代中子时间法求解点堆中子动力学方程

由于点堆中子动力学方程是个刚性方程,因此准确、快速、稳定地求解方程是困难的。得益于现代计算机技术的进步,本文直接采用代中子时间计算法求解点堆中子动力学方程,并用C++语言编制了计算程序。经过基准例题和动态-逆动态对比计算,验证了模型、程序计算的准确性和稳定性,而计算时间也是可接受的。     

A neutron point kinetic model for fusion relevant calculations

In the framework of research activities on fusion reactors a great effort is dedicated by the scientific community to the development of tritium breeding blankets. One of the main goals is to assess the neutronic behaviour of such devices to analyse their tritium breeding performance and to evaluate the required data for their thermal–mechanic and thermal–hydraulic design. Many papers have been published on this topic considering some stationary condition to calculate such important quantities as heating power, gas production and dpa rates, tritium breeding ratio, etc., but not much attention has been focussed to neutronic transport analyses in transient conditions. The present paper proposes a simple model based on the point kinetics approximation, which has been set up deriving an alternative formulation of the time-dependent neutron transport equation. This approach allows to define some physical characteristics that can be interpreted in a statistical way, making possible to calculate these quantities numerically by the Monte Carlo method. The adoption of the aforementioned numerical method has the great advantage that complex geometries (as the fusion reactor's blankets are) can be analysed with acceptable computational times. Some simple cases have been investigated to implement the theoretical model proposed with MCNP5 code and to show its potentiality. Then, applications to fusion reactor ITER blanket module and to the Helium Cooled Test Blanket Module, to be tested in ITER, have been taken into account in order to assess their neutronic time-dependent behaviour and the results obtained have been critically discussed.     

Numerical solution of the neutron transport equation using cellular neural networks

Various methods have been used for solving the neutron transport equation in the past, and a number of computer codes have been developed based on these solution methods. This paper describes a novel method for the solution of the steady-state and time-dependent neutron transport equation using the duality between neutronic parameters in the method of characteristic (MOC) and the electrical parameters in the cellular neural networks (CNN). The relevant electrical circuit can be simulated by professional electrical circuit simulator software, HSPICE. This software is used for numerical solution of the transport equation only by preparation of appropriate inputs. This method does not need inner and outer iterations, which is a necessary step in the other deterministic methods. One of the main applications of the proposed method may be the development of a new hardware by VLSI technology for online spatio-temporal calculations of the transport equation for nuclear reactor core. The accuracy and capability of this method are examined in a 2D steady-state problem for a BWR fuel assembly, and a 2D time-dependent TWIGL seed/blanket problem.     

基于三维MC-SN耦合方法的PWR压力容器快中子注量计算基准分析

蒙特卡罗(MC)-离散纵标(SN)耦合方法是解决同时具有复杂几何和深穿透特点的核装置屏蔽问题的有效方法。本文首次将三维MC-SN耦合方法应用于压水堆屏蔽计算。针对NUREG/CR-6115压水堆基准模型,选取热屏蔽内表面为公共交界面,将其分为几何复杂的MC模拟区和具有深穿透特点的SN模拟区。三维MC程序用于精确描述堆芯到热屏蔽精细模型,并记录穿过热屏蔽内表面的中子径迹信息。接口程序将中子径迹转换为SN计算所需的边界源,提供给三维SN程序进行热屏蔽到压力容器的计算。计算结果包括压力容器内表面、1/4壁厚处及焊缝处快中子注量(E>1.0 MeV)圆周方向分布。三维耦合方法计算结果与基准报告提供的MCNP、DORT结果符合良好,验证了该方法处理圆柱坐标系屏蔽问题的有效性和程序使用的正确性。     

Numerical analysis of pressure load in a PWR cavity in an ex-vessel steam explosion

Ex-vessel steam explosion may happen as a result of melting core falling into the reactor cavity after failure of the reactor vessel and interaction with the coolant in the cavity pool. It can cause the formation of shock waves and production of missiles that may endanger surrounding structures. Ex-vessel steam explosion ener- getics is affected strongly by three dimensional （3D） structure geometry and initial conditions. Ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is developed for simulating fuel-coolant interactions. The reactor cavity with a venting tunnel is modeled based on 3D cylin- drical coordinate. A study was performed with parameters of the location of molten drop release, break size, melting temperature, cavity water subcooling, triggering time and explosion position, so as to establish parame- ters＇ influence on the fuel-coolant interaction behavior, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. The most dangerous case shows the pressure loading is above the capacity of a typical reactor cavity wall.     

Identification of reactor internals'' vibration modes of a Korean standard PWR using structural modeling and neutron noise analysis

The vibration characteristics of a Korean standard PWR reactor internals have been estimated through a three-dimensional finite element analyses and verified by using the mode separated power spectral density functions obtained from the ex-core neutron noise signals. Also the natural vibration modes of the fuel assembly have been identified measuring both the ex-core and the in-core neutron noise signals which are close to each other. As a result, the fundamental bending mode frequency of the reactor internal structure is found to be around 8 Hz and the fundamental shell mode frequency 14.5 Hz, respectively. It is also shown that the fundamental bending mode frequency of the fuel assembly is 2.3 Hz and the 2^nd bending mode frequency 5.8 Hz, respectively. These results can be used for the supplements of the Korean standard PWR's CVAP (Comprehensive Vibration Assessment Program) data.     

Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity.     

Reactivity space analysis of PWR core depletion

Reactivity space analysis of PWR core depletion is used to investigate solutions to the core burnup and fuel utilization maximization problems. Results of few-region analysis of the SEQUOYAH PWR with the EPRI nodal code simulate-e show that there is an order of magnitude difference in the effect of fuel arrangement and reactivity control in the optimal solutions. It is concluded that emphasis in the core reload design should be given to achieving the optimal fuel arrangement. Furthermore, it is shown that the constant power, Haling depletion strategy is an effective method of isolating the arrangement problem from reactivity control considerations during the core design process.     

压水堆核容器密封分析(英文)

本文作为核容器密封性能综合研究中心课题之一,给出容器密封分析基本方程及程序系统。经多种试验校核证实程序可信。根据多个容器分析计算,提出了就密封性能而言的压力容器类型概念,这对容器设计选定合宜预紧系数、保证密封并改善主螺栓疲劳性能有重要意义。     

Seismic analysis of RCS with finite element model for advanced PWR

Reactor Coolant Pumps (RCPs) are very important to the safe operation of Nuclear Power Plants (NPPs), especially during the earthquake, which needs detailed seismic analysis of individual RCPs and the boundary conditions, for example, at the nozzles. In this paper, three-dimensional finite element model of Reactor Coolant System (RCS) is constructed from a systematic perspective to perform dynamic evaluation, in which the boundary conditions could be given. The seismic spectrum analysis with three orthotropic directions is performed to obtain the stress and displacement response, which shows that the maximum Tresca stress locates in the connection part of SG with RCP and the maximum displacement occurs at the surge line. Sensitivity analysis of spectrum input angle and stiffness of supports is performed, which may be useful to further design and analysis. Furthermore, direct integration method is used to perform time-history analysis, and the boundary conditions of RCP, the loads, acceleration and displacement at nozzles are obtained, which could support the detailed analysis of RCP components. Besides, the lumped mass model of RCS is also constructed to compare with three-dimensional finite element model, which means that for the complicated geometry the 3-D model is better than the lumped mass model.     

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U-ZrH_1.6 fuelled PWR cores are investigated and compared against those of the currently used UO₂ fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k_∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U-ZrH_1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 ^v/_o of the ZrH_1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.     

Nonstationary problems in the kinetic theory of neutron transport

The nonstationary kinetic equation for monoenergetic neutrons is solved by an application of a Fourier transformation of the unknown function and the source function, taking account of initial data. Detailed consideration is given to the problem of the distribution of neutrons from a moving or oscillating source. Formulas are given for the diffusion length as a function of the velocity of the source, and the neutron field near this is analyzed.Translated from Atomnaya Énergiya, Vol. 16, No. 6, pp. 504–509, June, 1964     

Numerical investigations of concrete attenuation effectiveness in various PWR shield configurations

The choise of a suitable concrete composition for a biological reactor shield often seems difficult because of the variety of available aggregates. To estimate the influence of these aggregates on concrete attenuation effectiveness, test calculations were made for power reactor shields, with extreme variations of concrete components. Concretes have been analysed with single natural (mineral) aggregates such as basalt, barite, magnetite, and serpentine; and with mixed natural aggregates such as basalt-magnetite, serpentine-magnetite, barite-magnetite, and serpentine-barite in varying proportions. The relative attenuation effectivenesses of different concretes in various shield configurations of a PWR-type reactor have been examined.     

压水堆核电厂稳压器波动管热分层现象数值分析

为分析评价压水堆核电厂稳压器波动管热分层现象对波动管结构完整性的影响,采用计算流体力学(CFD)分析方法,对稳压器波动管热分层现象进行了数值模拟.研究了波动管内的流体流动,得到了稳压器波动管的传热特性、流体流场和温度分布,分析了稳压器波动管波动热分层现象与波动流速之间的关系.研究结果表明:波动流速在一定范围内变化时,管道最大截面温差随着波动流速的增大而增大.并且得到了不同波动流速下管道最大截面温差及其出现的位置,指出了热分层现象发生时波动管的薄弱环节.     

压水堆下腔室流量分布数值分析

建立了压水堆下腔室流场的三维数值计算模型,计算了不同环腔厚度和环腔内冷却剂速度条件下,下腔室内冷却剂的流场,分析了环腔厚度和环腔内冷却剂速度对下腔室流向堆芯的流量分布的影响。入口速度不同或环腔厚度不同,在下腔内冷却剂流动形成漩涡的位置、大小和流动速度均会发生改变,导致通过流量孔板通孔的流量分布不同。入口速度较低时,流量孔板上所有通孔的流量分布比较均匀,在平均值附近波动,流量最高的通孔小组出现在边缘处;入口速度较高时,流量明显地呈现出中心高边缘低的特点。通孔小组的流量最大值随着环腔厚度增加由孔板的中心向边缘移动。     

Computation of fundamental time-eigenvalue of the neutron transport equation

A modified α − k power iteration method is presented for the prediction of time-eigenvalue(α) of the neutron transport equation. By developing a direct relationship between K-eigenvalue and α-eigenvalue, a new formula is introduced to estimate the value of α. Compared with the conventional method, it is not required to provide the initial values of α for the modified method. Since it is always difficult to guess the suitable initial values, the modified method is more convenient for solving time-eigenvalue problems. Computational experiences show that the accuracy of the modified method is the same as the conventional method.     

Thermo-hydraulic analysis for SCWR during power-raising phase of startup

The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor(SCWR).According to the given sliding pressure mode of SCWR,thermal analysis on temperature-raising phase and power-raising phase of startup are carried out.Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward.The results show that,during temperature-raising phase with core power increased only,the temperature of moderator,coolant and fuel cladding in inner assemblies are increased with little amplitude.During power-raising phase with core power and feed-water flow rate increased,the coolant temperature keeps unchanged,but the moderator temperature is decreased.With a greater variation of power,fuel cladding temperature shows a greater increase.Furthermore,considering the uneven distribution of radial power,thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies,which becomes serious as flow rate and power increase.By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup,the cladding temperature difference could be effectively reduced,which provides a certain reference for startup optimization of SCWR.