DF-Ⅵ堆中心平均裂变截面比~(238)■_f/~(235)■_f计算

计算和测量反应堆芯部可裂变材料的平均裂变截面比,使有可能估算出堆芯部中子能谱的硬化程度。计算和测量~(238)U和~(235)U的平均裂变截面比~(238)■_f/~(235)■_f,就能估算出堆芯部~(238)U裂变阈1.4 MeV以上的中子份额。用固态核径迹探测器法已测得DF-VI堆芯部     

一种探测裂变谱高能中子方法的分析研究

探讨了提高系统中子灵敏度和测量信噪比的技术措施,通过结构参数的优化设计,提高了探测高能中子灵敏度.对影响信号测量的内本底干扰水平进行了理论计算.结果表明,在强辐射内本底条件下,可实现较低强度下探测裂变谱高能中子.     

裂变元素中子共振反应的裂变和俘获截面的相关分析

我们曾用多能级展开的Adler-Adler公式对裂变元素的中子共振反应(裂变或俘获)截面作过分析,求出裂变或俘获截面的共振参数。本工作是进一步对同一核素的裂变和俘获截面两组实验数据作相关分析,以求取同一核素两种截面相应的共振参数。     

基于MCNP的缓发裂变中子衰减时间谱模拟计算

基于核裂变反应发生与缓发裂变中子发射这两个物理过程在时间上可相互分离的思想,利用MCNP程序计算得到了含铀地层0~100s范围内的缓发裂变中子衰减时间谱。模拟计算了球体几何模型中235 U与238 U核裂变反应率空间分布;借助SOURCE子程序依次对缓发裂变中子的发射位置、发射方向、发生核裂变反应的核素、缓发裂变中子所在的群组、发射时间及其初始能量等参数进行了抽样。结果表明,计算得到的缓发裂变中子衰减时间谱的时间特征与铀核素的缓发裂变中子群参数相符合。对缓发裂变中子总计数与地层铀含量之间的关系进行了线性拟合,拟合结果的相关系数平方值大于0.998 5,该结果从理论上验证了缓发裂变中子测井铀含量线性模型的正确性。     

组件燃耗对堆芯裂变中子源参数的影响

在中子输运计算中，准确的裂变中子源是保证计算结果可靠性的前提和基础：随着组件燃耗的加深．^235U等核素不断消耗，^239Pu、^240Pu、^241Pu等核素不断积累，导致每次裂变产生的平均中子数v和裂变释放的平均能量E也随之变化，裂变中子源的归一化因子随燃耗增加而增加。因此，在中子输运计算中，必须考虑组件燃耗对堆芯的裂变中子源参数的影响。     

^252Cf微型快裂变室

本文叙述的快裂变室主要用于~(252)Cf裂变中子飞行时间谱测量。该裂变室具有重量轻(1.65g)、上升时间快(6ns)、对裂变碎片探测效率>99％等特点。针对裂变中子能谱测量的特殊要求,详细论述了裂变室设计应遵循的基本原则及其对能谱测量可能带来的影响。     

用于裂变研究的多参数测量系统

介绍了一种用于理解变研究的多参数测量系统。描述了裂变源、微型裂变室、中子探测器、实验安排以及降低测量本底的措施。系统适用于精确测量自发裂变中子能谱和裂变碎片特性,也可用于其他类型的测量。     

中子裂变链统计涨落问题的数值计算方法

研究了弱中子源驱动下,裂变系统中子裂变链统计涨落问题的数值计算方法。进行了数值计算建模、数值方法分析、数值计算检验、一类问题概率分布函数的统计涨落特征量的数值计算示范。特例数值检验表明:只要数值解方程组阶数(截断)N足够大,则数值解满足归一(守恒)律、指数增长律并与精确解析解一致。建立了对非定常裂变系统中子裂变链统计涨落问题进行一维等效模型下数值模拟的方法。     

脉冲中子-裂变中子铀矿测井技术的蒙特卡罗模拟

脉冲中子-裂变中子铀矿测井方法（PNFN）是采用脉冲式中子源,利用-3He管中子探测器记录瞬发裂变超热中子或缓发裂变热中子,得到地层中铀矿含量信息的测井方法。利用MCNP程序模拟了不同铀含量、不同地层孔隙度地层条件下PNFN的响应,分析了瞬发裂变超热中子和缓发裂变热中子与地层铀含量和孔隙度的关系。结果表明,地层孔隙度对利用PNFN确定地层铀含量有影响,孔隙度越大,利用裂变中子直接计算得到的地层铀含量比真实含量越小。利用瞬发裂变超热中子或热中子时间衰减谱计算得到地层宏观俘获截面,对裂变中子进行校正,可以有效提高地层铀含量计算结果的准确度。     

24.4KeV中子诱发^235U裂变的产额测量

一、前言 对~(235)U热中子裂变已经作了充分的研究,裂变产物的产额作了广泛的测量,并对实验数据进行了多次编评。但是对于其它单能中子诱发~(235)U裂变研究还远远不够,尤其是keV能区的中子更是如此。J.G.Cuninghame等测量了130-1700keV中子诱发~(235)U裂变中一些核素的产额。但是低于130keV中子诱发的裂变研究,文献中未见过报道。为了研究产额随中子能     

Neutron Excess Generation by Fusion Neutron Source for Self-Consistency of Nuclear Energy System

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities.     

BF3长计数器效率刻度及其应用

利用²³⁸U裂变反应，通过测量裂变碎块的数量，能够较为准确地测量快中子注量。但这种方法在中子能量为1.6MeV以下（裂变道未开或处于裂变截面第1个台阶上升处）和6.0～7.0MeV能区（第2个台阶上升处）由于不确定度过大而不适用。本工作采用4种不同的核反应作中子源，对北京大学4.5MV静电加速器中子实验大厅的BF₃长计数器的相对效率进行了刻度。将²³⁸U(n,f)反应与相对效率已知的BF₃长计数器相结合，解决了在上述能区准确测量中子注量的问题。      

用薄膜~(252)Cf源测定ST451快中子探测器的相对效率

本文叙述了用飞行时间测量~(262)Cf瞬发裂变中子能谱来刻度ST 451探测器的快中子相对效率响应的方法。在飞行时间谱仪中,一个流气式的微型电离室作为裂变碎块的探测器。文中给出了对实验数据的处理和修正的过程。得到了阈值分别为0.420、0.625、0.825、1.168、1.565和1.882MeV中子能量从几百keV到10MeV范围内的相对探测效率。实验结果与用Monte Carlo方法计算的效率作了比较。     

Monte Carlo simulation of prompt neutron emission during acceleration in fission

The possibility of prompt neutron emission during acceleration of fission fragments (FFs) was examined by means of Monte Carlo method and statistical neutron emission model. Multimodal random neck-rapture model was used to describe the initial distribution of mass, charge, and total kinetic energy of the primary fragments. Statistical model was used to simulate the de-excitation process of the fragments from the moment of scission until full acceleration. By random number sampling, the fission process was simulated in order to obtain the basic physical quantities, and their correlations were analyzed to verify the adequacy of the model. It was found that, on the average, ～10% and ～16% of prompt neutrons for ²³⁵U(n _th,f) and ²⁵²Cf(sf), respectively, were emitted before reaching 90% of the final fragment kinetic energy.     

Development of an improved method for calculating the effective delayed neutron fraction of a core using a structure with only a few collapsed energy groups

In this study, an improved method for calculating the effective delayed neutron fraction of a core with a few collapsed energy groups has been developed. To accurately calculate the effective delayed neutron fraction of a core using the conventional method, a structure with several energy groups is needed for the fast energy region in order to reflect the difference in the fission spectra for prompt fission neutrons and delayed neutrons. On the other hand, a structure with only a few energy groups is needed for the criticality evaluation. Thus, the calculation cost increases for the effective delayed neutron fraction calculations owing to the need for a large number of energy groups. To solve this problem, in the present study, the error mechanism for the effective delayed neutron fraction calculation using a structure with only a few energy groups was studied, and it was found that the error results from the collapse of the fission spectra after the cell calculations without adjoint flux weighting. In addition, an improved method for the collapse fission spectra with an adjoint flux obtained by one-point calculation was developed. Using the proposed method, the effective delayed neutron fraction can be estimated with sufficient accuracy using a structure consisting of only a few collapsed energy groups. This result will contribute to reducing the calculation cost and/or improving the accuracy of effective delayed neutron fraction calculations.     

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。     

Simplified Method for Calculation of Neutron Capture Transformation Effects of Fission Products on Decay Power

A simplified method is proposed for the calculation of the effects of neutron capture transformations of fission products (FPs) on the decay power of FPs. The decay power of FPs after shutdown changes by the neutron capture transformations of FP nuclides during reactor operation. It is proposed to calculate the neutron capture transformation effects considering the production of the following 7 nuclides ¹⁰³Ru, ¹³⁴Cs, ¹³⁶Cs, ^148mPm, ¹⁴⁸Pm, ¹⁵⁴Eu and ¹⁵⁶Eu by the neutron capture reaction of the direct mother nuclide alone giving a cumulative fission yield for the mother nuclide. The present method was assessed by com-paring the calculation results with the rigorous calculation results for the thermal-neutron fission of ²³⁵U irradiated between 1 and 5 yr in a light water reactor with thermal-nentron flux between 3 x 10¹³ and 6 x 10¹³ n/cm²·s and for the fast-neutron fission of ²³⁹Pu irradiated between 1 and 5 yr in a fast breeder reactor with total neutron flux between 3 x 10¹⁵ and 6 x 10¹⁵ n/cm²·s. It has been clarified that the present method can calculate the neutron capture transformation effects within the accuracy of ±1% of the decay power for the irradiation of 1yr and cooling time less than 10⁹s irrespective of fission type and neutron flux. The accuracy varies little with neutron flux but considerably with irradiation time. For a irradiation of 5 yr the present method can calculate the capture effect within the accuracy of +1% and -5% of the decay power. The accuracy can be improved to ±1% of the decay power with the simple correction factors.     

先进裂变核能的关键核数据测量和CSNS白光中子源

在设计加速器驱动的次临界系统（ADS）、核废料嬗变装置及钍基熔盐堆时亟需一些关键核数据,当前核数据库受实验条件或中子能区的限制,存在核数据精度不高甚至少部分核素数据缺失的情况。本文综述了国内外相关的核数据研究和相应的白光中子源情况。基于中国散裂中子源(CSNS)的反角通道白光中子源实验终端的中子束流具有非常宽的能谱（0.01 eV~200 MeV）和很好的时间特性。模拟得到距靶80 m处的实验终端的中子注量率为9.3×10⁶cm^-2•s^-1,1 eV ~ 1 MeV能量间隔内的中子数占总中子数的53%;同时,加速器运行在双束团模式或单束团模式,时间分辨率均在0.3%～0.9%之间,适合开展核数据测量。     

252Cf自发裂变谱形对中子反射实验数值模拟的影响

针对裂变源中子反射实验中涉及到的中子源能谱问题,选取两套用公式拟合、被普遍认可的252Cf自发裂变能谱,分别将其应用于该类实验的数值模拟中。对实验测点的中子注量及裂变反应率进行了对比研究,结果表明,相对于Maxwell谱,Watt谱能更好地与实验结果符合。     

Point by Point model calculation of the prompt neutron multiplicity distribution P(ν) for spontaneous and neutron induced fission of actinides

The prompt neutron multiplicity distribution P(ν) is a very sensitive quantity which depends on the model calculation of the multi-parametric matrix ν(A,TKE) and on the fission fragment distributions.