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。     

核径迹热释中子探测器测量高中子通量密度

本文提出一种用于高中子通量密度测量的方法,即使用核径迹热释中子探测器测量中子通量密度,该方法在低中子通量密度测量方面已成功在微型中子源反应堆上得到验证。为了测试其在高中子通量密度测量方面的适用性,在中国先进研究堆辐照孔道内进行了应用研究。结果表明:孔道内中子通量密度相对分布总体趋势与MCNP的计算结果符合较好,此种方法测量高中子通量密度有效可行。     

环形燃料反应堆通量密度分布测量

相对中子通量密度分布是反应堆的重要物理参数之一,测量环形燃料零功率反应堆堆芯相对中子通量密度分布对了解环形燃料堆芯反应堆物理特性及开展安全分析具有指导意义。本文在环形燃料堆芯多边形装载下,采用箔活化法对辐照后燃料元件外表面不同位置金箔的γ活度进行测量,得到不同位置燃料元件轴向、径向的相对中子通量密度分布,并将测量值与蒙特卡罗理论计算值进行比对。结果表明：实验测量值与理论计算值最大相对偏差在12%以内,相对中子通量密度分布测量结果符合实验设计预期,现有蒙特卡罗分析手段可较好地分析堆内元件轴向通量密度分布情况。本文结果可为环形燃料的工程化应用提供重要的数据支撑。     

反应堆辐照制备超钚核素的产额估算

本文估算了利用反应堆中子辐照生产超钚元素时,~(248)Cm以前各核素的产额与中子通量及中子能谱的关系,并以图表形式描述了生产某一核素的最佳照射时间以及在生产链上各核素的相对比值。     

脉冲堆功率分布及绝对功率的测量

本文简要介绍了在脉冲堆零功率实验装置上,利用“中子通量自动测量装置”进行的功率分布、中子通量不均匀系数和绝对功率的测量工作。     

LiD中子物理蒙特卡罗模拟计算

本工作旨在模拟计算处于反应堆堆芯适当位置的杯形LiD靶室经热中子辐照后在腔室中形成的能量约为14MeV的中子场。通过采用蒙特卡罗多分支方法模拟中子-氚的联合输运过程,得到了腔室中的中子通量、能谱以及LiD靶室行处的沉积能量。 对于带有H2O反射层的装置,计算该层内的各区中子通量。经过比对,计算结果与实验数     

核压力容器材料辐照监督管的设置

本文简要介绍压水堆压力容器材料辐照监督管的设置,分别就结构设计、辐照材料试样,温度和中子通量的测量以及使用等方面作了具体说明.     

铝及铝合金拉伸机械性能的辐照效应

本文分析了辐照温度对辐照效应的影响。发现各种铝材在温度高于200℃辐照时都不呈现辐照损伤,而在低于100℃时辐照损伤是严重的。还分析了积分中子通量与拉伸性能之间的关系。发现在通量范围不太大时(30倍以内)幂函数σ=σ_0 (λD)~(1/n)及指数函数σ=σ_0 P(1-e~(-BD))~(1/m)都能给出与实验数据相当符合的结果。最后分析了不同成分铝合金与拉伸性能辐照效应之间的关系。由此估计了某些铝材在低于100℃下辐照到～10~(21nvt)积分中子通量时的拉伸性能数据。     

CFETR氦冷陶瓷增殖包层中子学分析

为满足中国聚变工程实验堆(CFETR)包层的应用要求,本文提出氦冷陶瓷增殖(HCCB)包层方案。为验证HCCB包层设计方案的合理性与可行性,采用三维蒙特卡罗粒子输运程序MCNP,计算和分析了HCCB包层方案的氚增殖比、中子壁负载、中子通量密度、核热、辐照损伤等中子学特性。结果表明,HCCB包层方案满足氚自持要求,中子通量密度和核热分布合理,屏蔽性能良好,基本满足设计要求。     

6Li(n,t)4He反应微分截面的实验测量

用屏栅电离室对1．85和2．67MeV中子^6Li(n,t)^4He反应的微分截面及截面进行了测量。使用氚固体靶通过T(p,n)^3He反应产生中子,利用BF3长中子管进行相对中子通量监测,绝对中子通量则用^238U(n,f)反应来刻度。测量结果与已有数据进行了比较。     

Albedo Analytical Method for Multi-Scattered Neutron Flux Calculation in Cavity

A simple formula which describes multi-scattered neutron flux in a spherical cavity was derived based on the albedo concept. The formura treats a neutron source which has an arbitrary energy-angle distribution and is placed at any point in the cavity.

The derived formula was applied to the estimation of neutron fluxes in two cavities, i. e. a spherical concrete cell with a 14-MeV neutron source at the center and the “YAYOI” reactor cavity with a pencil beam of reactor neutrons. The results of the analytical formula agreed very well with the reference data in the both problems. It was concluded that the formula is applicable to estimate the neutron fluxes in a spherical cell except for special cases that tangential source neutrons are incident to the cavity wall.     

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.     

用于核素在线瞬发伽马数据测量的热中子源设计

基于MCNP程序建立了西安脉冲堆热中子源设计的蒙特卡罗深穿透耦合屏蔽计算方法;采用MCNP临界源模型计算了热柱方腔前表面的中子、伽马平面源的参数,并与实验值进行了对比,给出了平面源的修正系数;基于中子、伽马等效平面源,采用新型硼铝复合材料以及铅、铋等材料,优化设计了热中子束流滤束装置,给出了热中子束流滤束装置的升级改造方案,得到热中子通量密度较原设计方案提高3倍、中子伽马通量密度比值大于10的平行热中子束,且束流外侧区域的中子、伽马本底剂量率接近0.025 mSv/h的辐射防护标准。     

Optimization of moderator assembly for neutron flux measurement: experimental and theoretical approaches

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF03(U10and 200) and3He(U0.500)neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at(1.17 ± 0.09) 9 105 and(1.19 ± 0.1) 9 105n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements.     

Destructive Analysis Determination of Neutron Emission from Spent RBMK Fuel

Absolute calibration of the UNIT-10m setup on the basis of the results obtained by destructive analysis is examined. The calibration procedure consists in performing measurements using the SKAT laboratory test stand at the Radium Institute with a model unirradiated assembly and neutron sources with known intensity, simulating neutron emission, comparing experimental data and calculations for certain reference points. The data obtained from destructive analysis can be used to establish a relation between the neutron flux of an irradiated assembly and the parameters being determined (fuel burnup, content of the main fissioning isotopes).A relation is presented between burnup and accumulation of the main neutron-emitting isotopes in spent RBMK fuel. The data were obtained by radiochemical and - and mass-spectrometric methods. The total neutron emission is determined taking account of the neutron yield from spontaneous-fission reactions of even isotopes of plutonium and curium and from (, n) reactions.     

离散纵标输运计算方法在压水堆核电厂41Ar活化源项分析中的应用

压水堆核电厂功率运行期间,反应堆压力容器外的环形空腔空气中所含的⁴⁰Ar被中子活化,形成具有放射性的⁴¹Ar。文章采用二维离散纵标输运计算程序DORT分析了反应堆堆腔区域的中子注量率分布情况,采用NJOY评价核数据处理程序,根据DORT分析得到的通量作为权重通量,利用基础评价核数据库ENDF/B-Ⅶ.0制作⁴⁰Ar中子俘获反应的微观截面,在此基础上,分析了百万千瓦级压水堆核电厂每台机组反应堆堆腔空气中⁴⁰Ar中子活化生成⁴¹Ar的生成率以及电厂⁴¹Ar的环境排放源项。文章给出的⁴¹Ar源项分析方法可作为压水堆核电厂设计中确定⁴¹Ar源项的最佳估算值的参考。     

Existence of the fifth unstable nuclided series

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中子大气传输特性的Monte Carlo 模拟

用Monte Carlo方法计算了中子通过大气传输到不同高度轨道探测器的中子注量和能谱。研究结果表明：到达不同轨道的中子的能谱结构相同．因此中子能谱的基本结构在大气传输过程中保持不变;在保持能谱基本结构不变的前提下．随中子的传输．其低能中子份额在缓慢增大,高能中子份额在缓慢减小;大气中的中子注量超过了自由空间中相应的中子注量;能谱及注量的研究结果同时证明了中子的大气传输主要受散射机制而不是吸收机制所控制。     

Variation of environmental neutron flux with altitude and depthof both water and soil

Applying the extreme low-level y-ray spectroscopic analysis the environmental neutron flux is measured using different moderator construction and environment through the reaction ^197Au (n, γ) ^198Au- The contribution of thermal and resonance neutrons is separated using the cadmium difference technique, while fast neutrons are measured by the paraffin moderator. The results of altitude dependence of the neutron flux are discussed. The thermal neutron flux near the surface of sea water is less than its value at 100 cm over ground near sea water, while the value over the surfaces of fresh water is higher than that near the surface of sea water. Also the thermal neutron flux at 5 cm soil depth increases, then decreases to its original value at 10 cm depth and still constant until 25 cm, then decreases rapidly to reach 27% of its original value at 60 cm depth. The soil compositions, corresponding neutron temperatures and effective absorption cross sections of earth are the most effective factors on the equilibrium region of thermal neutrons in the ground.     

Neutron and Gamma Sensitivities of Dynamic Detectors

The neutron and gamma sensitivities of nine detectors were studied in experiments performed at the Sandia Pulsed Reactor (SPR). Of the detectors tested, six are predominantly gamma detectors at the SPR and three produce a significant fraction of their total current due to the neutron flux during a normal SPR burst. None of the detectors studied were found to be predominantly sensitive to the neutron flux during a normal SPR burst. Tests performed at pure gamma sources on a P-Intrinsic-N (PIN) diode detector are also described. The neutron induced current in the silicon detectors is treated theoretically.