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高温气冷堆的余热排出系统为非能动式系统,是一回路舱室冷却系统的组成部分之一。本文建立了10 MW高温气冷实验堆(HTR-10)余热排出系统在反应堆舱室内结构的三维模型,模拟HTR-10运行过程中余热排出系统的工作状况。在HTR-10上进行余热排出系统试验,获得了HTR-10在最高热功率为3 MW条件下余热排出系统的相关数据。将试验数据与模拟结果进行比对,结果表明:模拟结果与试验数据存在偏差。通过分析,提出从模型设计、工况适应性两方面对模型进行优化。 相似文献
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HTR-10主氦循环风机的设计、试验和运行 总被引:3,自引:2,他引:1
主氦循环风机是10MW高温气冷实验堆(HTR-10)的关键设备,在250℃、3.0MPa的氦气气氛下将反应堆的热能输送到蒸汽发生器。针对反应堆的特殊要求,主氦循环风机的设计包括总体结构、叶轮型式、冷却系统.轴承,测量仪表、电气贯穿件和隔断阀对设计制造的主氦循环风机进行了出厂试验和安装后的冷、热态性能试验。按照反应堆的调试要求,主氦循环风机随反应堆的调试进行了初步运行。试验和运行结果表明,主氦循环风机达到了设计要求,能满足HTR-10的运行要求。 相似文献
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10MW高温气冷堆(HTR-10)技术规格书在线监督系统以HTR-10仪表与控制系统提供的数据为基础,采用智能模拟运行人员行为的方法.实时分析判断反应堆系统和设备的工作状态是否满足技术规格书的要求,同时自动按照规定的频度完成部分定期试验和检查工作.对于不能自动完成的检查项目及时提醒运行人员,弥补了人工执行技术规格书时因工作量大、容易出现漏项的不足。本文介绍了HTR-10技术规格书在线监督系统的模型、组成模块、技术开发特点以及具体应用中的注意事项和方法。 相似文献
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10 MW高温气冷堆(HTR-10)的分散控制系统(DCS)执行对HTR-10的运行监测和控制功能.原DCS在设备可靠性、历史数据存储和转换等方面存在不足.根据HTR-10各工艺系统及控制对象的要求,分析了DCS的系统架构、功能和性能指标等;对I/O通道进行配置;提出以施耐德Quantum 67160系列产品为主要模件的多重冗余硬件平台以及分别以UNITY和iFIX作为系统软件平台和组态工具的配置方案;采用PTO模块完成对棒控和装卸料系统控制,使用智能仪表对交流采样方案进行改进,在通信网络中加设逻辑网关的办法实现第三方通信功能.该设计方案可有效解决HTR-10原DCS存在的问题,满足HTR-10对DCS的要求. 相似文献
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吸收球停堆系统是10MW高温气冷实验堆(HTR-10)的第二停堆系统,于紧急事故停堆之后、重新开堆之前投入运行,利用负压输送过程将在紧急停堆时进入反应堆堆芯落球孔道内的中子吸收球输送到位于堆顶的贮球罐内,实现正常开堆或反应堆再临界。运用气力输送的密相输送理论,对回路各部件和各管段的气固两相流阻力进行计算,并在1:1模拟试验台架上,以空气和氦气为载体,真实硼吸收球为物料,进行了气力输送试验研究。试验数据与理论分析相符合,吸收球第二停堆系统的气力输送功能满足HTR-10工程的技术要求。 相似文献
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在10MW高温气冷堆(HTR-10)氦净化系统中,设计并建造了用于取样收集一回路放射性石墨粉尘的实验系统。结合国外已有的研究结果,根据HTR-10氦净化系统的运行参数进行了模拟计算。计算结果表明,该实验系统能有效过滤收集到的放射性石墨粉尘。所设计的取样过滤器便于拆卸和后期测量,可实现对放射性石墨粉尘进行长期系统的研究,给出反应堆不同运行工况下一回路氦净化系统中石墨粉尘及固体裂变核素活度的信息,将为HTR-10高温气冷堆裂变产物行为研究提供大量重要的实验研究数据。 相似文献
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基于NECP-X程序中已经研发的全局-局部耦合共振计算方法,研究了针对非棒状几何燃料的共振计算方法。首先,采用中子流方法计算真实问题的丹可夫修正因子,以处理全局的空间效应;其次,基于丹可夫修正因子等效获得小规模问题周围慢化剂的几何信息;最后,对于小规模问题燃料区的有效自屏截面的计算采用共振伪核素子群方法。将该方法应用于非棒状几何燃料数值计算,结果表明,该方法在处理非棒状几何燃料栅元的共振计算时,与蒙特卡罗结果程序相比,微观吸收截面偏差不超过1.8%,无限介质增殖因数偏差不超过110 pcm(1 pcm=10-5),具有较高的计算精度;在大规模问题的计算中,基于板状燃料的JRR-3M实验堆全堆在整个燃耗过程有效增殖因数偏差均在300pcm左右,组件功率偏差在整个燃耗过程不超过0.62%。因此,本研究提出的共振计算方法具有较高的正确率和精度。 相似文献
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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。 相似文献
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High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment. 相似文献
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T. Yamashita H. Akie Y. Nakano K. Kuramoto N. Nitani T. Nakamura 《Progress in Nuclear Energy》2001,38(3-4):327-330
Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO2 and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO2 fueled PWR under both reactivity initiated accidents and loss of coolant accidents. 相似文献
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W. Marshall J. Wagner 《Packaging, Transport, Storage and Security of Radioactive Material》2014,25(1):1-7
Commercial used nuclear fuel (UNF) in the USA is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high burnup values (>45 GWd t?1) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF are not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on the criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories, and specific configurations were evaluated to understand trends and quantify the consequences of worst case potential reconfiguration progressions. These results are summarised here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g. >20% Δkeff). However, for credible fuel failure configurations from ES or transportation following ES, the consequences are judged to be manageable (e.g. <5% Δkeff). The current work expands on the previous efforts by including part length rods in fresh boiling water reactor fuel assemblies and studying the effect of damage in varying numbers of fuel assemblies. 相似文献