用固体径迹探测器验证公式Φ_U（r）／_U≈Φ_M（r）／_M

利用固体径迹探测器测量反应堆不同位置燃料元件内的中子注量率，得到反应堆燃料元件内的中子注量率分布。与对应点慢化剂内中子注量率进行比较，对反应堆物理实验中一个近似假设公式ΦＵ（ｒ）／ΦＵ≈ΦＭ（ｒ）／ΦＭ进行了验证。给出了该公式成立的条件。     

用固体径迹探测器验证公式φU（r）／φ^—U≈φM（r）／φ^—M

利用固体径迹探测器测量反应堆不同位置燃料元件内的中子注量率，得到反应堆燃料元件内的中子注量率分布，与对应点慢化剂内啊子注量率进行比较，对反应堆物理实验中一个近似假设公式φＵ（γ）／φ＾－Ｕ≈φＭ（γ）／φ＾－Ｍ进行了验证。给出了该公式成立的条件。     

固体径迹探测器测量反应堆中子注量率的三角修正公式

给出了在用固体径迹探测器测量反应堆中子注量率时,反应堆启动过程对径迹计数影响的修正公式,即所谓的“三角修正”。该三角修正公式与反应堆功率上升周期和径迹探测器辐照时间有关。最后讨论了使该修整小于１％对周期和辐照时间的要求。     

可逆比热容法测量反应堆中子注量率的可行性及应用前景

热分析仪器和测量技术的迅速发展为通过测量受辐照材料热性质的变化测量中子注量提供了可能。本文提出采用调制差示扫描量热（MDSC）法测量反应堆辐照的含硼材料可逆比热容的变化，进而得到反应堆的中子注量率。从理论和实验两方面讨论了利用该方法测量反应堆中子注量率的可行性。介绍了可逆比热容法测量反应堆中子注量率的原理和实验方法。展望了这种测量方法在测量高注量反应堆中子注量率的应用前景。     

HFETR辐照孔道中子特性研究

采用蒙特卡洛(MCNP)程序对辐照孔道中子特性进行研究,包括各辐照孔道E1.0 Me V时的中子注量率(φE1.0 Me Vf(29))与E0.625 e V时的中子注量率(φE0.625 e Vf(29))的比值k,辐照孔道阳面、阴面样品中子注量率比值,辐照孔道样品最佳布置高度。研究结果表明:高通量工程试验堆(HFETR)辐照孔道k值随轴向和径向变化不同,但平均变化程度一致;9#孔道阳面、阴面材料快中子注量率比值达1.43,而G7、K11孔道阳面、阴面材料快中子注量率比值相对较小,分别为1.21和1.18。综合考虑,对于P15孔道,辐照试验段样品布置区的高度可达500 mm;对于9#孔道,样品布置高度可适当增至600 mm。     

CFBR-Ⅱ堆脉冲中子注量测量

为了监测CFBR-Ⅱ堆辐照空间内脉冲中子注量,根据能谱随空间缓慢变化的特点,选择28个特征点,使用239Pu电离室对S活化片进行相对刻度,修正了不同空间区域能谱对S活化片测量的影响,测得了特征点的脉冲中子注量。该方法引入的不确定度为5%;脉冲产额为1.10×1016裂变时,距辐照腔底50mm处脉冲中子注量为5.87×1013cm-2;合成标准不确定度为11%。     

CFBR-Ⅱ堆脉冲中子注量测量

为了监测CFBR-Ⅱ堆辐照空间内脉冲中子注量,根据能谱随空间缓慢变化的特点,选择28个特征点,使用239pu电离室对S活化片进行相对刻度,修正了不同空间区域能谱对S活化片测量的影响,测得了特征点的脉冲中子注量.该方法引入的不确定度为5%;脉冲产额为1.10×1016裂变时,距辐照腔底50mm处脉冲中子注量为5.87×1013 cm-2;合成标准不确定度为11%.     

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。      

环形元件超高通量堆堆芯初步概念设计

基于环形燃料元件，提出了一种超高通量堆(UFR)堆芯概念设计。UFR燃料组件设计采用61个燃料元件构成的六角形组件，堆芯采用52盒燃料组件、9盒控制棒组件和厚反射层设计。通过开展堆芯概念设计方案评价，给出了堆芯循环长度、中子注量率、中子能谱、中子空间分布等关键参数。结果表明，在当前的总体参数下所提出的UFR的最大中子注量率可达到1.0×10¹⁶ cm^-2·s^-1。     

活化法测量散裂靶中子能谱的实验验证

散裂靶中子的能谱对加速器驱动次临界系统的倍增因数和嬗变率等影响很大,计算表明散裂靶中子谱在MeV能区与裂变中子谱相近。本文利用活化法测量临界装置的泄漏中子谱和中子注量率,提出了用In、Al、Mg、Ti、Au、Zn、Ni、Rh、Fe和Co等活化箔测量散裂靶中子能谱和中子注量率的方案。结果表明,将活化箔在散裂靶中子场中辐照5h,中子注量最高达5×1014 cm-2量级,辐照后1h内取出活化箔,根据半衰期的长短安排测量顺序,可测量散裂靶的中子能谱和中子注量率。     

238U裂变电离室测量25.5MeV中子注量率

在飞行时间谱仪测量中子能谱的基础上，利用²³⁸U裂变电离室测量了中国原子能科学研究院HI-13串列加速器产生的25.5MeV中子注量率。为验证该裂变电离室测量快中子注量率的可靠性，在中国原子能科学研究院5SDH-2串列加速器上，利用该电离室和伴随α粒子装置同时测量14.8MeV中子注量率，结果在不确定度范围内一致。     

微型中子源反应堆中子谱参数测量

以Au、Zr和Fe为活化探测器,采用裸探测器法测量中国原子能科学研究院微型中子源反应堆的中子谱参数f、α、f_F和φ_th。内辐照座的α、f和f_F分别为－0.007±0.003、20.8±0.4、5.5±0.2。该方法对φ_th的测量结果与4πβ-γ符合法的一致,相对偏差小于2%。与SLOWPOKE相比,微堆有较高的α、f_F值。与已有测量数据的比较表明,微堆中子谱在很长一个时期内是稳定的,利用微堆作为中子源的k₀法中子活化分析不需中子注量率监测器,且比较器一经照射和测量后,可用于其后较长时间内所有分析的计算标准。     

0.144~1.2MeV单能中子参考辐射场的建立

本工作利用2×1.7MV串列加速器建立了0.144、0.250、0.565和1.2MeV单能中子参考辐射场。所需能量的中子通过⁷Li(p,n)⁷Be和³H(p,n)³He反应产生。利用TARGET程序计算了中子注量谱，利用自行设计的反复充气式反冲质子正比计数器绝对测量了上述能量点的中子注量，合成标准不确定度≤2.0%。2001年参加了由国际电离辐射咨询委员会第三分部组织的单能快中子注量测量国际比对，取得了较为满意的结果。     

14MeV中子引发SRAM器件单粒子效应实验研究

在中国原子能科学研究院的高压倍加器装置上开展了SRAM器件的14 MeV中子单粒子效应实验研究。介绍了中子的产生、中子注量率的测量和调节以及中子单粒子效应的测试等的实验方法,获得了HM628512BLP型和R1LV1616HSA型SRAM器件的14 MeV中子单粒子效应截面。前者与文献的单粒子效应截面在误差范围内一致,验证了实验方法的科学性和可行性。后者与由效应机制出发获得的理论分析结果在量级上一致,对实验结果给出了定性的解释。     

Development of the high-energy neutron fluence rate standard field in Japan with a peak energy of 45 MeV using the 7Li(p,n)7Be reaction at TIARA

In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the ⁷Li(p,n)⁷Be reaction using a 50-MeV proton beam from an azimuthally varying field (AVF) cyclotron of the Takasaki Ion Accelerators for Advanced Radiation Application (TIARA). The neutron energy spectrum was measured using an organic liquid scintillation detector and a ⁶Li-glass scintillation detector by the time-of-flight method, and using a Bonner sphere spectrometer by the unfolding method. The absolute neutron fluence was determined using a proton recoil telescope (PRT) composed of the liquid scintillation detector and a Si(Li) detector that was newly developed in the present study. The detection efficiency of the PRT was obtained using the MCNPX code. The peak neutron production cross section for the ⁷Li(p,n)⁷Be reaction was also derived from the neutron fluence in order to confirm the neutron fluence of the TIARA high-energy neutron field. The peak neutron production cross section obtained in the present study was in good agreement with those of previous studies. The characteristics of the 45-MeV neutron field in TIARA were successfully evaluated in order to calibrate high-energy neutron detectors and high-energy neutron dosimeters.     

Energy spectrum measurement and dose rate estimation of natural neutrons in Tibet region

In this work, natural neutron spectra at nine sites in Tibet region were measured using a multi-sphere neutron spectrometer. The altitude-dependence of the spectra total fluence rate and ambient dose equivalent rate were analyzed. From the normalized natural neutron spectra at different altitudes, the spectrum fractions for neutrons of greater than 0.1 MeV do not differ obviously, while those of the thermal neutrons differ greatly from each other. The total fluence rate, effective dose rate and the ambient dose equivalent rate varied with the altitude according to an exponential law.     

Unfolding the measured neutron spectra in the irradiation chamber of the UZrH reactor using iterative method

In the procedure of neutron fluence measurement in the whole energy range （10^-4 eV-18 MeV）, in the irradiation chamber of a UZrH reactor, the neutron energy spectra are unfolded using the method of minimizing directed divergence and SAND-II, which are used broadly at home and abroad. These methods belong to the iterative methods. In this article, the procedure of the spectra unfolding using the two methods is described in detail. The neutron spectrum distribution unfolded by the two methods agree well with each other. In the end, the major differences of the two iterative methods are compared with each other, and the main factors affecting the accuracy of the spectra unfolding with the iterative method are discussed.     

Evaluation of D（d,n）^3 He reaction neutron source models for BNCT irradiation system design

A mathematical method was developed to calculatc the yield.energy spectrum and angular distribution of neutrons from D(d,n)3 He(D-D)reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV.The data of energy spectrum and angular distribution wefe applied to set up the neutron source model for the beam-shaping-assembly(BSA)design of Boron-Neutron-Capture-Therapy(BNCT)using MCNP-4C code.Three cases of D-D neutron source corresponding to incident deuteron energy of 1000.400 and 150 kaV were investigated.The neutron beam characteristics were compared with the model of a 2.45 MeV mono-energetic and isotropic neutron source using an example BSA designed for BNCT irradiation.The results show significant differences in the neutron beam characteristics,particularly the fast neutron component and fast neutron dose in air,between the non-isotropic neutron source model and the 2.5 MeV mono-energetic and isotropic neutron source model.     

Evaluation of D(d,n)~3He reaction neutron source models for BNCT irradiation system design

A mathematical method was developed to calculate the yield,energy spectrum and angular distribution of neutrons from D(d,n)~3He(D-D)reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV.The data of energy spectrum and angular distribution were applied to set up the neutron source model for the beam-shaping-assembly(BSA)design of Boron-Neutron-Capture-Therapy(BNCT)using MCNP-4C code. Three cases of D-D neutron source corresponding to incident deuteron energy of 1000,400 and 150 key were inves- tigated.The neutron beam characteristics were compared with the model of a 2.45 MeV mono-energetic and isotropic neutron source using an example BSA designed for BNCT irradiation.The results show significant differences in the neutron beam characteristics,particularly the fast neutron component and fast neutron dose in air,between the non-isotropic neutron source model and the 2.5 MeV mono-euergetic and isotropic neutron source model.