Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.     

A Method of Isotopic Ratio Determination of Uranium by Ge(Li) Detector

The effect of the presence of a reentrant hole for extracting the neutron beam from within experimental systems of two different geometries is analyzed theoretically with use made of multi-group 2- dimensional discrete S_n method without resorting to bold assumptions for neutron transport nor drastic simplification of geometry. One of the two experimental systems is a rectangular light water prism 12 cm high of 40 × 40 cm² cross section, poisoned with Cd and/or In, and provided with a 1, 2 or 3 cm diameter reentrant hole. The other system is a 1″ thick natural uranium plate sandwiched between two layers of pure light water, each 4.6 cm thick, which also is provided with a 1cm diameter reentrant hole.

The following is concluded by comparing the angular neutron flux with and without the reentrant holes. With the first experimental system, perturbations of the order 10～25% is caused, which is particularly strong below about 0.3 eV, except when the hole diameter is 1cm. The perturbation effect increases as the reentrant hole becomes larger in diameter and shallower in depth. In the case of the second experimental system, the effect results in about 2% increase of the neutron flux at the bottom of the reentrant hole when the bottom is located in the natural uranium plate. On the other hand, if the bottom is in the light water region, the neutron flux is reduced by about 2～4% at the peak of the thermal neutron spectra.     

Critical heat flux for finned and unfinned geometries under low flow and low pressure conditions

This paper presents the experiment and analysis for the critical heat flux (CHF) in a vertical annulus with finned and unfinned geometries under low flow and low pressure conditions. To consider the fin effect on CHF, the tests were performed on both finned heater and unfinned heater having same dimension as finned heater without fins. An analytical model was applied to estimate the heat flux and temperature distributions along the periphery of the finned geometry. The physical phenomena observed during the experiments are discussed and the parametric trends of the obtained data are examined to investigate the CHF characteristics for the finned geometry. A new correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental data with an RMS error of 13.7%.     

Neutronic performance of the MEGAPIE spallation target under high power proton beam

The MEGAPIE project, aiming at the construction and operation of a megawatt liquid lead-bismuth spallation target, constitutes the first step in demonstrating the feasibility of liquid heavy metal target technologies as spallation neutron sources. In particular, MEGAPIE is meant to assess the coupling of a high power proton beam with a window-concept heavy liquid metal target. The experiment has been set at the Paul Scherrer Institute (PSI) in Switzerland and, after a 4-month long irradiation, has provided unique data for a better understanding of the behavior of such a target under realistic irradiation conditions. A complex neutron detector has been developed to provide an on-line measurement of the neutron fluency inside the target and close to the proton beam. The detector is based on micrometric fission chambers and activation foils. These two complementary detection techniques have provided a characterization of the neutron flux inside the target for different positions along its axis. Measurements and simulation results presented in this paper aim to provide important recommendations for future accelerator driven systems (ADS) and neutron source developments.     

先进三代核电机组反应堆达临界方式及特性分析

介绍了先进三代核电机组如何在低中子注量率的情况下通过堆外核测量系统源量程探测器监视反应堆达临界，并对其达临界过程中探测器的计数率变化进行比照、分析。通过分析发现，在低中子注量率情况下，利用反应堆启动率（或周期）的变化能够实现对反应堆临界实现与否的判断。同时，利用相对中子源不同位置的探测器计数率的变化规律，能够监测反应堆逼近临界的程度。这一反应堆达临界方式可以在诸如无源启动等低中子注量率情况下得到应用。      

Optimization study of some neutron imaging parameters in a reactor based neutron radiography facility

Neutron imaging technique can be used as a means of material Non-Destructive testing. One of common neutron sources for neutron radiography is nuclear research reactor. In this work, several neutron imaging parameters such as aperture distance and the radiography plane location from the neutron source as well as the aperture diameter have been computationally optimized to deliver a proposed neutron beam. According to the results, the aperture diameter of 3.5–4 cm which was located at 55–85 cm from the outer layer of reactor core and the position of image plane at 300–400 cm fulfills delivering of the suitable neutron flux and other required conditions. W, Fe and Pb walls with an identified length formed the convergent-divergent collimator and shielded the neutron and gamma out of beam path. Bi and Fluental filters with an optimal dimension were used to efficiently improve the neutron beam profile at a sample position.     

Irradiation experiments related to reactor safety in particular pressure vessel steels

The dependence of neutron induced embrittlement of reactor pressure vessel steels on irradiation temperature and neutron exposure was investigated for steels with different copper content. A pronounced increase of the ductile to brittle transition temperature shift with decreasing irradiation temperature was found and quantitatively determined. The influence of the neutron energy spectrum and flux density on the embrittlement was not significant.Rigs for irradiating assemblies of fracture mechanics specimens (CT and WOL) up to 100 mm thickness and also for irradiation experiments under cyclic loading were developed. Irradiation experiments with these rigs are in progress.Creep experiments on canning tubes under different load conditions (uniaxial load and biaxial load under internal and external overpressure) as well as an irradiation device for investigating defective PWR fuel rods are briefly reported.     

用于反应堆内相对中子通量密度在线测量的闪烁体光纤探测系统研制

为实现反应堆不同空间和能量的相对中子通量密度在线监测,本文研究开发了一套新型的用于狭小空间且位置灵敏的闪烁体中子探测系统。该套系统由5种探头、5路光子计数器、1台计算机及相应的软件组成。5种探头的主要构成物质分别为~6 LiF+ZnS(Ag)、~(232) ThO_2+ZnS(Ag)、~(238) UO_2+ZnS(Ag)、~9Be+ZnS(Ag)以及BGO晶体,故可测量不同能量的相对中子通量密度。其中,掺有~6 LiF的探头用于热中子的测量,BGO探头用于γ测量,其余3种探头用于快中子的测量。利用该系统进行了启明星1#装置内热中子及快中子的相对通量密度分布测量,并将测量结果与利用蒙特卡罗方法得到的理论分布结果进行了比较。考虑到理论设置参数与实际实验参数的差别,可认为测量结果是可信的。     

机械速度选择器标定技术及标定实验

机械速度选择器作为关键部件广泛应用于中子散射谱仪上，研发标定技术、研制标定设备及开展标定实验是机械速度选择器应用的前提。本文基于中国先进研究堆小角中子散射谱仪，设计了标定中子飞行时间设备的结构，确定了设备的参数。研究了漏计数对波长分辨率测量的影响，发现波长分辨率测量误差取决于死时间及高斯峰位计数率之积，若死时间不变，波长分辨率测量误差随高斯峰位计数率的增加而变大。开展了飞行时间法机械速度选择器标定实验，发现单色中子波长的理论计算结果与实验数据的高斯拟合结果非常接近；波长分辨率实验值随波长的增加而增加，与波长分辨率计算值有一定差距，这些变化和差距源自束流发散。使用漏计数对波长分辨率测量影响的规律分析了实验结果，计算出了样品位置中子通量密度上限；使用VITESS软件模拟得出了不同波长样品位置中子通量密度并验证了二维可调狭缝调节中子通量密度的效果。     

利用测热技术测量核反应堆中子通量密度

一种新型中子探测器被研究,其原理是利用带电离子在矿物中沉积的能量退火时会以热量的方式释放出来,通过测量释放的热量而确定中子通量密度。对新型中子探测器进行刻度,在反应堆内某位置测量的热中子通量密度为5.108×10¹¹ cm^-2•s^-1,与标定的热中子通量密度(5.000×10¹¹ cm^-2•s^-1)在2%内符合,说明该探测器可测量中子通量密度。本文方法制作的探测器体积小,可制作成不同形状,便于反应堆不同环境下的中子通量密度测量。选取相应中子能量反应截面较大的元素,该探测器还可测量不同中子能量的通量密度。     

双环路压水堆非对称入口条件下物理-热工特性研究

双环路压水堆存在反应堆入口流量、温度不对称的非正常运行工况。本文建立了基于CFD方法的反应堆整体三维流场模型,并耦合中子动力学计算程序和RELAP5程序,对这种非对称入口条件下的反应堆物理-热工特性进行了数值模拟。结果表明：反应堆入口流量不对称会加剧堆芯入口流量分配的不均匀性,并进一步导致局部功率变化,对反应堆安全不利;在入口温度不对称的条件下,冷却剂在下腔室的混合非常不充分,并导致堆芯入口温度分布不均匀,引起局部功率变化较大,对反应堆安全不利。     

The effect of temperature and control rod position on the spatial neutron flux distribution in the Syrian miniature neutron source reactor

The effect on the spatial neutron flux distribution for both of water and fuel temperature increase as well as the change in the control rod position are presented in the Syrian miniature neutron source reactor (MNSR). The cross-sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the spatial neutron flux distribution at different water and fuel temperatures and different control rod positions using four energy groups. This work shows that the increase in water and fuel temperatures during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. The change in the control rod position does not affect as well the spatial neutron flux distribution in the reactor except in the region around the control rod position.     

加速器中子源的中子注量测量方法

在用静电加速器中子源标定探测器的中子灵敏度实验中,采用“BF3长计数管 定标器”系统过渡,用^197Au中子活化分析方法达到了对中子注量在线、绝对监测的目的。这种方法给出与加速器束流不同角度、不同距离处的中子注量。介绍了这种中子注量测量方法。     

基于任意三角形网格的中子扩散变分节块法

传统的基于矩形和六角形几何的堆芯计算程序已不适用于具有复杂几何的新型反应堆堆芯计算,本文开展了基于任意三角形网格的多群中子扩散变分节块方法研究。首先,采用ANSYS软件对计算区域进行三角形网格剖分,并利用坐标变换将任意三角形变换为正三角形;其次,采用Galerkin变分技术建立包含节块中子平衡方程的泛函,将三角形节块内变量利用正三角形内正交基函数进行展开;最后,利用变分原理,获得中子通量密度与节块边界上分中子流的响应关系,并基于传统的源迭代法对其进行求解。基于上述理论模型开发了程序TriVNM,并采用不同几何基准题进行了验证。结果表明,TriVNM计算的堆芯keff和归一化功率分布与参考解吻合较好,该计算方法适用于复杂几何堆芯扩散计算。     

An alternative solution of the neutron diffusion equation in cylindrical symmetry

Alternative analytical solutions of the neutron diffusion equation for both infinite and finite cylinders of fissile material are formulated using the homotopy perturbation method. Zero flux boundary conditions are investigated on boundary as well as on extrapolated boundary. Numerical results are provided for one-speed fast neutrons in ²³⁵U. The results reveal that the homotopy perturbation method provides an accurate alternative to the Bessel function based solutions for these geometries.     

In-reactor tubular creep of Zircaloy-2 at 260 to 300 °C

Tubular specimens of Zircaloy-2, 23 mm diameter, have been creep tested in-reactor at 260 to 300°C (530 to 570 K). The specimens were biaxially stressed by internal pressure, with transverse stresses from 100 to 300 MN/m². Zircaloy-2 was tested in three conditions; 20% cold drawn, 70% tube reduced then stress-relieved and annealed.All creep curves, both in and out of neutron fluxes, can be represented by straight lines on log strain-log time curves. Fast neutron flux increased the slopes of the log-log creep curves of the cold-worked materials. These slopes increased from 0.24–0.27 for unirradiated specimens (and specimens in the thermal neutron flux) to 0.42–0.47 for specimens in a fast neutron flux. This means that creep rate does not diminish with time as rapidly in-reactor as out-reactor. The creep behaviour of the annealed Zircaloy-2 was little affected by fast neutron flux.     

Evaluation of Dancoff Factors in Complicated Geometry using the Method of Characteristics

A simple and efficient method to estimate the Dancoff factor in a complicated geometry, named “the Neutron current method,” is presented in this paper. In this method, Dancoff factors are evaluated from the flux values obtained by the method of characteristics (MOC). By setting appropriate neutron sources in the non-fuel regions of target geometry and then executing fixed source calculation by MOC, the neutron current method can evaluate Dancoff factors for complicated geometry. It was demonstrated that the neutron current method can easily be adopted for complicated geometries, such as a PWR fuel assembly or large-scale geometry that is difficult to handle by the traditional collision probability method. By utilizing the neutron current method instead of a traditional collision probability method, the calculation time of Dancoff factors in complicated large geometry is drastically reduced.     

ADS次临界系统中子时空动力学计算与瞬态分析

加速器驱动次临界反应堆(ADS)中子时空动力学计算需要考虑外中子源和空间分布的影响,比临界系统中子动力学计算要复杂得多。本文将改进准静态(IQS)近似与蒙特卡罗(MC)方法相结合,对于带外源的ADS次临界系统中子时空动力学过程,形状函数、动力学参数由MCNPX程序计算得到,幅度函数与集总参数热工反馈模型进行耦合计算,并开发了IQS/MC计算程序可视化操作界面。针对CIADS靶堆耦合系统参考方案物理模型,对引入束流瞬变及无保护失流工况过程进行瞬态模拟计算分析,给出了堆芯相对功率、燃料温度及冷却剂出口温度随时间的变化曲线。同时,将中子注量率进行分群计算,得到了堆芯分能群的相对中子注量率网格分布随时间的变化,模拟结果与理论分析一致。     

中子辐射屏蔽材料PVA/PEO水凝胶的制备及其作用研究

为研究一种新型中子辐射屏蔽材料水凝胶的制备及其对中子辐射的防护作用,应用物理交联法制备不同厚度的单纯和含有金属离子的PVA/PEO水凝胶;利用基于Monte Carlo模拟的SHIELD程序计算不同组分水凝胶对中子输运的影响,以期在理论上证实PVA/PEO水凝胶材料对2.45MeV中子辐射的屏蔽作用;采用BF3中子辐射探测器测量了K-400型高压倍加器发射的2.45MeV中子经过不同水凝胶后的中子通量变化。模拟计算结果显示,随着水凝胶厚度的增加,中子通量和能量逐渐减少;与单纯组比较,相同厚度含金属组中子数和能量减少更明显。BF3探测器测量结果显示,厚度为6—10cm的含金属组的中子通量计数减少的百分率显著高于单纯水凝胶组,辐射屏蔽效率与水凝胶厚度符合线性方程y=-4.51x+86.23,10m厚的含金属离子水凝胶中子通量计数的百分率可减低61.3?。结果表明,高分子聚合物PVA/PEO水凝胶对快中子辐射具有良好的屏蔽作用,含金属组的中子屏蔽效果明显优于单纯组。