A study of the pressure vessel steel of the WWER-440 unit 1 of the Kozloduy nuclear power plant

A comparison between highly neutron irradiated samples from the region of weld № 4 and low irradiated samples from weld № 1 taken from the pressure vessel of the WWER-440 Unit № 1 of the Kozloduy NPP has been performed. Measurements of the residual activity of samples from the outer surface of the reactor pressure vessel bottom corpus reveal very low activity of ⁶⁰Co. Insofar as there the base and weld metal appear to be exposed to a very low neutron fluence, the samples from these locations can be considered as practically not affected and may serve as a reference basis for comparison with highly irradiated pressure vessel regions. The Mössbauer parameters isomer shift (IS) and quadrupole splitting (QS) were found to be absolutely irradiation insensitive. A stepwise reduction of the internal hyperfine magnetic field B_hf, each by about 2.6 T, was observed. This can be attributed to the replacement of one or two surrounding iron atoms as first nearest neighbors by non-iron alloying atoms. The Mössbauer experimental line widths for irradiated and non-irradiated samples are practically the same, which is a quite unexpected result. The area fraction ratio for the three main Zeeman sextet subspectra S1:S2:S3 shows very high irradiation sensitivity. For the bottom low irradiated region of the reactor vessel the values are S1:S2:S3 = 50.1:40.0:9.4. After seven years of operation between the pressure vessel annealing in 1989 and the autumn of 1996 when the samples from weld № 4 were taken the ratio changes strongly to S1:S2:S3 = 56.4:34.7:8.5. A possible explanation of this result is that neutron irradiation gives rise to a precipitation process involving predominantly alloying atoms as Ni, Mn, Cr, Mo and V which become mobile and precipitate in the form of carbides and/or P-rich phases and alloying atom aggregates. This “refinement” process lowers the partial area of subspectra S2 and S3 where alloying atoms are involved and leads to a higher area fraction of the pure iron component S1, which is the major experimental result. For a more complete Mössbauer investigation on the processes of generation of structure defects caused by the neutron fluence, a new series of measurements will be performed by using a set of so-called surveillance specimens with different irradiation histories which are available only for the WWER-1000 reactors of the Kozloduy NPP.     

Main results of study on the interaction between the corium melt and steel in the VVER-1000 reactor vessel during a severe accident performed under the MASCA project

The interactions that take place in the corium melt in the reactor vessel in the case of a severe accident at a nuclear power plant were investigated in accordance with the MASCA international program. Results of the interaction between the oxide melt and iron (steel), partition of the main components [U, Zr, Fe (stainless steel)] between the oxide and the metal phases of the melt, partition of low-volatile simulators of fission products between the phases of the stratified core melt pool, and impact of the oxidizing atmosphere on the melt stratification are presented. The results obtained were used for prediction of thermodynamic properties of the melts belonging to the U-Zr-Fe-O system.     

Application of neutron stress diffractometry for studies of residual stresses and microstrains in reactor pressure vessel surveillance specimens reconstituted by beam welding methods

The high-resolution neutron-diffraction technique is used to determine the residual stresses and microstrains in unirradiated reactor pressure vessel surveillance specimens reconstituted by means of different welding methods. Comparative analysis of the results demonstrates that the lowest level of residual stresses is observed in the specimens reconstituted via electron-beam welding. The level of microstrains thereof is maximum, indicating a high dislocation density in the material.     

聚变实验堆真空室壳体成型残余应力分析

为了了解聚变实验堆真空室壳体表面残余应力的分布以及退火工艺对残余应力的影响,通过模拟分析和实验检测两种方式对不锈钢316LN冷压曲面和热压曲面残余应力进行研究,获得退火前后曲面表面残余应力的大小,得到冷压曲面和热压曲面残余应力的分布以及退火工艺对残余应力分布的影响。研究结果为分析成型工艺提供数据支撑,对中国聚变工程实验堆真空室的研究与制造具有重要意义。     

Allowance for the burnup of a rhodium self-powered neutron detector in a VVER-1000 reactor

Methods for determining the dependence of the burnup of a rhodium self-powered neutron detector (SPND) from the charge flow in it are described.     

Micro-segregation model calculation of residual tin in boiler and pressure vessel steel

Tin is a typical residual element in steel and mainly originates from Sn-containing complex iron ore and steel scrap. The segregation of Sn in steel is harmful to the performances of steel. In this paper, the micro-segregation of residual element Sn during the solidification process of boiler and pressure vessel steel by micro-segregation model was studied. The results showed that the micro-segregation degree of Sn reduces apparently with the increase of cooling rate and remarkably deteriorates during the solidification process. When the initial content of C is higher than 0.1%, it will cause the solidification transform of the solid phase converting from the ferrite phase to austenite phase and the significant increase in the micro-segregation degree of Sn. However, increasing the initial contents of Si, Mn, P and S separately has non-significant effects on the micro-segregation degree of Sn. In addition, the improvement of initial content of Sn will lead to the micro-segregation degree decrease of Sn and has an inapparent impact on zero strength temperature and zero ductility temperature of the boiler and pressure vessel steel.     

A full-field method for measuring residual stresses in optical fiber

A method for measuring the axial residual stress profile in axisymmetric optical fibers is presented. The procedure is based on integrated photoelasticity and a fringe shifting technique is used to measure the optical retardation. The radial distribution of the axial residual stress is reconstructed using the inverse Abel transform. The paper describes the operating principle, the experimental setup and the results obtained on a multimode fiber are also reported. The influence of the measurement errors is finally discussed.     

Irradiation hardening of reactor pressure vessel steels due to the dislocation loop evolution

The irradiation hardening of reactor pressure vessel steels due to the formation of dislocation loops is analyzed. The analysis is based on the original model for the nucleation and subsequent evolution of dislocation loops in irradiated materials. The loop formation in displacement cascades is taken into account, along with the homogeneous clustering of point defects. The loop evolution is shown to contribute mainly to the athermal component of the yield stress, which is determined by interaction of gliding dislocations with strong barriers. Irradiation-induced hardening is evaluated as a function of irradiation dose and temperature, dose rate, material parameters and initial microstructure. The model results are compared with experimental data for neutron irradiated pressure vessel steels of various grades and with empirical low power expressions of the yield stress increase with increasing irradiation dose.     

真空规管对低温容器残余气体影响的实验研究

以高真空多层绝热低温容器为研究对象,利用残余气体分析仪研究冷阴极真空规管、热阴极真空规管以及热阴极真空规管不同放置位置对低温容器真空夹层内残余气体的影响.结果表明:冷阴极真空规管对低温容器真空夹层内残余气体的影响几乎是可以忽略不计的;热阴极真空规管对低温容器真空夹层内残余气体的影响较大,热阴极的灯丝和残余气体之间存在化...     

大型压水堆装载50%MOX燃料方案压力容器辐照安全计算

对含MOX燃料堆芯的压力容器(RPV)快中子注量率计算进行了初步研究,探讨了适用于MOX堆芯方案屏蔽计算的堆芯源项处理方法。采用三维离散纵标程序TORT,针对CAP1400型堆芯装载50%MOX燃料方案开展了RPV快中子注量计算,结果表明：堆芯装载50%MOX燃料可满足RPV屏蔽安全设计要求;对比分析含MOX堆芯方案和全UO2堆芯方案的RPV快中子注量率的特性差异,从RPV辐射防护最优化的角度,后续燃料管理方案优化时可重点关注关键位置处组件的布置。     

大型压水堆装载50%MOX燃料方案压力容器辐照安全计算

对含MOX燃料堆芯的压力容器(RPV)快中子注量率计算进行了初步研究,探讨了适用于MOX堆芯方案屏蔽计算的堆芯源项处理方法。采用三维离散纵标程序TORT,针对CAP1400型堆芯装载50%MOX燃料方案开展了RPV快中子注量计算,结果表明:堆芯装载50%MOX燃料可满足RPV屏蔽安全设计要求;对比分析含MOX堆芯方案和全UO2堆芯方案的RPV快中子注量率的特性差异,从RPV辐射防护最优化的角度,后续燃料管理方案优化时可重点关注关键位置处组件的布置。     

聚变实验堆真空室壳体成型模具设计及实验研究

参考ITER真空室制造规范,在常温下对中国聚变工程实验堆(CFETR)真空室壳体成型进行工艺预研。先根据成型经验公式计算出所研究壳体的成型模具尺寸范围,再利用有限元软件选择3组公式值附近的尺寸模拟了该成型工艺最佳模具尺寸参数,分析出最佳型面尺寸误差在±1.5mm内,并以此指导成型实验。测试了成型实验后工件减薄量、回弹量、变形率等参数,与模拟分析结果吻合度较高,验证了该成型模具设计的合理性。     

基于FLUKA计算压力容器快中子注量的可行性研究

基于知识产权的考虑,通过与蒙特卡罗程序MCNP计算结果比对,研究使用FLUKA程序替代MCNP程序进行反应堆压力容器快中子注量计算的可行性。通过修改和调用子程序对次级粒子堆栈进行操作,解决了关闭裂变中子这一关键问题,FLUKA程序的计算结果与MCNP程序的计算结果相对偏差在5%以内,符合得较好,证明使用FLUKA程序替代MCNP程序用于计算反应堆压力容器快中子注量在技术上是可行的。     

核反应堆压力容器模拟钢中富Cu原子团簇的析出与嵌入原子势计算

利用原子探针层析技术(APT)和热处理时效方法,研究了合金元素Ni对核反应堆压力容器模拟钢中富Cu原子团簇析出的影响.实验结果表明,添加合金元素Ni(0.84wt%)的样品中析出富Cu原子团簇的数量密度高于不添加Ni的样品,富Cu原子团簇内以及团簇和基体界面处都有Ni元素的富集现象,这说明合金元素Ni会促使富Cu原子团簇的析出.从多体势的角度出发,利用嵌入原子势理论,基于纯金属元素Fe,Cu,Ni的多体势参数,建立了Fe-Cu二元和Fe-Cu-Ni三元体系的嵌入原子多体势.计算结果表明,当模拟合金中存在1at%Ni时有利于富Cu原子团簇的析出,这与实验结果相符.     

Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.     

Neutron irradiation effects on magnetic minor hysteresis loops in nuclear reactor pressure vessel steels

Magnetic minor hysteresis loops have been measured on A533B-type nuclear reactor pressure vessel steels with various combinations of Cu and Ni contents after neutron irradiation to a fluence up to 3.32 × 10¹⁹ n cm^?2. A strong compositional dependence of minor-loop properties, which are indicators of internal stress, was found. The properties of high-Cu and high-Ni steel show a large increase in the low fluence regime below 0.4 × 10¹⁹ n cm^?2, followed by a slow decrease, while those for low-Cu or low-Ni steel show a sudden decrease. The changes are roughly in linear proportion to the yield strength changes. The results were explained from the viewpoint of the formation and growth of Cu-rich precipitates and/or fine scale defects in the matrix and along pre-existing dislocations.     

Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.