Accelerator driven subcritical system as a future neutron source in Kyoto University Research Reactor Institute (KURRI) —Basic study on neutron multiplication in the accelerator driven subcritical reactor—

A basic study on the nuclear characteristics in the accelerator driven subcritical reactor (ADSR) was performed through a series of neutronics calculations in view of a future neutron source in Kyoto University Research Reactor Institute (KURRI) for the joint use program among researchers of Japanese universities. In this series of calculations, it was assumed that three kinds of monoenergetic neutrons were isotropically generated at the center of spherical and homogeneous cores with different moderator-to-fuel volume ratios in order to examine the spectrum mismatching effect between injected neutrons and fission neutrons born in the subcritical core. The results of calculations clearly showed the spectrum mismatching effect on the neutron multiplication in the ADSR.     

Preliminary Experiments on Accelerator-Driven Subcritical Reactor with Pulsed Neutron Generator in Kyoto University Critical Assembly

A series of preliminary experiments on an accelerator-driven subcritical reactor (ADSR) with 14 MeV neutrons were conducted at Kyoto University Critical Assembly (KUCA) with the prospect of establishing a new neutron source for research. A critical assembly of a solid-moderated and -reflected core was combined with a Cockcroft-Walton-type accelerator. A neutron shield and a beam duct were installed in the reflector region for directing as large a number as possible of the high-energy 14MeV neutrons generated by deuteron-tritium (D-T) reactions to the fuel region, since the tritium target is located outside the core. And then, neutrons (14MeV) were injected into a subcritical system through a polyethylene reflector. The objectives of this paper are to investigate the neutron design accuracy of the ADSR with 14MeV neutrons and to examine experimentally the neutronic properties of the ADSR with 14MeV neutrons at KUCA. The reaction rate distribution and the neutron spectrum were measured by the foil activation method for investigating the neutronic properties of the ADSR with 14 MeV neutrons. The eigenvalue and fixed-source calculations were executed using a continuous-energy Monte Carlo calculation code MCNP-4C3 with ENDF/B-VI.2 for the subcriticality and the reaction rate distribution, respectively; the unfolding calculation was done using the SAND-II code coupled with JENDL Activation Cross Section File 96 for the neutron spectrum. The values of the calculated subcriticality and the reaction rate distribution were in good agreement with those of the experiments. The results of the experiments and the calculations demonstrated that the installation of the neutron shield and the beam duct was experimentally valid and that the MCNP-4C3 calculations were accurately carried out for analyzing the neutronic properties of the ADSR with 14MeV neutrons at KUCA.     

托卡马克实验混合堆 FEB 嬗变 MA 可行性研究

研究了在聚变实验混合堆ＦＦＢ设计中，嬗变长寿命放射性少锕系（ＭＡ，ＭｉｎｏｒＡｃ－ｔｉｎｉｄｅｓ）核废物的可行性。应用改进的一维中子输运和燃耗计算程序ＢＩＳＯＮ３．０，完成了嬗变中子学与核素贫化计算。研究了核废物的嬗变率与辐照时间、包层厚度和废物装载量的关系，并对系统有关参数的选择进行了优化设计。结果表明，该设计（ＭＡ＋Ｐｕ）可年嬗变处置来自５５座相同功率的ＰＷＲ卸出的ＭＡ核废物，同时输出热功率５．４ＧＷ（ｔｈ）。     

基于Dragon与Donjon程序的液态熔盐实验堆临界计算与分析

基于组件输运程序Dragon与堆芯节块法程序Donjon,对包含有上下熔盐腔室、控制棒、实验孔道与中子源孔道的液态熔盐实验堆堆芯进行了计算与分析,给出了液态熔盐实验堆不同组件的等效均匀化模型。根据液态熔盐实验堆特性将中子能群划分为5种少群能群结构,基于所划分的每一种少群能群结构,对单根控制棒与不同控制棒组插入堆芯后的有效增殖因数和控制棒价值进行了计算分析。结果表明,7群能群结构具有更好的计算结果。基于7群能群结构开展了堆芯径向与纵向功率分布,以及控制棒拔出后堆芯的温度反应性系数计算分析,其计算结果与MCNP5计算结果相近,证明了模型等效的合理性以及Dragon和Donjon程序对液态熔盐实验堆的适用性。     

中子核截面在线多普勒展宽方法研究

由中子截面多普勒展宽带来的反应性温度效应对反应堆中子学计算结果具有重要影响。基于自由气体模型和对靶核速度随机抽样的在线多普勒展宽方法,可使用0 K温度下的中子截面对给定温度的问题进行蒙特卡罗计算,摆脱对专用多普勒展宽程序的依赖。本文通过对在线多普勒展宽方法的程序实现,针对典型算例进行了验证和分析,证明了该方法能处理反应性温度效应,并对其适用性和未来发展前景进行了评价。     

A zero dimensional model for simulation of TRIGA Mark II dynamic response

Aim of this work is to reproduce the dynamic behavior of the TRIGA Mark II reactor of the University of Pavia on the entire operative power range (i.e. 0–250 kW) using a zero dimensional approach. In this work the coupling between neutronics and thermal-hydraulics in natural circulation has been considered. In specific, a point reactor kinetics model with one energy group and six delayed neutron precursors groups has been adopted while for thermal-hydraulics modeling two regions have been defined (i.e. the fuel and the coolant). The nonlinear system of coupled Ordinary Differential Equations has been solved by means of MATLAB Simulink^®, which represents a reliable tool for dynamic and control analysis. The model has then been validated through the comparison with a set of experimental data collected in four different reactor power transients, obtaining a very satisfying agreement. Finally, the linear stability analysis of the TRIGA reactor has been performed by means of the root locus, finding out that the power level at which reactor is operating deeply influences the position of the poles of the transfer function between control rod height and neutron density. These results can then be employed as a reliable starting point in designing an automatic device for reactor power control.     

压水堆核电站堆芯物理/热工水力耦合特性研究

采用RELAP5-HD作为堆芯耦合计算程序,以秦山核电二期工程反应堆堆芯为研究对象,建立堆芯活性区的物理/热工水力耦合模型,在此基础上进行了稳态计算和掉棒事故仿真研究。结果表明,使用RELAP5-HD计算得到的结果与电厂实测值符合较好,获得的掉棒事故参数曲线能准确反映事故工况下的参数变化趋势。稳态和事故工况的计算结果均符合堆芯物理/热工水力反馈效应的理论分析,证实了所建立的堆芯耦合模型的准确性,为下一步进行核电站系统的仿真分析提供基础。     

中国聚变工程试验堆包层的核热耦合效应研究

本文以中国聚变工程试验堆（CFETR）的氦冷固态包层和水冷固态包层为研究对象,基于蒙特卡罗程序MCNP和计算流体力学程序FLUENT,利用3D-1D-2D耦合方法和伪材料方法,分别对200 MW的氦冷固态包层和水冷固态包层及1.5 GW的水冷固态包层方案进行了核热耦合计算分析。研究结果表明,金属铍的热散射效应和轻水密度是聚变包层核热耦合效应的主要来源,核热耦合效应对氦冷固态包层的影响可忽略,对水冷固态包层的氚增殖比和温度分布有一定程度的影响。     

反应堆堆芯先进中子学模拟软件SCAP-N研发

堆芯中子学计算是反应堆设计分析的基础,为提高堆芯中子学计算的模拟分辨率与计算精度,开发了反应堆堆芯先进中子学模拟软件(SCAP-N)。该程序首先根据轴向特征对堆芯进行分层,并逐层进行二维堆芯非均匀输运计算,再采用超级均匀化方法(SPH)获得栅元等效均匀化截面,最后进行三维堆芯逐棒(pin-by-pin)输运计算,获得堆芯有效增殖因子与精细棒功率分布。为提高程序计算效率,采用分布式/共享式(MPI/OPENMP)混合并行方式对程序进行了并行化开发。利用虚拟反应堆(VERA)系列基准例题及美国先进非能动压水堆(AP1000)启动物理试验实测数据对程序进行了测试验证。结果表明,相比于商用核设计程序系统,SCAP-N程序采用的逐棒输运技术能够提高堆芯中子学的计算精度。与同类型高精度中子学程序相比,SCAP-N具有更高的计算效率,可进一步提高核电厂的经济性及运行灵活性。     

热管反应堆多物理耦合平台初步研究

为实现高精度、高置信度的核能系统先进数值模拟技术,探究核能系统内部真实的物理过程,本文开发了中子物理-固体导热-应力分析的三维高精度核热固多物理耦合计算平台MPCH,可开展核反应堆的中子输运、热扩散和热膨胀的多物理耦合计算。该程序基于Picard迭代的外耦合框架,整合了开源蒙特卡罗程序OpenMC、有限元程序Nektar++和SfePy。本文以新型空间热管反应堆KRUSTY为对象,在核热固耦合的计算框架下对其进行计算分析。多物理耦合计算结果表明,该耦合平台能够有效预测KRUSTY反应堆的有效增殖因子变化、功率分布、温度分布及热膨胀现象;在4 kW的堆芯热功率下,全堆局部温差为21.6K,热应力导致的形变率为2.47%,核热固耦合的作用会使堆芯的温度分布更加均匀。该多物理耦合计算程序的设计对新堆设计、研发和校核具有重要作用。      

新概念熔盐堆的固有安全性及相关关键问题研究

新概念熔盐堆是6种第四代反应堆中唯一的液体燃料反应堆,在固有安全性、经济性、核资源可持续发展及防核扩散等方面具有其它反应堆无法比拟的优点。针对熔盐堆的特点,建立通用的物理分析、热工水力分析和安全分析模型,并采用隐式方法实现物理热工的耦合。将建立的数学模型应用于锕系元素再循环嬗变熔盐堆（MOSART）的计算,对其堆芯物理特性、热工水力特性和安全特性进行了系统分析,考察了入口温度、速度及燃料盐在堆芯外运行时间的影响。     

基于CFD程序的物理热工耦合计算不确定性分析

基于计算流体力学（CFD）程序FLUENT的用户自定义函数（UDF）,耦合中子动力学计算模型、燃料棒热传导计算模型、不确定性分析程序SIMLAB,开发了物理热工耦合计算不确定性分析平台CFD/PFS,并开展了小型自然循环铅基快堆SNCLFR-10的无保护超功率（UTOP）事故的不确定性量化,最后对计算结果进行不确定性分析和敏感性分析。研究表明,CFD/PFS平台的物理热工耦合计算具有良好的可靠性、精确性;总反应性峰值、功率峰值等瞬态安全参数的名义值均处于95/95双侧容忍限值内,且名义值与限值相对偏差小于3.95%;燃料多普勒系数是主要不确定性来源,对反应堆安全影响最大。     

CERMET-SNRE堆芯物理计算分析

核火箭发动机功率高、寿命长、比冲大,在执行深空探测和星际航行任务时具有不可替代的优势。小型化是核火箭发动机的一个重要趋势,基于此提出了一种使用钨基金属陶瓷燃料的小型核火箭发动机（CERMET-SNRE）堆芯方案,并采用蒙特卡罗程序（MCNP）进行了精确建模,计算了相关物理参数。计算分析结果表明：CERMET-SNRE堆芯能谱硬,燃耗浅,后备反应性足够,功率分布合理,控制鼓与安全棒价值足够,发射掉落事故下有效增殖因数小于0.98,堆芯方案合理,满足设计要求。     

先进反应性测量系统的研发与应用

相比于传统的反应堆控制棒价值测量方法,快速的动态棒价值测量方法要求反应性测量设备具有更高的精度和性能,以准确获取和处理堆外探测器的电流信号,并需通过额外的堆芯中子学计算对试验过程中的空间效应进行修正。为此本研究开发了一套包含先进物理试验测量仪（APTC）和动态棒价值测量软件包（LIGHT）的先进反应性测量系统（SMART）,并对SMART开展了一系列验证试验。结果表明,SMART具备完整的物理启动试验功能,其精度和性能能够满足包括动态棒价值测量在内的物理启动试验的要求;在300 MW压水堆核电厂中的成功应用也充分验证了SMART的工程应用能力。      

Gallium-cooled liquid metallic-fueled fast reactor

We have examined the effects on core characteristics of using two different types of Pu-based metallic alloy fuels in the gallium-cooled fast reactor core. In the proposed concept, the liquid metal fast nuclear reactor uses metallic fuel in the liquid phase and gallium coolant at high temperature (inlet 1700K, outlet 1900K). The liquid fuel is continuously supplied to the reactor during operation at full reactor power. The reactor power is controlled by rotational control drums with absorber material. The aim was to evaluate reactor core neutronics and safety characteristics demonstrating a feasibility of the reactor system. Although gallium has large absorption cross section in the high neutron energy region, we can design the core with rather good neutronics performances. The large negative reactivity feedback induced by the thermal expansion of liquid metallic fuel ensures the core's inherent safety against the unprotected loss-of-flow transient.     

Experimental Study of Nuclear Characteristics of Large Axially Heterogeneous Core Using Fast Critical Assembly

The physics characteristics of large axially heterogeneous liquid-metal fast breeder reactors (LMFBRs), particularly the parameters for use in design and safety assessment, were examined using the JAERI fast critical assembly facility, arranged in Assembly XH-1, a partial mock-up of axially heterogeneous LMFBR. The properties measured were (1) criticality, (2) reaction rates and reaction rate ratios, (3) material sample worths, (4) sodium-void worths and (5) B4C control rod worths.

The results were compared with those of prior experiments with assemblies representing conventional homogeneous core. Confirmation was obtained of the typical nuclear characteristics attributed to axially heterogeneous LMFBRs, including flattening of the axial distribution of power and of the differential worth of control rod, as also lower sodium void worth.

Theoretical analyses paralleling the experiments, using JENDL-2 cross section library and JAERI standard calculation code system for fast reactor neutronics, resulted in some discrepancies, particularly for the internal blanket, in respect of plutonium sample worth, fission rate and fission rate ratio.     

A neutronics–thermalhydraulics model for preliminary studies on TRADE dynamics

In the last few years the possible role of accelerator driver systems (ADS) for effective transmutation strategies with fully closed cycles has received increased attention due to their potential to improve the flexibility and safety characteristics of transmutation systems. The substantial difference between the neutron kinetics and dynamic behavior of ADS and conventional critical reactors has given rise to a wide international consensus on the need of an experimental program to improve their knowledge and to validate calculation methods. To this end the international cooperation TRADE proposed a sub-critical experiment based on the coupling of a TRIGA reactor in sub-critical core configuration with a proton accelerator (cyclotron) by means of a neutron spallation target. The experiment was initially conceived in the RC1-TRIGA reactor located at the ENEA Center of CASACCIA (Rome, Italy) to demonstrate the feasibility of the accelerator driven system (ADS) concept at a representative power. This article presents a preliminary study performed with the RELAP5/PARCS code on the dynamic behavior of such a system in order to demonstrate the code capability to support the design of the experiment and the safety analysis. The specific code version used joins the well known capability of RELAP5 to treat light water reactors with the potentiality of PARCS modified by ENEA to simulate the three-dimensional neutronics of sub-critical systems, i.e. to treat external neutron sources. PARCS modifications are preliminary assessed against a simple analytical solution of the sub-critical neutronics of the experiment based on the kinetics pseudo-potentials method. A quite detailed model for the coupled code is developed in order to realistically evaluate both the thermal feedback effects, the control rod action and the external source strength changes. A wide range of operational and accidental transients of the sub-critical reactor are simulated with the coupled model in order to obtain a first system response to a number of reactor elementary events at different subcriticality levels. The calculation results show a high qualitative agreement with the sub-critical system physical theory underlining how the numerical model developed could be a useful tool for the definition of the operational procedures and the investigation of accidental conditions; moreover the accidental transient trends highlight the inherent safety behavior of the TRIGA research reactors that makes them extremely suitable for the coupling of the different components with a quite simple licensing procedures.     

Effect of Nuclear Data Libraries on Tritium Breeding in a (D–T) Fusion Driven Reactor

In design a Deuterium–Tritium (D–T) fusion driven hybrid reactor, neutronics and nuclear data libraries have an essential role for reliable neutronics calculations. Therefore, nuclear data libraries are very important to calculate of the neutronic parameters and selection of tritium breeder materials to be used in the blanket. In this study tritium breeding performances of candidate tritium breeding materials, namely, Li₂O, LiH, Li₂TiO₃, Li₂ZrO₃ and Li₄SiO₄ in a (D–T) driven fusion–fission (hybrid) reactor is investigated based on three dimensional (3-D) and one dimensional (1-D) neutronic calculations. 3-D and 1-D neutron transport calculations are performed with Monte Carlo transport code (MCNP 4C), SCALE 5 and ANISN nuclear data codes to determine the tritium breeding ratio (TBR) of the blanket. The effects of different nuclear data libraries on TBR are examined and TBR calculation results are comparatively investigated.     

Development of three-dimensional reactor analysis code system for accelerator-driven system,ADS3D and its application with subcriticality adjustment mechanism

In order to perform the parametric survey for an accelerator-driven system (ADS) core with the subcriticality adjustment mechanism, a new calculation code system, ADS3D, was developed on MARBLE which is a comprehensive and versatile framework for reactor analysis. The application of ADS3D was also demonstrated on the neutronics design of ADS operated by control rod (CR) movement. Through the neutronics calculation, it was shown that the maximum proton beam current was decreased from 20.5 to 11.6 mA due to the switch from beam-operated to CR-operated core.