Summary of the PISC II project: PISC II report No. 1—June 1985

In assuring light water reactor safety, it is vital to have confidence that no leaks or breaks will develop in the reactor pressure vessel and associated piping, which together constitute the primary coolant circuit. Initially small defects in these thick steel components can grow under the stresses arising from repeated pressure and temperature changes and the embrittlement of the metal caused by the radiation emanating from the reactor core.

Ultrasonic testing is widely used for detecting, locating and sizing flaws in primary circuit elements at various stages of plant life. The successive PISC projects have constituted the most notable sustained international effort to assess the effectiveness of these inspection techniques.

The Plate Inspection Steering Committee (PISC I) programme (1976–1980) was intended to establish the capabilities of the 1974 ASME Code Section XI ultrasonic procedure. The Programme for the Inspection of Steel Components (PISC II, 1981–1986) constitutes a more profound evaluation of the best performance obtainable by modern ultrasonic techniques under optimal conditions.     

European developments in NDE for steam generators

A high standard of integrity is demanded of the steam generator tubing in nuclear power stations to minimise the cost of any loss of availability due to a need for repairs or plugging. If any deterioration should occur it is desirable to identify it before failure in order to permit preventive maintenance to be considered and, particularly for prototype plant, provide early guidance as to the possible need for changes in the design of later plant.

These requirements have resulted in the continuing development of inspection techniques for application at the fabrication stage and subsequently in-service. For the latter it is desirable that inspections do not extend the length of shutdowns beyond that required for core refuelling and routine maintenance and the associated statutory plant inspections.

The above needs apply to all reactor types but the in-service inspection undertaken is very much affected by the type of plant and the available operating experience with similar units. Thus the acceptable leak rate between sodium and water in the fast reactor steam generator is measured in grammes per hour while kilogrammes per hour would be a more appropriate unit for PWRs. For the latter there is an additional requirement to lower the radiation exposure during inspection.

This paper discusses reported European developments in the field and concentrates on the problems of inspecting the tubing and tube-to-tubeplate joints. Also, although the ultimate integrity of a plant unit is dependent on the quality of the original fabrication with its associated inspection, most attention is given to ISI.

Apart from the continuing development of eddy current techniques for austenitic tubing, the past five years have seen an increasing interest in ultrasonic techniques for detailed tubing examination while digital data handling has been demonstrated in the field. In the fast reactor area the development of improved NDT procedures for both fabrication and ISI of explosive welds is helping to provide the basic technological know-how which is essential if explosive welding is to be considered as a serious design option for large scale application for the tube-to-tubeplate joints in large steam generators.     

UK developments in the ultrasonic examination of pressure vessels

This paper reviews developments in the UK of ultrasonic inspection methods for both in-service inspection of existing, older plant and also for the proposed PWR at Sizewell.

In the case of existing plant there are problems which stem from the fact that ultrasonics was not in widespread use at the time the reactors were designed and constructed. Consequently weld geometries, surface finishes and access pose difficulties which must be overcome by developments in the way ultrasonics is applied.

For the PWR, the reactor has been designed from the outset with inspection in mind. The safety case requires a demonstration that certain parts, including the reactor pressure vessel, cannot fail and this, in turn, involves a demonstration of high inspection reliability. Such a demonstration has necessitated a range of developments to understand and quantify the way sound reflects from a variety of metallurgical defects.

The ultrasonic inspections carried out for Sizewell B involve extensive use of automated techniques during fabrication. These have also been the subject of intensive development in the UK.     

Master curve techniques to evaluate an irradiation embrittlement of nuclear reactor pressure vessels for a long-term operation

Irradiation embrittlement is a limiting condition for the long-term safety of a nuclear reactor pressure vessel (RPV). The first PWR in Korea is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. The RPV was designed by an old construction code and its beltline circumferential welds have suffered from an irradiation shift problem like other Linde 80 welds. The master curve method is considered as the most promising tool to characterize irradiated structural steels by using a fracture mechanics basis. In order to implement the master curve method for the assessment of an irradiation embrittlement of old power reactors for a continued long-term operation, three practical issues were emphasized in this investigation, which are the specimen geometry effects on the master curve results, the specimen reconstitution techniques in an old existing surveillance program, and the index temperatures of an irradiation embrittlement when compared with the conventional Charpy data.     

Evaluation of induction heating stress improvement (IHSI) treatment applied to nuclear primary piping

A new technique, IHSI (induction heating stress improvement), has been developed to reduce the tensile welding residual stress in the surface of steel piping for nuclear power plants, so reducing the susceptibility to stress corrosion. It is shown in this paper that the technique is effective and safe even if it is applied to piping which contains a small initial crack which is impossible to detect by ultrasonic inspection.     

Combining discrete-time Markov processes and probabilistic fracture mechanics in RI-ISI risk estimates

Based on the research conducted in the Finnish SAFIR project, which is a national nuclear energy research program, discrete-time Markov processes and probabilistic fracture mechanics (PFM) methods are further developed and applied in this paper. The purpose of this work is to increase the accuracy of risk estimates used in RI-ISI, and to quantitatively evaluate the effects of different inspection strategies on risk. Piping failure probabilities are obtained by using PFM analyses. PFM has the advantage that its results are not affected by existing in-service inspection (ISI) activities at the nuclear power plants (NPPs), unlike failure probabilities assessed from existing failure data. The PFM results for crack growth are used to construct transition matrices used in a discrete-time Markov process. The application of Markov process allows the examination of effects of inspections on the failure probabilities. Finally, the developed method and results are showcased by applying them to a selected piping system in an existing Finnish NPP.     

Pipe cracking experience in Swedish BWRs

There are twelve commercial nuclear reactors in Sweden of which nine are boiling water reactors (BWR) of ABB Atom design. The five oldest are equipped with external recirculation loops while the four youngest have internal pumps. Cracking has appeared in reactor water systems such as the residual heat removal system (RHR), hydraulic scram system, reactor water clean-up system (RWCU), feed water system (FW) and lately the recirculation system. Within the reactor pressure vessel (RPV) some piping has been affected in the core spray system. Some defects resulted in replacement of parts of the piping system while others just led to repairs of the involved component itself. The cracking mechanisms experienced during nearly twenty years of operation are intergranular stress corrosion cracking (IGSCC) and thermal fatigue.     

Evaluating the NESC-I test and the integrated approach to structural integrity assessment

The NESC-I spinning cylinder test was designed to simulate selected conditions associated with an ageing reactor pressure vessel (RPV) subjected to severe pressurised thermal shock (PTS) loading and containing hypothetical flaws. It formed the focal point of the first project of the Network for Evaluation of Structural Components (NESC), with the objective of validating the combination of non-destructive inspection and structural mechanics assessment procedures for evaluating the integrity of such an aged structure containing postulated flaws. The huge amount of data generated over the seven-year project has been evaluated and is now available to the international community. The test demonstrated that, for the specific conditions considered, defects of up to 74 mm depth in material related to that of an ageing RPV would not propagate to cause catastrophic failure under a severe PTS-type thermal shock. This outcome was fully in line with the pre-test analysis forecasts, which combined the defect-size information supplied from blind inspections trials, a comprehensive materials data set, and a range of structural analysis tools.     

Irradiation embrittlement of model alloys and commercial steels: analysis of similitude behaviors

The deep understanding of the irradiation embrittlement of the pressure vessel of nuclear reactors is a key issue for the plant lifetime assessment and life extension through mitigation methods like annealing and much effort have been done in the last decades to tackle such complex issue. The reactor pressure vessel (RPV) material of nuclear power plants is exposed to neutron irradiation during its operation. Such exposure is generally inducing a degradation of the mechanical and physical properties of the materials, e.g. an increase of the ductile to brittle, DBT, transition temperature and a decrease of the upper shelf energy.The different response of materials to neutron irradiation, even many factors are also playing significant role, is mainly due, for a given exposure, to the chemical composition of the materials. In particular, for the RPV steel, elements like phosphorus, P, copper, Cu, and nickel, Ni, are playing a key role.A parametric study of the response to neutron irradiation of 32 different model alloys with parametric variation of elements (Ni from 0.004 to ∼2 wt%, P from 0.001 to 0.039 wt%, Cu from 0.005 to ∼1 wt%) has been recently completed within the frame of the European Network AMES and EC-JRC AMES Institutional project [1].Such study on model alloys reveals to be a fundamental tool to understand the individual role of each element and synergisms.To demonstrate the usefulness of the study to commercial RPV steels, an analysis of the results and the similitude of behavior between model alloys and available RPV commercial steels has been carried out and the results are presented in this paper.     

Improvement of the calculation of the stress intensity factors for underclad and through-clad defects in a reactor pressure vessel subjected to a pressurised thermal shock

The analysis of the stability of a defect in a cladded reactor pressure vessel (RPV) of a nuclear pressure water reactor (PWR) subjected to pressurised thermal shock (PTS) is one main elements of the general safety demonstration. Recently, CEA proposed several improved analytical tools for the analysis of the PTS. First, an analytical solution for the vessel through-thickness temperature variation has been developed to deal with any fluid temperature, taking into account the possible presence of a cladding, in the case of an internal PTS. The associated thermal stress expression has been simplified and a complete linearised solution is given for the thermal loading and also for internal pressure, depending on the main vessel material and on the cladding properties. Finally, a complete compendium is also given for the elastic stresses intensity factor calculation.     

国核压水堆示范工程再热管道布置方案分析

文章介绍了国核压水堆示范工程再热管道的布置方案,然后进行了再热管道的应力分析,并对管道施加在汽缸和MSR管口的载荷进行了校核,结果表明该方案满足工程要求。     

Japanese developments in the ultrasonic examination of pressure vessels

In Japan, nuclear power plant has grown to such an extent that it now provides a fundamental part of total power generation. In consequence, the assurance of structural integrity is increasingly important. The activities of three representative organizations, the Nuclear Power Engineering Test Center (NUPEC), the Japan Pressure Vessel Research Council (JPVRC) and the Japan Power Engineering and Inspection Corporation (JAPEIC) are briefly overviewed in this paper.

NUPEC is carrying out the proving test program of in-service inspection. Manual ultrasonic inspection testing was completed in early 1984; factors affecting defect detection and sizing were examined and some of these test results are introduced here. Mechanized automatic inspection testing with devices now in use in recent operating plants is underway. JPVRC is participating in the international round robin test, PISC-II, and in the test between US-PVRC to investigate the reliability of ultrasonic examination, and some test results are introduced. JAPEIC recently expanded its activities into research and development in the area of structural integrity, robotics, mechanized inspection and non-destructive inspection. Some topical items for these programs are overviewed.     

Experience in the replacement of safety related piping in German boiling water reactors

In the German boiling water reactors (BWR) of the 69 series and their forerunner plant, high-strength low-alloy ferritic materials were used for a large number of pipings both inside and outside the pressure boundary (PB). The choice of this type of material led to comparatively thin-walled piping which, at that time, had been designed and manufactured in accordance with the codes and standards applying in the Federal Republic of Germany.

Due to material properties resulting from production in a conventional manner, design features which did not sufficiently meet the requirements for nondestructive testability, and defects caused during processing, mainly in the area of circumferential welds, the ferritic pipings inside the PB were replaced in the course of a plant upgrading by new piping designed and manufactured according to the basis safety concept.

The improvements and experience gained during backfitting of five German BWR plants are part of the German safety strategy and can be summarized as follows:

1. (1) Exclusion of large fractures on the basis of an optimized quality level for the piping.

2. (2) Elimination of need for the pipe whip restraints which existed in the former piping.

3. (3) Limited reduction of the former scope of inservice inspections, mainly as a consequence of improved weld quality and optimized weld performance.

4. (4) Reduction of personnel radiation exposure, e.g. by reduced number of welds and by manufacture of welds using automatic equipment, as well as by improved nondestructive testing.

5. (5) Availability values for backfitted BWR comparable to German PWR values.

The pipings made of stabilized austenitic materials, which are arranged inside and outside the containment of the BWR plants, were not replaced since their quality level has been proved to be sufficient even on the basis of the present standards.     

A comparison of Western and Eastern nuclear reactor pressure vessel steels

The task was essentially to compare the irradiation response of `East' and `West' steels. Since the plates and forgings of pressure vessels must be welded together, it is obvious that the strength requirements of the welds and heat affected zones (HAZ) can be no less demanding than those of the plates and the forgings themselves, particularly as experience has shown that the most likely location for flaws is in the welds or their HAZs. These and the highly stressed regions of the reactor pressure vessel (RPV) are important because neutron irradiation degrades the mechanical properties of steels.After comparing the various designs, manufacture and materials of the various RPVs, a comparison was made of the irradiation response of these different steels. The role of mitigating the change in mechanical properties on irradiation by thermal annealing was also considered.Particular codes/guides could only be used for the predicting results underpinning their own database because a major difference between these national codes/guides is that the elements conferring irradiation sensitivity are different for the two cases considered, i.e. Russian codes [1] (PNAE G-7-002-86) and the USNRC guide [2] (RG 1.99 Rev. 2). In the former, copper and phosphorus are significant, while copper and nickel are identified as significant in the latter case.Predictions were compared for `real' materials used in NPPPVs whose compositions were known. The irradiation response of these steels is coincidentally similar. The essential difference in behaviour is in the lifetime fluence. Eastern steels are irradiated to a much higher fluence than Western steels. Differences in the predictions of the Eastern–Western codes/guides are a reflection of differences in the concentration of deleterious elements and pessimisms of the various codes/guides, particularly at low concentrations of deleterious elements where they are most conservative. Thirdly, and on a `fitness for purpose' basis, the shift in transition temperature produces a limitation to the lifetime of the earlier Eastern RPVs. However, by thermally annealing the RPV to mitigate the effect of neutron irradiation, where the conditions to recover the mechanical properties of both Eastern and Western steels are nearly the same, the operational life of these older Eastern plants has been extended. Life assurance of these plants has, therefore, become practicable.This aspect of RPV technology, which is currently being considered in the US, could extend the operational life of nuclear power plants and thereby reduce the cost of the electricity generated.     

French developments in the ultrasonic examination of pressure vessels

In 1985, 34 Pressurized Water Reactors (PWRs) were already in service in France and 21 were being constructed. The manufacturer (Framatome) and the owner (Electricité de France, EDF) have consequently gained great experience in the ultrasonic examination of the various components and particularly of the reactor vessel.

The development work initiated concerns:

• —the improvement of the knowledge of methods used;

• —the beginning of automation in fabrication control;

• —the continuation of the technical implementation of the signal and image processing.

At the fabrication stage, work was carried out to demonstrate the total adequacy of those examinations required by the RCCM Code.

In addition, Framatome is developing the automation of ultrasonic inspection in shop.

Inservice inspection of the vessel is carried out with ‘MIS’ inspection equipment.

Data analysis is improved by special software allowing the performance of different types of UT indications to be presented.

Diffraction echoes from focused probes are used to size underclad defects on reactor nozzles.     

核电示范项目设计方案变化管控研究

设计方案变化管理是核电项目设计变更管理的组成部分,方案变化控制对示范项目质量、费用、进度控制都有重要意义。结合高温堆示范工程设计方案变化管理实践,总结了核电示范项目设计变更管理中方案变化管理的难点,分析其原因。针对核电示范项目设计方案多变、涉及专业多等特点,提出了方案变化管理的相关建议和工作流程,以供后续类似工程参考。     

RIMAP demonstration project. Risk-based life management of piping system in power plant Heilbronn

In the framework of EU project RIMAP [Risk Based Inspection and Maintenance Procedures for European Industry (2000)] a new European Guideline for optimized risk based maintenance and inspection planning of industrial plants (RBLM, Risk Based Life Management) is being developed. The RIMAP project consists of the three clustered projects

• •development (RTD)
• •demonstration (DEMO)
• •thematic network (TN)

Current work and future, planned work in RIMAP demonstration project on applications of the RIMAP methodology in power plants are presented briefly in the first part of the paper. Also presented in the paper are the results of a preliminary analysis of piping system in power plant Heilbronn using the concept of risk-based monitoring as part of overall concept of risk-based life management. Shortly the following issues are discussed in the paper

• •identification of critical components
• •application of a multilevel risk analysis (…from ‘screening’ to ‘detailed analysis’)
• •determination of PoF (Probability of Failure)
• •determination of CoF (Consequence of Failure)
• •optimation of inspection and maintenance plan

From our experience with the application of the RIMAP methodology the following conclusions can be drawn: The use of risk-based methods in inspection and maintenance of piping systems in power plants gives transparency to the decision making process and gives an optimized maintenance policy based on current state of the components. The results of the work clearly show the power of the proposed method for concentration on critical items: out of 64 monitored components 5 were selected for intermediate analysis and only 1 for the detailed analysis (probabilistic high temperature fracture mechanics).     

对于厚度70mm〜140mm焊缝超声检测的讨论

超声检测作为一种面积型缺陷的有效检测方法,被广泛应用在气化炉、除氧器、汽包等的焊缝检测上。随着近年来电厂、化工厂对质量的要求提升,确保焊缝质量就成了生产制造厂的重点。对于这些重型容器,焊缝厚度普遍在70mm~140mm,在这个厚度区间,采取的多道焊接,工艺控制不严很容易产生缺陷。无损检测标准的制定只是简单的对此厚度划分两个区间,并未给出划定此区间的依据。本文通过实际检测气化炉、除氧器、汽包等焊缝来讨论超声检测在厚度70mm?140mm区间的划定依据。     

承压设备焊缝超声相控阵检测读谱

介绍按ASME法规最新版（2009增补版）要求，用超声相控阵技术检测承压设备焊缝时，根据相控阵扇形（S－）扫描图像结合相应探测布置截面图，对典型焊接缺陷检测图谱进行识别和评定的应用案例．所涉及的焊接接头为相比于一般X型坡口双面焊是有一定检测难度的V型坡口单面焊和T型接头组合焊，焊接缺陷包括：焊趾裂纹、焊道下裂纹、内表面开口裂纹、坡口未熔合、根部未焊透、以及密集气孔等。同时，还比较了相控阵检测的定量结果与实际缺陷尺寸。目的是依据相关法规和标准要求，为承压设备焊缝相控阵超声检测图像识别和结果评定提供有用借鉴。     

Cracking in feedwater pipework of light water reactors: Causes and remedies

The present paper deals with cracking in ferritic pipework of light water reactor (BWR) feedwater systems whose causes are unsuitable design and manufacture and, at least occasionally, unsuitable water quality. Thus in two BWR plants in the German Federal Republic large circumferential cracks have been found in the circumferential welds in the main feedwater pipelines immediately adjacent to the reactor pressure vessel, and in a further BWR large longitudinal cracks have been found in pipe bends of the reactor water purification pipework connected to the main feedwater pipelines.The piping regions near the reactor pressure vessel feedwater nozzles represent a boundary region of varying thermodynamic states of the pressurised water. The reactor pressure vessel is heated to saturation temperature by the radioactive decay heat, and then cooler water, and sometimes (during start-up) cold water is injected into the reactor pressure vessel in order to maintain a constant water level about 2 m above the upper edge of the feedwater nozzles. In order to improve the state of knowledge regarding the stressing conditions under prevailing operating conditions, extensive strain and temperature measurements have been carried out. The results of these measurements carried out in several BWRs confirm the occurrence of rapid temperature changes in the feedwater pipework in the regions where it connects to the reactor pressure vessel, leading to varying stresses which at times reach the plastic region. These processes are triggered by special operating states such as start-up and shut-down or hot standby operation with feedwater flows smaller than 6% relative to normal operation under full load.