基于瞬态基准实验的CAACS程序验证

核临界安全是核工业发展的特殊安全问题,其中临界事故的分析评价具有重要的学术意义和工程价值。CAACS程序是自主开发的圆柱形溶液系统临界事故分析程序,可以实现计算临界事故的裂变次数,裂变功率、温度随时间的变化等功能。为了验证自主开发的CAACS程序,本文利用瞬态基准实验对程序进行验证,并于其他程序作了对比。结果表明:CAACS程序的计算结果与实验测量值符合较好,与国外同类程序具有相当的技术水平和精度,可为后处理厂建设提供事故分析的技术手段,为后续的临界瞬态研究奠定基础。。     

基于蒙特卡罗均匀化理论与有限体积方法的溶液系统临界事故分析方法

基于蒙特卡罗均匀化理论与有限体积方法,建立了适用于瞬发临界事故分析的三维扩散时空动力学模型。将三维扩散时空动力学模型与非稳态传热模型、辐照裂解气泡模型耦合,对计算程序GETAC-S进行了升级,使其具备了对溶液系统任意几何与材料条件下的瞬态分析能力。使用国际上已有的瞬态装置TRACY的实验数据对GETAC-S进行了验证,结果符合良好。使用GETAC-S对日本的JCO临界事故进行了事故进程反演,证明GETAC-S具备了对复杂溶液系统下的临界事故后果进行评价与反演的能力,为核临界事故的预防、评估和屏蔽提供了理论支持。     

利用修正准静态方法评估溶液系统临界事故功率

随着我国核电的快速发展,对后处理能力有了越来越大的需求。更大的后处理能力,意味着对于临界安全提出了更高的要求。因此需要对核临界事故分析开展研究。本文使用修正的准静态方法计算了溶液系统临界事故的平均功率,并且使用CRAC实验、TRACY实验和JCO事故进行了验证。在非沸腾工况期间和沸腾工况期间,该方法得到了较好的计算结果。验证表明,该方法适用于反应性引入的临界瞬态,可快速且较为准确的估算总功率,并用于裂变次数的估计,为临界事故后果评价分析提供技术手段,为后续的临界事故研究奠定基础。     

核燃料溶液系统临界事故分析的辐解气体模型改进

对瞬态临界事故的准确模拟是核燃料溶液系统临界安全评估的关键因素。现有的辐解气体模型经验参数较多,导致功率特性预测存在较大偏差。为提高模拟精度和避免对模型中经验参数取值的依赖,需对辐解气体模型进行改进。基于对溶液中辐解气体行为的分析和简化假设,建立了包含辐解气体浓度、辐解气泡单位体积物质量和气泡数量密度的守恒模型,并将其与点堆中子动力学模型和二维导热模型相耦合,开发了溶液系统二维瞬态分析程序,通过日本TRACY实验进行了验证。结果表明,程序模拟值与实验数据符合较好,程序模型能够准确模拟溶液系统临界事故的功率变化。     

圆柱形溶液系统临界事故分析程序研制与验证

圆柱形溶液系统临界事故的分析评价具有重要的学术意义和工程价值。为实现该溶液系统临界事故的模拟与分析,本文研究了核临界事故的发生发展机理,自主研制了圆柱形溶液系统临界事故分析程序(CAACS),为后处理厂改造及后续商用后处理厂建设提供事故分析的技术手段,为后续的临界瞬态研究奠定基础。     

铀氢锆瞬态实验装置计算程序的研发

本文针对新型核反应堆电源系统物理特性研究的铀氢锆瞬态实验装置,结合点堆动力学方程、非稳态热传导方程和热弹性动力学方程,研发了专用于铀氢锆瞬态实验装置实验过程模拟的瞬态计算程序。以保健物理研究堆（HPRR）为例,利用研发的程序进行模拟,得到HPRR的功率、反应性、温度和径向位移随时间的变化。程序的计算结果与Fuchs-Hansen模型的结果较为一致,验证了该瞬态计算程序的正确性。     

乏燃料溶液系统临界安全分析计算方法研究

核临界安全分析是保证乏燃料后处理厂安全性的关键技术，而现有核临界安全事故分析程序中，或在几何适用范围上受限，或由于计算效率低而工程实用性差。因此，亟需研发一套适用范围大、计算精度高的临界安全分析方法，提高对核临界事故的分析精度，为乏燃料后处理厂提供技术保障。为此，本文针对乏燃料溶液系统特性，基于零维超细群截面制作与全问题并群方法、预估-校正准静态中子动力学计算方法和二维轴对称热工-辐解气体模型，开发了相应的计算程序模块，最终形成了一套具备并行功能的三维乏燃料溶液系统临界安全分析程序hydra-TD。进一步利用该程序对法国SILENE实验装置进行了验证，结果显示：第一裂变功率峰、倍增时间、总裂变次数等关键参数的误差较小，证明hydra-TD程序正确模拟了燃料溶液系统临界过程中的多物理过程，具备临界安全分析的能力。     

跨临界泄压瞬态传热特性的模型敏感性分析

在停堆或失水等事故工况下,超临界水冷堆将经历跨临界泄压过程,系统压力从超临界状态降到拟临界点22.064MPa以下。而对于次临界区,临界点附近的临界热流密度值很低,极易发生沸腾临界,导致加热棒壁面温度迅速升高,因此跨临界泄压过程是超临界水冷堆失水事故安全分析的关键。目前,跨临界泄压瞬态过程可以通过系统程序进行计算,但依然缺乏有效的实验验证。故本文依托上海交通大学的超临界流体多功能实验回路(Supercritical WAter MUltiPurpose loop,SWAMUP)跨临界泄压过程的实验,利用德国核安全中心(GRS)开发的系统程序ATHLET3.0进行建模计算,分析跨临界泄压过程传热特性。通过调节次临界区临界热流密度、最小膜态沸腾温度、骤冷前沿模型等相关参数,对计算模型进行敏感性分析,为跨临界泄压瞬态过程的准确计算提供参考。计算结果表明,加热棒壁面是否发生温度飞升取决于所选用的临界热流密度和最小膜态沸腾温度的值;骤冷前沿模型的使用可以实现壁面再湿润,降低壁面温度。     

TP2008核数据库验证

TP2008是新研制的用于TPFAP程序的69群核数据库,本文利用IAEA压水堆棒状燃料组件基准问题和零功率临界实验结果对TP2008核数据库进行了验证分析。结果表明,燃料组件无限增殖因数k_∞与机构TUR的符合相对好;棒状燃料组件相对功率分布计算结果与参考程序的符合较好。零功率临界实验的堆芯有效增殖因数k_eff的相对偏差大部分在-0.5%以内,符合较好。     

超功率下金属燃料钠冷快堆的动态仿真

编制了计算金属燃料钠冷快堆在超功率事故下的动态过程的仿真程序ＭＦＴＯＰ，并对它对美国池式钠冷快堆ＥＢＲ－Ⅱ在启动和功率运行工况下的反应引入事故瞬态进行了大量的分析计算，所得结果与国外大型程序ＮＡＴＤＥＭＯ的相应预测结果符合良好。本程序可用于其它钠冷快堆的超功率瞬态计算。     

Improvement in estimation of first peak power based on non-linear temperature feedback reactivity in criticality accident with instantaneous reactivity insertion

A simple equation for the first peak power in a criticality accident due to instantaneous reactivity insertion into nuclear fuel solution system has been developed to improve the accuracy in the estimation of the first peak power keeping the easiness of calculation.

The equation is based on the assumption that temperature feedback reactivity is a second-order function of an increase in fuel temperature. Peak power estimated using the equation was in a range between about a half and twice of experimental value. Its applicability to a wide range of initial reactivity and accuracy of estimation have been confirmed in the comparison to one-point kinetics numerical calculation.

The expression suggests the first peak power increases with the square of small initial reactivity and three-halves power of large initial reactivity.     

核临界事故报警系统布置分析方法研究

本文阐述了布置核临界事故报警系统的意义和原则,分析了核临界事故可能发生的机理,初步建立了一套核临界事故情景假设分析方法。研究了最小临界事故源项计算方法以及三维剂量场分布计数的方法,采用各设备最小临界事故剂量场分布最小值等高线图的方法来从众多剂量场分布图中优化选取合适的核临界事故报警系统布置点位,以确保其可以覆盖到每个具有核临界事故风险的设备,并对核临界事故报警系统探头类型选择的原则和方法进行了分析。     

基于SCALE的RFA改进型燃料组件堆内贮存临界安全分析

RFA改进型燃料组件是西屋公司设计的能应用于大功率先进压水堆的改进型燃料组件。SCALE计算程序是一款在国际上得到广泛认可的综合性建模及模拟程序包,可用于核设计与核安全分析。基于SCALE计算程序,针对大功率先进压水堆的乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,分析乏燃料水池正常贮存及事故工况下的临界安全。计算结果表明一区正常贮存工况keff值为0.901 29,组件跌落事故工况下,有效增值因子为0.907 93。二区正常贮存工况下,计算模型keff值为0.909 98,新燃料组件误插入事故工况keff值为0.924 07。先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

CEFR-MOX新燃料组件运输货包临界安全计算

易裂变材料运输过程中重要的安全问题之一是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤对临界安全影响、最佳水慢化条件等因素。本文采用MCNP 程序针对CEFR-MOX新燃料组件运输货包进行了临界安全计算。计算结果表明:MCNP程序(采用核截面库为ENDF/B-V库)对本问题的次临界限值为0.924 6;正常运输条件下无限个运输货包的最大k_eff值为0.574 4,运输事故条件下无限个运输货包的最大k_eff值为0.659 7。根据临界安全指数的定义,确定CEFR-MOX新燃料组件运输货包的临界安全指数为0。     

铀加工与燃料制造设施核临界事故所致瞬发剂量计算研究

建立了基于蒙特卡罗（MCNP）程序建模的铀加工与燃料制造设施核临界事故工况下瞬发剂量的计算方法，并将该计算方法与EJ/T 988—96规定的计算方法进行了比较分析。以我国某核燃料元件研发厂址为例，采用MCNP程序建模计算了该厂址核临界事故对厂界公众所致的瞬发剂量。结果表明，EJ/T 988—96的计算方法过于保守的估计了核临界事故工况下的瞬发剂量；基于MCNP程序建模的计算方法，因其求解算法的科学性和模型对屏蔽介质的准确描述，以及结果误差的可控性，使得计算结果更准确。因此，建议采用基于MCNP程序建模的方法计算铀加工与燃料制造设施核临界事故下的瞬发剂量。      

Modification of thermal hydraulic code,THEATRe, with fuel rod behavior models and its application to reactivity-initiated accident (RIA)

During reactor operation, many complex changes occur in fuel rod which affects its thermal, mechanical and material properties. These changes also affect the reactor response to the transient and accident situations. Realistic simulation of fuel rod behavior under transients such as reactivity-initiated accident (RIA) is of great significance. In this study, thermal hydraulic analysis code THEATRe (Thermal Hydraulic Engineering Analysis Tool in Real-time) has been modified by addition of fuel rod behavior models for dynamic simulation of nuclear reactor. Transient changes in gas-gap parameters were taken into account by modeling the gas-gap behavior. Thermo-mechanical behavior of fuel rod is modeled to take into account the thermal, elastic and plastic deformation. To simulate RIA, point reactor kinetics model is also incorporated in the THEATRe code. To demonstrate the transient fuel rod behavior, AP1000 reactor is modeled and three hypothetical RIA cases are simulated. The RIA is considered at three different reactor power levels, i.e. 100, 50 and 1% of nominal power. The investigated parameters are fuel temperature, cladding stress and strain, fuel and cladding thermal conductivity and heat transfer coefficient in gas-gap. Modified code calculates the fuel rod temperatures according to updated fuel, clad and gas-gap parameters at the onset of steady-state operation and during the transient. The modified code provides lower steady-state fuel temperature as compared to the original code. Stress and strain analyses indicate that the hoop and radial strain is higher at high power locations of the fuel rod; therefore, gap closure process will initially occur in the central portion of the fuel rod and it should be given more emphasis in the safety analysis of the fuel rod and nuclear reactor during accidents and transients.     

Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code ‘FEMAXI-FBR’

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code ‘ASTERIA-FBR’ in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code ‘FEMAXI-6’, FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P ₀/s (P ₀: steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions.     

VVER-1000 neutronics calculation with ENDF/B-VII data

The initial release of the ENDF/B-VII nuclear data library is verified for VVER-1000 reactors. For neutronics calculation, the MCNP code based on the Monte-Carlo method is applied. Continuous-energy cross-sections for use with MCNP are calculated with the NJOY code. Isotopics for burned fuel is calculated with the WIMSD code. Calculated criticality, pin-to-pin power distribution, time-dependent critical concentration of soluble boron, worth of the control rods, average fuel assembly powers and time-dependent axial power distribution are compared to the corresponding experimental values for both zero-power VVER-1000 model, created at the LR-0 experimental facility, and the first fuel cycle of a real VVER-1000 reactor. For all of these parameters, neutronics calculation with ENDF/B-VII is in good agreement with the measurement. Moreover, for VVER-1000 neutronics calculation, ENDF/B-VII provides better results than ENDF/B-VI.