中国实验快堆缓发中子探测系统的瞬态模拟与分析

根据中国实验快堆缓发中子探测系统的结构特点和探测原理,构建了缓发中子探测系统的计算模型.基于该模型,开发了计算机模拟程序.针对不同工况和不同燃料元件包壳破损时刻,进行了缓发中子探测信号的模拟计算.计算结果基本反映了计算情况下缓发中子探测信号的发展趋势.同时,还对燃料温度和燃料燃耗对缓发中子探测信号的影响进行了物理分析.     

利用裂变径迹探测器刻度缓发中子监测仪的探测效率

一、前言本文所指的缓发中子监测仪是探测燃料元件破损的仪器,探头为BF_3计数管,当燃料元件破损后,裂变产物所释放的缓发中子经慢化后成热中子,进入BF_3计数管产生~(10)B(n,α)~7Li反应。记录其脉冲,监督燃料元件的破损。     

熔盐堆物理热工耦合程序开发及验证分析

熔盐堆(Molten Salt Reactor,MSR)是第四代反应堆6种堆型中唯一的液态燃料反应堆,与固态燃料-液体冷却剂反应堆相比,原理上有较大不同。在熔盐堆中,流动的熔盐既是燃料又是冷却剂与慢化剂,中子物理学与热工水力学相互耦合;由于熔盐的流动性,缓发中子先驱核会随燃料流至堆芯外衰变,造成缓发中子的丢失,导致堆芯反应性降低。正是由于熔盐堆的这些新特性,造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同,需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响,研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序;利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算,验证熔盐堆相关实验数据;并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况,用于分析熔盐堆的安全特性。计算结果表明,程序能够对熔盐反应堆实验(Molten Salt Reactor Experiment,MSRE)的相关实验数据进行较好的模拟计算,并且验证了熔盐堆的固有安全性。     

反应堆燃料元件破损当量诊断方法

破损当量是衡量反应堆燃料元件破损严重程度的重要指标,但破损当量无法直接测量,在决策应用中不具有可操作性,需要建立与破损当量对应的可监测指标。本文结合实践经验,分析确定了可用于燃料元件破损诊断的典型核素,建立了反应堆一回路冷却剂中裂变产物核素活度浓度与燃料元件破损当量之间的传递关系;给出了一回路冷却剂取样分析实验方法,并指出实验过程中应注意的问题;建立了采用监测一回路冷却剂中典型裂变产物核素活度浓度诊断破损当量的方法,并分析了诊断中不确定度的主要影响因素。本研究为反应堆燃料元件破损当量诊断提供了技术方法。     

压水堆核电厂燃料元件破损诊断方法

在核电厂运行管理中, 如果在停堆前知道燃料棒的性能和状态,采用合适的燃料检测管理策略,可减少反应堆的停运时间.本文以燃料元件破损后裂变产物向冷却剂释放的理论为基础,开发了一种通过分析反应堆冷却剂中裂变产物放射性活度估算破损燃料元件的数量、破损尺寸和位置的方法.用大亚湾核电站1号机组第2循环的运行跟踪数据对U1C2燃料组件进行了破损诊断.结果表明,诊断结果与停堆后的实测结果基本吻合.     

中国实验快堆缓发中子监测系统的建模与计算

中国实验快堆（CEFR）缓发中子监测系统属于燃料破损探测系统的一个子系统，主要用于监测一回路钠中的缓发中子强度，以判断燃料包壳是否发生接触性破损，以便采取必要措施，防止破损的进一步发展。该系统的设计主要参考了俄罗斯的技术方案，直接将缓发中子探测站布置在中间热交换器附近，利用一回路钠的流动将可能泄漏出来的缓发中子先驱核从破损燃料处输运到中间热交换器并释放缓发中子而进行探测的。     

中国实验快堆燃料破损探测系统建模、计算与分析

在借鉴国外研究成果的同时,结合中国实验快堆(CEFR)燃料破损探测系统的设计特点,建立了CEFR燃料破损探测系统的计算模型,并根据所建计算模型,利用LabWindows/CVI开发了CEFR燃料破损探测系统计算分析程序。用该程序进行了缓发中子探测系统可探测最小破损面积的计算,并对裂变产物的释放产生比进行了计算验证。     

MC模拟中子探测器监测脉冲堆燃料元件破损响应特性

为评估缓发中子探测器监测脉冲堆燃料元件破损的响应特性,建立了MCNP中子源模型和3He球形计数管探测器模型,应用MC方法计算给出了最佳监测位置和该位置处不同聚乙烯厚度下探测器对缓发中子的探测效率曲线。结果表明:最佳监测位置紧贴换热器外表面且偏向入口端,偏移距离和冷却剂贯穿换热器耗时有关,耗时越长,偏移越大;探测效率曲线的变化趋势取决于平均反应截面与中子注量率的积,采用聚乙烯慢化层在提高平均反应截面的同时会降低中子注量率,存在最优厚度4 cm使探测器对缓发中子探测效率最大,其值约为1.1×10-5。     

缓发超临界过程燃料元件温度场动态分析与快速计算

针对反应堆缓发超临界过程特点,本工作建立了物理、热工及热量传递模型,采用准稳态方法对缓发超临界过程反应堆平均通道单根燃料元件温度场进行了计算分析,并与某反应堆仿真平台计算结果进行对比。分析表明,在功率变化不剧烈的反应堆动态过程,采用准稳态计算方法可以较准确地计算出燃料元件温度场随时间的变化规律。且本工作模型相对简单,计算速度快,对于反应堆动态运行研究及船用反应堆事故分析均有重要意义,对于反应堆热工设计也具有重要的参考价值。     

西安脉冲堆中子价值计算与分析

中子价值是反应堆中的重要参数,其物理意义是中子对反应堆功率的贡献大小。首先采用栅元程序WIMS,计算了西安脉冲堆各栅元的6群群常数,然后采用堆芯扩散程序CITATION,计算西安脉冲堆的中子价值分布。采用连续能量点的蒙特卡罗程序MCNP,对CITATION扩散计算的正确性进行了验证。分析了西安脉冲堆的中子价值随空间和能量的变化,以及中子价值对动态参数缓发中子有效份额和中子代时间的影响。结果表明:在燃料栅元中,中子价值随能量的增加而降低,在控制棒和水栅元中,中子价值随能量的增加而增加。缓发中子有效份额大于缓发中子份额的主要原因是燃料栅元中缓发中子的中子价值较瞬发中子的大。中子代时间小于瞬发中子寿命的主要原因是燃料栅元中的中子价值较其他栅元中的大。     

压水堆核电站乏燃料元件γ扫描测量

在压水堆核电站乏燃料元件检验中,完成了4根完整元件棒、4根破损元件棒的γ扫描测量,元件燃耗分布在9600～45000 MW•d/t（U）之间,获得了完整元件轴向相对燃耗分布、破损元件¹³⁷Cs分布及迁移流失情况。结果显示,破损元件均存在不同程度的Cs迁移流失,破口处存在¹³⁷Cs计数突变（降低）。破损元件¹³⁴Cs/¹³⁷Cs原子比分布与相邻完整元件基本一致,表明¹³⁴Cs、¹³⁷Cs流失比例近似相等,可用¹³⁴Cs/¹³⁷Cs原子比表征其相对燃耗分布;破口处可通过低挥发性核素¹⁵⁴Eu计数水平判断燃料芯块是否缺失。检验结果可为燃料元件破损原因分析及堆内行为分析提供重要依据。     

LOCA工况下环形燃料元件内包壳外压屈曲试验研究

环形燃料一种安全高效的新型核燃料。为对环形燃料元件冷却剂丧失事故（LOCA）下整体受压失效形式的问题进行研究,将环形电加热棒、模拟芯块和试验件组装成试验装置,在空气环境中,以环形电加热棒外加热的方式,对环形燃料元件内包壳进行了外压屈曲试验,并将试验屈曲压力与Bresse?Bryan公式计算结果和特征值屈曲数值模拟分析结果进行了对比分析。结果表明：Bresse?Bryan公式计算结果除以安全系数m=2?5得到的结果高于试验结果而不够保守,试验结果分布于特征值屈曲数值模拟分析结果的1/5?1/3之间。本文结果可为环形燃料元件安全评价及后续工程化提供基础数据。     

压水堆核电厂冷却剂碘同位素活度比值131I/133I与燃料完整性关系研究

压水堆核电厂正常运行期间燃料元件破损会造成一回路裂变产物活度升高,碘同位素活度比值131I/133I是行业内最常用的判断燃料破损情况的指标之一。本文介绍了压水堆正常运行期间冷却剂131I和133I的产生来源和迁移过程,建立模型估算了燃料完整、小破口和大破口情况下131I/133I范围,并通过在运CPR1000型压水堆核电厂的运行监测数据对计算模型进行了验证,两者符合得较好。     

LaBr_3(Ce)γ谱仪在燃料元件破损监测中的应用研究

为了实现用LaBr_3(Ce)γ谱仪实时监测压水堆燃料元件的破损,对该谱仪系统在燃料元件破损监测中的几个关键问题进行了研究。通过实验测试与蒙特卡罗(MC)模拟计算,提出了使用LaBr_3(Ce)γ谱仪测量一回路冷却剂中裂变产物~(135)Xe和~(88)Kr的活度浓度来判断燃料元件是否发生破损的方法,并对该方法进行了验证。对某反应堆一回路冷却剂进行测量的结果表明,基于LaBr_3(Ce)γ谱仪的燃料元件破损监测方法可有效避免监测中的干扰因素的影响,降低了定量测量中的不确定度。     

Thermal-hydraulics analysis during fuel element failure in an operating PWR

The fuel element failure in an operating pressurized water reactor (PWR), including fuel element breaks, has an effect on the operation safety of PWR. In this paper, the RELAP5 model of the fuel element failure is established for the safety analysis. The RELAP5 time step sensitivity analyses for the element pre-break steady and post-break transient simulation are carried out. And the variations of main thermal-hydraulics parameters related to the fuel element break are quantitatively studied, which include the internal gap pressure and the maximum fuel pellet temperature as well as the releasement of noncondensables in the gap. It is found that (1) the results by the RELAP5 code is very sensitive to the time step in a volume system with the noncondensables, and the time step sensitivity analysis is necessary if the effective time step range is unknown, (2) the larger the break area is, the more quickly the gap pressure increases and the maximum pellet temperature reaches to the stable value, (3) when the gap pressure increases and reaches to the coolant pressure, at the break the liquid inflow from coolant to gap will be turned to the vapor outflow from gap to coolant, (4) during the failure transient, the gap thermal conductivity experiences a sharp decrease in the break instant, which results in the decrease of heat transferred to cladding and the sharp decrease of cladding temperature as well as the sharp increase of minimum departure from nucleate boiling ratio (MDNBR). These conclusions can provide the basic for the operation safety analysis of PWR during the fuel element failure.     

Advanced fuel elements for passive pressure tube light water reactors

This paper focuses on the development of advanced fuel elements for innovative pressure tube light water reactors. Considerations and constraints that affect the design process and various possible options are discussed. The two most promising fuel designs, which can survive a loss-of-coolant accident without primary coolant replenishment, while having sufficient margins to fuel design limits, are proposed, described and evaluated. It is demonstrated that this key objective can be achieved, provided that reliable SiC cladding or coating, which can withstand operating and accident conditions without failure, can be manufactured. Recent advances in ceramic coating technologies and experimental tests of coated specimens indicate that the attainment of this goal is feasible.     

Prediction of Nongaseous Fission Products Behavior in the Primary Cooling System of High Temperature Gas-Cooled Reactor

In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified.

This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model.     

Studies on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (II) Radial Propagation of Sodium Boiling by Reactivity Effect

A computer code BORE was developed, with which analyses were performed on channel plugging accidents that would occur on a 1,500 MWe LMFBR. The BORE code calculated the dynamic characteristics of coolant boiling and fuel failure propagation radially in the core, and the requirements of core instrumentation systems were also analyzed. The results show that coolant boiling and/or fuel failure in a channel plugging accident are propagated only to a limited number of adjacent channels when sensors are installed that detect anomalies in channel flow, channel outlet temperature, boiling or reactivity. It is also concluded that the coolant void effect is not serious from the standpoint of safety when the time required for boiling propagation to adjacent channels can be made longer than 0.15 sec.     

Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

In the Shutdown Heat Removal Testing (SHRT) Program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elements, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date.