核电站反应堆辐射屏蔽程序系统

核电站反应截辐射屏蔽程序系统包括源项程序、离散座标输运程度、蒙特卡罗和反照蒙特卡罗程序、点核积分程序、最佳化程度、温场程序、大气散射和结构壁屏蔽效应分析程序、数据库以及加工程序和耦合程序，本程序系统程序类型比较齐全，程序和参数配大，在核电站反应堆以及其它类型反应堆和核设施辐射屏蔽设计和安全分析中得到了广泛应用。     

MCNP4C在HTR-PM设备γ射线屏蔽设计中的应用及与QAD-CGA对比研究

本文基于球床模块式高温气冷堆核电站(HTR-PM)反应堆设备间γ射线辐射屏蔽设计工程实例,建立了乏燃料中间贮存系统的蒙特卡罗模型,给出了蒙特卡罗方法计算程序MCNP4C和点核积分方法计算程序QAD-CGA的计算剂量值,并通过对二者进行分析和对比,得知点核积分程序QAD-CGA的计算结果较蒙特卡罗程序MCNP4C的计算结果偏大.     

MCNP4C在HTR-PM设备γ射线屏蔽设计申的应用及与QAD-CGA对比研究

本文基于球床模块式高温气冷堆核电站（HTR-PM）反应堆设备间γ射线辐射屏蔽设计工程实例，建立了乏燃料中间贮存系统的蒙特卡罗模型，给出了蒙特卡罗方法计算程序MCNP4C和点核积分方法计算程序QAD-CGA的计算剂量值，并通过对二者进行分析和对比，得知点核积分程序QAD-CGA的计算结果较蒙特卡罗程序MCNP4C的计算结果偏大。     

中子反照实验的理论模拟

中子反照实验的目的是检验和标定工程设计使用的程序、参数和方法。本工作采用蒙特卡罗粒子输运程序模拟中子反照实验。模拟结果表明，铁球壳外表面各测点处的中子活化反应率及反照系数在实验数据不确定度范围内符合。     

钴源γ射线天空反照剂量率的模拟计算

本文介绍用蒙特卡罗模拟方法计算大强度~(60)Co γ射线源的天空反照剂量率。计算结果给出了天空反照剂量率与一些参数的定量依赖关系。并且对这些关系曲线进行了函数拟合,给出了拟合公式。     

中子反照效应实验的理论研究

为深化中子反照效应研究,获得宏观测量数据,开展了中子反照效应的理论研究。对中子反照效应实验进行理论设计,分析了可体现反照性质的实验状态,采用蒙特卡罗程序模拟了不同实验状态下的注量和反应率空间分布。结果表明,设计的实验中测点注量能谱可体现中子反照效应,给出的实验布局方案是有效的。获得实验测量数据可给出中子反照效应更多的定量信息。     

加速器驱动的次临界系统质子束管产生的辐射剂量

利用MCNP程序,根据IAEA提供的ADS———NEUTRONICBENCHMARK中的模型和数据,对加速器驱动的次临界系统质子束管中中子和光子泄漏产生的辐射进行了计算。通过计算与分析可知:有较大份额的中子和光子直接从质子束管中泄漏出去,且泄漏的中子和光子的能量相当高;在质子束管周围的辐射剂量有80%以上是由于质子束管中中子和光子辐射产生的,只有相当少的辐射剂量来自堆本身。     

液体闪烁探测器中子γ射线测量研究

液体闪烁体对中子的响应函数,利用O5S蒙特卡罗程序进行计算,γ射线响应函数利用MATHA蒙特卡罗程序进行计算。分别对液体闪烁探测器(n,γ)分辨品质测量、Am-Be中子源的中子谱和γ射线能谱的测量、铁球在D-T中子照射下泄漏中子和伴生γ射线能谱的测量、水球在D-T中子照射下泄漏中子和伴生γ射线能谱的测量等内容进行了介绍,并对结果进行了分析和讨论。     

MCMG程序用于多群截面检验的可行性研究

以三维多群蒙特卡罗输运程序MCMG模拟计算了5个一维快临界基准实验装置和1个三维快临界基准装置的有效增殖因数和注量谱或泄漏谱,并与确定论多群离散纵坐标程序ANISN和连续截面蒙特卡罗程序MCNP的计算结果进行对比分析。结果表明,MCMG计算结果与ANISN结果有较好的符合,并接近MCNP计算结果,初步验证了MCMG应用于多群截面检验的可行性。     

压力容器环形空腔上部材料屏蔽计算分析

用离散纵座标SN法与反照蒙特卡罗AMC法相耦合,对秦山600MW核电厂反应堆压力容器环形空腔上部密封环区、褐铁矿混凝土及电离室入口轻质保护材料的中子和γ射线注量率、剂量率和释热率进行计算分析.计算结果表明,对于停堆后人员可能进入的部位,增加轻质保护材料能有效降低中子辐射;采用SN-AMC计算技术能较好地完成反应堆的大尺寸空腔和复杂孔道屏蔽设计计算.     

The neutron flux in a spherical void

Using the basic theory developed in our earlier work (Cassell, J.S., Williams, M.M.R., 2005. Particle flux in an annular gap about a sphere, Annals of Nuclear Energy 32, 457, we have evaluated the neutron flux across a spherical void due to a point source in a moderating and absorbing medium. Neutron motion in the moderator is described by diffusion theory and that in the void by the free streaming Boltzmann transport equation. An explicit solution is obtained in the form of an infinite series. This is evaluated numerically for a number of practical cases and comparison is made with an exact transport calculation using a Monte Carlo code. The hybrid method is seen to be highly accurate.     

Development of a physics analysis procedure for the prismatic very high temperature gas-cooled reactors

A new physics analysis procedure has been developed for a prismatic very high temperature gas-cooled reactor based on a conventional two-step procedure for the PWR physics analysis. The HELIOS and MASTER codes were employed to generate the coarse group cross sections through a transport lattice calculation, and to perform the 3-dimensional core physics analysis by a nodal diffusion calculation, respectively. Physics analysis of the prismatic VHTRs involves particular modeling issues such as a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment and state parameters. Double heterogeneity effect was considered by using a recently developed reactivity-equivalent physical transformation method. Neutron streaming effect was quantified through 3-dimensional Monte Carlo transport calculations by using the MCNP code. Strong core-reflector interaction could be handled by applying an equivalence theory to the generation of the reflector cross sections. The effects of a spectrum shift could be covered by optimizing the coarse energy group structure. A two-step analysis procedure was established for the prismatic VHTR physics analysis by combining all the methodologies described above. The applicability of our code system was tested against core benchmark problems. The results of these benchmark tests show that our code system is very accurate and practical for a prismatic VHTR physics analysis.     

Measurement of dose rate profile and spectra through a cylindrical duct vis-à-vis Monte Carlo simulation studies for optimisation of reactor shield design

In the design of a nuclear reactor, penetrations are provided in the top shield to carry out some essential operations. Radiation streaming is envisaged through such penetrations. To avoid radiation streaming, complementary shielding is provided. Optimisation of complementary shielding is carried out by performing calculations using MCNP code. Uncertainties in the calculations are taken care of by incorporating a safety factor. The assumption of the safety factor, while designing the reactor shielding, has been validated by undertaking experimental measurements on a similar geometry vis-à-vis the computed values obtained using MCNP code. The results of the present work agree with the safety factor of two assumed during the shield design. The details of gamma spectral measurements carried out with high purity germanium detector to understand the pattern of the scattered spectrum are also presented.     

Anomalous Isotope Fractionation in Uranium Enrichment Process

A benchmark calculation for a deep penetration problem of 14 MeV neutrons through a 3m thick iron slab was carried out by using a vectorized continuous energy Monte Carlo code MVP with the JENDL-3 and ENDF/B-IV cross sections. Reference solutions for neutron spectra and averaged cross sections were obtained at various locations through the iron slab with good statistics owing to a high computation speed of the code. The accuracy of multigroup calculations with the JSSTDL/J3 library was investigated by comparison with the obtained reference solutions.

Both calculations with JENDL-3 and ENDF/B-IV showed a similar attenuation of total fluxes from thermal to 14 MeV through the slab, while differences of one order at the maximum were observed in the calculated fluxes in the resonance energy region. The multigroup calculations with the JSSTDL/J3 295- and 125-group libraries underestimate the streaming effect through the cross section minima above the well-known 24 keV window, which resulted in the underestimation of fluxes above this window by more than two decades at 3 m penetration compared with the continuous energy method. Taking into account the spatial dependence of averaged cross sections, the underestimation was reduced to about one decade. It was found, however, that an accurate prediction of streaming effect is fairly difficult by the multigroup method.     

Application of importances or weight windows in MCNP4C to a geometry similar to an ITER equatorial port

Neutron streaming through the gap between a port plug and the port walls will be a major contributor to the shutdown dose rate in ITER, and calculating this streaming requires very effective variance reduction. The article describes a way of estimating importances for a calculation of the fast neutron flux, which will then generate weight windows. If necessary, these can be refined iteratively. Once the fast neutron calculation works well, the weight windows can be extended to lower energies and to photons using appropriate factors for which some values are suggested here. Again, these weight windows can be improved iteratively. Test runs in a simplified geometry performed fairly well, though optimizing performance simultaneously for the gamma flux at the end of the port and above it is difficult.     

Experiments and Calculations of 14 MeV Neutron Streaming through Multi-Layered Slit Assembly

A streaming experiment using a D-T neutron source was carried out to verify the calculational technique for neutron transport in a shield assembly with multi-layered slits. Reaction rate distributions of a small spherical NE213 scintillation detector to fast neutrons were measured in the slits made of 304SS and in the mortar surrounding the slits. The energy spectrum of fast neutrons in the slit was also measured with the same detector. These measurement were compared with calculations using the continuous energy Monte Carlo transport code MCNP. The calculated reaction rates in the slits agreed with the measured ones within experimental and calculational errors. Besides, it is suggested that the attenuation of fast neutrons in the mortar is significantly different from that in the slits and the behavior is nearly traced by the calculation with the MCNP code. The measured and calculated spectra at a position close to the exit inside the lower slit agreed within the both errors.     

Development of Discrete Ordinates SN Code in Three-Dimensional (X,Y, Z) Geometry for Shielding Design

A discrete ordinates transport code ENSEMBLE in (X, Y, Z) geometry has been developed for the purpose of shielding calculations in three-dimensional geometry. The code has some superior features, compared with THREETRAN which is the only code of the same kind so far developed. That is, the code can treat higher order anisotropic scattering and employs a coarse mesh rebalancing method. Moreover it has a negative flux fix-up routine using a variable weight diamond difference equation scheme and has a ray-effect fix-up option using a fictitious source based on S_N→P_N-1 conversion technique. Formulations for these advanced features in three-dimensional space have been derived.

As the demonstration of the capabilities of the code, several numerical analyses and an analysis of an annular duct streaming experiment in JRR-4 at Japan Atomic Energy Research Institute, have been performed.

As a result of these analyses, confirmation has been obtained for the prospect of applicability of ENSEMBLE to practical shielding design.     

A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor,the Fokker-Planck equation(FPE)governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper.Besides,the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found.As for the results,the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found.Also,different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod,i.e.Krypton,Xenon,Iodine,Bromine,Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution.The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution.The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap.It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.     

Nuclear analysis of a laser fusion reactor,SENRI-I

A neutronic analysis of the laser-driven inertial-confinement fusion reactor SENRI-I is presented. Three-dimensional Monte Carlo calculations were performed to examine the effects of laser beam ports on the flux distribution, tritium breeding ratio, thermal energy deposition in the blanket, and radiation streaming. A Monte Carlo code was also used for the time-dependent radiation-damage analysis accounting for the time of the flight spread of neutrons and the results are compared to the analysis for the HIBALL design. Induced radioactivity was estimated, based on the one-dimensional transport calculation and depletion analysis. The calculated results reveal the advantages of the SENRI-I design with a thick Li layer compared to other reactor systems employing a dry-wall scheme.     

A simple treatment of increased gap due to fuel assembly bowing through correction of cross sections

Increased fuel assembly gap due to bowing in commercial light-water reactors (LWRs) has an impact on local pin-power distribution due to increased local moderation. In order to consider the effect of increased assembly gap without explicit consideration of increased gap width, a correction method of cross sections in the gap region is proposed. In this cross section correction method, the average chord length of gap region is preserved to capture the effect of increased gap width. The validity of the present method is confirmed by verification calculations in a single assembly and 5 × 5 assembly geometries using the GENESIS code, which is a transport code based on the method of characteristics.