VVER-1000型反应堆压力容器热老化分析评估

本文系统介绍了VVER-1000型反应堆压力容器(RPV)的温度监督情况,针对田湾核电站1#机组RPV的温度监督测试结果进行分析,评价运行3年后RPV力学性能(包括拉伸、冲击、断裂韧性)变化行为及热老化脆化机理,评估了当前田湾RPV服役运行后的热老化脆化状态和温度监督的时间安排。结果表明,温度监督样品经过堆内高温环境考验后,焊缝材料表现出一定程度的脆化特征,但母材、热影响区脆化不明显。与康采恩模型的结果和俄罗斯数据相比较后,认为田湾核电站1#机组RPV热老化脆化情况在合理范围内。     

Fracture toughness prediction for highly irradiated RPV materials: From test results to RPV integrity assessment

New fracture toughness data are represented for highly irradiated RPV materials that were obtained by testing standard compact specimens with thickness of 12.5 mm and 25 mm and pre-cracked Charpy specimens machined from the RPV decommissioned. Two advanced engineering methods, the Master Curve and the Unified Curve, are applied for treatment of the test results. Application of the dependence of fracture toughness K_JC on test temperature T predicted with the Master Curve and the Unified Curve methods on the basis of surveillance specimens testing is discussed for RPV integrity assessment when the reference K_JC(T) curve is recalculated to the crack front length of the postulated flaw that is considerable larger than thickness of surveillance specimens. The prediction of the K_JC(T) curve transformation caused by neutron irradiation is considered.     

Reactor radiation dosimetry using fused quartz

The characteristics of quartz glass for separate dosimetry of γ radiation from a VVR-SM reactor in the presence and absence of neutron fluxes are investigated. Comparing the absorption and photoluminescence spectra of samples irradiated with γ rays from a stopped reactor and mixed with neutron and γ radiation from an operating reactor shows that in both cases oxygen defects are produced and the dose dependences are linear. The dosimetric bands are stable with respect to light and temperatures up to 400°C. The stationary γ-ray flux from the reactor, after the reactor is stopped, is calibrated according to a known source of γ-ray source ⁶⁰Co and is ≈15 Gy/sec. __________ Translated from Atomnaya énergiya, Vol. 100, No. 3, pp. 216–220, March, 2006.     

ASTEC and ICARE/CATHARE modelling improvement for VVERs

ASTEC and ICARE/CATHARE computer codes, developed by IRSN (France) (the former with GRS, Germany), are used in RRC KI (Russia) for the analyses of accident transients on VVER-type NPPs. The latest versions of the codes were continuously improved and validated to provide a better understanding of the main processes during hypothetical severe accidents on VVERs.This paper describes modelling improvements for VVERs carried out recently in the ICARE common part of the above codes. These actions concern the important models of fuel rod cladding mechanical behaviour and oxidation in steam at high and very high temperatures. The existing models were improved basing on the experience in the field and latest literature data sources for Zr + 1%Nb material used for manufacture of VVERs fuel rod claddings.Best-fitted correlations for the Zr alloy oxidation through a broad temperature range were established, along with recommendations on model application in clad geometry and starvation conditions. A model for the creep velocity was chosen for the clad mechanical model and some cladding burst criteria were established as a function of temperature.After verification of modelling improvements on Separate Effect Tests, validation was carried out on integral bundle tests such as QUENCH, CODEX-CT, PARAMETER-SF (the application to the CORA-VVER experiments is not described in the present paper) and on the Paks-2 cleaning tank incident. The comparison of updated code results with experimental data demonstrated very good numerical predictions, which increases the level of code applicability to VVER-type materials.     

Fracture assessment of a brittle RPV under cold loading

The oldest Swedish reactor is a boiling water reactor (BWR) with a vessel made of A302 Grade B with rather high Cu and Ni content. These elements have intensified the irradiation embrittlement in the beltline region so that RT_NDT of certain welds may exceed 100 °C at the end-of-life condition. A preliminary study of the fracture risk for the beltline region showed that the limiting loading case would be the cold over-pressurization of the reactor. The objective of this study was to develop a reliable methodology for fracture assessment of the aged reactor vessel under cold loading scenarios. The test program covered experiments on standard SEN(B) specimens and clad beams under uniaxial and biaxial loading. The test material was a reactor vessel steel prepared with a special heat treatment to simulate fracture toughness properties of the aged reactor. No significant effects of shallow crack and biaxial loading were observed on cleavage fracture toughness in different clad specimens. While the ASME K_Ic reference curve was shown to be overly conservative, the Master Curve methodology satisfactorily predicted the experimental outcomes of the test program. The Master Curve methodology indicated that a 20-mm deep surface crack was acceptable in the beltline region under a cold over-pressurization scenario. This value was three times greater than what a methodology based on the ASME K_Ic reference curve yielded.     

Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement.     

基于RCC-M规范的反应堆压力容器快速断裂分析与讨论

对2000版和2007版RCC-M附录ZG中的快速断裂分析评价方法进行对比分析。结果表明:2007版附录ZG的适用范围更广,分析评价过程相对简化,对材料老化的考虑也更加全面。以堆芯筒体段的快速断裂分析评价为例,对基准裂纹的方位参数进行敏感性分析,将2个版本的评价结果进行对比。结果表明:采用2007版附录ZG得到的评价结果安全裕量更大,同时增大了反应堆压力容器压力-温度(P-T)曲线的限制范围,扩大了核电厂运行操作空间。     

Failure assessment of RPV nozzle under loss of coolant accident

In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the K_IC-reference curve and the J_R curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner.     

高温气冷堆压力容器开孔形状优化的边界元法研究

大开孔应力集中是高温气冷堆反应堆压力容器结构完整性分析的重点关注的课题之一。应力集中水平取决于开孔形状的设计。本文将三维边界元法用于高温气冷堆核电站示范工程反应堆压力容器的大开孔形状优化研究。边界元法只需在边界划分网格,在开孔形状优化过程中仅移动边界网格,避免了体积网格畸变问题。采用基于自适应生长的形状优化算法,通过应力集中区域沿法线方向生长有效降低应力。数值算例表明,该算法可方便得到具备低应力水平的优化开孔形状。     

反应堆压力容器承压热冲击(PTS)分析

在反应堆运行过程中发生严重的失水事故(LOCA)时,应急堆芯冷却系统启动,冷的安注水从安注接管注入反应堆压力容器(RPV)中,此时压力容器还维持较高压力,这种瞬态就称为承压热冲击,即PTS(Pressurized ThermalShock).按照10CFR50,61[2]和RCC-M规范[1],对安注接管、焊缝和堆芯筒体三个区域,进行了PTS工况评估,分析结果表明,在发生PTS时,压力容器的完整性是能够保证的.     

反应堆压力容器CRDM管座设计改进

对控制棒驱动机构(CRDM)管座压力试验后管座内径、垂直度等关键配合尺寸变大问题,基于目前采用的CRDM管座设计结构和制造工艺,从CRDM管座焊缝结构设计、不锈钢材料特性、管座设计强度、压力试验实施等方面对CRDM管座内径尺寸变化和垂直度变化原因进行分析,确定了尺寸变化的原因,并提出效控制关键参数的措施。     

核压力容器材料国产化的可行性评述

文中扼要介绍了我国首次生产的 RPV 用 A508-3钢锻件的工艺和性能。通过对生产经验、试验研究、国内外文献和国内现有及新添设备的分析给出:实现600MW 核电站压力容器国产化在实际上具备了可行性和现实性。     

反应堆压力容器承压热冲击中的PSA方法研究

在瞬态过程中,当处于承压状态下的反应堆压力容器(RPV)的内表面被快速冷却时,即为承压热冲击(PTS)。由此,反应堆压力容器可能出现贯穿裂纹而失效。为分析PTS事件导致RPV出现裂纹的频率,需要进行概率安全评价(PSA)。通过PSA模型确定可能引起PTS的事件序列,并结合这些序列的热工水力分析结果,为PTS概率断裂力学分析提供支持。     

Approach to the lifetime assessment of the bayonet back plate for IFMIF target

The IFMIF facility is aimed at the production of high flux (10¹⁸ n/m²/s) of 14 MeV neutrons to test the candidate Fusion materials under significant neutron damage, up to 50 dpa/year. The conceptual configuration of the IFMIF target, based on the bayonet back plate (BP), has been developed in the past years by several authors. The appropriate engineering design of the back plate, to be developed in the EVEDA (Engineering Validation and Engineering Design Activities) phase, would require a very high level of knowledge on the materials behaviour under irradiation, that will be acquired only after some years of IFMIF experimental activities. For this reason the back plate, which is primarily invested by the highest IFMIF neutron flux, has to be considered a sacrificial component. In spite of its systematic replacement, the engineering design has to be optimised and the lifetime analysis has to be made carefully, in order to credibly estimate the expected replacement frequency. Since the replacement time interval must be conservatively shorter than the back plate lifetime and, at each replacement, the facility has to be stopped for, at least, one week and subjected to risky and uncomfortable operations, it is necessary to perform a trustworthy analysis of the lifetime. To this purpose the various interconnections between the main damaging causes are discussed in order to evidence the most plausible reasons of back plate malfunctioning. Due to the lack of knowledge in some fields and the early stage of design, the analysis is only semi-quantitative. The analysis, which accounts for erosion/corrosion, hydraulic stability, neutron damage and thermo-mechanical stress as the main damaging causes, evidences also the research areas which deserve foremost attention during the EVEDA phase. The considered malfunctions are: lithium boiling, burning/piercing of the back plate, non-sufficient neutron flux, brittle rupture of the back plate, creep rupture, loss of tightness of the back plate sealing.     

反应堆压力容器在典型事故工况下的结构完整性评估

以反应堆压力容器(RPV)堆芯带区和入口接管为研究对象,建立断裂力学有限元分析模型,以典型事故瞬态的详细热工水力分析结果作为输入条件,对其进行瞬态温度场分析和应力分析。结合RPV辐照脆化计算结果,采用确定性断裂力学分析方法,对RPV在4种典型瞬态下的结构完整性进行了分析评估。分析结果表明,40年寿期内,关注区域不会发生脆性断裂失效,但要关注冷却剂温度变化速率大的瞬态。     

Study on models for RPV material irradiation embrittlement prediction

反应堆压力容器材料辐照脆化预测模型的研究对保证核反应堆安全运行、并预防重大灾难性事故的发生具有非常重要意义.本文基于RPV材料中Mn-Mo-Ni及Cr-Mo-V两种钢系的应用实际,分析了适用于RPV两类不同材料辐照脆化预测模型,研究了这些模型的物理思想和建模方法.首次提出了参数化模型和结构化模型的概念,充分肯定了参数化模型在反应堆压力容器材料辐照脆化预测方面的重要作用,并对结构化模型的发展前景及深入研究所面临的问题进行了讨论.     

秦山核电二期工程反应堆压力容器支承设计

根据总体设计要求和设计输入要求,进行了秦山核电二期工程反应堆压力容器支承的结构设计和力学设计.介绍了该反应堆压力容器支承结构的材料及性能指标.通过分析计算表明,该设计满足总体设计的要求.     

反应堆压力容器快中子注量计算方法研究

依据美国核管会有关压力容器中子注量确定方法的管理导则,在二维SN程序DORT的基础上,通过自行开发X-Y与R-θ坐标系间的中子源转换程序,建立起了一套基于综合法的压力容器三维快中子注量计算方法。NUREG/CR-6453和NUREG/CR-6115基准问题检验结果表明,本文所建立的方法能够提供满足工程精度要求的压力容器注量空间和能谱分布。     

Strategies and dismantling solutions for RPV internals of Trino NPP

The last step of nuclear power plants’ life is the decommissioning phase. Many strategies have been developed in the last decades; the main options consists in: immediate dismantling, safe storage and entombment.In Italy, due to the premature shutdown of nuclear power plants (NPP) as a consequence of the 1986 referendum following the Chernobyl accident, all the NPPs were shut down. Therefore, currently decommissioning activities are under way. In this work specifically the dismantling procedures of the “Enrico Fermi” NPP reactor pressure vessel (RPV) are dealt with. The attention is so focused on the optimization of number, position and length of the Internals cuts, optimization of volume and number of waste containers, and of the dose rates, according to the imposed site and the transport requirements.The systematic approach developed to identify the optimum dismantling solution is presented and discussed. To the intent three solutions have been identified as more advantageous. In addition, the evaluation of dose rate outside the container has been performed by means of VISIPLAN software, in order to guarantee the respect of the limits imposed by National and International regulations. The obtained results suggest that the cutting of Zircalloy elements in correspondence of the gap between two non-consecutive central stumps, with the possibility to re-arrange the element, led to a significant reduction of the number of containers, with consequent decrease of the stowed RWs volume and overall supplying costs of the activity.     

脉冲堆堆水池的建造

堆水池的建造质量对脉冲堆的安全起着重要作用。在堆水池的建造过程中,通过对不同的砼配比进行试验,以普通砼代替了重砼。另外,在砼的浇灌时还采取了一系列措施,保证了水池的精确定位与不变形。建造质量符合设计要求。