固定式中子剂量仪器现场标校方法研究

本文利用小型可控中子源和中子剂量参考仪器组成的核设施现场标校系统,通过空间位置对称条件下的待校准仪器与参考仪器对中子管出射中子响应之比,得到校准系数;为了完成仪器的周围剂量当量校准系数从国家标准计量检定中心参考辐射场到核设施现场的传递,必须对其进行能量修正。通过Am-Be中子源模拟核设施内部辐射场,在其中进行中子剂量仪器现场标校实验,并与校准过的多球中子谱仪测量结果相比较。结果表明:使用未经修正的校准系数对待校准仪器的测量结果进行校准,周围剂量当量率相对偏差为17.0%;使用修正后的校准系数对待校准仪器的测量结果进行校准,周围剂量当量率相对偏差为-2.4%。     

多球中子谱仪的周围剂量当量转换研究

以辐射场全能段中子能谱获取及周围剂量当量转换为目的,针对一套以3He正比计数器为探测器的多球中子谱仪,介绍了采用脉冲幅度分析法进行中子-γ甄别,准确测量中子计数率的方法;测量并获取了实验室本底环境中子、Am-Be中子源、氘氘中子管连续出射中子三种辐射场的中子能谱,并予以分析;基于中子能谱进行周围剂量当量转换,分析了不同能量段中子所占的剂量比重,发现快中子是主要贡献者;在比较测量中,对三种单慢化体中子剂量仪器产生的误差进行了理论分析,并给出了依靠多球中子谱仪测准中子周围剂量当量的三点建议。多球中子谱仪可以直接测算周围剂量当量,可以指导其它中子剂量仪器的标校,还可以针对具体能量段进行中子防护设计,精准有效指导外照射剂量防护工作的开展。     

BH3105型高灵敏中子剂量当量仪

考虑到在传统的“吸收筛”设计原理框架内难以设计成高灵敏度中子剂量当量测量仪,研制了一种基于全新的中子剂量生物等效调整原理——“吸收棒”法的BH3105高灵敏度中子剂量当量仪。该仪器的灵敏度达10cps/(μSv·h~(-1)),比国外同类仪器高18～40倍,比国产FJ342型中子雷姆仪高约10~4倍。其量程范围为0.1～1Sv/h,比同类仪器扩展1～2个     

BH3105型高灵敏中子剂量当量仪

考虑到在传统的“吸收筛”设计原理框架内难以设计成高灵敏度中子剂量当量测量仪,研制了一种基于全新的中子剂量生物等效调整原理——“吸收棒”法的BH3105高灵敏度中子剂量当量仪。该仪器的灵敏度达10cps/(μSv·h~(-1)),比国外同类仪器高18～40倍,比国产FJ342型中子雷姆仪高约10~4倍。其量程范围为0.1～1Sv/h,比同类仪器扩展1～2个量级。其它性能,如耐γ性能、能量响应及方向性等,皆达到与超过国外同类产品水平。     

一种新型中子剂量当量仪的电子学系统设计

介绍了一种新型中子剂量当量仪的电子学系统的设计,该系统由电源电路、放大器电路、脉冲幅度甄别电路和仪器控制电路等部分组成.该系统能够满足该新型中子剂量仪的功能要求,能够实现该新型中子剂量仪的信号采集、结果计算和仪表控制等功能.     

中子等剂量当量探测实验方法研究(英文)

对中子剂量当量仪实现中子生物等效探测的原理进行了研究。讨论了“吸收筛法”与“多探测器法”两种传统的实验方法。在此基础上,提出一种新的实现中子生物等效探测的原理——“吸收棒法”,以期实质性地提高该类仪器的中子灵敏度,克服两种传统方法的缺陷。并基于新原理,研制成功一种新型的中子剂量当量仪。该仪器的中子灵敏度达10cps/(μSv·h~(-1)),比国内外现有的同类仪器指标(0.23～0.56cps/(μSv·h~(-1))),高18至40倍。其它技术指标也皆达到或超过国际上现有的同类产品水平。因而证实了“吸收棒法”实现中子生物等效探测原理的科学性、先进性与实用性。     

《辐射防护仪器 中子周围剂量当量(率)仪》国家标准审查会圆满结束

正2018年8月15日至17日,全国核仪器仪表标准化技术委员会辐射防护仪器分技术委员会(SAC/TC30/SC3)在厦门组织召开了《辐射防护仪器中子周围剂量当量(率)仪》国家标准审查会。会议由秘书长李运文主持,主任委员肖雪夫主审。标准编制组首先汇报了《辐射防护仪器中子周围剂量当量(率)仪》国家标     

高灵敏中子剂量当量率仪宽能区响应函数曲线的校准

参考国家相关标准和国际标准化组织(ISO)及国际电工委员会(IEC)对中子剂量当量(率)仪校准的相应标准,对已设计加工组装完毕的一台长圆柱型、较大体积高灵敏中子剂量当量率仪在宽能区(热中子～20MeV)内进行了能量响应和剂量灵敏度的实验校准。能谱响应函数曲线实验校准采用参考仪器和监测仪器归一校准法,利用实验和理论计算方法确定了NH-1型中子剂量当量率仪在较宽能区(热中子～20MeV)内的能量响应曲线;并给出了仪器的综合灵敏度48.9cps/(μSv.h-1)以及在3种特征能谱中的剂量响应。     

中子周围剂量当量率测量技术的发展与现状

中子周围剂量当量率的监测对辐射防护有重要的意义。通过对中子探测元件和中子剂量当量率仪发展过程的调研可知,根据其原理和结构可将其分为:单探测器单慢化体剂量当量率仪、单探测器多慢化体剂量当量率仪和多探测器集成剂量当量率仪,并对γ峰修正探测器的研究现状和优缺点进行了探讨。本文对这4种类型仪器的发展成果以及存在问题进行了总结与分析。     

BH3105E型中子剂量当量仪研制

研制成功BH3105E型中子剂量当量仪.仪器采用镉吸收棒法实现热中子至14 MeV区间等剂量当量测量;选用6Li玻璃闪烁体作为中子探测器;探头输出信号送至硬件电路进行处理.信号经主放大器、脉冲幅度甄别器、整形器由处理器进行计数采集.仪器系统软件设计实现了测量数据存储,采用串口协议与单片机进行通讯.     

采用无源符合中子法测量含钚材料

研究了无源符合中子法测量不同含Ｐｕ材料的方法和技术。采用高计数率符合中子计数器测量材料中的２４０Ｐｕ等核素自发裂变产生的裂变中子，通过计算等效２４０Ｐｕ量来确定材料中的总含Ｐｕ量。在现场实测了含Ｐｕ炉渣等十多类各种物理形态的材料。讨论了测量数据的修正及误差来源。测量误差一般在８％—１１％，大部分结果与天平称重估算法在１０％内符合。     

能谱变化对239Pu裂变室测量快中子注量的影响

热中子和共振区的中子在快中子临界装置中所占的份额很小,但是由于其相对大的截面,在慢化物存在的情况下,热中子和共振中子份额的微小变化,对^239Pu裂变室测量中子注量的结果影响很大。通过测量^239Pu裂变电离室在包镉和包硼、周围有无慢化物等情况下的反应率,Au、In活化片的镉比,S活化片在能谱变化下与^239。Pu的反应率比等,分析了快中子临界装置中热中子和共振区中子的分布,讨论了中子能谱变化对^239Pu裂变室测量快中子注量的影响及解决办法。     

Research and Improvement of Pu Sample Property Measurement Based on Fast Neutron Multiplicity Counter

Fast neutron multiplicity counting (FNMC) analysis method can effectively measure the properties of samples. Based on the fourth-order FNMC analytical equations, a set of three-layer fast neutron multiplicity counters with six liquid scintillators per layer was constructed for Geant4 simulation, and the values of related parameters were determined. Metal Pu sample with 1 cm iron, aluminum, carbon, and stainless steel packaging material was externally simulated, and the sample satisfied the assumption by the equation adaptive analysis. The measurement parameters such as detection efficiency and multiplicity counting rate were simulated. When the mass of Pu sample is less than 500 g, the increase of sample solution mass deviation is less than 1.20% with carbon as packaging material, and the influence of iron material and stainless steel material is less. According to the measurement results, the self-multiplication factor was corrected for the sample without shell, and the third-order polynomial fitting equation was obtained and the goodness of fit is 0.933. The corrected solution mass deviation of sample with mass less than 1 kg is less than 6.00%. The results show that the medium-heavy metal with thickness of 1 cm has little effect on the measurement of Pu samples. The combination of the fast neutron multiplicity counter and the coefficient correction method can achieve more accurate measurement of the sample properties.     

基于快中子多重性计数器的Pu样品属性测量研究及改进

快中子多重性计数（fast neutron multiplicity counting, FNMC）分析方法能有效实现对样品属性的测量。本文在研究四阶FNMC分析方程的基础上，利用Geant4模拟搭建了1套3层、每层6个液闪的快中子多重性计数器，并确定了相关参数的数值。模拟设置外部具有1 cm厚的铁、铝、碳和不锈钢包装材料的金属Pu样品，通过方程适应性分析，该放射源基本满足设定的假设。对探测效率和多重计数率等测量参数进行模拟，当Pu样品质量在500 g以内时，由于增加碳作为包装材料，使得样品求解质量偏差增大的幅度小于1.20%，铁材料和不锈钢材料影响较小。根据测量结果，对无外壳条件下的样品进行增殖系数修正，得到三阶多项式拟合方程，拟合优度为0.933，质量在1 kg以内的样品，修正后的求解质量偏差小于6.00%。研究结果表明：1 cm厚的中重金属对Pu样品的测量影响较小，模拟搭建的快中子多重性计数器和系数修正相结合的方式实现了对样品属性的较准确测量。     

Analysis and improvement of the influence of sample shape on Pu measurement results

ABSTRACT

The analysis method of Fast Neutron Multiplicity Counting (FNMC) plays an increasingly important role in the measurement of nuclear material properties. Based on the assumption of point model, fast neutron multiplicity measurement equation is derived which can be used to measure the mass of Pu sample. However, the deviation of the simulated measurement of 1 kg Pu sample reaches 16.6% and increases with mass. Because nonpoint source samples of different shapes do not fully satify the hypothesis. To correct this deviation, a set of fast neutron multiplicity counters was built by Geant4 to simulate and study the mass attribute of Pu samples.The cylindrical sources of different shapes and different masses were simulated, the self-multiplication factor and α coefficient were corrected.And the corresponding third-order polynomial fitting equation was obtained, the goodness of fit was greater than 0.970. In the same way, the spherical and spherical shell source samples in the mass range of 0–5 kg were analyzed, the corrected mass deviation of samples was less than 10% in this interval. The results show that the combination of the fast neutron multiplicity counter and parameter correction can accurately measure the sample mass attribute.     

Implementation of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

An accelerator-driven system (ADS) combined with a subcritical molten salt reactor (MSR) is a type of hybrid reactor originally designed to use Th/U (or U/Pu ) fuel cycles. In most accelerator-driven molten salt reactor (AD-MSR) concepts, the salt material is also used as a target for inducing spallation neutrons. Although a neutron source is an important component in the design of ADS, only a few studies have addressed the effects of the neutron spallation source in the AD-MSR. Incidentally, there is no quantitative study on how much the beam power can be reduced by installing a spallation target in a sodium chloride-based fast reactor. We studied the proton and the neutron source efficiencies of an AD-MSR with chloride fuels by considering an Lead Bismuth Eutectic (LBE) spallation target. This LBE target is found to increase the proton source efficiency significantly. The required beam power for an AD-MSR can be reduced by 33 % and 16 % for NaCl-Th/²³³U and NaCl-U/Pu fuels, respectively, relative to the AD-MSR without the LBE spallation target by keeping the same k_eff. The energy gain can be increased up to 1.5 times and 1.2 times for NaCl-Th/²³³U and NaCl-U/Pu fuels, respectively. Thus, incorporating a spallation target module in an AD-MSR can significantly reduce the burden on the accelerator.     

Determining Pu isotopic composition and Pu content of PuBe sources by neutron coincidence technique

The paper describes a pure neutron method for determining both Pu content and Pu isotopic composition of PuBe neutron sources by neutron coincidence technique, without using gamma-spectrometry. The new procedure based on the R/T-T relationship is a developed version of the R/T-method based on R/T-M_Pu calibration curve described in [C.T. Nguyen, J. Bagi, L. Lakosi, A novel method of quantitative assay of PuBe neutron sources by neutron coincidence technique, Nucl. Instr. and Meth. B 246 (2006) 409], utilizing Pu isotopic correlations; here R, T, M_Pu are double count rate, single count rate and total Pu content, respectively. Accuracy of the method was found to be about 2-3% and 15% for ²³⁹Pu component and Pu content, respectively. Measurement time as a function of detector efficiency is treated in detail. It is shown that in a system of frame, a transuranium neutron source can be characterized by a pair of co-ordinates [R/T,  T].     

活化法测量CFBR-II堆中子注量和中子能谱

采用活化法研究了CFBR-II堆中子能谱、中子注量分布和辐照样品对中子场的扰动。建立了用于求解中子能谱的SAND-II解谱程序。对实验结果的分析表明,活化法得到的中子注量率与裂变室得到的结果是一致的,辐照样品对中子能谱有一定的软化。     

Improved Los Alamos model applied to the neutron induced fission of 239Pu and 240Pu and to the spontaneous fission of Pu isotopes

The Los Alamos model with multiple fission chances upgraded with (a) the linear relation between the average prompt gamma ray energy and the average prompt neutron multiplicity and (b) the dependence of the average fission fragment kinetic energy on the incident neutron energy, is used for the n+²³⁹Pu and n+²⁴⁰Pu reactions, and also for the spontaneous fission of ^237–241Pu isotopes. In the case of ²⁴⁰Pu fissioning nucleus the variation of the average energy released versus the incident neutron energy is also taken into account. The calculated prompt fission neutron spectra and average prompt neutron multiplicity well represent the experimental data, proving a better predictive power of the improved Los Alamos model.     

Computation of neutron dose distribution in a cylindrical phantom of tissue equivalent materials

This work gives the total dose and the spectral distribution at any specific point inside a tissue equivalent cylindrical phantom due to Pu–Be neutron source. The absorbed dose through the material are coming from 17 reactions between the neutrons and the phantom constituents: H₂,0₂,N₂. and C. The changes in the neutron spectra are due to energy loss as a result of neutron slowing down. A computer program is used to carry out these calculations using the Monte-Carlo technique.