Analysis of Heavy-Water-Moderated,Cluster-Type Fuel Lattices by Cluster Physics Code ‘MESSIAH’

This paper presents a comprehensive analysis performed by a new cluster analysis code ‘MESSIAH’ on reactor physics constants measured in the critical facility for a pressure-tube-type, heavy-water-moderated reactor. The MESSIAH code system utilizes the method of the collision probability to solve the neutron transport equation. The effective space dependent cross sections are calculated in the thermal and resonance energy range before the eigenvalue calculation for the whole energy range. With use of these cross sections, the multi-group, space dependent transport equation is solved, and the flux distribution, spectrum and k _eff are obtained to the input bucking. In the above three steps the method of the collision probability is used consistently and extensively. The treatment of leakage neutrons from lattices in MESSIAH is also confirmed by an independent method using a Monte Carlo calculation. The calculated reactor physics constants, especially the micro-parameters and the activation traverse of Dy, agreed fairly well with the experiment. The diffusion calculation with use of the group constants calculated by MESSIAH predicts the reactivity of 0% void core excellently (<0.12%). However, for a 100% void core, the calculated reactivity was slightly lower than the experiment (~0.74%), which was attributed to over prediction of the diffusion constants.     

Improvement of Monte Carlo Lattice Burnup Calculation Performance with the Correlated Sampling Method

For next generation reactor designs, which are attempting wide variations of assembly configurations, the flexibility Monte Carlo method holds is attractive, but still costly for repetitive design study works. This paper presents an advanced correlated sampling (ACS) method which was developed to speed up Monte Carlo lattice burnup calculations. The ACS method is the combination of the correlated sampling method and a pseudo-scattering technique. All burnup steps are considered as consecutive perturbed problems using the same neutron collision history, which is pre-calculated based on a selected unperturbed problem. Since neutron weights can be adjusted on every collision point, rather than along paths between them, the perturbed calculation is very fast and the neutron collision history is light enough to be stored in memory or physical storage, which is an indispensable feature for consecutive perturbed calculations. The presented theory shows that the ACS method has good potential to work for a wide range of neutron absorption variations, the dominant perturbation in the lattice burnup. In an example calculation on a BWR lattice, the ACS calculation results of 600,000 neutrons/step agree well with the independent Monte Carlo runs of 20,000,000 neutrons/step within 0.1%dk/k in terms of k_? throughout 95 steps (~50GWd/t). Average calculation time of neutron tracking with the former method is 3.4 s/step with 600,000 neutron histories on a single processor of an Alpha21164-600 MHz, and the speed-up factor against the Monte Carlo calculation turns out to be about 100.     

薄膜和纤维塑料闪烁体中子能响特性的Monte Carlo数值计算

根据中子与塑料闪烁体的作用机制及塑料闪烁体的荧光转换特性,研究了塑料闪烁体薄膜和塑料闪烁体纤维中子能响特性的Monte Carlo数值计算方法,给出了不同厚度的塑料闪烁体薄膜和不同直径的塑料闪烁体纤维对裂变能区中子的灵敏度计算结果,并对计算结果及其在电流型坪响应裂变中子探测器方面可能的应用进行了初步探讨。     

~(115)In中子非弹性散射截面的实验测量及蒙特卡罗修正

~(115)In是一种重要的活化材料,准确测量它的中子非弹性散射截面数据对中子注量监测具有重要意义。在四川大学原子核科学技术研究所2.5 MV静电质子加速器上,利用核反应D(d,n)~3He产生的单能中子,以~(197)Au作为标准,采用活化法测量了2.95 Me V、3.94 Me V、5.24 Me V能点的~(115)In中子非弹性散射截面。用Monte Carlo程序MCNPX(Monte Carlo N-Particle eXtended)对靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及注量率衰减效应等进行了修正计算,得到最终结果与Loevestam的计算值符合较好,并且实验中可通过减小靶管、靶底衬、水层及样品的包层材料等厚度来减小多次散射效应和自屏蔽效应的影响。     

Benchmark experiment on titanium with DT neutron at JAEA/FNS

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li₂O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of ²³⁵U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′_cont) reaction cross section data and resonance parameters were improved.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Comparison of calculated and measured reaction rates obtained through foil activation in the subcritical dual spectrum facility YALINA-Booster

Reaction rates were measured by the foil activation technique to obtain neutron spectrum information in a subcritical core driven by an external neutron source. The experimental results are compared with Monte Carlo calculations in order to examine the capability of the Monte Carlo code MCNP together with ENDFB-6.8, JEFF-3.1.1 and CENDL-3.1 neutron cross section libraries to predict the neutron spectrum dependent reaction rates correctly in a subcritical core. The focus lies on fast neutrons. A discrepancy is found in the calculated-to-experimental values of the reaction rates and an inaccurate cross section is identified in CENDL-3.1.     

MCMG蒙特卡罗多群-连续截面耦合中子输运计算

本文针对多群蒙特卡罗计算省时但共振自屏处理存在缺陷,以及连续截面蒙特卡罗输运计算精度高但计算费时的问题,发展了一种多群-连续截面耦合计算方法。该方法在自主研发的三维中子-光子耦合输运蒙特卡罗程序MCMG中得到应用,通过多个模型的计算验证了方法的有效性。MCMG耦合计算取得了与连续点截面MCNP程序一致的结果,其计算速度较MCNP的提高了1倍左右。     

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample.     

The total cross section and resonance parameters for the 0.178 eV resonance of Cd

It has been suggested that the capture and scattering cross sections of natural cadmium are not well described by the resonance parameters that are given in the evaluated data files. In particular, doubts on the parameters of the first resonance of ¹¹³Cd at 0.178 eV have been raised. This resonance is of high importance in the interpretation in many integral experiments, such as neutron activation analysis, in which cadmium foils are used to shield from thermal neutrons. A new set of experiments has been designed and performed at the neutron time-of-flight facility GELINA, to determine the total cross section and to extract a set of resonance parameters. The covariance information of the experimental data is propagated and the correlation between the resonance parameters is derived. The obtained parameters are then compared to the data available in the literature. Finally a set of criticality experiments from the international handbook of evaluated critical safety benchmark experiments is used to quantify the influence of the change in the resonance parameters.     

全陶瓷微密封燃料有效多群截面计算方法研究

全陶瓷微密封（FCM）燃料是一种弥散颗粒燃料。由于弥散颗粒燃料存在双重非均匀性,传统的确定论方法及蒙特卡罗方法皆难以处理这种双重非均匀效应以获得有效多群截面。本文基于超细群方法建立FCM燃料的有效多群截面计算方法。为描述燃料棒内TRISO颗粒的非均匀性,在共振能量段,通过采用超细群方法求解包含TRISO颗粒的一维球模型得到超细群缺陷因子,通过超细群缺陷因子修正所有核素的超细群截面即可将颗粒和基质均匀化。由于TRISO颗粒在热能区也存在较强的自屏效应,在热能区,利用穿透概率及碰撞概率等价得到多群缺陷因子,通过多群缺陷因子修正所有核素的多群截面将燃料和基质均匀化。均匀化后的FCM燃料组件即可视为普通压水堆燃料组件进行共振计算。利用丹可夫修正因子等价得到FCM燃料组件各燃料棒的等效一维棒模型,对一维棒模型求解超细群慢化方程从而得到共振能量段的有效自屏截面。数值结果表明,该方法能有效处理FCM燃料的双重非均匀性,得到精确的有效自屏截面。     

Effect of Resonance Scattering of Sodium on Resonance Absorption of U-238

The interference effects of the resonance scattering of Na on the resonance absorption of ²³⁸U were investigated. The results for self-shielding factor obtained from exact treatment were compared with those from the usual conventional method, as well as with results based on the assumption that φ(u)σ_t=const. (φ(u): flux, σ_t: total cross section).

In the exact treatment, the neutron spectrum used is obtained with accurate treatment of the slowing down by the elastic scattering of light and medium elements. In the φ(u)σ_t = const, assumption, the energy dependence of the Na resonance scattering cross section is taken into account as in the exact treatment, whereas only the averaged value is used in the conventional method.

Self-shielding factors and their temperature dependence were calculated for several compositions and energy regions. It was found that the conventional method is not satisfactory, while the assumption that φ (u)σ_t= const, compares well with the exact treatment.     

Automatic Plotting of Isoflux Contours with On-Line Printer

The effect of neutron streaming upon the neutron multiplication factor in a fast critical assembly FCA-VI is estimated by considering the anisotropy of the diffusion coefficient. In this paper, the Benoist formula is adopted to obtain the diffusion coefficient. In his original expression, many effects such as those of neutron angular distribution, multiple collision (correlation of different neutron passes) and finiteness of lattice system are included. In order to estimate these effects, the generalized first-flight collision probability method is adopted as in the previous paper, and an improvement is brought over the previous method so as to calculate the effective diffusion coefficient in a practical slab lattice cell which is asymmetric about the center of the cell.

Using the method of calculation described above, the anisotropic effect of neutron streaming in the lattices TA, TB, TC and TD used in the FCA VI-1 assembly is estimated. The effect of anisotropy of the diffusion coefficient upon the multiplication factor for the TA, TB, TC and TD lattices is found to be ?0.156, ?0.181, ?0.242 and ?0.330% δk/k, respectively.

Usually, the effects of neutron distribution, multiple collision and anisotropic scattering are neglected in evaluating the diffusion coefficient when using the Benoist formula. Among these factors, the effect of multiple collision of neutrons has the largest effect upon the diffusion coefficient and upon the neutron multiplication factor. For the TA lattice, 30% of the overall effect of anisotropy is attributable to multiple collision.     

A cold neutron beam for the experiment in grenoble

The neutron beam for the experiment is produced at the HFR reactor of the ILL Institute in Grenoble. Cold neutrons are transported by means of a slightly bent neutron guide with a 6 × 12 cm² cross section 60 m long to the experimental area, where they are injected into a divergent guide specially designed in order to reduce the beam divergence. The measurements of the beam intensity and profile are compared to the Monte Carlo calculations.     

Application of Multiband Method to Pin Cell Calculations

Abstract

In order to accurately calculate effective neutron cross sections in the resonance energy region, the multiband method has been applied to cell calculations. Cell calculations for UO² and MOX fuels of light water reactors have been performed and the results were compared with those of a continuous energy Monte Carlo code VIM and the conventional self-shielding method using the Dancoff factor.

The k∞values calculated by the multiband method agreed with those of the VIM calculations within 0.20% Δk for the UO² fuel cell and within 0.30% Δk for the MOX fuel cell, respectively, whereas the Dancoff factor method yielded about l.l%Δk errors for the two cells. The element- wise contribution to this error was investigated, and it was found that the effective microscopic cross sections, particularly those for the giant resonances of ²³⁸U, calculated by the multiband method were in good agreement with those of VIM. It was also found that interference effect between ²³⁸U and ²³⁵U resonances in the UO² fuel and that between ²³⁸U and ²³⁹Pu resonances in the MOX fuel made about 0.20%Δk contributions to k∞ in both fuel cells.     

Detailed spatial measurements and Monte Carlo analysis of the transportation phenomena of thermal and epithermal neutrons from the 12-GeV proton transport line to an access maze

In order to investigate the neutron transportation from a beam-line tunnel to an access maze at a 12-GeV proton accelerator, we measured the spatial distribution of thermal and epithermal neutrons by using the Au activation method in detail. Gold foils were placed at about 70 positions in the maze in the case of the insertion (or extraction) of a copper target of 1 mm thickness into (or from) the beam axis in front of the maze. After the end of accelerator operation, relative activities of the Au foils were simultaneously measured by using an imaging plate technique and the radioactivity of one reference foil was also measured with a HPGe detector to convert to the absolute activities of all foils.It was found that the neutrons reach to the depth of the maze in the case of the insertion of the copper target. This result reflects higher proportion of high-energy particles from the copper target to that from other beam loss points and high-energy particles become the successive source of low-energy neutrons. Furthermore, it was found that several circumstances such as door walls and electric wire cables obviously affect the absorption effect of thermal neutrons. The reaction rates obtained in this study were also used for the benchmark of the Monte Carlo simulation code, MARS15 (version of February 2008). The results of the MARS15 calculations precisely reproduced experimental results and significant effects of the electric wire cables and door walls.     

海底原位PGNAA探测含水率对中子分布影响研究

运用MC法模拟14 MeV快中子进入沉积物后的物理过程,得到了不同深度下沉积物中子能谱分布,分析了0~1 eV中子在沉积物中横、纵向分布规律以及含水率与Cl~-浓度对中子分布的影响,探讨了模拟条件下0~1 eV中子扩散的最大深度范围。结果表明:沉积物含水率对中子能谱分布和0~1 eV中子横、纵向分布均有显著影响,进行中子活化的最佳深度为2 cm处;0~1 eV中子扩散的最大深度范围为20~40 cm;海水中Cl~-浓度对中子吸收有影响但影响不显著。     

反应堆中子注量率相对分布测量系统铅屏蔽计算分析

为批量化测量探测片活性,设计了多道中子注量率相对分布测量系统。每一个NaI探测器测量一个探测片,探测器之间采用铅屏蔽。为了消除其他通道探测片对本通道探测片测量结果的影响,通过设计保证邻近通道活化探测片对本通道干扰小于0.05%。本文采用点源积分方法计算(经蒙特卡罗模拟验证),并确定了满足屏蔽要求的铅屏蔽体尺寸。     

Applicability of non-analog Monte Carlo technique to reactor noise simulation

Reactor noise simulations have been performed with the analog Monte Carlo technique in the past. The applicability of the non-analog Monte Carlo technique, which uses “weighing” for the purpose of variance reduction, to reactor noise simulations has been discussed. The joint probability of a pair of counts and Feynman variance-to-mean ratio in the non-analog Monte Carlo technique are formulated for one-speed neutron random walk in an infinite homogeneous medium. Unlike the analog Monte Carlo technique, the fission-related correlation exists even for the number of fission neutrons ν = 1 because the neutron causing a fission survives and will contribute to subsequent detections. As a results, the joint probability and the variance-to-mean ratio has the same function of time as the analog Monte Carlo technique. The probability of an uncorrelated pair of counts for a coincidence detection within one detector is different from the analog Monte Carlo technique, which introduces an extra unknown parameter in Feynman-α method. In the two-detector system, the extra parameter does not arise and the conventional Feynman-α formula can be applied to non-analog Monte Carlo simulations. The formulations derived in this work are verified by the fact that the theoretical values agree well with the numerical results.     

Fission neutron spectrum of Cf

Average cross sections for the unmoderated spontaneous fission neutron spectrum of ²⁵²Cf have been measured for 10 different threshold reactions by activation method. On the basis of the measured data and of the excitation functions the Maxwellian temperature for ²⁵²Cf neutron spectrum has been deduced and a value of T = 1.41 ± 0.02 MeV was found between 2.5 and 15 MeV. The temperature has been determined also from the average energy which was deduced from the age of epithermal neutrons in water measured by gold foil. The high σ values obtained by the low threshold reactions and the age method confirm the assumption on the excess of neutrons in the low energy region of the spectrum.