Critical Facility for lattice physics experiments for the Advanced Heavy Water Reactor and the 500 MWe pressurized heavy water reactors

Bhabha Atomic Research Centre (BARC), Mumbai, is embarking on a broad based program for thorium utilization in power production to achieve all-round capability in the entire thorium cycle. As a step in this direction, a low power Critical Facility is under construction at BARC. The facility will greatly contribute to the understanding and validation of the calculational models and nuclear data used in the design of thorium based Advanced Heavy Water Reactor. The facility is also designed to cater to the experimental requirements of future lattice studies related to 500 MWe pressurized heavy water reactors. This paper covers the basic design features, safety aspects and the planned experimental program of the new facility.     

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.     

Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

重水堆核电厂典型严重事故氢气风险分析

核电厂严重事故下的氢气控制一直是核电厂关注的热点问题之一。本文采用重水堆一体化事故分析程序建立了主热传输系统(PHTS)模型、排管容器及端屏蔽系统、堆腔以及安全壳模型。分别选取代表高压熔堆和低压熔堆的全厂断电及出口集管大破口失水事故始发严重事故序列,从堆芯氧化产氢以及系统热工水力行为出发,对重水堆产氢特性及点火器的消氢效果进行了研究。分析表明:严重事故下随着堆芯冷却恶化,排管容器内发生锆水反应而产生氢气,排管容器和堆腔内的水对氢气产生有较长时间的抑止作用,随着排管容器和堆腔内水的逐渐烧干,排管容器蠕变失效,熔融堆芯落入堆腔发生堆芯熔融物与混凝土的相互作用而产生大量氢气。当氢气点火器失效时,安全壳隔间内氢气体积份额持续增加,存在燃爆风险;点火器开启时,隔间中的氢气混合气体在较低浓度下点燃,氢气燃烧模式处于慢速燃烧区。     

压水堆核电站锆水反应微观机理

压水堆核电站严重事故下的氢气行为研究需建立氢气生成的动力学模型,氢气生成反应的微观机理和反应速率常数是建立动力学模型的基础。本工作采用量子化学理论,应用量子化学软件包Gaussian03,在HF/3-21G理论模型上研究了压水堆严重事故下锆水反应的微观机理,并计算了反应速率常数。计算结果表明,锆水反应是由4个基元反应组成的总包反应。第2步基元反应的正反应速率最小,是锆水反应的速控步。在微观上研究减少或控制氢气生成的措施应从第2步基元反应入手。文中的计算结果偏于保守,以该方法建立起的动力学模型模拟压水堆核电站严重事故下的氢气行为是安全的。     

Particle Swarm Optimization applied to the nuclear reload problem of a Pressurized Water Reactor

The concept of Swarm Intelligence is based on the ability of individuals to learn with their own experience in a group as well as to take advantage of the performance of other individuals, which are social–collaborative aspects of intelligence. In 1995, Kennedy and Eberhart presented the Particle Swarm Optimization (PSO), a Computational Intelligence metaheuristic technique. Since then, some PSO models for discrete search spaces have been developed for combinatorial optimization, although none of them presented satisfactory results to optimize a combinatorial problem such as the Nuclear Reactor Reload Problem (NRRP). In this sense, we have developed the Particle Swarm Optimization with Random Keys (PSORK) to optimize combinatorial problems. PSORK has been tested for benchmarks to validate its performance and to be compared to other techniques such as Ant Systems and Genetic Algorithms, and in order to analyze parameters to be applied to the NRRP. We also describe and discuss its performance and applications to the NRRP with a survey of the research and development of techniques to optimize the reloading operation of Angra 1 nuclear power plant, located at the Southeast of Brazil.     

压水堆核电机组一回路抽真空排气技术

压水堆核电机组一回路在调试启动期间,一回路系统充水时需要排气,与传统的排气方法相比,采用一回路抽真空排气技术不仅可以缩短试验工期,还能大大降低主泵首次启动时的风险。一回路抽真空排气技术在国内中国改进型百万千瓦级压水堆(CPR1000)机组调试启动中应用广泛,经过多次试验和不断改进,该技术已经成熟。本文对一回路抽真空排气技术进行了介绍。     

Full-scale aircraft impact test for evaluation of impact force

Previously, estimates of the force caused by aircraft impact into rigid structures have been determined using theoretical methods based on the aircraft's calculated mass and strength distribution. However, these methods required many assumptions to be made and they left many questions unanswered. Because of the uncertainty involved in these analytical predictions of impact force, a full-scale aircraft impact test was performed and an extensive set of response measurements was analyzed to evaluate the impact force against a rigid target. The analysis and evaluation gave an accurate impact force-time curve under the test condition and confirmed the practical use of an existing analytical method. An analysis with a lumped mass spring model also provided good agreement with the test results.     

Preliminary evaluation of aircraft impact on a near term nuclear power plant

For the assessment of the safety and durability of a nuclear power plant (NPP), the containment building behaviour shall be evaluated, under various service and extreme conditions, both natural or produced by natural accident or vicious man activities, like September 2001 jet aircraft crashes.The aim of this paper is to preliminary evaluate the effects and consequences of the energy transmitted to the outer containment walls (according to the international safety and design code guidelines, as NRC or IAEA ones) due to a military or civil aircraft impact into a nuclear plant, considered as a ‘beyond design basis’ event.To perform reliable analysis of such a large-scale structure and determine the structural effects of the propagation of this types of impulsive loads (response of containment structure), a realistic but still feasible numerical model with suitable materials characteristics were used by means of which relevant physical phenomena are reflected. Moreover a sensitivity analysis has also been carried out considering the effects of different containment wall thickness and reinforced/prestressed concrete features. The obtained results were analysed to check the NPP containment strength margins.     

Functional reliability analysis of Safety Grade Decay Heat Removal System of Indian 500 MWe PFBR

Passive systems are increasingly deployed in nuclear industry with an objective of increasing reliability and safety of operations with reduced cost. Methods for assessing the reliability of thermal-hydraulic passive systems, that is systems with moving working fluid, address the issues in natural buoyancy-driven flow that could result in a failure to meet the design safety limits under accident scenarios. This is referred as design functional reliability. This paper presents the results of functional reliability analysis carried out for the passive Safety Grade Decay Heat Removal System (SGDHRS) of Indian Prototype Fast Breeder Reactor (PFBR). The analysis is carried out based on the overall approach reported in the Reliability Methods for Passive System (RMPS, European Commission) project. Functional failure probability is calculated using Monte-Carlo method and also with method of moments.     

秦山重水反应堆钒探测器首次更换后物理验证试验的设计与实施

随着钒探测器中子辐照时间的增加,探测器的灵敏度将降低,甚至失效。因此,在运行一定时间后,需要进行钒探测器的整体更换。文章主要对CANDU6反应堆钒探测器首次更换后相关物理验证试验的设计与实施展开讨论和分析。通过验证和分析,最终结果符合要求。     

大功率压水堆PCS水分配试验系统设计与实现

针对大功率非能动安全壳基准事故下的水流特征,采用和原型安全壳相同尺寸比例及切片形式,设计了椭球扇面试验台架装置和相应的测量系统以研究安全壳穹顶水膜覆盖率和延迟时间等关键参数与冷却水流量之间的关系。同时开发了大空间曲率表面的视频测量系统,通过电容探针及其三维可调节支架系统实现了本体各处的水膜厚度非接触式测量,并对关键测量系统进行了标定。初步分析结果表明,试验本体及回路设计合理可行,获得了水膜覆盖率和相对延迟时间随雷诺数的变化关系。     

压水堆核电站硅含量异常升高原因分析

采用静态浸泡法研究了玻璃纤维滤芯在硼锂水溶液中SiO2的浸出行为及浸出机理,探讨了温度、pH值、过滤精度对SiO2浸出行为的影响。实验结果表明:SiO2浸出率随浸泡时间的延长而迅速减小,60d后逐渐趋于稳定值,表明滤芯表面生成一层胶体保护膜。浸出速率随温度的升高显著增大,过滤精度为0.6μm的滤芯在浸泡最初的4d,298K时的浸出速率(0.009 8g/(m2·d))较343K时的浸出速率(0.204g/(m2·d))约低1个数量级,说明SiO2浸出过程不是单一的离子扩散反应或网络溶解反应,而是两者共同参与或多种机制共同控制的复杂过程;相同条件下,玻璃纤维的过滤精度越高,浸出SiO2的速率越大。     

先进压水堆核电厂保护系统需求分析的层次结构

为提高核电厂的安全性和运行裕量，本工作在已有技术的基础上自主进行核电厂数字化保护系统需求分析,完成需求分析报告。需求分析报告采用1种三等级的金字塔式层次结构，该结构可直观阐明先进压水堆核电厂数字化保护系统的设计特性和逻辑实现。     

中子能谱对压水堆慢化剂温度系数的影响分析

在压水堆中,水铀比和²³⁵U富集度是影响中子能谱分布的重要参数。本工作在不同水铀比、²³⁵U富集度下分析两群中子能谱随燃耗的变化。利用中子能谱分布对慢化剂温度系数的变化进行分析,结果表明：在给定²³⁵U富集度条件下,随着水铀比的变化,堆芯存在一慢化剂温度系数绝对值最大值;²³⁵U富集度的增加、燃耗的加深,不一定导致慢化剂温度系数绝对值增大。     

先进压水堆"C"形环研究

"C"形环是一种用于压力容器法兰密封的密封环,目前国内使用的为进口"C"形环.为了研制出性能良好的国产"C"形环,本课题对试制的"C"形环密封性能进行了深入研究,并在"C"形环的制造工艺上取得了突破.研制的样环通过了冷热态综合性能试验,结果表明其研制的工艺合理,试制的"C"形环密封性能良好.此项研究结果为研制工程用"C"形环打下了基础.     

压水堆核电厂装换料操作参数化建模研究

为了优化压水堆核电厂装换料工艺流程,提高装换料效率,分别建立了装卸料机、水下燃料运输系统、燃料抓取机等设备的运动学模型,并根据模型推导出计算装换料操作总时间的初步方程,形成反应堆装换料操作流程参数化分析的方法。通过对某典型堆型进行参数验证,证明了该模型和方法的正确性和有效性。     

大型压水堆机械补偿控制策略仿真研究

机械补偿（MSHIM）运行的优点之一是实现了堆芯功率和轴向功率偏移（AO）在控制手段方面的部分解耦,但原始控制策略设计并未充分利用该优点。本研究通过理论分析提出了一种新的改进型MSHIM控制策略,同时基于节点反应堆模型开发了MSHIM控制系统仿真平台,并利用该平台对西屋公司原始控制策略、西屋公司Drudy的改进控制策略和本研究提出的改进控制策略进行仿真研究和比对。结果表明,本研究提出的改进型MSHIM控制策略能够显著地提高AO的控制精度,并能减少控制棒的移动,明显地改善了AP1000核电机组的运行效果,可在工程中参考使用。      

燃耗对船用压水堆温度系数的影响效应分析

根据船用压水堆临界棒位、固体可燃毒物以及核燃料物理性能随燃耗的变化规律,分析了这些参数变化对反应堆温度系数的影响,得出船用压水堆温度系数随燃耗的变化规律,即在整个燃耗寿期内,船用压水堆具有负的温度系数,但随燃耗的加深温度系数的绝对值将逐渐减小.