共查询到19条相似文献,搜索用时 62 毫秒
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研究遗传算法在HFETR堆芯优化设计中的应用。计算程序中部分参数采用岷江试验堆堆芯优化程序中已采用的参数,计算方法上引入了对称变异算子与虚拟堆芯技术。虚拟堆芯技术的引入减少了获得优化方案的计算时间,而对称变异算子的应用增加了优化方案的工程实用性。从优化结果可以看出,转换比与同位素总产量都可提高10%以上,优化堆芯装载方案对称实用。满足工程要求。 相似文献
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反应堆堆芯入口流量分配是反应堆水力性能研究的重要内容之一,其与堆芯热裕量和燃料组件燃料棒的流致振动密切相关,从而影响反应堆的运行。CAP1400反应堆堆芯入口流量分配试验是验证CAP1400反应堆结构设计与分析的一个重要环节,旨在验证CAP1400反应堆堆芯入口流量分配的均匀程度。本文通过1/6比例模型试验,获得无均流板结构工况和带均流板结构3种工况(均匀流量工况、非均匀流量工况、偏回路流量工况)下CAP1400反应堆堆芯入口流量分配结果,并进行了各工况下流量分配均匀程度的分析。试验结果表明,CAP1400反应堆堆芯入口具有较好的流量分配效果。 相似文献
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基于PAnySimu仿真支撑系统对PWR核电站一回路堆芯部分进行建模与仿真分析.通过研究分析岭澳二期3/4号机组堆芯实际结构,将其分为功率计算、堆芯传递计算、控制棒引起的反应性、反应性反馈、毒物计算五个模型.在此基础上,分析堆芯中子通量,考虑控制棒位置、燃料和慢化剂温度、氙和钐中毒、硼浓度等因素对中子通量的影响.利用P... 相似文献
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【日本《原子能视野》1999年10月刊第64~67页报道】1.整个堆芯采用MOX燃料的ABWR堆芯概要1.1基本考虑整个堆芯都采用MOX燃料的ABWR的燃料与堆芯设计的基本方针是,不改变以前ABWR的热功率、燃料组件的件数、控制棒的根数等基本规格,也不对以前的装铀燃料堆芯做大的变更。MOX燃料组件的基本构造与到目前为止已取得很好实绩的高燃耗8×8铀燃料(II级燃料)一样,在燃料组件的中央配置一根大口径挤水棒,在挤水棒周围配置排成8列8行的60根燃料棒。关于堆芯装MOX燃料组件问题,起初先装0~264根MOX燃料,然后阶段性地逐渐增加MOX燃料的比… 相似文献
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对压水反应堆分别采用反应堆平衡循环寿期末和燃耗包络两种计算方法计算堆芯积存量,对比结果的差异。结果表明:各种核素受计算方法的影响程度不同,83mKr、135Xe和138Cs等10余个核素受影响较大,燃耗包络法计算结果更为保守,其余核素受影响较小。核素放射性活度随着反应堆运行时间的增长可分为核素活度逐渐增加、核素活度先增大后趋于稳定、核素活度逐渐减小和核素活度先增大后减小等不同的变化规律。 相似文献
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介绍了在实验性PWR堆上完成的深燃耗条件下测量反应性概况。用实验结论剖析了在国外核电站堆芯上应用噪声分析法对慢化剂温度系数作全燃耗期监测研究中出现的测量结果与常规方法相差2~5倍的现象。从测量公式和堆芯扰动模型图入手所作的分析结果说明,没有消除随燃耗不断增长的强自发裂变中子源干扰是产生差异的根本原因。事实说明:在多种噪声分析技术中,只有能够清除自发裂变中子源干扰的方法才能成功地应用于燃耗后堆芯的反应性测量。 相似文献
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蒙特卡罗程序已经广泛应用在裂变反应堆设计和验证过程中,快速获得高效的计算模型可以有效缩短反应堆的设计周期。本研究提出并实现了一种裂变堆芯快速蒙特卡罗建模的方法,该方法基于参数可视化和层次化两种建模思想快速构建出精细裂变堆芯计算机辅助设计(Computer Aided Design,CAD)模型且将其快速转换成蒙特卡罗计算模型,同时采用一种新的堆芯分段管理方法实现了大规模裂变堆模型流畅交互。基于此方法快速构建了加速器驱动次临界反应堆(Accelerator Driven Sub-critical System,ADS)的精细堆芯模型,通过与蒙特卡罗程序计算的参考结果进行对比,证明了此建模方法的高效性和可靠性。 相似文献
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基于国际经典的压水堆全堆芯Hoogenboom基准模型,对超级蒙特卡罗核计算仿真软件系统SuperMC(Super Monte Carlo Simulation Program for Nuclear and Radiation Process)进行了校验。对有效增殖因数keff、功率等反应堆关键参数进行了计算与正确性校验,对并行效率进行了分析。结果显示,SuperMC的计算结果与MCNP(Monte Carlo N Particle Transport Code)吻合较好,在使用640核计算时并行效率高达98.7%,初步验证了SuperMC在全堆芯计算中的准确性及高效性。 相似文献
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提出了一种新型的超临界水堆概念设计:混合能谱超临界水堆,它包括慢谱区和快谱区两部分.其慢谱区燃料组件采用双排燃料组件,快谱区采用简单的正方形栅元燃料组件.慢谱区与快谱区的燃料组件都采用同向流动方式来简化堆芯设计.慢谱区的冷却剂出口温度远低于整个堆芯的出口温度,这大大降低了慢谱区包壳的温度峰值.此外,由于快谱区冷却剂密度很小,流速很高,故可采用较大的栅元结构,这有效地降低了包壳周向局部传热不均匀性.所以混合堆在充分继承慢谱、快谱堆芯优点的基础上,弥补两者的不足. 相似文献
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Verification of nuclear data and codes is highly recommended for reactor safety analyses. In this research, MTR-PC package and MCNP 4C are used as deterministic and Monte Carlo simulation code, respectively. They are taken into account safety parameters as a function of control rods. Control absorbers classified in two different types. The first called Shim Safety Rods (SSR) made of an Ag–In–Cd composition, and the second one is the Fine Regulating Rod (FRR) made of Stainless Steel. One startup test of a 5 MW Material Testing Reactor (MTR) is simulated throughout the 3-D core modeling. It checks the overall simulating performance. Reactivity worth effects of control rods are calculated and benchmarked against rod-drop experimental results. Both types of codes are taken into account the integral and differential behavior of reactivity worth effects. The startup state has been simulated very carefully, and also total reactivity worth effect of control rods differs from the measured value less than 1.1% (in pcm). Results are in good agreement and highly reliable to calculate the total reactivity worth effect of control rods, and to simulate the startup state of MTRs as Best Estimate (BE) tools. 相似文献
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In the fine-grain TL dating the full α dose must be converted into the equivalent β dose.The conversion is finished by Keff-value,which is an effective α effectiveness.But the Keff can not be measured directly for each sample and only the external radiative efficiency K3.7 can be measured.In order to obtain the Keff a special study for the conversion factor of Keff to K3.7 has been made using the ultrathin TLD.The results show that the conversion factor of the TLD for archaeological samples is 0.847,which is in agrcement with calculated value 0.85. 相似文献
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The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out. Crucial development issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. As a result, it has been confirmed that the sodium-cooled FR concept is highly suited to the development targets and R&D issues are related enhancing the economy with certain perspectives for realization. A flexible and robust development program for the FR cycle system will be proposed taking account of the characteristics for each FR concept until the end of the Phase II study. 相似文献
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《核技术(英文版)》2016,(3):196-202
The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out. 相似文献