Comparative assessment of theoretical and experimental measurement of photo neutron flux level in GHARR-1 after shutdown

Direct photo-neutron source strength was evaluated for the Miniature Neutron Source Reactor (MNSR) in subcritical condition in the GHARR-1 facility. Two different static methods were applied for comparison. A theoretical method based on the use of MCNP code and an experimental method based on foil activation technique. The latter has been found to be most convenient method for neutron flux measurement. The method depends only on the activity of a bare and cadmium covered foil if the irradiation positions are known. Photo-neutron flux level was determined theoretically using MNCP after measuring neutron flux at shutdown; and experimentally using Neutron Activation Analysis (NAA) technique also at shutdown with great care. The values obtained from the theoretical and experimental measurements are tabulated in Table 2. The results recorded were validated using biological peach leave and a geological rock sample. The results after validation for Mn concentration in the samples were 87 ± 1 μg/g and 432 ± 23 μg/g, respectively. Results for the two methods were in good agreement. Realization of photo-neutron source existence due to beryllium reflector was also experienced.     

次临界能源堆安全性研究

次临界能源堆由中心的托卡马克装置和围绕其的裂变包层组成。本文根据物理和热工专业分析计算得出的一种针对其裂变包层的燃料和冷却剂通道布置方式,分析设计的包层结构安全性和工程应用中的安全性。包层结构安全性分析使用CFD方法,计算了正常运行工况和冷却剂通道堵管的情况,得到堵管发生后包层的局部状况。通过RELAP程序模拟了裂变包层参与核电厂发电运行过程中,其本身所具有的固有安全性。本文通过计算发现了其安全上的薄弱环节,并提出了改进措施,为以后改进次临界能源堆安全性提供参考。     

泳池式轻水反应堆内电磁线圈可控温辐照装置设计

本文设计了在泳池式轻水反应堆（简称泳池堆）内在线测量电磁线圈电性能的可控温辐照装置。采用MCNP程序进行中子物理计算,对泳池堆、线圈骨架的结构尺寸与物质组分进行了精细全尺寸模拟,得出辐照装置的发热功率和中子注量率。通过初步估算,使用ANSYS CFX进行了数值模拟,得出辐照装置内线圈在堆运行时的温度,并提出温度控制的方法。辐照装置采用铝材加工制造,并进行了垂直度测试、气压测试、检漏测试。增加了绝缘设计,将辐照装置与泳池堆之间进行绝缘。在线圈处预埋铠装热电偶,对线圈温度进行实时监测。在泳池堆内对电磁线圈进行辐照试验,结果表明,本文设计的辐照装置能满足电磁线圈在泳池堆孔道内进行辐照试验的要求,并可对电磁线圈进行实时温度控制。     

Coupled neutronics/thermal–hydraulics analysis of CANDU–SCWR fuel channel

At supercritical pressure condition, the thermal–hydraulics behavior of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal–physical properties across the pseudo-critical line. A coupling analysis of neutronics and thermal–hydraulics has become important for SCWR, because of the strong link between the water density and the neutron spectrum and subsequently the power distribution. The neutronics code Monte Carlo N-Particle code (MCNP) and the subchannel code Advanced Thermal–Hydraulics Analysis Subchannel (ATHAS) are used in a coupled way to better understand the design characteristics of a pressure tube type SCWR fuel channel. The results show that: the developed coupled code system can be used to analyze pressure tube type SCWR fuel bundles; improved radial fuel enrichment profile will optimize the coolant and cladding temperature distribution to meet the design criteria; smaller pressure tube pitch will result in more flatten axial power distribution and more uniform radial power distribution.     

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.     

In-core fuel management practice in HANARO

The HANARO (High-flux Advanced Neutron Application Research reactOr) is a newly created research reactor. Its initial criticality was achieved on February 8, 1995. In its design stage, the HANAFMS, which is the nuclear analysis and fuel management code system for the HANARO, was verified using the results from the MCNP4A full core model because there was no similar research reactor in the world. It was needed to verify the HANAFMS with reactor physics experiments, which were performed during the reactor commissioning and power operation. The calculated results for the criticality, power distribution and control absorber rod (CAR) worth were compared with the measured ones. In the criticality calculation, the clean and depleted cores were applied and in the comparison of power distribution, the gamma scanning data of the fuel assemblies were used. The CAR worth was calculated following the measurement positions and then compared with the measurements. The calculated results for verifying the HANAFMS are in good agreement with the measured ones.     

加速器驱动系统的燃料选择

对装载不同增殖材料的现实加速器驱动系统(ADS)的安全及嬗变超铀核素特性进行研究。分别 以(U,TRU)O2和(Th,TRU)O2作为堆芯燃料,先用LAHET和MCNP程序对ADS进行稳态模拟计 算,再耦合MCNP和ORIGEN2程序计算燃耗过程中的核素密度变化。结果显示,装载钍基燃料的 ADS对超铀核素的嬗变效果较好,且在燃耗过程中其反应性和质子流强波动较小;装载铀基燃料的 ADS则具有更安全的多普勒效应和缓发中子有效份额。总体来看,如果需要堆长时间安全嬗变超铀核 素,装载钍基燃料会取得更好的效果。     

MCNP在次临界装置实验中的应用研究

MCNP3B程序是一个大型的多功能蒙特卡罗程序,可用于屏蔽设计和应用孔道的设计,以及某些临界装置的设计计算。文章介绍了应用MCNP3B程序对DF堆次临界装置的几个实验布置进行的物理参数计算,并将计算结果与实验结果进行了比较。通过对DF次临界装置的研究,以求MCNP3B程序能为零功率实验提供指导性意见,保证零功率实验的安全。     

ADS实验大厅中子防护工程技术研究

为了保障加速器驱动次临界系统(ADS)散裂靶与反应堆耦合特性及影响验证实验的顺利进行,以原子能院现有的临界实验装置为基础,对堆厅部分墙体进行屏蔽改造。建造由聚乙烯、镉、铅、钢以及混凝土等材料构成的屏蔽装置,以防止临界装置产生的射线外泄,使工作人员受到的照射保持在合理水平。通过MCNP模拟计算,完成了屏蔽结构的优化设计。基于槽钢支撑结构、铅屏蔽层、镉屏蔽层和聚乙烯屏蔽层等材料组成的组合屏蔽结构建立简化模型,采用ANSYS有限元分析程序计算分析得出各部分应力小于许用应力,稳定性符合要求。最后通过工程实践,完成对屏蔽性能理论计算结果的验证。     

加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

运用MCNP与ORIGEN2耦合计算程序COUPLE，对加速器驱动的次临界系统（ADS）钠冷金属燃料快堆堆芯进行稳态与燃耗计算，比较分析次锕系核素（MA）非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明，对比均匀布置，非均匀布置具有更高的MA嬗变率与嬗变支持比，在反应性参数方面导致多普勒效应与有效缓发中子分额降低，钠空泡效应增大，在堆芯功率分布与加速器束流功率方面没有明显变化。     

CFBR-Ⅱ堆典型辐照位置的中子能谱计算

建立基于MCNP程序的中子能谱及平均中子能量计算方法,模拟计算了CFBR-Ⅱ堆典型辐照位置的中子能谱及平均中子能量随空间位置的变化关系。结果表明,各典型辐照位置的中子能谱集中分布于0.05–3MeV(～90%);去耦盒与辐照孔道轴线上各点的平均中子能量随距离大致呈S形变化趋势,做辐照效应研究时要考虑能谱分布空间不均匀性的影响;去耦罩45°纬线圈到顶部较大范围内平均中子能量波动较小,是较理想的辐照区域。     

ADS次临界实验装置设计方案验证

根据设计要求．使用MCNP／4C程序计算了多种次临界反应堆均匀化堆芯布置方案的临界问题。确保keff在0．92～1．00之间。为加速器驱动的洁净核能系统(ADS：Accelerator Driven system)的次临界实验装置设计提供了初步数据。     

中子残余应力谱仪静态屏蔽体表面辐射剂量率初算

中子残余应力谱仪静态屏蔽体主要用于对谱仪装置的附加闸门、中子导管等组件的辐射剂量的屏蔽,使装置操作人员可以安全地在装置周围活动。通过MCNP5程序对谱仪装置静态屏蔽体的屏蔽能力进行了计算,可为该方案的改进、优化提供依据,以便最终制造出满足辐射剂量要求的屏蔽体。     

Investigation of thermohydraulic parameters during natural convection cooling of TRIGA reactor

Important steady-state thermohydraulic parameters of the TRIGA research reactor operating under natural convection mode of coolant flow were investigated using NCTRIGA computer code. Neutronic parameters used in preparing the input of NCTRIGA were taken from the analysis performed by 3-D Monte Carlo code MCNP4C. Benchmarking of the NCTRIGA calculated results were performed against the experimental data measured by the thermocouples in the instrumented fuel element (IFE) during the steady state operation of the reactor under natural convection mode of coolant flow. Various thermohydraulic parameters like the coolant velocity, flow rate and mass flow rate were generated for the hot channel as well as for the two channels comprising instrumented fuels. Calculated peak fuel temperatures at different power levels were compared with the measured values and also with the calculations performed by PARET code. Axial temperature profile at the fuel centreline, fuel surface and coolant in the hot channel were generated. Fuel surface heat flux, heat transfer coefficient and Reynolds’s number for the hot channel were also calculated. The effect of fuel-cladding gap and the influence of fuel rod spacing were investigated to validate the performance of NCTRIGA code. The investigated results were found to be in good agreement with the experimental values, which indicates that the NCTRIGA code can be used with confidence for TRIGA reactor analysis.     

ADS次临界反应堆的中子共轭方程

与临界反应堆相比,ADS次临界反应堆的外源中子和裂变中子的空间分布具有严重的不均匀性,对应的中子价值也不同。本工作对次临界反应堆的稳态输运方程作分群扩散近似,得到了多群方程,进一步推导出按堆芯功率归一化的中子共轭方程表达式和与功率相关的中子价值函数表达式,给出了次临界反应堆中子价值的物理意义。由稳态中子共轭方程组出发,给出了两种带外加中子源的次临界反应堆增殖因数的表达式。     

基于蒙特卡罗-离散纵标双向耦合方法的堆坑粒子注量率计算

为有效解决大型复杂核设施屏蔽计算问题,研究了三维蒙特卡罗(MC)-离散纵标（S_N）双向耦合方法,通过自主开发接口程序实现MC粒子概率分布与S_N角通量密度之间的相互转换,实现MC-S_N双向耦合计算。将基于MC-S_N双向耦合方法的程序用于某反应堆堆坑底部粒子注量率计算。利用MC程序建立堆芯及堆坑处的精细模型进行计算,三维S_N程序用于堆芯下表面与压力容器底面之间区域的计算。通过MC-S_N-MC两步耦合计算,给出堆坑通道及小室内的中子和光子注量率。三维MC-S_N双向耦合方法计算结果与单一MCNP程序结果吻合较好,初步验证了该方法是解决大型复杂核装置屏蔽问题的有效工具。     

物理-热工耦合计算方法在熔盐堆稳态分析中的应用

针对美国橡树岭国家实验室（ORNL）熔盐堆（MSR）实验的堆芯设计,采用物理分析程序MCNP进行三维堆芯功率分布计算。针对以石墨作为慢化剂的堆芯结构,开发了并联多通道程序来进行堆芯热工水力分析。在此基础上,把物理和热工分析程序进行耦合,用ORNL技术报告中的相关内容来验证物理 热工耦合分析的可行性和准确性。结果表明,本工作的耦合计算方法可获得熔盐堆堆芯功率分布、温度分布、压降和流量分配。熔盐堆耦合程序的研发对熔盐堆概念设计、运行分析有重要意义。     

原型微堆辐照座物理特性参数模拟测定

文章给出了原型微堆辐照座同的某些物理特性参数；相对中子通量密度分布，绝对中子通量密度，能谱能数（镉比、超热指标和中子温度），某些样品在辐照座内对反应性的影响以及各辐照座之间的相互关系，实验研究在原型微堆的零功率实验装置上完成。     

Corrosion damage to Kh18N9T steel after 25 years of operation in the steam-water medium of a VK-50 boiling-water reactor

The results of an investigation of the material used for the VK-50 measurement channel, which showed no signs of unsealing when it was extracted after 25 years of operation, are presented. The results of metallographic and electron-microscope investigations and Auger-spectroscopy determination of the elemental composition are presented. Intercyrstallite corrosion is found in the main metal of the measurement channel tube at the core-bottom level. A network of nonthrough, mainly longitudinal, cracks of an intercrystallite character, formed at the core-center and core-top levels and immediately under the reactor cover. The investigations show that the degree of corrosion damage to the channel material depends on the coolant density along the core height. Neutron irradiation can be a provoking factor but it is not the main factor making the main metal more likely to undergo corrosion cracking. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 31–35, January, 2005.     

Validation of the Monte Carlo model supporting core conversion of the Portuguese Research Reactor (RPI) for neutron fluence rate determinations

The Monte Carlo code MCNP was used to calculate absolute values of thermal, epithermal and fast neutron fluence rates in the new RPI core using fresh LEU fuel. Discrepancies smaller than 20% were obtained between calculated results and activation foil measurements. A previous knowledge of general characteristics of the neutron energy spectra, provided by the MCNP reactor model itself, has been fundamental to determine the conditions yielding a proper comparison of simulated and measured results. An excellent agreement (6%) was also obtained for the relative neutron fluence rate profiles along the fuel height. The MCNP model of the reactor core was therefore validated for a tri-dimensional determination of neutron fluence rates in the fuel assemblies and neighbouring irradiation positions.