低温核供热反应堆安全壳

低温核供热反应堆采用一体化布置。主热交换器、控制棒驱动机构和乏燃料与堆芯一起置于反应堆压力壳内。主冷却剂靠自然循环流经堆芯和主热交换器。压力壳的冷却剂穿管 都是小口径的。反应堆压力壳构成了主冷却剂压力边界的主体。这样,反应堆就可以采用紧贴式的小体积金属安全壳。该安全壳的设计原则和功能与压水堆的安全壳相同,但由于它的体积很小,可以带来一些有利之处,如制造周期短、造价低、容易实现密封等,并且在安全壳内出现主冷却剂压力边界破口时,不会导致反应堆失冷事故。     

低温核供热反应堆安全壳(英文)

低温核供热反应堆采用一体化布置。主热交换器、控制棒驱动机构和乏燃料与堆芯一起置于反应堆压力壳内。主冷却剂靠自然循环流经堆芯和主热交换器。压力壳的冷却剂穿管都是小口径的。反应堆压力壳构成了主冷却剂压力边界的主体。这样,反应堆就可以采用紧贴式的小体积金属安全壳。该安全壳的设计原则和功能与压水堆的安全壳相同,但由于它的体积很小,可以带来一些有利之处,如制造周期短、造价低、容易实现密封等,并且在安全壳内出现主冷却剂压力边界破口时,不会导致反应堆失冷事故。     

IRIS反应堆核电厂总平面布置

IRIS(国际革新安全反应堆)是一种轻水冷却、电功率为335MW的堆型，由美国能源部核能研究组领导下的—个国际联盟进行设计。IRIS所特有的一体化反应堆容器，包容了所有的反应堆主冷却剂系统设备(包括反应堆堆芯、冷却剂泵、蒸汽发生器和稳压器)。这种一体化设计方案取消了大型冷却剂环路管道，消除了失冷事故(LOCA)以及分体式设备的压力容器及支撑。此外，IRIS被设计为长寿命堆芯，并提高安全性来满足美国能源部为第四代反应堆所确定的要求。Bechtel公司在西屋公司的咨询帮助下，对IRIS电站进行了布置研究，本文将对此设计努力的结果进行介绍。     

AP1000核电厂反应堆冷却剂系统布置设计

AP1000核电厂反应堆冷却剂系统布置设计,在满足系统功能的前提下,充分考虑了屏蔽防护、核级部件在役检查、模块化设计、内部灾害防护等方面的要求。反应堆冷却剂系统主设备及主回路采用了紧凑型的布置方式,改善了环路配置的经济性,波动管布置在考虑足够柔性的基础上采用了大倾斜角连续上坡的方式,降低了波动管在运行过程中出现热分层的可能性,稳压器安全阀及ADS第1、2、3级集中布置在稳压器顶部,组合成一体化的模块Q601,改善了反应堆冷却剂系统布置结构。     

海阳核电一期工程严重事故下非能动安全壳冷却能力分析

采用一体化事故分析程序建立了包括主冷却剂系统、专设安全设施、安全壳系统和非能动安全壳冷却系统(PCS)的海阳核电一期工程核电厂模型,对核电厂压力容器直接注射(DVI)管线破裂、冷段双端断裂、自动卸压系统(ADS)误启动、热段2英寸破口等严重事故序列进行了模拟计算,分析反应堆系统的热工水力行为。并通过安全壳系统的压力和温度响应,分析了非能动安全壳冷却系统在严重事故工况下的冷却能力。计算表明,对于分析的严重事故工况,在72h内,PCS的冷却能力能够保持安全壳内压力和温度处于较低水平,可以保障安全壳完整性。分别针对PCS水膜覆盖率以及环境温度对PCS冷却效果进行了敏感性分析,表明水膜覆盖率降低和环境温度升高均会使PCS冷却能力降低,安全壳内压力升高,但均未超出其设计压力。     

余热排出系统隔离阀配电方案研究

"华龙一号"工程项目的余热排出系统从安全壳内移到安全壳外,从而在母管位置增加了安全壳隔离阀.由于安全壳外的隔离阀设置在母管上,而安全壳内的隔离阀和与反应堆冷却剂系统的隔离阀都设置在支管上,因此这些阀门如何选择供电列保证安全是一个重要的问题.本文详细分析余热排出系统5个隔离阀的在不同的工况下需要执行的功能.为了满足这些功...     

国际革新与安全反应堆的蒸汽发生器

IRIS(国际革新与安全反应堆）是一种轻水冷却、335MWe动力堆，一个国际联盟正进行设计，它是美国能源部（DOE）NERI项目的一部分。IRIS的特点是具有一体化的压力容器，它容纳了反应堆的所有主要冷却剂系统部件，包括堆芯、冷却剂源、蒸汽发生器和稳压器。这种一体化设计取消了大的冷却剂管路系统，因而消除了大破口失水事故（LOCAs），并去掉了一些独立部件的承压壳及其支撑。另外，IRIS被设计成长寿命堆芯并增强了安全性，以达到美国DOE对第四代反应堆定义的要求。反应堆压力容器内置蒸汽发生器的设计，是一体化IRIS概念开发的一项主要设计尝试。本文的主题是正在进行的蒸汽发生器的有关设计活动。     

秦山核电二期工程反应堆及反应堆冷却剂系统设计

秦山核电二期工程电功率为2×600MW,反应堆为压水堆,两环路结构,A模式运行;堆芯平均线功率密度为161W/cm;换料方式采用年换料四分之一.反应堆冷却剂系统采用对称布置,以反应堆容器为中心,两条环路两边对称;主冷却剂系统额定流量为每条环路各24290m3/h.中国核动力研究设计院(NPIC)承担了反应堆及反应堆冷却剂系统及相关的控制、保护、仪控系统的设计与技术服务任务,并承担有关的设计验证工作.工程实行院长领导下的项目负责制,建立分工明确的组织管理机构.以中国的核安全法规、工程合同和业主要求为基础,制定质量保证大纲和设计文件清单.设计中主要采用法国RCC系列规范,系统中重要的设计结果都经过了试验的验证.各种实测值与设计分析计算值的比较表明,秦山核电二期工程反应堆及反应堆冷却剂系统设计的理论计算值与实堆的实测值符合良好.试验结果表明设备性能完善,能够满足核电站正常和事故工况下的运行要求.     

秦山核电二期工程反应堆及反应堆冷却剂系统的仪表和控制

秦山核电二期工程反应堆及反应堆冷却剂系统的仪表和控制设计参考了大亚湾核电站的设计,但作了冷却剂系统三环路改二环路的适应性修改.本文总结了秦山核电二期工程反应堆及反应堆冷却剂系统仪表和控制的设计、重要仪表控制设备的研制.具体介绍了反应堆保护系统保护变量的选取、反应堆控制系统对堆芯的控制和监测以及提高核电厂可利用率的设计,并着重介绍了重要仪表控制设备的国产化研制过程.1号机组的成功运行证明设计和研制是非常成功的.     

“华龙一号”安全特性分析

"华龙一号"采用177组先进燃料组件、先进的堆芯测量系统和反应堆冷却剂系统,提高了核电厂的固有安全性和堆芯热工裕量。在系统设计方面,配置了能动和非能动相结合的安全系统,核电机组具有完善的超设计基准事故、严重事故应对措施。"华龙一号"采用单堆布置、双层安全壳,实现了布置优化和实体隔离,有效降低了安全系统共模失效问题。这些设计使得"华龙一号"安全性达到了三代核电技术的先进水平。     

Containment for the low temperature district nuclear-heating reactor

An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happened during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety.     

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.     

Design and test of hydraulic vacuum breaker check valve for simplified boiling water reactor

In the current design of the simplified boiling water reactor, the vacuum breaker check valve is an important safety component. The vacuum breaker check valve is the only key safety components which is not passive in nature. Failure of this mechanical valve drastically reduces the passive containment cooling system cooling capability and hence containment pressure may exceed the design pressure. To eliminate this problem novel vacuum breaker check valve was developed to replace the mechanical valve. This new design is based on a passive hydraulic head, which is fail-safe and is truly passive in operation. Moreover this new design needs only one additional tank and one set of piping each to the wetwell and drywell. This system is simple in design and hence is easy to maintain and to qualify for operation. The passive vacuum breaker check valve performance was first evaluated using RELAP5. Then the passive vacuum breaker check valve was constructed and implemented in the PUMA integral test facility. Its performance was studied in a large break loss of coolant accident simulation test performed in PUMA facility.     

严重事故下氢气风险及氢气控制系统的初步分析

采用一体化严重事故分析工具,对600MWe压水堆核电厂严重事故下氢气风险及拟定的氢气控制系统进行分析。结果表明：相对于小破口失水始发事故和全厂断电始发事故工况,大破口失水始发严重事故堆芯快速熔化,在考虑100%锆 水反应产氢量的条件下,大破口失水始发事故氢气风险较大,有可能发生氢气快速燃烧;在氢气控制系统作用下,发生大破口失水始发严重事故时,安全壳内平均氢气浓度和隔间内氢气浓度低于10%,未达到氢气快速燃烧和爆炸的条件,满足美国联邦法规10CFR中关于氢气控制和风险分析的准则,认为该氢气控制系统是可行、有效的。     

Shielding Design to Obtain Compact Marine Reactor

The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water- filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the AN ISN, DOT3.5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses.     

Thermal-hydraulic and stress analysis of AP1000 reactor containment during LOCA in dry cooling mode

Some kinds of break in the reactor coolant system may cause the coolant to exit rapidly from the failure site,which leads to the loss of coolant accident (LOCA).In this paper,a stress analysis of an AP1000 reactor containment is performed in an LOCA,with the passive containment cooling system (PCCS) being available and not available for cooling the wall's containment.The variations in the mechanical properties of the wall's containment,including elastic modulus,strength,and stress,are analyzed using the ABAQUS code.A general two-phase model is applied for modeling thermal-hydraulic behavior inside the containment.Obtained pressure and temperature from thermal-hydraulic models are considered as boundary conditions of the ABAQUS code to obtain distributions of temperature and stress across steel shell of the containment in the accident.The results indicate that if the PCCS fails,the peak pressure inside the containment exceeds the design value.However,the stress would still be lower than the yield stress value,and no risk would threaten the integrity of the containment.     

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.     

A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations.An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance.This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels.     

华龙一号非能动安全壳冷却系统热工水力分析

本文采用不可压缩流体均匀流模型对华龙一号(HPR1000)的非能动安全壳冷却系统（PCS）进行数值模拟,在反应堆冷却剂系统（RCS）大破口丧失冷却剂事故（LOCA）工况下对PCS进行热工水力分析,并对PCS设计工况进行性能分析计算。结果表明：PCS的非能动运行特性与事故进程具有很好的匹配能力,能在事故早期极快启动,并在24 h内将安全壳的温度和压力稳定在安全范围内。通过PCS设计工况的换热性能分析,PCS在运行5 h后进入两相流传热阶段,当换热水箱介质达到饱和温度后仍能长期稳定运行,导出安全壳内热量。     

Hydrogen and steam distribution following a small-break LOCA in large dry containment

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations. Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident （LOCA） without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump （RCP） compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.