钍基先进CANDU堆(TACR)钍-铀燃料功率影响研究

在CANDU堆燃料栅元物理的研究中,通常选择堆芯平均的燃料比功率对栅元进行计算模拟,而在TACR中,由于使用了钍燃料,比功率的不同就可能对核反应产生影响,并通过影响棒束栅元的基本截面参数而影响到全堆计算的结果.本文对不同定功率条件下,含全铀燃料和钍-铀燃料棒束的栅元截面参数随辐照值的变化以及钍燃料棒束中233Pa和233U的质量份额进行了计算分析,认为功率会对钍燃料的栅元宏观截面产生影响,在全堆计算中,栅元基本参数应尽量使用基于历史的局部参数法.     

熔盐堆钍铀、铀钚燃料循环核数据不确定度分析

反应堆物理设计不确定度是第4代核能系统的QMU（quantification of margins and uncertainties）有效性认证所必须的参数之一,核数据不确定度是其重要来源。基于自主开发的耦合程序BUND（burnup uncertainty of nuclear data）,将SCALE程序TRITON和TSUNAMI-3D模块耦合,完成了熔盐堆钍铀燃料循环、铀钚燃料循环核数据引起的有效增殖因数k_eff不确定度分析,并与ENDF/B-Ⅶ.1协方差数据库计算结果进行了对比。结果显示：初始时刻,两种燃料循环模式下,核数据导致的k_eff不确定度分别为0.490%和0.582%。随燃耗的增加,核数据引起的k_eff不确定度增加。寿期末,两种燃料循环模式下,对k_eff不确定度影响显著增加的反应道分别为²³⁹Pu(nubar)、(n,f)、(n,γ)、¹⁰⁵Rh(n,γ)、¹³⁵Xe(n,γ)和²³⁴U(n,γ)、¹⁴³Nd(n,γ)、^131,135Xe(n,γ)等。     

聚变堆氦冷固态包层结构和6Li富集度对产氚率的影响

在聚变堆固态包层基本参数基础上，建立简化20°模型，包层分第1壁装甲、第1壁冷却板、氚增殖区和支撑结构。分别选择Li₄SiO₄和Li₂O做增殖材料，应用MCNP程序，研究第1壁结构布置和6Li富集度对产氚率的影响。结果表明：₆Li富集度适宜选择在30％~80％之间；第1壁选择Be装甲可提高产氚率；冷却管板的厚度应取3cm以下，以避免对产氚造成不利的影响。     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

聚变堆交叉冷却固态包层中子学设计优化

针对聚变堆固态包层设计路线,提出了一个交叉排列氦冷固态包层概念。设计采用Be、Li₂TiO₃分层球床。两种尺寸的氦气冷却管道交叉排列,分两个回路同时冷却,以增加系统安全可靠性。分析比较了4种⁶Li富集度布置方案。结果表明：径向远离第一壁降低⁶Li富集度较为合理,靠近第一壁的增殖层⁶Li富集度不能过低,以减少长期运行中Li的消耗对氚增殖性能的影响。借助蒙特卡罗程序MCNP建立11.25°对称模型,全堆包层氚增殖率为1.176,包层寿期内产氚性能稳定,在包层寿命运行时间内的燃耗分布相对均匀。     

基于CATHARE程序的熔盐堆热工物理耦合模拟计算

针对熔盐堆燃料和慢化剂的特点,对压水反应堆热工水力计算程序CATHARE进行了适应性改造。使用改造后的程序进行熔盐堆单通道流体的物理和热工耦合计算分析。分析计算表明,改造后的CATHARE程序可以模拟反应性引入事故和流量引入事故等瞬态工况的堆芯热工物理耦合过程。     

Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations.     

Optimization studies of fuel loading pattern for a typical Pressurized Water Reactor (PWR) using particle swarm method

A three dimensional multi-energy group computer model PRISHA, which solves the neutron diffusion equations using finite difference method is developed for Pressurized Water Reactor (PWR). This computer code can find an optimum loading of a group of fresh fuel assemblies along with fuel assemblies of different exposures. The successive line over relaxation (SLOR) method is used to solve neutron diffusion equations. After validation of this part of computer code against an IAEA – PWR benchmark problem with 177 fuel assemblies in the core, particle swarm optimization (PSO) method is incorporated in the code for finding the optimum fuel loading pattern. A typical PWR core with 157 fuel assemblies, where 289 fuel pins are arranged in 17 × 17 rectangular arrays in a fuel assembly, was analyzed using this computer model for two cycles using PSO method. Different numbers of particles and iterations were used in PSO method. The results are found to be not very sensitive to either the number of particles or the number of iterations used in PSO method for considered case. However, a number of experiments have to be performed to arrive at the best global fitness parameter. Reasonably low power peaking factors were obtained for both the cycles.     

Optimization of in-core fuel management strategy of Tehran Research Reactor (TRR) using MCNP-4C

In order to optimize fuel utilization in TRR, the method of fuel management is modified using MCNP-4C code system. An important parameter of fuel management is the uniformity of neutron flux distribution in the core region, which is obtained efficiently in the present strategy. This strategy is based on calculation of position factors and power densities utilizing MCNP simulations. This study shows that the core life time and average extracted burn up of spent fuel elements of TRR are improved significantly.     

Evaluating the effect of using different sets of enrichment for FAs on fuel management optimization using CA

In nuclear reactor core design, achieving the optimized arrangement of fuel assemblies (FAs) is the most important step towards satisfying safety and economic requirements. In most studies, nuclear fuel optimizations have been performed by using a finite number of different types of FAs. However the effect of FA numbers with different enrichments and the difference between their maximum and minimum enrichment values can be important and should be evaluated in the optimization process.     

On the Tritium Breeding Capability of Flibe, Flinabe, and Li20Sn80 in a Fusion-Fission (Hybrid) Reactor

In a commercial (DT) driven fusion reactor, the tritium breeding ratio per incident fusion neutron must be greater than 1.05 to maintain tritium self-sufficiency for the driver. In this study tritium breeding capability of three different coolants, namely Flibe (LiF·BeF₂), Flinabe (LiF·NaF·BeF₂), and Li₂₀Sn₈₀ in a (DT) driven fusion-fission (hybrid) reactor was investigated for different refractory alloys (W-5Re, TZM, T111, and Nb-1Zr) as structural material. Neutron transport calculations were conducted with the help of SCALE 4.3 SYSTEM by solving the Boltzmann transport equation with code XSDRNPM. The contribution of Flibe, Flinabe, and Li₂₀Sn₈₀ with respect to ⁶Li enrichment in their lithium content to overall TBR was investigated. In addition, the effect of structural material type on TBR was examined.     

棒状氢化锆慢化钍基熔盐堆燃料组件稳态核热耦合程序开发

基于蒙特卡罗粒子输运程序MCNP与自主开发的子通道热工水力学程序SubTH,开发了棒状氢化锆慢化钍基熔盐堆燃料组件稳态核热耦合程序MCNP-SubTH,解决核热耦合程序因网格类型不同难以耦合的问题,程序具有普适性。MCNP-SubTH通过外耦合的方式进行MCNP和SubTH之间的数据交换,将MCNP计算得到的功率场加载到SubTH的求解文件中,然后将SubTH计算得到的密度和温度场更新到MCNP的输入卡中,实现程序迭代计算。分模块验证了MCNP-SubTH的准确性,并用MCNP-SubTH对棒状氢化锆慢化钍基熔盐堆燃料组件进行了稳态核热耦合计算,验证了核热耦合方法的有效性。     

Research on Coupling of Neutronics and Thermal-hydraulics for Fuel Assembly of Thorium Molten Salt Reactor Moderatedby Zirconium Hydride Rod

Based on Monte Carlo particle transport code MCNP and self-developed sub-channel thermal-hydraulic code SubTH, a code system MCNP-SubTH coupling neutronics with thermal-hydraulics was developed, which was suitable for steady state analysis for a thorium molten salt reactor moderated by zirconium hydride rod (ZrH-MSR). It solved the difficulties in the neutronics and thermal-hydraulics coupling code due to different mesh types, and has a general validity. MCNP-SubTH exchanged data between MCNP and SubTH by an external coupling. The power density field obtained from MCNP was provided as a SubTH solution file to give a user-specified source term, and then the density and temperature field from SubTH was updated and as a new MCNP input file by MCNP-SubTH to realize iterative calculation. The accuracy of MCNP-SubTH was verified by each relatively independent module. MCNP-SubTH application in the fuel assembly of ZrH-MSR was studied, and its validity was verified.     

产氚聚变堆概念设计系统研究

提出了基于球环类型的先进氚生产堆概念设计,它是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,设计中利用了球形环的先进等离子体物理性能和紧凑的结构特征,并尽量利用真空室内的空间安置氚生产包层以减少氚泄露而增加氚增殖率,达到年生产氚1000 g的目标,相应的堆利用因子为40％。在2D中子学计算的基础上提出了较为完整的初步概念设计。逐项进行了分析,同时对设计的风险、不确定性和后备方案也做了概括的解释。为下一步更详细、具体的概念设计提供了直接的依据和重要的参考价值。     

基于RCC-M规范的反应堆压力容器快速断裂分析与讨论

对2000版和2007版RCC-M附录ZG中的快速断裂分析评价方法进行对比分析。结果表明:2007版附录ZG的适用范围更广,分析评价过程相对简化,对材料老化的考虑也更加全面。以堆芯筒体段的快速断裂分析评价为例,对基准裂纹的方位参数进行敏感性分析,将2个版本的评价结果进行对比。结果表明:采用2007版附录ZG得到的评价结果安全裕量更大,同时增大了反应堆压力容器压力-温度(P-T)曲线的限制范围,扩大了核电厂运行操作空间。     

Development of Reflux Vapor Trap for Sodium Cooled Fast Breeder Reactor, (I)

With the view of obtaining basic data required for designing durable reflux vapor traps of high performance for use in sodium-cooled FBR's, experiments were conducted to (a) select a suitable packing, and (b) to examine the effect brought on trapping performance by changing the gas flow rate, packing material, packing density and trap outlet temperature. The results indicated that:

(1) As trap packing, plane weave stainless steel mesh proved to ensure lower pressure drop through trap.

(2) Using the plane weave mesh packing, and with the trap outlet temperature kept at 130°C, the reflux vapor trap efficiency was found to exceed 99.6% in the range of vapor trap gas velocity below 1.3 m/s, and packing density below 0.07 g/cc. The efficiency decreased at outlet temperatures above 130°C.     

Effects of fuel salt composition on fuel salt temperature coefficient(FSTC) for an under-moderated molten salt reactor(MSR)

With respect to a liquid-fueled molten salt reactor(MSR),the temperature coefficient of reactivity mainly includes the moderator temperature coefficient(MTC)and the fuel salt temperature coefficient(FSTC).The FSTC is typically divided into the Doppler coefficient and the density coefficient.In order to compensate for the potentially positive MTC,the FSTC should be sufficiently negative,and this is mostly optimized in terms of the geometry aspect in pioneering studies.However,the properties of fuel salt also directly influence the FSTC.Thus,the effects of different fuel salt compositions including the~(235)U enrichment,heavy metal proportion in salt phase(HM proportion),and the~7Li enrichment on FSTC are investigated from the viewpoint of the essential six-factor formula.The analysis is based on an undermoderated MSR.With respect to the Doppler coefficient,the temperature coefficient of the fast fission factors(a_T(ξ))is positive and those of the resonance escape probability(a_T(p)),thermal reproduction factor(a_T(g)),thermal utilization factor(a_T(f)),and total non-leakage probability(a_T(λ))are negative.With respect to the density coefficient,a_T(p)and a_T(g)are positive,while the others are negative.The results indicate that the effects of the~(235)U enrichment and HM on FSTC are mainly reflected in a_T(e)and a_T(p),which are the dominant factors when the neutron spectrum is relatively hard.Furthermore,the~7Li enrichment influences FSTC by a_T(f)and a_T(λ),which are the key factors in a relative soft spectrum.In order to obtain a more negative FSTC for an under-moderated MSR,the possible positive density coefficient,especially its a_T(p),should be suppressed.Thus,a lower~(235)U enrichment(albeit higher than a certain value,5 wt%in this article)along with a lower HM proportion and/or a higher~7Li enrichment are recommended.The analyses provide an approach to achieve a more suitable fuel salt composition with a sufficiently negative FSTC.     

稀土氧化物Gd2O3、Nd2O3、Sm2O3和Dy2O3在KCl-LiCl-Li2O熔盐中的电解

采用循环伏安法和计时电位法研究了Li₂O在KCl-LiCl熔盐中的电化学行为,并利用卷积伏安法计算了923 K下O₂^-在KCl-LiCl熔盐中的扩散系数（D),得到D=0.5×10^-5 cm²/s。以Gd₂O₃、Nd₂O₃、Sm₂O₃和Dy₂O₃为阴极,在KCl-LiCl-Li₂O(w=1%)熔盐中进行电解（恒电压3.40 V、电解温度923 K、电解时间25 h）。通过X射线衍射分析(XRD),证实稀土氧化物被部分还原为金属,并分析了电解过程中可能发生的反应。同时利用PRS模型（该模型可将固态阴极内离子的极限扩散速率与固态氧化物孔隙P、金属/氧化物摩尔体积R、阴极还原后的体积收缩率S等参数关联）分析了这些稀土氧化物的电解还原模型,得到Gd₂O₃、Nd₂O₃、Sm₂O₃和Dy₂O₃的最优孔隙率分别为18.7%、24.2%、30.6%、16.7%,最短电解时间分别为133、157、143、119 h,将这些结果与电解实验结果进行对比,发现阴极的孔隙率和电解时间均不满足金属氧化物完全被还原的要求,并给出了相应的解释。     

可变量液体样品的Ge（Li）探测器全能峰效率刻度

用多核素混合标准溶液制备了６种不同重量的圆柱体，井形体标准源，作了Ｇｅ（Ｌｉ）探测器对不同样品量的全能峰效率刻度。应用最小二乘法对两种不同形状的液体源的效率与γ射线能量及样品量关系分别作函数拟合。对两种几何形状的液体源作了相同样品量时的效率与探测限的比较。     

~2H(~6He,~7Li)n反应角分布和~7Li的质子谱因子

在中国原子能科学研究院HI-13串列加速器次级束流线上,使用6He次级束首次测量了质心系能量为9.1 MeV的2H(6He,7Li)n反应角分布,并用扭曲波波恩近似(DWBA)进行理论分析,导出了7Li的质子谱因子为0.40±0.02。