专利简讯

低温供热堆或研究堆超设计基准事故停堆方法和停堆系统【公开日】2005.03.02【分类号】G21C9/02【公开号】1588559【申请号】200410069185.8【申请日】2004.07.07【申请人】中国核动力研究设计院【文摘】本发明属于核反应堆控制领域,具体说是一种低温供热堆或研究堆超设计基准事故的停堆方法和停堆系统。其特征是在堆芯内靠近中央处布置至少两个由中子吸收截面小的材料制成的空腔栅元,反应堆正常运行时,该空腔栅元内充氮气,在发生超设计基准事故时,向空腔栅元注入含硼水或去离子水,引入负反应性,实现停堆。与注含硼水的停堆方法相比,其显著…     

浮动堆抗地震设计问题

为了明确浮动堆的抗震设计准则,分析了法规标准中工程抗震设计准则的内涵和应用范围,建议采用基于性能的方法确定浮动堆的抗震设计基准。以地震资料相对丰富的渤海地区为例,计算了区域内部分场点的地震危险性。基于计算结果,初步统计了基岩水平方向峰值加速度随地震动超越概率的变化情况,为确定浮动堆设计抗震基准提供了参考。     

氟盐冷却球床堆与HTR-10和MSRE的定量相似性分析

氟盐冷却球床堆是当前国际上一种新的研究堆型,尚无已经建造完成的反应堆,因此,选择相似且具有运行经验的反应堆作为基准题有助于堆芯核设计软件适用性分析。利用国际上常采用的相似性分析软件,可对熔盐实验堆(Molten Salt Reactor Experiment,MSRE)及10 MW高温气冷堆(10 MW high-temperature gas-cooled test reactor,HTR-10)与氟盐冷却球床堆的相似性进行分析,定量判断它们作为基准题的合理性。分析结果表明,MSRE和氟盐冷却球床堆的能谱峰位能量接近且堆内元素种类相近,二者相似程度较高;常温临界HTR-10和氟盐冷却球床堆冷却剂不同,且能谱峰位能量差异较大,二者相似程度较低。因此,MSRE是氟盐冷却球床堆中子物理设计软件较理想的基准题。     

一维扩散程序NDP及其对快基准装置的分析

可靠而精确的计算方法是进行核数据基准检验必不可少的手段。一维扩散程序1DX曾准确而有效地用于ENDF/B核数据库的快堆基准检验。为此,我们在1DX基础上建立了NDP程序。可靠而精确的积分实验是核数据基准检验的积分依据。美国截面评价工作组将快堆基准做为检验ENDF/B库的主要积分依据。这不仅是因为当今核动力堆中快堆是其发展方向,而且不同能谱的基准快堆几乎复盖了核截面的整个能区,最有利于对核截面在全能区进行检验。我们用NDP程序和苏联26群邦达连柯群常数计算了美国检验ENDE/B-4     

基于SPND跳堆功能研究

反应堆保护系统的功能是保护三大核安全屏障（即燃料包壳、一回路压力边界和安全壳）的完整性。在发生设计基准工况DBC2～4工况下,反应堆保护系统自动启动,执行跳堆功能,使反应堆达到可控状态。目前在建的EPR反应堆跳堆功能,偏离泡核沸腾比低（LDNBR）和线功率密度高（HLPD）均是基于自给能中子通量探测器（SPND）测量的中子通量计算的结果。本文对EPR核电厂基于SPND跳堆功能进行了研究,进一步分析和研究反应堆保护功能的要求,以分析此设计是否满足标准法规对核电厂安全运行和审评的要求。分析结果表明,现有设计能满足标准法规的要求。     

快堆基准检验程序系统FRBT

一、引言核数据的基准检验是核数据评价工作的重要环节之一。基准检验结果是评价核数据的识分依据,是核数据不断更新的重要基础,也是核装置设计的重要依据。美国ENDF/B库各版的基准检验都以快堆基准为主要依据。这不仅因为快堆是核动力堆的发展方向,而且不同能谱的多个快堆几乎复盖了核截面的整个能区,最有利于     

裂变产额数据对铅基快堆RBEC-M燃耗计算影响研究

第四代核能系统是一种具有更好安全性、经济竞争力、核废物减少,以及防止核扩散的先进核能系统,代表了先进核能系统的发展趋势和技术前沿。铅基快堆是第四代核能系统中重要堆型之一。目前国际上通用的反应堆程序,比如MCNP+ORIGEN、RMC或者Serpent,很多研究主要针对压水堆,国际上也有研究发现针对铅基快堆基准题RBEC-M,确定论方法和蒙卡方法计算结果有较大偏差。本文深入研究了蒙卡程序使用的裂变产额对计算结果的影响。首先对反应堆蒙特卡罗程序RMC自带和燃耗库中的部分核素的裂变产额数据进行了更新,采用国际上著名RBEC-M基准题和OECD/NEA发布的快堆Pu循环燃耗基准题进行了验证分析,计算得到了裂变份额数据对快堆燃耗计算的影响。计算结果表明:更新后的裂变产额数据对系统的有效增殖因子和主要重核的质量变化影响较小,但对部分裂变产物的质量变化影响较大,部分核素偏差超过86%。对于快堆Pu循环燃耗基准题,长寿命高放废物~(133)Cs和~(129)I的计算结果偏差分别可达22.4%和47.8%,这将对长寿命高放废物的嬗变效率和核燃料循环有重要影响。     

低温供热堆或研究堆超设计基准事故停堆方法和停堆系统

本发明属于核反应堆控制领域，具体说是一种低温供热堆或研究堆超设计基准事故的停堆方法和停堆系统。其特征是在堆芯内靠近中央处布置至少两个由中子吸收截面小的材料制成的空腔栅元，反应堆正常运行时，该空腔栅元内充氮气，在发生超设计基准事故时，向空腔栅元注入含硼水或去离子水，引入负反应性，实现停堆。     

不同核评价数据库中Pb和Bi的截面对铅基快堆k_(eff)的影响

铅基快堆是GIF官方发布的第四代核能系统堆型之一,不同的核评价数据库中铅截面的较大差别会影响铅基快堆物理设计计算的准确性。本文利用国际上最新发布的核评价数据库JENDL-4.0、JEFF-3.2、ENDF/B-Ⅶ.0和BROND-3.1,制作了关键核素Pb、Bi的连续点截面,利用国际基准题评价手册中的PMF035和国际原子能机构发布的铅基快堆RBEC-M基准题以及cosRMC程序,对Pb和Bi的截面对系统有效增殖因数的影响进行了详细研究。对于PMF035带Pb反射层的临界基准题,上述所有核数据库的新版本较旧版本的计算偏差均有所减小,其中BROND的改善最为明显。对于RBEC-M基准题,使用ENDF/B-Ⅶ.0核数据库的计算结果与基准报告中结果符合较好;使用上述其他新版本数据库中截面数据替换计算结果表明,采用不同库中的Pb、Bi截面数据会使计算结果出现不同的偏差,其中,JENDL-4.0中Pb截面对计算结果的影响较Bi截面的影响大。     

西安脉冲反应堆超设计基准事故动态特性分析

为填补以往西安脉冲反应堆(脉冲堆)超设计基准事故研究的不足,利用RELAP5/SCDAP/MOD3.4程序对脉冲堆系统进行了建模计算,给出了脉冲堆在断电ATWS事故和大破口失水ATWS事故下的瞬态响应特性。计算结果表明:发生断电ATWS事故后,在无人为干涉情况下,反应堆部分燃料可能熔毁;发生大破口失水ATWS事故后,破口位置和尺寸对事故后果的严重程度有重要影响,破口位置在堆池底部时,燃料最高温度低于1 800℃,而破口位置高于堆芯下栅板时,将导致燃料元件熔毁。根据脉冲堆在超设计基准事故下的动态响应,针对两种事故工况分别提出了相应的缓解措施。     

研究堆运行许可证有效期延续申请审查策略

生态环境部第8号令《核动力厂、研究堆和核燃料循环设施安全许可程序规定》对核动力厂、研究堆和核燃料循环设施运行许可证件延续事项作出了新的规定。为推动我国研究堆老化管理标准体系建立,分析了我国研究堆延寿审查策略发展历程,结合高通量工程试验堆等研究堆运行许可证有效期延续申请审查工作中的几个关键问题,提出了以定期安全审查为主、重点依据老化管理并兼顾技术规格书审查及差异性审查的审查策略,研究成果为我国研究堆老化管理法规标准的建立提供了实践经验及理论指导依据。      

Research and Development Work on Nuclear Reactors for Power Production in Russia

The directions and content of research and development work on nuclear reactors for civilian power production, conducted in the USSR and in Russia presently, are presented. The development of technological directions of channel and vessel water-cooled reactors, fast reactors with liquid-metal coolant, gas-cooled reactors, low-capacity nuclear power for isolated users is studied.Continuity and development of new directions as factors in the present-day trends of nuclear power in Russia are discussed.     

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance.

The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

空间核反应堆电源用核燃料研制进展

核燃料是空间核反应堆电源的主要材料之一，由于空间核反应堆电源的运行条件明显有别于地面反应堆，空间核反应堆电源用核燃料的类型和技术要求也明显不同于地面反应堆。国际上空间核反应堆电源用核燃料研制取得了长足的进展，多种核燃料材料在工程应用中得到了检验，并在持续开发新型核燃料。我国在亚化学计量二氧化铀芯块、铀钼合金、铀氢锆合金、碳化铀芯块、氮化铀芯块等多种具备在空间堆中应用的燃料材料上开展了一定的研究，并掌握了部分材料性能数据。本文就上述内容展开论述，同时针对与国际相应领域明显落后的实际情况，提出了我国后续核燃料研究的初步设想。     

SIMMER model of a low-enriched uranium non-power reactor

IRSN has started using the coupled neutronics–fluid dynamics code SIMMER [Tobita, Y., Kondo, Sa., Yamano, H., Morita, K., Maschek,W., Coste, P., Cadiou, T., 2006. The development of SIMMER-III, an advanced computer program for LMFR safety analysis, and its application to sodium experiments. Nucl. Technol. 153 (3), 245] to study core-disruptive accidents induced by insertions of large reactivities to produce very short period power excursions in fuel plate-type and water-moderated experimental research reactors. Until now, French safety analyses retain a bounding thermal energy released and mechanical yields, deduced from analysis of destructive in-pile test programs, to study the behavior of such reactors and design their structures and containment.Contrary to this approach, the present research program aims at modeling the design basis accident of research reactors with a low-enriched fuel using a CFD code. The objective is to analyze the effects of reactivity feedbacks and how they would limit the generated thermal energy released in the fuel. These aspects require a close coupling of the neutronics to the fluid dynamics analysis. The consequences of the nuclear power excursion, the changes of state of the fuel and the coolant, and ultimately the mechanical energy released are calculated by SIMMER. For large step-wise reactivity introductions, the Doppler effect and, at a lower extent, the fuel element thermal dilatation, which generates locally a decrease of the moderator to fuel ratio, limit the power excursion before the energy released is high enough to melt a large part of the fuel. Moreover, it has been shown that imposing an external reactivity as a step-wise or time-dependent reactivity introduction yields results quite different from those of the physical movement of control rods.     

Performance of UO2 ceramic fuel in low-power research reactors

The Low Enriched Uranium UO₂ fuel performance in low-power research reactors is assessed in this paper. The usability of this fuel has been demonstrated in some research reactors in the world (SLOWPOKE-2). The fuel proved to be usable in the miniature neutron source low-power research reactors when about 50 fuel rods were substituted by as many dummy rods, while in SLOWPOKE reactors the number of fuel pins reduced by 98. About 3.8531 mk reactivity was rendered available at reactor start-up in MNSRs. The power of MNSRs needed to be increased by about 19%. Shut-down margin, effective shut-down margin, and control rod worth all decreased.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

使用分级方案应用研究堆应急准备与响应要求研究

我国当前的研究堆应急管理没有对不同类别研究堆的应急准备与响应作出差异性要求,分级方案是根据与反应堆相关的潜在危害正当应用这些安全要求的良好手段。按照分级方案的步骤,基于我国当前研究堆安全分类准则、国际原子能机构（IAEA）功率相关应急威胁分类准则以及应用IAEA应急准备与响应要求的分级方案的依据,提出了我国研究堆的应急管理分类准则以及对不同应急管理类别研究堆应急状态分级和应急计划区（EPZ）要求,这为简化低功率研究堆营运单位应急预案的内容和细节的范围、程度和水平以及建立与不同类别研究堆危害评定结果相称的我国研究堆应急管理系统提供了依据。      

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As an integral effect test database for major design basis accidents has been accumulated, a domestic standard problem (DSP) exercise using ATLAS was proposed in order to transfer the database to domestic nuclear industries and to contribute to improving the safety analysis technology for pressurized water reactors (PWRs). As the third DSP exercise, a double-ended guillotine break of the main steam-line at an 8% power without loss of off-site power was decided as a target scenario. Seventeen domestic organizations joined this DSP exercise. They include universities, government, and nuclear industries. The participants of DSP-03 were classified into three groups and each group has focused on the specific subject related to the enhancement of the code assessment; (1) scaling capability of the ATLAS test data by comparing with the code analysis for a prototype, (2) multi-dimensional thermal–hydraulic phenomena anticipated during the steam-line break transient, (3) effect of various models in the one-dimensional safety analysis codes.