临界漂移分析

使用临界漂移分析程序(Cri SAA)对TRACY装置上的匀速注入核燃料溶液(RF)实验进行数值模拟,研究核燃料溶液装置开放临界漂移的瞬态特征,模拟结果是预测性的。Cri SAA给出实验的第一次峰时间、第一次峰功率、第一次峰释能和总裂变能等物理量的随机性特征,分析了反应性添加速度、铀溶度对临界漂移的影响。提出适用于开放临界漂移分析的改进的Fuchs模型,该模型有助于理解开放临界漂移过程。     

Fatigue crack propagation properties from small sized rod specimens

Mechanical properties characterization is needed in many industrial applications yet sufficient amount of material for fabricating standard-sized testing specimens is often not available. Techniques for testing miniaturized specimen must be adopted. Much effort has been made to develop techniques for impact, fracture toughness and tensile properties of sub-sized specimens. Work on the testing of fatigue properties is more limited. In this study, fatigue crack propagation behavior is evaluated from the growth of surface crack in a cylindrical rod under tension. Rods of various lengths and diameters were tested. As the size of the rod specimen is reduced, the fatigue crack growth rate tends to increase when correlated using the stress intensity factor range. This increase is explained largely by the decrease in the degree of premature crack closure in the small specimens. Valid fatigue crack growth data can be obtained among the specimens examined except on the crack growth on the surface of the smallest specimen, which has a length of 26 mm and diameter of 8 mm. Even so, valid data can still be elucidated on the latter specimen if the interior growth is considered. The dimensions of the latter specimen allow fatigue properties to be evaluated using broken remnants from impact or other test specimens.     

Effect of Added Scatterers on Nuclear Criticality

Experience involving basic concept of nuclear criticality has shown that there are many complex effects involved, some of which have been little studied and are not well understood. One of these involves the effect of added scatterers on criticality.

A study has been made of the effect of added scatterers (O, C, N) on criticality by means of transport theory calculations utilizing the DTF-IV codes. This study pertains to plutonium compounds likely to be encountered throughout the nuclear industry — in fuel processing and fabrication operations. It is shown that in some cases of undermoderated thin slab with hydrogenous reflectors, the effect of added scatterers can cause an increase in the critical dimension contrary to the usual expectation that the size should be reduced in such cases.

Interpretation of the results. obtained for water-reflected and nickel-reflected slabs, including scatterers added homogeneously to the core region, contribute to a better understanding of the observed effects. The dominant factor causing the increase in slab thickness is the decrease of neutron leakage into the moderating reflector.     

Criticality analysis of pulse core and laser module coupled small reactor with low enriched uranium

Nuclear-pumped laser can directly convert nuclear energy to optical energy. A coupled reactor which consists of two pulse cores with highly enriched metallic uranium and a subcritical thermal laser module with highly enriched metallic uranium is one of the reactors for nuclear-pumped laser. In this paper, criticality analysis of a coupled reactor which consists of pulse cores with 20% enriched metallic uranium and a subcritical thermal laser module with 20% enriched metallic uranium was performed by Monte Carlo calculation. The result of criticality analysis showed the following three points. First, a coupled reactor with 20% enriched metallic uranium can achieve criticality. Second, using eight pulse cores in axial direction is effective to achieve flattened axial power distribution in the laser module. Third, less than 20% of the energy released from fissions in the whole coupled reactor has the possibility to be converted to optical energy for a coupled reactor with 100% enriched uranium, and less than 7% for a coupled reactor with 20% enriched uranium.     

Transferability of post yield fracture mechanics properties from small scale specimens to large scale specimens and components

This contribution describes a method for the determination of the J-integral as a function of the load-line displacement for arbitrary specimen geometries.A correspondence could be found between the approximation method and the results determining with the Rice integral by means of a FE-calculation. Using the initiation values of the J-integral as a fracture mechanics parameter determined from the J_R-curve, correspond with failure values of double-édged notched tensile specimens and circumferentially notched round tensile specimens of which crack initiation was tantamount to instability. Consequently, it could be proved that the J-integral is a transferable parameter that may be ascertained from simple determinable deformation values. The application to real components seems to be promising, due to these good results.     

DF-VI快中子临界装置2号堆芯临界实验

DF-VI快中子临界装置在改造完成、堆芯发生了变化以后,进行了重新启动和一系列的实验测量。测量内容有：根据29次临界实验的数据对2号堆芯平均临界元件数和临界质量进行了计算;应用周期法和棒补偿法对控制棒价值进行了刻度;用逆动态反应性计对安全棒和安全块的价值进行了测量;对单根边缘元件价值和径向元件价值分布进行子测量。通过以上实验测量,确定了DF-VI快中子临界装置2号堆芯的主要安全运行参数。     

来自日本东海临界事故国际研讨会的信息

1　会议简况1 999年 9月 30日 ,在日本茨城县东海的铀转换设施发生了一起临界事故 ,3名工人严重受照于中子和γ射线 ,最终发展成急性辐射综合征( Acute Radiation Syndrome,ARS)。对其中 2名皮肤严重损伤的患者实施了干细胞移植 ,最终死于多器官衰竭。另 1名存活者面临着许多来自社会方面的困难。2 0 0 0年 1 2月 1 4～ 1 5日 ,在日本千叶县文化馆召开了关于日本东海临界事故的国际研讨会。笔者当时正在千叶放射线医学综合研究所研修 ,有机会参加了这次研讨会。来自美国、法国、中国及日本等国家的从事辐射防护、剂量学、医学等研究领域…     

Criticality assessments using polyurethane foam

Abstract

Rolls-Royce has designed a package to transport and store fresh fuel assemblies and anticipates approval from the regulators for the new package design in the near future. The space between the inner and outer steel shells is filled with shaped blocks of rigid polyurethane foam, of two different densities. The criticality safety case for the fresh fuel package had to consider single packages and arrays of packages under routine, normal and accident conditions. IAEA regulatory requirements state that the criticality assessment must include investigations on the effect on the neutron multiplication factor k _eff due to impacts, flooding and fire. Sensitivity studies must also be carried out to determine the effects on the k _eff due to any uncertainties in the composition of the fuel and container materials. An important part of the criticality safety case is the treatment of the foam. The approach adopted to model the polyurethane foam is the subject of this paper. The following were investigated: (1) the effect on the k _eff of varying the elemental composition of the foam, including the removal of hydrogen; (2) the experimental analysis of burnt foam; (3) the effect of addition of water to the foam to simulate water absorption; (4) a simple representation of crushed foam to simulate knock-back in the package; (5) extreme representations of burnt foam, such as replacing foam with solid carbon or as randomly distributed spheres of carbon to represent soot. These investigations were most informative and should be considered in any criticality assessments of transport packages containing large amounts of foam in the future.     

基于微扰计算的蒙特卡罗临界搜索方法研究

基于微扰计算,提出一种新的蒙特卡罗临界搜索方法。该方法仅需要对初始系统进行一次临界计算,利用临界计算结果及微扰计算结果,对有效增殖因子变化曲线进行估计,并以此为基础得到临界搜索的结果。计算结果表明,该方法可有效、准确地对临界搜索问题进行计算,并有效地克服了传统方法的不足。     

临界装置设计安全规定的编制思考

我国在临界装置的设计方面缺乏明确的、有针对性的法规,对临界装置的设计进行规范,填补核安全法规的空白,对监管部门的管理和科研单位的工作都很有意义。根据临界装置和临界相关物理实验的特点,针对临界装置的设计,对目前核安全法规的现状和不足进行了相应的分析,以及对临界装置设计安全规定也做了一些分析和思考。     

CENDL-NP热谱临界基准检验

为了解决CENACE-1.0库热谱临界实验中高估有效增殖因子(keff)的问题,选取多种类型的热谱临界基准实验,对制作CENACE-1.0库的基础评价数据库CENDL-NP-1.0进行基准检验。通过检验结果的趋势分析和相关性分析,发现热能区的235U裂变截面、辐射俘获截面和裂变平均中子数可能是导致高富集度铀金属热装置keff计算结果偏高的直接原因。     

高通量工程试验堆临界蒙特卡罗计算

用新研制的三维多群P3 中子输运蒙特卡罗程序MCMG ,通过与栅元均匀化WIMS程序耦合 ,计算反应堆临界 燃耗问题。高通量工程试验堆 (HFETR)临界计算取得了与MCNP程序和实验一致的结果 ,且在相同计算精度下 ,MCGM计算时间较MCNP计算时间少数倍。     

基于BEAVRS全堆基准题的RMC临界计算验证

BEAVRS是基于20世纪60年代美国商用反应堆建立的基准题模型,具有精细的堆芯配置参数和详细的运行测量数据,是新型的反应堆分析工具验证基准题。本研究利用自主堆用蒙卡程序(RMC)对BEAVRS进行建模,计算反应堆在不同硼浓度和控制棒步数条件下的临界有效增殖系数、控制棒价值、温度系数、仪表管里的235U裂变率等一系列重要参数。结果与实测数据及同类软件结果在可接受范围,验证了RMC进行反应堆临界分析的精确性和可靠性。     

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。     

Anisotropic Diffusion Effect on Criticality of Plate Lattice Fast Assembly

The neutron diffusion in plate lattice is generally somewhat anisotropic. In case of usual plate cells for the mockup of LMFBR composition, the diffusion coefficient for parallel direction to lattice plate, based on Benoist's theory, proves to be larger by 2～4% than that for perpendicular direction, which is considered to affect the criticality of plate lattice fast assembly.

A practical treatment of the anisotropic diffusion effect on criticality has been proposed, in which, like the transport correction, the anisotropic diffusion effect is treated as a correction term to be applied to the conventional isotropic diffusion calculation. The method is applied to actual plate lattice critical assemblies, already built in FCA, ZPR or ZEBRA. The anisotropy correction on criticality turns out to amount to the order of -0.2～-0.4%Δ k/k for the normal plate lattice-core. The amount of anisotropy correction is further enhanced in case of an assembly consisting of plate lattice-blanket or sodium-voided core. The anisotropic diffusion effect on criticality is, therefore, important for the analysis of criticality of plate lattice assembly, and should be corrected in addition to the conventional heterogeneity effect. The present method, based on the perturbation theory, is practical and useful.     

临界实验中本底计数对倒数外推临界质量的影响

介绍了在硝酸铀溶液核临界安全实验装置上进行不同浓度铀溶液的临界实验时，用倒数外推方法确定硝酸铀酰的核临界质量的过程；着重论述了本底计数对倒数外推方法的影响。实验表明，溶液在不同液位下的本底计数是不一样的，在外推临界时，中子计数若不减本底则外推的临界质量将偏大，临界实验过程是危险的；若减恒定本底，则实验过程偏保守，增加了实验时间。应用本文介绍的本底扣除法将避免上述缺点。     

Criticality Safety Evaluation of Materials Concerning Pyroprocessing

Criticality evaluation of materials concerning pyroprocessing has been performed employing Monte Carlo techniques. Fresh UO₂ fuel and spent PWR fuel have been employed in an electroreducer and, according to the pyroprocessing material flow, their metallic products are also evaluated correspondingly. TRU and Pu metals are evaluated concerning the electrowinner products. Totally, 6 kinds of reflectors and 4 stuffing materials are introduced to simulate the complexity during the pyroprocessing. With the optimized water intrusion models, certain subcritical data have been determined: 33.6 kg for UO₂ fuel in electroreducer, 617.5 kg for spent fuel in electroreducer, 613.6 kg for metallic spent fuel in electroreducer, and 40.0 kg for U in electrorefiner. The conservative cases for electrowinner indicate the subcritical masses of 6.8 kg for Pu and 7.9 kg for TRU. However, with regard to the dilution effect of Cd, the Pu mass needed to reach the subcriticality changes from 11.1 kg with 1.2 kg of Cd to 500.0 kg with 2000.0 kg of Cd. The TRU mass needed to reach subcriticality changes from 13.0 kg with 1.4 kg Cd to 714.0 kg with 2856.0 kg of Cd. TSUNAMI has been employed to clarify the contributions of each nuclide quantitatively. The data obtained would therefore be used in the conceptual design study of pyroprocessing facilities.     

AP1000机组首次临界试验验证计算

首次临界试验是压水堆核电厂调试启动过程的关键环节,旨在确认核反应堆堆芯能按照设计要求达到预期的临界运行状态。本文利用西安交通大学自主研发的NECP-Bamboo程序系统对AP1000机组堆芯的首次临界试验的设计结果进行了验证计算,并与AP1000堆芯的核设计结果进行了比较。计算结果表明：预估临界状态下的硼浓度的偏差为-15 ppm,控制棒积分价值的最大偏差为-52 pcm,硼微分价值的偏差不超过0.2 pcm/ppm,反应性温度系数的偏差不超过1 pcm/K。本文计算结果的精度与高保真计算程序KENO（概率论方法）和VERA（确定论方法）的计算精度相当,为确保AP1000堆芯调试启动阶段的核安全提供了进一步的数据支撑。