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.     

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。     

启明星1#实验装置ks、keff和ψ*的模拟计算

用MCNP程序对启明星1#实验装置(Venus 1#)的ks、keff和ψ*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ks、keff和ψ*进行1次计算,共增加了13层,最终得到keff为0.962 46,满足了Venus 1#的设计要求.元件层数增加,ψ*先增后降,当增至12层时,ψ*又明显增大.外源位置和能量对ψ*有影响,外源在轴向离中心越近、能量越高,ψ*越大.     

加装辅助停堆系统的加速器驱动系统次临界实验装置初步设计

针对现有加速器驱动核能系统概念设计中存在的安全隐患,采用MCNP程序设计了加装辅助停堆系统的次临界实验装置;通过改变热区栅距、热区厚度和控制棒的下插深度来改变堆芯参数,对反应堆有效增殖系数进行计算和比较分析,实现了不同堆芯参数下keff值可调;对几种不同热区栅距和热区燃料棒根数所对应的停堆深度和控制棒价值进行了计算,结果表明,辅助停堆系统提高了加速器驱动系统(ADS)的安全性.     

启明星1#次临界实验装置上的外推实验

启明星1#次临界实验装置是“973”项目——加速器驱动洁净核能系统(ADS),02课题为研究次临界反应堆中子学,进行次临界反应堆中子学实验,校核理论计算程序,检验核数据而设计建造的实验平台。2005年,在启明星1#上进行了一系列工作:上半年,进行了启明星1#实验装置的安装调试和实验仪器的准备;7月18日举行了首次装料仪式;之后,用Am-Be中子源驱动次临界装置,测量启明星1#的中子学特性;于10～11月,实现了高压倍加器与次临界反应堆的对接。外推实验是已进行的一系列实验中的一项。先后用Am-Be源、D-T源、D-D源做外推实验,以得到在不同能谱的外…     

启明星1号装置中子通量能谱与中子平均能量计算

启明星1号装置是我国研究ADS次临界中子学的一个快热耦合系统。本文用离散坐标法的程序TWODANT对启明星1号装置能谱进行分析计算。计算结果表明，启明星1号装置具有比较硬的中子能谱，可用以进行有关ADS的研究。     

ADS次临界实验装置--启明星1#

文章介绍加速器驱动次临界系统（ADS）中次临界实验装置——启明星1#的设计目的、要求、结构和可开展的工作。启明星1#是由快中子能谱区和热中子能谱区耦合组成的堆芯和由高压倍加器氘-氚反应中子源来驱动的次临界系统。快中子能谱区处在堆芯内部，该区提供快中子谱，还可放大外中子源，以驱动热区；热中子能谱区处在堆芯外部，主要用来能量放大，以维持装置的链式裂变反应。     

"启明星1#"单向耦合快区外中子源放大系数研究

计算分析"快-热"耦合加速器驱动次临界实验装置"启明星1#"在252Cf,Am-Be和氚--氚(2H-3H)中子源驱动下,单向耦合时快区泄漏中子数、净中子产额和泄漏中子谱随快区层数与外中子源能量的变化关系.结果表明,从反应堆产能和嬗变角度考虑,"启明星1#"实验装置快区燃料元件单向耦合时存在最优化的元件装载量.     

启明星1#次临界装置内不同位置探测器的中子计数率变化

在启明星1#次临界装置上进行了次临界外推实验,外中子源分别采用Am-Be中子源和252Cf中子源,放置在启明星1#次临界装置中心,中子探测器放置在次临界装置内不同位置,研究相对中子计数率的变化。实验测量结果表明:在启明星1#次临界装置不同位置的探测器测量得到的中子计数率变化不同,但对外推结果影响不大。     

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…     

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。     

高温熔盐泵散热器的数值优化与热分析

为提高高温熔盐泵散热器的散热效果,对影响散热器散热面积的重要因素——散热孔的直径和数量进行了数值优化,并将数值模拟得到的热载荷加载到结构体上,分析了热载荷下的热变形。结果表明:散热孔直径为12 mm、散热孔数量为6时,散热流体的流动均匀稳定,泵轴和散热器的综合散热效果较好,最终泵轴顶端的平均温度可降到440 K,散热器顶端的平均温度可降到417.5 K,满足散热需求;泵轴热变形量较小,最大值为0.46 mm,散热器的变形量较大,最大值为2.37 mm,加工时必须考虑材料的热膨胀和热变形。     

混合能谱超临界水堆堆芯设计分析

提出了一种新型的超临界水堆概念设计:混合能谱超临界水堆,它包括慢谱区和快谱区两部分.其慢谱区燃料组件采用双排燃料组件,快谱区采用简单的正方形栅元燃料组件.慢谱区与快谱区的燃料组件都采用同向流动方式来简化堆芯设计.慢谱区的冷却剂出口温度远低于整个堆芯的出口温度,这大大降低了慢谱区包壳的温度峰值.此外,由于快谱区冷却剂密度很小,流速很高,故可采用较大的栅元结构,这有效地降低了包壳周向局部传热不均匀性.所以混合堆在充分继承慢谱、快谱堆芯优点的基础上,弥补两者的不足.     

Toughness investigation on simulated weld HAZs of SQV-2A pressure vessel steel

In order to get detailed information about weld HAZs toughness of SQV-2A steel and determine the optimum welding and heat treatment parameters, the toughness of simulated CGHAZs (coarse grained heat affected zone) and CGHAZs (intercritically reheated CGHAZ) were systematically investigated. The influence of tempering thermal cycles on weld ICCGHAZs toughness was clarified. The effect of post weld heat treatments (PWHT) on weld CGHAZs toughness was also determined. The results showed that high toughness (absorbed energy >200 J) of weld HAZs could be achieved by selecting the optimum welding and PWHT parameters (cooling time Δt_8/5: 6–40 s, PWHT: 893 K, 3.6–7.2 ks). Tempering thermal cycles with peak temperature of above 573 K could remarkably improve the toughness of deteriorated ICCGHAZs and reduce the hardness, when cooling time Δt_8/5(2) of the reheating thermal cycle was 6 s, which implies that welding of SQV-2A without PWHT is possible, provided that low heat input welding is adopted and welding procedure is correctly arranged. Metallography and fractography revealed that M–A constituents in weld HAZs played an important role in controlling weld HAZ toughness.     

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.     

Physical studies of minor actinide transmutation in the accelerator-driven subcritical system

The accelerator-driven subcritical system(ADS)with a hard neutron energy spectrum was used to study transmutation of minor actinides(MAs). The aim of the study was to improve the efficiency of MA transmutation while ensuring that variations in the effective multiplication factor(k_(eff)) remained within safe margins during reactor operation. All calculations were completed using code COUPLE3.0. The subcritical reactor was operated at a thermal power level of 800 MW, and a mixture of mononitrides of MAs and plutonium(Pu) was used as fuel.Zirconium nitride(ZrN) was used as an inert matrix in the fuel elements. The initial mass composition in terms of weight percentages in the heavy metal component(IHM)was 30.6% Pu/IHM and 69.4% MA/IHM. To verify the feasibility of this MA loading scheme, variations in k_(eff), the amplification factor of the core, maximum power density and the content of MAs and Pu were calculated over six refueling cycles. Each cycle was of 600 days duration, and therefore, there were 3600 effective full power days.Results demonstrated that the effective transmutation support ratio of MAs was approximately 28, and the ADS was able to efficiently transmute MAs. The changes in other physical parameters were also within their normal ranges.It is concluded that the proposed MA transmutation scheme for an ADS core is reasonable.     

Coupled thermal-hydraulics and neutron-physics analysis of SCWR with mixed spectrum core

The new SCWR conceptual design (SCWR-M) is proposed on the basis of a mixed spectrum core consisting of a thermal spectrum zone and a fast spectrum zone. This new core design is considered to be the hybrid of the existing thermal SCWR and fast SCWR cores. It combines the merits of both thermal and fast SCWR cores, at the same time minimizes their shortcomings. For the thermal zone, the difficulties in the mechanical design and the maximum cladding temperature can be reduced as far as possible by the co-current flow mode; and for the fast zone, a sufficiently large negative coolant void reactivity coefficient and breeding ratio can be achieved by the multi-layer arrangement of fuel rods.The performance, including the burn-up behavior, of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. The results obtained so far have shown that the mixed spectrum SCWR concept (SCWR-M) is feasible and promising.     

Critical neutron heating in the control drums of a dual purpose thermionic space reactor for power and propulsion

The critical neutron heating in the reflector control drums is investigated for a fast incore thermionic space craft reactor for power and nuclear propulsion. The reactor is fueled with uranium carbide (UC) and controlled with the help of rotating B₄C drums imbedded into the beryllium reflector. While the neutron heating in the drums would not require a cooling mechanism in the power phase, the heat generation during the thrust phase obliges cooling for a nuclear thermal thrust around F = 5000 N by a specific impulse of 670 s⁻¹ at an hydrogen exit temperature around 1900°K. With a beryllium reflector without extra cooling measures, thermal thrust must be kept F < 2500 N to relieve the thermal load in the reflector. On the other hand, a reflector made of BeO may withstand a thermal load for a nuclear thermal thrust of F = 5000 N. The neutronic analysis has been conducted in S₁₆-P₃ and S₈-P₃ approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. A reactor control with boronated reflector drums (drum diameter = 14 cm) at the outer periphery of the radial reflector of 16 cm thickness would make possible reactivity changes of Δk_{eff = 13.55%}—amply sufficient for a fast reactor—without a significant distortion of the fission power profile during all phases of the space mission. Calculations are conducted for a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels around 50 kW_el.     

核动力厂温排水混合区范围的可接受性论证的初步研究

基于国内现有的温排放控制标准可执行性不强的现状,以某滨海电站SNP（采用一次直流冷却循环）厂址为例,开展了温排水混合区范围的可接受性论证研究。通过对该电站厂址海域生态调查站位的春、秋季调查结果的聚类分析,得出如下结论：1） SNP一期（1~2号机组）运行工况下,温排水混合区的设定对海域整体生态功能影响较小,其范围可接受;2） SNP一、二期（1~4号机组）同时运行工况下,温排水混合区的设定,春季对海域整体生态功能影响较小,其范围可接受,秋季对海域整体生态功能影响非常显著,如无有效的缓解措施,其影响是不可接受的。