本质,提出了计算方法。以我国某核电厂厂址多方案概率地震危险性分析（PSHA）结果和核电厂地震响应分析给出的最小割集为例,展示了计算方法的应用过程,分析了地震动参数和置信度参数对定量化计算结果的影响。结果表明,针对置信度参数进行拉丁超立方采样,采样次数较小时即可给出地震导致的核电厂堆芯损坏频率（SCDF）的稳定估计值;通常情况下,设备失效对SCDF的贡献最大,厂房失效的影响相对较小;地震动年发生率对SCDF的贡献需要根据工程场地的位置进行具体分析。      

Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A “analysis of flaw indication” for the application to a PWR primary piping. Results of the analysis are discussed in detail.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.     

基于蒙特卡罗抽样和最大-最小法的地震易损度算法研究

本文探索并研究了一种新的地震易损度算法,基于蒙特卡罗（MC）抽样和最大-最小法计算了单个设备和多个设备组合的最小割集的易损度。最小割集包括3种类型：纯地震失效最小割集、包含非事件的最小割集、地震失效和随机失效混合割集。对于仅包含地震失效的事故序列,可直接采用基于蒙特卡罗抽样和最大 最小法的易损度算法进行计算。涉及地震失效和随机失效混合的事故序列,可采用极限近似方法（MCUB）或其他割集定量化算法进行计算。经对比,基于蒙特卡罗抽样和最大 最小法的地震易损度算法计算结果与理论值一致,为工程应用中的地震易损度计算提供了另一种可行的算法。     

Comments on the probability of leakage in piping systems as used in PRA's

In risk analysis of power reactors the leakage or failure of piping structures has to be taken into account as a possible cause of loss of coolant.As part of the SMiRT post conference seminar on “PRA of NPP for External Events” the present practice of selecting pipe failure rates as initiating or related events for PRA's has been discussed.For pipe failures as initiating events an approach has been developed in the framework of the risk study for German PWR's. As compared to NUREG-1150 some significant differences are identified.For external events the effect of seismic induced loads on pipe failure has been a subject of considerable efforts in research. Several studies have demonstrated that for moderate siting conditions the effect of seismic induced loads on pipe failure rates of large diameter high pressure piping does not lead to significant contributions to the overall risk.The main subject for future research on pipe failure mechanisms is the detailed assessment of the influence of the water chemistry conditions.     

A risk approach to the management of boiler tube thinning

A large set of industrial thickness inspection data covering four boiler units of a power station over a period of five years was made available to the authors. The measurements were made in regions of the boiler where corrosion/erosion was the major cause of failure of the boiler tubes. There were over 40,000 separately measured data points in the data and all were collected with some care and expense.In the development of maintenance strategies for equipment, this type of data is typical of the data that must be collected and assessed. This data thus represents an opportunity to evaluate the ability to generate a useful risk approach to the management of the tubing.An important example of a risk-based approach is the American Petroleum Institute (API) Risk Based Inspection (“RBI”), API 581. A variety of problems were encountered applying this to boiler tubes. The problems include irrelevant API 581 corrosion rate tables, lack of information on how to analyse inspection data, difficulty of dealing with multiple inspection categories and lack of suitable direction for programming inspection intervals.     

Neutron noise measurements on pressurized water reactors

Noise analysis contributes to increase significantly understanding of safety and monitoring of PWR. The difficulties of a correct interpretation of noise signal in a power reactor encourage a deeper insight into the theoretical model. Following, this paper is dealing with 4 topics:

• - theoretical model

• - measurements made in PWR's, and evolution of power spectral densities,

• - experimental test of models,

• - future study in this range : methodology for early detection of failure and to indicate incipient failure.

Using the expressions for linear system and for feedback loop, we obtain a simplified diagram for PWR with reactivity, temperature, velocity of fluid inputs and movement of internal structures.

Neutron noise measurements are performed periodically on the “Centrale Nucléaire des Ardennes”. An investigation is also performed, in order to detect core barrel movements.

The change of neutron noise at several power levels is shown, and used to check the model.

The development of signal analysis and models for PWR is investigated in the last chapter.     

Quantification of the reliability required of non-destructive inspection of PWR pressure vessels

This paper presents a method of quantifying the reliability required of non-destructive inspections of PWR pressure vessels. It gives a strategy for improving the effectiveness of ultrasonic non-destructive testing in assuring the integrity of a PWR vessel and allows targets of inspection reliability to be set in order to achieve the requisite level of vessel integrity. To do this the failure rate of PWR pressure vessels is predicted on the basis of a probabilistic fracture mechanics model. We use various models of the reliability of non-destructive inspection to discover the minimum level of reliability which is consistent with the desired integrity of the structure and to demonstrate how improvements can be made most effective.The reliability of inspection is usually modelled by a function giving the probability of leaving an unacceptable defect in the vessel. This function B(a) is really the “unreliability” of inspection and so 1 - B(a) gives the usual reliability. A reliable inspection is one which detects and correctly classifies defects according to some criterion usually based on size. A reliable inspection must use a technique which is intrinsically capable of detecting and sizing defects in the required size range and it must be reliably applied in practice.We find that, based on certain stated assumptions, that an inspection reliability of 80% of detecting and correctly sizing a defect of 15 mm through-wall extent yields a predicted failure rate of 10⁻⁷ per vessel year. The failure rate includes a frequency of a major accident such as a large loss of coolant (LOCA) of frequency 10⁻⁴ per vessel year. The predicted failure rate can be reduced to 10⁻⁸ per vessel year if the sizing accuracy of the technique is improved so that the chance of undersizing a 15 mm defect falls from 0.19 to about 0.01. However, the failure rate of the vessel is not predicted to decrease further with any subsequent improvement in sizing accuracy unless there is also an improvement in the asymptote of the reliability of inspection. This asymptote is due to factors beyond the capability of the technique such as, for example, human error.     

Development and application of filters for air cleaning in nuclear power plants

The development of filter systems for air cleaning in nuclear power plants will be briefly described. The result of research work on iodine filters was the basis for the use of gasketless deep-bed filters and of the multiway sorption filter in German reactor stations. The composition of the iodine release to the environment was validated with the “discriminating iodine species sampler”. The main sources for the release of elemental iodine from BWR and PWR were discovered and finally suppressed. The mechanical strength of HEPA filters is being tested at high temperatures and humidities. Prototype HEPA filters have been developed with much higher resistance against humidity, differential pressure, and corrosion. A filter for the removal of particles during extreme operating conditions such as containment venting was developed using stainless steel fibers for the filter medium. The first filter of this type has already been built and installed in a modern German PWR as a part of the containment-venting system for serious accidents.     

The SQUG program for resolution of USI A-46 — Status and implementation plans

In response to U.S. Nuclear Regulatory Commission (NRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment and Operating Nuclear Power Plants”, the Seismic Qualification Utility Group (SQUG), with the support of the Electric Power Research Institute (EPRI), developed a comprehensive program to verify the seismic adequacy of equipment in operating nuclear power plants. The primary thrust of the program has been the development of procedures, criteria, and data to apply actual experience on the performance of equipment during earthquakes to the verification of seismic ruggedness of similar equipment in nuclear plants. While the use of such experience data continues to play a primary part in the SQUG program for resolution of USI A-46, the overall SQUG program includes a number of other significant elements which, taken together, provide a comprehensive approach for verification of the seismic adequacy of equipment in nuclear plants. These elements of the SQUG program include the assimilation and use of seismic shake table data in a generic way; the development of simplified analytical tools and criteria for evaluation of equipment anchorage, tanks, heat exchangers and cable trays; and the development of procedures for identifying and evaluating electrical relays, which are essential to plant shutdown in response to an earthquake. Procedures and data bases for performing and documenting the various seismic evaluations and plant walkdowns, and a program for training the large number of engineers who will be required to implement the SQUG methodology, have also been developed. This paper describes the main elements of the SQUG program for resolution of USI A-46 and provides a status report on the plans for their implementation in SQUG member plants.     

压水堆核电厂高压熔堆严重事故序列分析

压水堆核电厂的高压熔堆事故覆盖了大部分的严重事故序列,且具有很大的潜在威胁。根据我国900MW压水堆核电厂的概率安全分析（PSA）结果选取了丧失厂外电、未能紧急停堆的预期瞬态、二回路管道破口、一回路小破口和蒸汽发生器传热管破裂5种典型的高压熔堆严重事故序列,并使用SCDAP/RELAP5程序对这些事故序列的进程和后果进行了计算分析。计算结果表明：5种典型高压熔堆事故序列可能导致高压熔喷和安全壳直接加热风险,可能引起安全壳早期失效,很有必要采取相应的一回路卸压措施。     

ATWS analysis for PWR using the coupled code system DYN3D/ATHLET

The ATWS transient “Loss of main feed water supply” in a generic four-loop PWR at the nominal power of 3750 MW was analyzed using the coupled code system DYN3D/ATHLET. A variation of the MOX-fuel-assembly portion in the core has an effect on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behaviour of the coolant pressure, which is safety-relevant. It has been demonstrated that the pressure maximum decreases with an increasing portion of MOX. For all core loadings considered, both primary-circuit mechanical integrity and sufficient core cooling are guaranteed.     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

On-line monitoring the incore power distribution by using excore ion-chambers

For most of nowaday pressurized water reactors (PWR), excore ion-chambers (EIC) are the sole real-time sensors responding to the incore power distribution (IPD). Conventionally, it is supposed that the EIC could not carry out information of the IPD in radial direction. As a result, there is an excess conservatism when they are being used for the core monitoring so far, and some flexibilities of the reactor operation are sacrificed. Though progresses have been made, there are still some inevitable obstacles about newly developed monitoring methods: mounting and maintaining a large amount of fixed incore detectors are costly; and the IPD calculated by on-line simulators is not the “measured” one. Fortunately variations in the IPD must be driven by reasonable physical causes, and the corresponding variations in the readings of the EIC are notable enough to be identified. So with a specific referenced IPD measured from periodical flux mappings, it is possible to on-line monitor such variations by using the EIC. Aimed at this, a harmonics grouping method and an algorithm of higher order perturbation variables are developed, and an influence equation for control rods is also proposed. Then a precise restriction equation describing the reasonable variations in the IPD is deduced for example of the Heating Reactor. Combined with a proper-configured EIC, a systematic “excore-to-incore” method is realized finally.     

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.