WWER-1000燃料组件特点及棒弯曲分析

本文根据WWER-1000反应堆的设计特点及其运行实践,阐述了WWER-1000燃料组件的设计特点,并与西方压水堆燃料组件进行了相应的比较.重点分析论述了WWER-1000反应堆燃料棒弯曲的特点,以及在热工水力和燃料组件设计中是如何考虑棒弯曲效应的,进行了燃料棒弯曲对临界热流密度影响实验的研究.结果表明:WWER-1000燃料组件在整个运行寿期内的性能是可以保证的.     

AP1000燃料组件设计特性

AP1000型燃料组件是西屋公司在40多年的燃料组件没小重行经验的基础上,改进开发的用于AP1000反应堆的高性能燃料组件。本文介绍了西屋压水堆燃料组件的设计发展,重点描述了AP1000型燃料组件的设计特点。     

WWER-1000反应堆中对燃料棒弯曲的考虑

WWER-1000反应堆无论是在俄罗斯还是在国际上都算是比较成熟的堆型，它和西方PWR反应堆都属于压水堆的范畴。但在堆芯设计的理念上，WWER-1000与PWR堆之间又存在着差异。本文结合WWER-1000反应堆的运行实践及其燃料组件的特点，来阐明燃料棒弯曲在WWER-1000反应堆堆芯设计，尤其是在堆芯热工设计中是如何考虑的，并使之与西方PWR的设计特点进行针对性比较，从而来说明WWER-1000燃料组件的性能在整个运行寿期内的适应性。     

WWER-1000型核电站SGTR事故分析

使用热工水力瞬态分析程序DINAMIKA－97，模拟WWWER－1000型核电站蒸汽发生器传热管破裂（SGTR）的事故过程，以及操作员所采取的缓解事故的动作，给出计算结果和结果分析。     

三维六角形组件压水堆堆芯燃料管理计算及程序系统研究

介绍所研制的WWER型压水堆堆芯燃料管理计算程序系统TPFAP-H／CSIM-H，六角形组件均匀化计算程序TPFAP-H是在压水堆正方形组件程序TPFAP的基础上，采用穿透概率法与响应矩阵方法相结合计算六角形组件内中子能谱分布，并考虑六角形栅元特点改造开发而成的CSIM-H是以先进六角形节块扩散程序为基础．参照SIMULATE程序功能而研制的物理-热工水力耦合的三维六角形节块PWR堆芯燃料管理程序两者通过接口程序LINK连接起来，可以考虑燃耗，功率、慢化剂密度变化．控制棒、氙等参数的多种反馈效应对IAEA的WWER-1000型Kalinin核电厂基准问题的校算的结果表明，临界硼浓度、功率和燃耗分布等结果与国际各研究机构的结果吻合良好，偏差均在工程要求之内。     

ACP1000中MOX燃料组件管理初步研究

使用SCIENCE程序包对MOX燃料组件进行了初步设计和研究。在此基础上,对采用部分MOX燃料组件的ACP1000堆芯开展燃料管理研究,得到由全堆装载UO₂燃料组件向部分MOX燃料组件堆芯过渡的燃料管理方案,并对MOX燃料组件和部分MOX燃料组件堆芯的安全参数及其他重要参数进行分析和比较。分析结果表明,各种安全参数均满足设计要求,证明在ACP1000堆芯应用MOX燃料是可行的,并为进一步研究提供了参考。     

ANSI／ANS—57．5《轻水堆燃料组件机械设计和评价》

本文对ANSI／ANS－57．5前后三个版本的规定进行了评述，从而为修订EJ／T629－92和制定我国压水堆核电厂高性能燃料组件设计标准提供了依据。     

WWER—1000反应堆替代燃料组件的发展

已为ＷＷＥＲ－１０００反应堆开发出燃料组件的燃料设计方案，目的是为了确保在运行期间堆芯的几何稳定性，这种燃料组件可以增强安全，并在ＷＷＥＲ－１０００燃料循环经济学方面的提供重大改进。     

CSR1000堆芯流量分配的研究

以中国百万千瓦级超临界水冷堆(CSR1000)堆芯为研究对象,建立热工水力计算模型,计算出冷却剂和慢化剂温度分布、堆芯功率分布、燃料组件出口压力及流量分配等参数。计算结果表明,适当增加堆芯内部燃料组件流量比例,可以有利于径向功率展平,内外燃料组件通道出口压降,呈现"N"型变化,增大内部燃料组件的堆芯入口功率,内部组件内的流量分配也将减少,而外部燃料组件通道中的流量将增加,适当调整堆芯入口流量初始分配比例,可以使各通道功率分布展平。     

AP1000型燃料组件水平放置对格架弹簧影响分析

核电厂建设的工期延误和组件制造厂的燃料组件存贮场地不足,不能按期交付的首炉燃料组件被迫存放在组件运输容器内水平放置,有的燃料组件在运输容器内水平放置时间约1年以上.针对燃料组件在运输容器内长期水平放置是否对组件格架弹簧的力学特性有影响,以AP1000型燃料组件为例,对格架弹簧进行了力学特性影响分析.分析认为,燃料组件在...     

Development of alternative fuel assembly for WWER-1000 reactor

An alternative design of fuel assembly has been developed for the WWER-1000 reactor with the aim of assuring a geometrical stability of the core during operation. The fuel assembly provides enhanced safety and substantial improvement in the WWER-1000 fuel cycle economics.     

The properties of WWER-1000 type materials obtained on the basis of a surveillance program

The surveillance test results of the reactor pressure vessels (RPV) of three Russian WWER-1000 units designated unit-1, -2 and -3 are given and the embrittlement rates compared to those predicted by the Russian Regulatory Guide. Dependence of the radiation behavior of WWER-1000 type RPV steels on metallurgical variables and the damage dose is considered. The trend curves for the steels under investigation are proposed.     

Cladding rupture life control methods for a power-cycling WWER-1000 nuclear unit

Using the cladding creep energy theory, taking into account the WWER-1000 fuel assembly four-year operating period transposition algorithm, as well as considering the disposition of control rods, the location of the axial segment limiting the fuel cladding operation time, at day cycle power maneuvering, has been found. It has been shown that the WWER-1000 fuel element cladding rupture life, at normal variable loading operation conditions, can be controlled by an optimal assignment of the coolant temperature regime and the fuel assembly transposition algorithm.     

Irradiation embrittlement monitoring of WWER-440/213 TYPE RPVs

In the paper the main goals and progress of the surveillance specimen programme for the RPVs WWER-440/213 in Jaslovské Bohunice V-2 and Mochovce NPPs are presented. At Jaslovské Bohunice V-2, the standard surveillance specimen programme (SSSP) was finished and so-called ‘Extended Surveillance Specimen Programme’ (ESSP) was prepared on the base of its critical analysis. For first two units of the Mochovce NPP completely new programmes of irradiation embrittlement monitoring called ‘Modern Surveillance Specimen Programme’ is prepared. It is based on the experience with SSSP and ESSP as well as the recommendations of IAEA experts. This programme will serve for Mochovce NPP during all planned service life. The experience of ESSP application on the 3rd and 4th units in Jaslovské Bohunice V-2 NPP are presented in the paper too.     

WWER-1000 base metal reference steel and its characterisation

The surveillance programmes of western power reactors include, in many cases, standard reference materials in addition to actual pressure vessel steels. These are specimens cut from standard steel plates (Heavy Section Steel Technology, JRQ, etc.) that are similar in composition and heat treatment to the base material in the respective reactor pressure vessels, and are supposed to serve as a monitor by comparing the radiation embrittlement of the plant-specific material to the reference material, and to detect anomalies in the radiation environment of the surveillance capsules.A correlation monitor material for the eastern WWER-1000 (similar as the JRQ for western reactors) is needed in order to determine the reliability of accelerated data for the validation of reactor pressure vessel surveillance data. Reference materials should be well characterised in terms of irradiation behaviour (transition temperature shift, non-destructive signal, etc.). The magnitude of the sensitivity to irradiation for this material should be measurable for the selected exposures. In this subject the IAEA is launching a new co-ordinated research programme. Material is already manufactured, and the JRC-IE has become its custodian. A detailed plan for characterisation of the reference steel is set up, including irradiation conditions, post-irradiation testing techniques and implementation plan. It is expected the participation of several research institutes worldwide in a round robin, which will allow a better comprehension of WWER-1000 steel's behaviour and will be considered as a benchmarking between different laboratories.The JRC-Institute for Energy in collaboration with the Russian Research Centre – Kurchatov Institute is performing the “as received” material characterisation by both destructive methods and non-destructive techniques.The non-destructive techniques used at the JRC-IE premises are novel methods specially developed for non-destructive assessment of the embrittlement state of materials, as the STEAM method and the measurement of magnetic properties. The STEAM technique (Seebeck and Thomson effects on aged material), is based on the measurement of the Seebeck coefficient. The magnetic properties evaluation is done through Barkhausen noise and permeability measurements.This paper presents a preliminary analysis of the results obtained by all involving laboratories.     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

WWER -1000型核电站N-16监测系统探测效率的模拟计算

主要针对核电站蒸汽发生器传热管破损的氮-16监测系统工程实践中遇到的探测效率无法用常规计算和实验得出的困难,以WWER - 1000堆型为基础,建立了光子与NaI探测器介质作用的简化数学物理模型,采用EGS5蒙特卡罗程序对探测器效率进行了数值计算.计算结果与工程设计资料提供的数据符合较好,为今后研究几何位置变更与探测效...     

WWER-1000堆型安全壳通风放射性监测

介绍了田湾核电站WWER-1000堆型安全壳通风系统放射陛监测的通道设置及相应的功能。通过蒸汽发生器间循环冷却系统监测通道1KLA20CR002报警的实例,表NWWER-1000堆型安全壳通风系统放射性监测具有及时警报、准确定位报警区域、方便确认报警的真实性、便于维护监测通道的特点。     

Leak Before Break application for Primary Coolant Loop and Surge Line of VVER-1000/320 Plant: TACIS Project R2.09/96

Activities were carried out within the frame of TACIS-96 Program—Project R2.09/96 named “LBB Applicability Review and Basic Implementation Engineering for Primary Coolant Loop & Surge Line of VVER-1000/320 Nuclear Power Plant”. Main objective of the Project was to perform a Leak Before Break assessment for the Main Coolant Pipes and Pressurizer Surge Line of the Reactor Coolant System of a selected “reference” Unit of a VVER-1000, type V-320 Nuclear Power Plant, which was assumed the Balakovo Unit 2.This has included the following:

• •definition of a “reference” methodology for primary piping systems with cladding;
• •verification of the Leak Before Break concept applicability to primary piping, in accordance with the “reference” methodology defined;
• •preparation of necessary materials data and validation files;
• •evaluation of effectiveness and reliability of the Leak Detection systems and the In-service Inspection Programs (methodologies and qualification);
• •preparation of a recommendations list to improve the existing information and diagnostic systems;
• •transfer of know–how from Consortium to Sub-Contractor, integrating the feedback of Western experience.

One of the main highlights of the Project was the definition and execution of a comprehensive Material Testing Program in order to establish the actual tensile and fracture properties of the primary piping materials: base materials, weld metals, and “heat affected zones” were tested. Compact tests and standard specimens for J–R curves determination were used; special attention was paid to the dissimilar typical joints of the Surge Line pipes, being made by carbon steel base metal with stainless steel weld. Fracture properties as J–R curves were determined both at room and operating temperatures.The “reference” methodology defined has taken into account both Western and Russian Leak Before Break approaches. The analysis has shown that criteria for Leak Before Break concept applicability to the primary piping of Balakovo VVER-1000/320 Unit 2, presented in the reference methods developed in the frame of this Project, are met.