5 启明星1#中子通量密度测量实验

启明星1#是由1个快中子能谱区／热中子能谱区耦合组成的堆芯和由外中子源来驱动的次临界系统。快中子能谱区在堆芯内部，热中子能谱区在堆芯外部，快区不仅能够提供快中子能谱，还可放大外中子源用于驱动热区，热中子能谱区主要用来能量放大以维持装置的链式裂变反应。在此装置上开展通量测量实验是为了了解这种新型快热耦合装置跟其它装置相比有何异同，分别用高压倍加器驱动产生的氘氚反应和镅铍外中子源研究整个堆的通量分布，以便于开展核嬗变研究工作。     

启明星1#中子通量密度测量实验

启明星1#是由1个快中子能谱区/热中子能谱区耦合组成的堆芯和由外中子源来驱动的次临界系统。快中子能谱区在堆芯内部,热中子能谱区在堆芯外部,快区不仅能够提供快中子能谱,还可放大外中子源用于驱动热区,热中子能谱区主要用来能量放大以维持装置的链式裂变反应。在此装置上开展通量测量实验是为了了解这种新型快热耦合装置跟其它装置相比有何异同,分别用高压倍加器驱动产生的氘氚反应和镅铍外中子源研究整个堆的通量分布,以便于开展核嬗变研究工作。在此装置上进行4种情况下的中子通量密度分布实验:1)空装载时的快区中子通量分布(利用镅铍…     

启明星1#次临界装置的裂变率分布

启明星1#次临界装置是我国为开展加速器驱动的次临界系统(ADS)研究而建立的国际上第1个具有快-热耦合结构的次临界反应堆实验装置。启明星1#次临界装置在确定的装载下、由不同能量的外中子源作用时,利用MCNP程序分别对装置快中子能谱区、热中子能谱区燃料元件的径向及轴向裂变率分布进行模拟计算,所使用外中子源的中子能量分别为2.5、5、14MeV。计算结果表明:在外中子源源强相同的情况下,源中子能量越高,裂变率越大;在源中子能量相同的情况下,次临界反应堆的轴向裂变率分布为中间高、两端低,径向裂变率分布在快中子能谱区先减小后增大,而热中子能谱区则是先增大后减小,然后,随着接近反射层又逐渐增大。该裂变率分布计算结果为后续实验测量和探测器布置提供了参考。     

ADS原理验证装置快热耦合次临界堆设计

作为加速器驱动洁净核能系统(ADS)原理验证装置"启明星"一号的次临界驱动堆,堆芯采用快-热耦合方式组成,由天然金属铀组成快中子能谱区能有效地嬗变锕系元素(MA),低浓铀元件组成热中子能谱区能有效地嬗变裂变产物(FP).使用MCNP程序对次临界实验装置进行设计计算,确保keff在0.90～1.00之间.     

ADS原理验证装置燃料布置方案

"启明星"的次临界驱动堆是加速器驱动洁净核能系统(Accelerator Driven System,ADS)的原理验证装置,采用快-热耦合的堆芯组成方式,由天然金属铀组成快中子能谱区,低浓铀元件组成热中子能谱区.使用MCNP程序的U卡和Fill卡对次临界实验装置进行设计计算,根据keff在0.90～1.00之间的设计要求,确定了热区的燃料栅格.     

4 启明星1#次临界装置上第2阶段实验研究

启明星1#次临界装置建成后，在第1阶段的实验研究即用Am-Be稳态外中子源驱动启明星1#次临界装置，Am-Be稳态外中子源的平均中子能谱约4MV，初步测量了其中子学特性后，又于2005年10月到11月进行了第2阶段的实验，即用高压倍加器产生的脉冲外中子源驱动启明星1#次临界装置。     

利用“启明星1号”开展嬗变的可行性分析研究

“启明星1号”是一个专门用于ADS研究的快热耦合装置。从堆芯向外依次为：快中子能谱区，超热中子能谱区，热中子能谱区。从理论设计上希望分别可嬗变MA、LLFP及输出能量。     

启明星1#次临界装置上第2阶段实验研究

启明星1#次临界装置建成后,在第1阶段的实验研究即用Am-Be稳态外中子源驱动启明星1#次临界装置,Am-Be稳态外中子源的平均中子能谱约4MV,初步测量了其中子学特性后,又于2005年10月到11月进行了第2阶段的实验,,即用高压倍加器产生的脉冲外中子源驱动启明星1#次临界装置。高压倍加器产生的脉冲外中子源能谱为14MV,测量其中子学特性,包括:1)验证第1阶段的实验,外推临界质量;2)跳源方法测量启明星1#的不同装载时的次临界度;3)用脉冲外中子源驱动启明星1#次临界装置,测量不同装载下不同空间位置的中子衰减特性等。其中,外推临界质量与第1阶段…     

跳源法在启明星1#次临界装置中的应用

简要介绍了跳源法在启明星1#次临界装置上测量次临界度的原理、外源驱动的次临界中子学实验装置、堆芯布置及中子源驱动系统。主要研究了中子源在堆芯轴向中心位置、不同装载情况下的反应性变化,并给出不同的有效倍增系数keff。实验测量结果与理论计算结果进行了比较,结果符合较好。     

启明星Ⅱ号双堆芯零功率装置

启明星Ⅱ号是针对我国新型先进核能系统基础性研发及工程化设计验证而研制的双堆芯零功率装置。启明星Ⅱ号拥有两个堆芯，水堆堆芯侧重于开展热中子能谱环境下的原理性验证实验研究，铅堆堆芯侧重于重金属冷却的快中子反应堆及加速器驱动的次临界系统（ADS）等先进核能系统的中子物理特性实验研究。启明星Ⅱ号通过一套仪控系统实现了两个堆芯的集成化控制和测量数据采集，每个堆芯均配备了多套非能动安全停堆系统，固有安全性强。在启明星Ⅱ号上获取了多种堆芯的基准性临界实验数据，可为我国轻水堆的技术创新、重金属冷却反应堆工程化设计及新型核能系统的集成研发提供支持。     

启明星1#实验装置ks、keff和φ*的模拟计算

用MCNP程序对启明星1#实验装置（Venus1#）的k_s、k_eff和φ^*进行模拟计算。在装置的源区、快区、反射层、屏蔽层已定条件下,逐层增加热区燃料元件,每增加1层,对k_s、k_eff和φ^*进行1次计算,共增加了13层,最终得到k_eff为0.96246,满足了Venus1#的设计要求。元件层数增加,φ^*先增后降,当增至12层时,φ^*又明显增大。外源位置和能量对φ^*有影响,外源在轴向离中心越近、能量越高,φ^*越大。     

启明星1#次临界装置热中子能谱区裂变率分布测量

启明星1#是我国专门为开展加速器驱动次临界系统研究而建立的国际上第1个具有快-热耦合结构的次临界反应堆实验装置。采用MCNP程序对堆芯裂变率分布进行指导性计算,并参考计算结果布置探测片,用固体核径迹探测器测量了堆芯热区裂变率分布。测量结果显示:堆芯有反射层一端的裂变率比无反射层一端的高;轴向加装反射层末端的裂变率明显增大。测量结果对确定热区的裂变功率提供了数据。     

快热耦合系统的中子学计算工具——采用确定论方法

对于既具有高能量外中子源又包含热中子区域的快热耦合系统，本文将离散坐标法Sn与栅格计算程序相结合，给出可用于快热耦合系统的中子学确定论计算工具。用该方法计算了启明星1号次临界装置的k_eff，并给出了计算结果。     

ADS次临界反应堆的中子共轭方程

与临界反应堆相比,ADS次临界反应堆的外源中子和裂变中子的空间分布具有严重的不均匀性,对应的中子价值也不同。本工作对次临界反应堆的稳态输运方程作分群扩散近似,得到了多群方程,进一步推导出按堆芯功率归一化的中子共轭方程表达式和与功率相关的中子价值函数表达式,给出了次临界反应堆中子价值的物理意义。由稳态中子共轭方程组出发,给出了两种带外加中子源的次临界反应堆增殖因数的表达式。     

Analysis of fuel management in the KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is ∼360 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.     

Core and sub-channel analysis of SCWR with mixed spectrum core

The SCWR core concept SCWR-M is proposed based on a mixed spectrum and consists of a thermal zone and a fast zone. This core design combines the merits of both thermal and fast SCWR cores, and minimizes their shortcomings. In the thermal zone co-current flow mode is applied with an exit temperature slightly over the pseudo-critical point. The downward flow in the thermal fuel assembly will provide an effective cooling of the fuel rods. In the forthcoming fast zone, a sufficiently large negative coolant void reactivity coefficient and high conversion ratio can be achieved by the axial multi-layer arrangement of fuel rods. Due to the high coolant inlet temperature over the pseudo-critical point, the heat transfer deterioration phenomenon will be eliminated in this fast spectrum zone. And the low water density in the fast zone enables a hard neutron spectrum, also with a wide lattice structure, which minimizes the effect of non-uniformity of the circumferential heat transfer and reduces the cladding peak temperature.     

核临界安全中的源倍增法研究

文章对核临界安全研究中通常采用的现场测量技术———源倍增法进行研究。从有源扩散理论出发,导出了与keff不同的有源次临界中子有效增殖因子ks的表达式,并在次临界系统上进行了验证研究。验证实验研究证实了所导出的ks 的正确性。源倍增法测量的参数实际上是次临界系统在外源作用下的有源次临界中子有效增殖因子ks,而不是以往的中子有效增殖因子keff,这就解决了长期困扰人们的有关源倍增法测量的参数问题。文章讨论了ks 与keff间的差别和关系以及它们对核临界安全的影响。     

Comparison of calculated and measured reaction rates obtained through foil activation in the subcritical dual spectrum facility YALINA-Booster

Reaction rates were measured by the foil activation technique to obtain neutron spectrum information in a subcritical core driven by an external neutron source. The experimental results are compared with Monte Carlo calculations in order to examine the capability of the Monte Carlo code MCNP together with ENDFB-6.8, JEFF-3.1.1 and CENDL-3.1 neutron cross section libraries to predict the neutron spectrum dependent reaction rates correctly in a subcritical core. The focus lies on fast neutrons. A discrepancy is found in the calculated-to-experimental values of the reaction rates and an inaccurate cross section is identified in CENDL-3.1.     

Effects of Spatial Harmonics on Power Distribution in Accelerator Driven Subcritical Cores

A method for solving the time-energy-dependent diffusion equation was devised to appreciate the effect of spatial harmonics on the power distribution in a subcritical graphite-moderated core with a pulsed fast neutron source. An instantaneous power peaking factor (IPF) was calculated using the thermal neutron flux thus obtained, to characterize the power distribution in the core. In case of a 50-cycles-per-second injections of fast neutron pulses of 3 ms pulse width, it was indicated that the value of IPF increased by 4.1 times with the decrease in the multiplication factor from 0.9951 to 0.9762, accompanying fuel burnup. To appreciate the dependence of the core power distribution on the spectrum hardening of the thermal neutron flux caused by injections of pulsed fast neutrons, comparison was made between the value of IPF and that derived from the time-dependent-one-group diffusion theory where the spectrum hardening of the thermal neutron flux was ignored. It was indicated that the disregard of the spectrum hardening of the thermal neutron flux resulted in the low estimate of IPF that changed from ?1% to ?14% with the decrease in the pulse width of pulsed fast neutrons from 3 ms to 0.15 ms.