Annealing of nuclear reactor pressure vessels

The neutron embrittlement that occurs in the beltline of reactor pressure vessels (RPV) can be managed by various techniques such as fuel management, but only thermal annealing can reverse the effects and result in a restoration of RPV beltline material toughness. The US Nuclear Regulatory Commission has recently revised the Code of Federal Regulations to include the use of thermal annealing of RPV for recovery of material toughness. The Annealing Rule, 10 CFR Part 50.66, has an associated Regulatory Guide 1.162 that describes the format and content of a thermal anneal report that must be submitted to the NRC prior to performing an anneal. This paper will describe the thermal annealing process including regulatory requirements in 10 CFR Part 50.66, techniques for predicting and measuring the toughness recovery, and NDE requirements. Although 14 Russian-designed RPVs have been annealed, there are sufficient differences between the Russian and US designs to question the ease of thermal annealing without producing any unwanted dimensional changes in the RPV and associated piping. The paper will discuss the ongoing annealing demonstration project supported by the Department of Energy which performed a thermal anneal on a canceled pressured water reactor at Marble Hill, Indiana. The associated NRC programs also will be described. This annealing demonstration will be used to bench mark the expected thermal and stress distributions created by thermal annealing and minimize the possible dimensional changes in the RPVs. The paper also will discuss the first possible implementation of thermal annealing for a US commercial nuclear power plant and some important issues that will need to be addressed.     

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVO's Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.     

承压热冲击下DVI接管动态应力特性研究

压力容器直接注入（DVI）接管在热冲击下的动态应力特性对于反应堆压力容器（RPV）结构完整性评估具有重要意义。建立了含DVI接管的RPV压力壳热流固耦合数值计算模型,并进行了验证分析;然后研究了蓄压安注箱（ACC）和堆芯补水箱（CMT）安注时RPV筒体和DVI接管热工水力特性;最后分析了热冲击下RPV筒体和DVI接管连接高应力区的温度分布、等效应力和等效塑性应变分布特性。研究结果表明,ACC安注阶段RPV筒体和DVI接管连接区存在较大的温度梯度和等效应力,且发生了局部塑性变形。若发生承压热冲击事件,应控制好DVI接管连接区温差,确保反应堆压力容器的结构完整性。本文开发的热冲击下热流固耦合数值计算模型和计算方法可用于核岛内DVI接管与RPV筒体的安全性评价,也可用于类似承压结构在热冲击下的动态应力特性分析。      

Three-dimensional transient sealing analysis of the bolted flange connections of reactor pressure vessel

The failure of sealing system of the bolt flange connections is the primary failure mode of the nuclear reactor pressure vessel (RPV). For the safety and integrity of RPV, it is important to predict the sealing behaviour of the bolt flange connections under various loading conditions. Based on the finite element (FE) method for coupled thermal elastoplastic contact problems, a three-dimensional (3D) transient sealing analysis program of nuclear reactor pressure vessels is developed with the consideration of the non-linearity from both surface and material, transient heat transfer and multiple coupled effects. A contact correction approach is proposed to simulate the loading of the bolt connection under the condition of pre-stressing. An automatic pre-processing program is developed for FE modelling of RPVs. Using these programs, a 1:4 scaled model of a 300 MW RPV is analyzed under the loading conditions including pre-stressing, pressurization, heating and cooling. The computational results obtained are in a good agreement with the data of experimental tests. These programs are also successfully used in analyzing the full-scale model of the RPV in a nuclear power plant.     

The effects of the stainless steel cladding in pressurized thermal shock evaluation

Fracture mechanics analysis is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV), such as the pressurized thermal shock (PTS) analysis and P–T limit curve construction. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, the simple analytical treatment schemes to calculate the stress and resulting stress intensity factor at the tip of the flaws in the RPV with stainless steel cladding are introduced. For a reference thermal–hydraulic boundary condition, the effects of cladding thermal conductivity and thermal expansion coefficients on the stress intensity factor of surface flaws were examined. Also, the effects of cladding plasticity and thickness were quantitatively examined. The analysis results showed that the existence of the stainless cladding had significant impacts on the RPV failure probabilities.     

承压热冲击下材料特性对反应堆压力容器结构完整性的影响

介绍了承压热冲击(PTS)分析的背景和研究现状,阐述了基于确定性断裂力学的反应堆压力容器(RPV)结构完整性分析方法.分析了材料性能模式(线弹性和弹塑性)和辐照效应对PTS下RPV结构完整性的影响.     

反应堆压力容器辐照脆化状态评估

基于大量相似辐照脆化试验测试数据和实际辐照监督测试数据,采用统计分析的方法,选出适合于某核电厂反应堆压力容器（RPV）的辐照脆化评估公式。以该核电厂已经完成的辐照监督管测试数据为输入,对RPV当前的辐照脆化状态进行了评估,并推算、分析了RPV在寿期末的结构完整性;基于辐照脆化计算结果,绘制了各运行阶段RPV的压力-温度限值曲线（P-T曲线）,并给出运行建议。      

Critical crack depth diagram of reactor vessel for pressurized thermal shock

During the operation of a pressurized water reactor, a certain type of transients could induce rapid cooldown of the reactor pressure vessel (RPV) with relatively high or increasing system pressure. This induces a high tensile stress at the inner surface of the RPV, which is called the pressurized thermal shock (PTS). The structural integrity of the RPV during PTS should be evaluated assuming the existence of a flaw at the vessel. For the quantitative evaluation of the vessel failure risk associated with PTS, the probabilistic fracture mechanics (PFM) analysis technique has been widely used. But along with PFM analysis, deterministic analysis is also required to determine the critical time interval in the transient during which mitigating action can be effective. In this study, therefore, the procedure for the deterministic fracture mechanics analysis of RPV during PTS is investigated using the critical crack depth diagram and the computer program to generate it is developed. Four transients of typical PTS, steam generator tube rupture, small break loss of coolant accident and steam line break are analyzed, and their response characteristics such as critical crack depth and critical time interval from the initiation of the transient are investigated.     

LOCA下具有表面裂纹的反应堆压力容器承压热冲击分析

失水事故（LOCA）瞬态下,具有半椭圆形表面裂纹的反应堆压力容器（RPV）承压热冲击（PTS）问题被研究。采用有限元方法计算瞬态过程的热-应力响应;采用影响函数法计算应力强度因子,分别对母材和堆焊层内的应力进行分解,从而解决了由于堆焊层存在造成的应力拟合困难带来的计算偏差。编制了相应的断裂分析程序,对LOCA下RPV的结构完整性进行了分析。结果表明,在研究的LOCA下,整个瞬态过程中RPV应力强度因子均未超过材料断裂韧性,压力容器结构安全。本文研究为RPV在PTS下的结构完整性评估提供理论指导。     

Structural integrity assessment of aging nuclear reactor pressure vessels

Irradiation embrittlement reduces both the cleavage fracture toughness and the ductile tearing toughness of reactor pressure vessel (RPV) steels. Extensive research programs have investigated the fracture behavior of heavy-section vessels containing flaws. Information obtained from that research has been used to develop regulatory guidance for evaluating the structural integrity of irradiated RPVs. Additional research programs have developed fracture analysis methods, and generated the data required for their implementation. Regulatory guidance employs fracture analysis technology to assure that adequate fracture-prevention margins for RPVs are maintained throughout the licensed operating period of nuclear power plants.     

Summary report on new regulatory codes and criteria and guidelines: IAEA guidelines for WWER reactor coolant system integrity assessment

In the frame of the nuclear safety programme to assist the countries of Central and Eastern Europe, the IAEA identified and ranked in total 263 safety issues for WWER-440/230, WWER-440/213 and WWER-1000/320 nuclear power plants, related to both design and operation. In the area of reactor coolant system integrity, 24 safety issues were identified and 15 of them ranked as of high safety significance. These include: reactor pressure-vessel integrity and related aspects, primary and secondary high-energy piping integrity, steam generator integrity and reliability of the non-destructive testing for in-service inspection. In addition to obtaining international consensus on possible solutions to address the safety issues identified and to reviewing completeness of proposed safety improvements, the IAEA initiated development of guidelines to address the issues of highest safety concern. In the area of the reactor coolant system integrity, guidelines for the leak before break concept application, for the pressurized thermal shock analysis and for the in-service inspection systems qualification were developed. Further activities of the IAEA were focused on the implementation of guidelines developed in the Member States concerned. With this objective, a Co-ordinated Research Programme ‘Round-robin Exercise on WWER-440 Reactor Pressure Vessel Embrittlement, Annealing and Re-embrittlement’, a ‘WWER-440/213 Pressurized Thermal Shock Analysis Benchmark Exercise’ and a pilot study to implement the qualification approach to a real power plant component have been initiated by the IAEA and are well under way at present. In this paper, an overview of these IAEA activities related to reactor coolant system integrity is provided and the main principles and elements of guidelines developed discussed.     

熔融物堆内滞留条件下RPV长期结构完整性分析

熔融物反应堆压力容器(RPV)内滞留(IVR)是三代核电厂重要的严重事故缓解措施，而防止RPV的热工失效和结构失效是实现IVR的前提。本文建立考虑内壁面熔蚀的RPV有限元模型，在温度场分析的基础上，开展蠕变计算，得到不同时刻下的应力应变响应，通过选取典型评定路径并利用基于Larson-Miller参数的累积损伤理论进行蠕变损伤计算及评价。分析结果表明：在考虑一定内压的IVR条件下，RPV不会发生蠕变断裂，长期结构完整性可保证。本文的研究方法可为后续核电厂RPV在IVR条件下的结构完整性分析提供参考。     

Structural integrity assessment of the reactor pressure vessel under the external reactor vessel cooling condition

Since the suggestion of external reactor vessel cooling (ERVC), the effects of melting and cooling on the response of structural integrity of the reactor pressure vessel (RPV) under core melting accident conditions have been investigated. To investigate the initial behavior of RPV lower head and the effects of analysis conditions on the structural integrity of RPV, the transient analysis is utilized considering the transient state. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The temperature and stress analyses are performed for core melting accident by using ABAQUS. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. In addition, the results by transient analysis are compared with those by steady state analysis and the effects of analysis conditions on structural integrity are reviewed.     

反应堆压力容器外水冷条件下贯穿件完整性分析

严重事敝下堆芯熔融物坍塌到反应堆压力容器(RPV)下封头时,可能造成贯穿件因高温熔融物热侵袭而失效,使压力容器丧失完整性,熔融物进入到反应堆堆腔中,导致熔融物堆内滞留(IVR)失效.在分析贯穿件脱落和熔融物流入贯穿件两种失效模式基础上,分别运用VTA程序和修正的整体凝固模型(MBF)计算贯穿件焊缝的熔化程度、热膨胀产生的摩擦力,估算贯穿件内熔融物流动的距离.结果表明,在成功实施反应堆压力容器外水冷(EVVC)措施条件下,300 MW压水堆核电厂压力容器的下封头不会因贯穿件失效而丧失完整性,堆芯熔融物小能通过贯穿件失效向堆腔迁移.     

Neutron embrittlement of VVER reactor pressure vessels—recent results, open issues and new developments

Decisions regarding the verification of design plant lifetime involve a determination of the component and circuit condition. Neutron embrittlement of reactor pressure vessels (RPV) becomes a crucial consideration for continued safe plant operation. Since 1991, the European Commission (EC) has financed a significant number of projects in this area, in particular through the TACIS and PHARE programmes. In Nuclear Safety support programmes the countries mainly concerned are Russia, Ukraine, Armenia, and Kazakhstan for the TACIS programme, and Bulgaria, Czech Republic, Hungary, Slovak Republic, Lithuania, Romania and Slovenia for the PHARE programme. The corresponding results of these projects on neutron embrittlement have contributed to a better understanding of the available data and their uncertainties. Recent confirmations of the irradiation temperatures of the surveillance specimens in the operating VVER nuclear power plants show that the surveillance specimens can be used further for the validation of the current and the expected neutron embrittlement. Taking into account the latest results as a baseline for the development of the necessary material data to be used for upgraded RPV integrity assessment, two new TACIS projects are being launched, jointly with Russia and Ukraine, whose scope is:

• Upgrading the surveillance databases.

• Acquisition of new impact test and toughness results on reconstituted surveillance specimens, including the evaluation of the Mater Curve Approach.

• Further validation tests of the shape of the fracture toughness curve and the base and weld metal and characterization of the cladding.

• Preparation of some selected upgraded RPV integrity assessments, with insights on the latest approved methodology.

The paper summarizes the major conclusions of the recent completed EC projects, reviews the remaining major open issues in the field of reliable determination of fracture toughness properties of the operating VVER RPVs, and details the scope of the new projects.     

Investigation on constraint effect of reactor pressure vessel under pressurized thermal shock

In recent years, the integrity of reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) accident has been treated as one of the most critical issues. Under PTS condition, the combination of thermal stress due to a steep temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the RPV wall. As a result, cracks on the inner surface of RPVs can experience elastic-plastic behavior that can be explained using the J-integral. In such a case, however, the J-integral may possibly lose its validity due to the constraint effect. The degree of constraint effect is influenced by the loading mode, the crack geometry and the material properties. In this paper, three-dimensional finite element analyses are performed for various surface cracks to investigate the effect of clad thickness and crack geometry on the constraint effect. A total of 36 crack geometries are analyzed and results are presented by the two-parameter characterization based on the J-integral and the Q-stress.     

The HTTR reactor pressure vessel and its integrity against a PTS event

Before manufacturing the real steel to be used in the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) the vessel manufacturer and materials supplier made a sample steel by the same procedure as for the real steel (2.25Cr-1 Mo) and conducted many tests to obtain material strength data for its base and weld metals. The test results showed that the sample steel satisfied the HTTR design requirements. Vessel cooling panels are set on the inner surface of the biological shielding concrete around the RPV, and are circulated with cooling water at 0.5 MPa and 40°C to cool the shielding concrete during normal operation of the reactor. By supposing that the cooling panel breakes and the water discharges to the RPV outer surface heated at 400°C, the stress distribution generated in the vessel wall by a pressurized thermal shock (PTS) event can be calculated using a finite element method code. This paper describes some of the results obtained from the material testing of the sample steel and the estimated result using the scheme developed for a light water reactor pressure vessel, to clarify the integrity of the HTTR-RPV under a PTS event.     

Regulatory strategy for effective application of periodic safety review system

In less than 10 years, the first commercial pressurized water reactor (PWR) plant in Korea will reach its official design life. As part of safety activities, developed countries have already implemented periodic safety review (PSR) or equivalent programs to check and improve the safety of operating nuclear power plants (NPP) during their plant life. At the end of 1999, it was decided by the Korean Atomic Energy Safety Committee to adopt the PSR program and to apply it to Korean operating NPP. Since Kori Unit 1 started the review for the first tentative application of PSR as a model case in May 2000, it is now progressing well. Management of aging is one of the major factors to be considered in PSR and life extension of a nuclear plant. This paper is intended to introduce the regulatory aspect and strategy of Korean PSR. The background and scope of basic PSR guidelines are described, and a summary of technical criteria for aging management, which shows a regulatory direction for PSR, is also presented.     

Apparative developments for inservice inspections of reactor pressure vessels

Inservice inspections of primary circuit components are important preventive measures to guarantee nuclear power plant integrity, satisfying simultaneously reactor safety and economy in plant operation. Emphasizing pressurized water reactor pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implementation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system.