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.     

HPLWR equilibrium core design with the KARATE code system

The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of neutronic core design within the framework of the "HPLWR Phase 2" FP-6 and the Hungarian “NUKENERG” projects. As the coolant density along the axial direction shows remarkable change, coupled neutronic-thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the assemblies. A parametrized diffusion cross section library was prepared for the HPLWR assembly with the MULTICELL neutronic transport code. The parametrized cross sections are used by the KARATE program system, which was verified by comparative Monte Carlo calculations. Preliminary loadings of the HPLWR core were assessed, which contain insulated assemblies with Gd burnable absorbers. The fuel assemblies have radial and axial enrichment zoning to reduce hot spots.     

Design and analysis of a thermal core for a high performance light water reactor

The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25 MPa with a core outlet temperature of 500 °C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 °C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat-up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. The concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility. Coupled neutronic/thermal-hydraulic analyses are defining the initial distribution of enrichment, control rod positions and the use of burnable poisons. Sub-channel analyses predict the coolant mixing inside assemblies, and a porous media approach simulates the flow of moderator water between assembly boxes. Finally, structural analyses of the assembly boxes are needed to minimize deformations during operation. Even though the core design cannot yet considered to be final, this state of the art review shall summarize the progress achieved so far and outline the remaining challenges.     

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C.In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Design for a high power-density Astron reactor

A liquid lithium blanket surrounding the plasma volume is described. The liquid lithium flows along magnetic flux tubes at a high speed. There is no vacuum wall between the blanket and the plasma. The E-layer of relativistic particles within which the plasma is confined serves as a vacuum wall protecting the plasma from the lithium vapor, which is continuously produced at the surface of the blanket, by ionizing the lithium atoms and ejecting the same along open magnetic lines. The heat load at the surface of the blanket generated by 14 MeV neutrons can be several hundred MW per square meter.Work performed under the auspices of the U.S. Atomic Energy Commission.Deceased September 24, 1972.     

On the diffusion coefficient calculation in two-step light water reactor core analysis

This paper presents consistent and rigorous accuracy assessments of various methods for calculating the diffusion coefficients in a two-step reactor core analysis of light water reactors (LWRs). The diffusion coefficients are significantly affected by the transport correction and critical spectrum calculations. There are various methods for the transport corrections (inflow/outflow/hybrid corrections) and critical spectrum calculations (B₁/P₁/CASMO-4E methods) so that it is necessary to decide the best combination to achieve a high accuracy in the transport/diffusion two-step analysis. Numerical tests are performed step-by-step to search for the best combination of the methods by comparing each other the transport one-step results, transport/diffusion two-step results, and Monte Carlo results. Numerical test results with a large and a small LWR core show that the combination of inflow transport correction and CASMO-4E critical spectrum calculation is most accurate than the other combinations in terms of eigenvalues and assembly power distributions.     

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.     

Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors

The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out. Crucial development issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. As a result, it has been confirmed that the sodium-cooled FR concept is highly suited to the development targets and R&D issues are related enhancing the economy with certain perspectives for realization. A flexible and robust development program for the FR cycle system will be proposed taking account of the characteristics for each FR concept until the end of the Phase II study.     

Thermal hydraulic analysis of supercritical water reactor cooled by TiO2 nanofluid

Heat transfer study of nanofluids as coolant in SCWRs core has been performed at Helwan University. A thermal hydraulic code has been produced to study the effect of TiO₂ nanofluid water based as a coolant with comparison with pure water as a coolant. Various volume fractions of nanoparticles TiO₂ (2, 6 and 10%) were used in order to investigate its effects on reactor thermalhydraulic characteristics. Based on Parameters of a SCW Canadian Deuterium Uranium nuclear reactor (CANDU), the fuel assembly was modeled to study the effect of nanoparticles volume fraction on thermos-physical properties of basic fluid and the temperature distribution of fuel, cladding surface and coolant in axial direction. The theoretical results showed that the density, viscosity and thermal conductivity of the coolant increases with the increase of nanoparticles volume fraction, contrasting to specific heat, which decreases with the increase in nanoparticles volume fraction.     

Investigation of burnable poisons effects in reactor core design

Burnable absorber rods (BAR) and chemical shim are the main control poisons that are used in the core for improving the reactor behavior and satisfying the safety criteria during the core life time. These poisons have several constraints, criteria, advantages and also disadvantages from the safety and operation points of view; and these characteristics depend on the concentration and distribution of mentioned poisons in the reactor core. Therefore, understanding their effects on the reactor core behavior, especially the mutual interaction between them, is a crucial issue in reactor core design procedure. In this study, the influences of the burnable poisons on the main parameters of the reactor such as multiplication factor, burnup, soluble poison concentration, moderator temperature coefficient and power peaking factor over the reactor life time are investigated. The VVER-1000 reactor was selected for this investigation.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

基于GDT的聚变裂变混合堆堆芯参数初步设计研究

基于Gas Dynamic Trap(GDT)装置的实验进展,提出了用于驱动聚变裂变混合堆包层的聚变堆芯参数设计。基于零维堆芯物理模型,计算分析给出了一套聚变功率为50MW的初步堆芯参数方案。利用GDT装置的实验结果对该物理模型进行计算对比校验,显示该物理模型和设计参数的可靠性。     

Application of homogenization techniques for inflow transport approximation on light water reactor analysis

The transport cross-section based on inflow transport approximation can significantly improve the accuracy of light water reactor （LWR） analysis,especially for the treatment of the anisotropic scattering effect.The previous inflow transport approximation is based on the moderator cross-section and normalized fission source,which is approximated using transport theory.Although the accuracy of reactivity is increased,the P₀ flux moment has a large error in the Monte Carlo code.In this s...     

Requirements for high temperature reactor fuel particle design assessment

In the framework of a large Research and Development programme devoted to High Temperature Reactors (HTR) and set up in the CEA from 2000 on, we will address ourselves to the issue of coated fuel performance and design. Although HTR fuel main features have been established for a long time, we need today to reassess the fuel design to make sure that it meets the requirements linked to the most recent projects of High Temperature Reactors. Thus, in collaboration with Framatome and in connection with the Gas Turbine - Modular Helium Reactor (GT-MHR) international project, we are planning to perform parametric thermal and mechanical studies, regarding different particle design options (kernel diameter, layers composition and thickness) and seeking optima concerning particle leak tightness and fission product retention. But to initiate such studies, we have first of all to define the design bases and the requirements for HTR fuel, in terms of kernel composition (fissile element, oxide stoechiometry, enrichment), particle and compact geometry (dimensions, particle volume fraction in the graphite matrix), power density, cooling gas temperature and irradiation conditions (burnup, fast fluence).     

聚变反应堆含氚废水处理系统的设计

以联合电解催化交换-气相色谱(CECE-GC)为技术路线基础,对聚变反应堆(International thermonuclearexperimental reactor,ITER)含氚废水处理系统(Water detritiation system,WDS)进行了总体设计和主要子系统的设计.与目前的重水提氚演示系统相比,ITER-WDS的不同之处在于不使用氢氧复合器,不采用碱性电解池而使用固体聚合物电解池(Solid polymer electrode,SPE),增加了Pd/Ag膜渗透系统进行氚的回收.     

Current understanding of radiation-induced degradation in light water reactor structural materials

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.     

高温气冷实验堆燃料元件双向探测器的研制

介绍了高温气冷实验堆燃料元件双向探测器的基本原理和实现方法。它以两个并联的感应线圈为敏感元件,通过双通道法采集信号,以８９Ｃ５１单片机为处理核心,系统软件采用循环扫描输入端口的方式获取过球信号,经智能分析、判断,实现了燃料元件的双向检测。     

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.     

Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.