Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States.     

ASME规范概率评价方法对细颗粒石墨的适用性研究

石墨由于其高中子散射截面和低中子吸收截面特性,被广泛应用于第四代高温气冷堆中作为慢化剂、反射层和堆芯结构,故保证其结构完整性对反应堆的安全运行非常重要。由于石墨材料强度分散,概率论方法评价其失效较常用的确定论评价方法更为合适。目前,美国ASME规范采用的概率方法主要针对NBG-18这种大颗粒石墨,对我国高温气冷堆核电站工程项目采用的细颗粒石墨IG-110的适用性未知。同时,我国成都碳素生产的高温堆备选石墨NG-CT-01颗粒大小与IG-110相似,也为细颗粒石墨。因此,文章研究ASME规范概率方法对细颗粒石墨的适用性,并通过实验数据加以验证。结果表明,对于细颗粒石墨,ASME规范过于保守,低估了材料的强度性能。     

Photon interaction with semiconductor and scintillation detectors

Molten salt is used as primary coolant flowing through graphite moderator channel of a molten salt reactor.Working at high temperature under radiation environment,the pore network structure of nuclear graphite should be well understood.In this paper,X-ray tomography is employed to study the 3D pore structure characteristics of nuclear grades graphite of IG-110,NBG-18 and NG-CT-10,and permeability simulation through geometries are performed.The porosity,number of pores and throats,coordination number and pore surface are obtained.NGCT-10 is of similar microstructure to IG-110,but differs significantly from NBG-18.The absolute permeabilities of IG-110,NG-CT-10 and NBG-18 are 0.064,0.090 and0.106 mD,respectively.This study provides basis for future research on graphite infiltration experiment.     

Relation between static and dynamic Young's modulus of nuclear graphites and carbon

In order to evaluate precisely the stresses and strains generated in the graphite structures of high temperature gas-cooled reactors (HTGRs), it is necessary to use not only correct stress-strain relationships but also proper values of the Young's modulus of HTGR graphites. In this study the relation between the Young's modulus obtained from the slope of the stress-strain curve at the origin and that measured by the ultrasonic wave propagation method was examined on two grades of HTGR graphites (IG-110 and PGX) and a grade of carbon materials (ASR-ORB). One of the main conclusions obtained here is that the ratio of the static Young's modulus to the dynamic one depends upon the accuracy of the strain measurement. If the Young's modulus which is evaluated from the secant at 0.01 to 0.05% strain is taken as the static modulus, the dynamic Young's modulus measured using a 5-MHz transducer is approximately equal to the static one.     

Present status of the High Temperature Engineering Test Reactor (HTTR)

In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis.The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR.The installation permit of the HTTR was issued by the Government in November 1990.     

Improved Graphite Damage Model for Predicting Property Changes of HTGR Graphites under Isothermal and Nonisothermal Irradiations

Graphite components of a High Temperature Gas-Cooled Reactor (HTGR) core will be subjected to fast neutron irradiation-induced damages causing changes in engineering properties which are dependent significantly on irradiation temperature and neutron fluence. A Graphite Damage Model (GDM) elaborated to interpret observed isothermal irradiation behavior of various polycrystalline graphites has been improved in order to apply it to nonisothermal irradiations. The paper outlines the physical and mathematical formulations of the improved GDM together with some comparisons between predictions and measurements. On the basis of nonlinear fittings of the model functions to measured data, a total of 28 global modeling parameters have been determined successfully. Furthermore, a recurrence formula has been devised to permit a nonisothermal irradiation behavior to be predicted in terms of the new GDM. This has been proved by comparison of the predicted changes in linear dimension and thermal conductivity with the measured ones of some graphite materials irradiated at constant and changing temperatures.     

Behaviors of fine (IG-110) and ultra-fine (HPG-510) grain graphite irradiated by 7 MeV Xe26+ ions

Developing a molten salt reactor needs molten salt-impermeable nuclear graphite. Ultra-fine grain graphite is a good choice as it is better in permeability than fine grain graphite. In this paper, ultra-fine grain graphite(HPG-510) and fine grain graphite(IG-110) samples are irradiated at room temperature by 7 MeV Xe ions to doses of 1 × 10~(14)-5 × 10~(15) ions/cm~2. Scanning electron microscopy, transmission electron microscopy(TEM), Raman spectroscopy and nano-indentation are used to study the radiation-induced changes. After irradiation of different doses, all the HPG-510 samples show less surface fragment than the IG-110 samples. The TEM and Raman spectra,and the hardness and modulus characterized by nano-indentation, also indicate that HPG-510 is more resistant to irradiation.     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手,对5个循环方案进行比较分析。综合考虑各种工程因素,上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的,其中下位循环在高温气冷堆范围是亚临界参数循环,在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化,且反应堆入口温度在反应堆压力壳材料允许的范围内,具有足够的安全性。     

高温气冷堆用石墨材料的腐蚀实验与软件模拟

在高温气冷堆进水进空气事故下,空气和水蒸气会与堆内的石墨材料发生化学腐蚀反应,从而可能影响反应堆的安全。为研究高温气冷堆内石墨材料的氧化腐蚀特性,本文利用气相色谱法实验测量了IG-110石墨在不同温度和不同气体组分配比情况下的腐蚀速率及腐蚀产物,并利用THERMIX/REACT软件对整个石墨腐蚀过程进行了模拟。研究结果表明：反应温度对石墨腐蚀的影响最为显著,腐蚀速率随着温度的升高而增大,同时随着温度升高,CO与CO₂的含量比也逐渐增大。通过与实验结果对比分析,验证了THERMIX/REACT软件用于高温气冷堆安全分析的可靠性。     

Anisotropy of coefficient of thermal expansion of nuclear graphite under compressive stresses

The coefficient of thermal expansion (CTE) of nuclear graphite IG-110 and NBG-18 under compressive stresses of 20 MPa, 30 MPa and 40 MPa has been measured by strain gauge method and corresponding anisotropies of CTE under stresses were investigated. With the increasing compressive stresses, the CTE of IG-110 and NBG-18 parallel and perpendicular to the loading directions increased significantly and decreased gradually respectively. The corresponding CTE anisotropies of IG-110 and NBG-18 almost maintain below 1.05 and keep their original near-isotropic properties under compressive stresses maybe due to the homogeneous sensitivity of CTE to the stresses, perfect crystallites in the grains and homogeneous alignment of grains in graphite. The constant isotropic properties of graphite IG-110 and NBG-18 under stresses are beneficial for the integrity and safety of the graphite used in the reactor.     

基于SAXS方法的核石墨微孔结构的温度影响

X射线小角散射(Small Angle X-ray Scattering,SAXS)是研究纳米尺度微观结构的重要手段。本文利用同步辐射SAXS技术测量了25oC、100oC、200oC、300oC和400oC时,IG-110和NBG-18核石墨在纳米尺度范围内孔隙的数量分布及其分形特征的变化。实验结果表明,IG-110和NBG-18核石墨的微观结构中存在微小尺寸上的不均匀区域,且核石墨孔隙的固气结构具有明锐的界面。但随着温度的升高,固气界面的变化并没有呈现出明显的规律性。此外,在纳米尺度上,IG-110和NBG-18核石墨的孔隙数量随温度呈现增加的趋势,且IG-110核石墨孔隙数量的增加幅度大于NBG-18核石墨,其平均孔隙尺寸的减小幅度大于NBG-18核石墨。在核石墨的微孔结构内,其固气界面的分形维数随温度升高逐渐减小,且NBG-18核石墨分形维数的变化幅度小于IG-110核石墨。这表明核石墨的分形结构随温度的升高逐渐光滑。     

Experimental study on the reaction between nuclear graphite IG-110 and carbon dioxide

With nuclear graphite IG-110, we measured various kinetic parameters and reaction rates of the C/CO₂ reaction. As a result, its activation energy is 295 ± 8 kJ/mol and the order of reaction is 0.9. It turns out that the rate of C/CO₂ reaction is much smaller than the rate of the C/O₂ reaction which is dominant in HTGR air-ingress below 1400 °C. Finally, we propose the following rate equation for the C/CO₂ reaction of IG-110:

     

Transient thermal–hydraulic simulations of direct cycle gas cooled reactors

This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to “active” heat extraction systems.     

Fracture mechanical properties and neutron irradiation effects of fuel compacts for the HTTR

To simulate the nuclear fuel for High Temperature Engineering Testing Reactor (HTTR), fuel compact models using SiC-kernel coated particles instead of UO₂-kernel coated particles were prepared under the same conditions as those for the real fuel compact. The mechanical and fracture mechanics properties were studied at room temperature. The thermal shock resistance and fracture toughness for thermal stresses of the fuel compact were experimentally assessed by means of arc discharge heating applied at a central area of the disk specimens. These model specimens were then neutron irradiated in the Japan Material Testing Reactor (JMTR) for fluences up to 1.7 × 10²¹n/cm² (E ·> 29 fJ) at 900°C ± 50°C. The effects of irradiation on a series of fracture mechanical properties were evaluated and compared with the cases of graphite IG-110 used as the core materials in the HTTR.     

Strength Degradation of Oxidized Graphite Support Column in VHTR

Air-ingress events caused by large pipe breaks are important accidents considered in the design of Very High Temperature Gas-Cooled Reactors (VHTRs). A main safety concern for this type of event is the possibility of core collapse following the failure of the graphite support column, which can be oxidized by ingressed air. In this study, the main target is to predict the strength of the oxidized graphite support column. Through compression tests for fresh and oxidized graphite columns, the compressive strength of IG-110 was obtained. The buckling strength of the IG-110 column is expressed using the following empirical straight-line formula: σ _cr,buckling = 91.34—1.01 (L/r). Graphite oxidation in Zone 1 is volume reaction and that in Zone 3 is surface reaction. We notice that the ultimate strength of the graphite column oxidized in Zones 1 and 3 only depends on the slenderness ratio and bulk density. Its strength degradation oxidized in Zone 1 is expressed in the following nondimensional form: σ/σ₀ = exp(—kd), k = 0.114. We found that the strength degradation of a graphite column, oxidized in Zone 3, follows the above buckling empirical formula as the slenderness of the column changes.     

压缩应力下石墨IG-110热膨胀系数的测量及验证

采用应变电测法测量压缩应力状态下石墨IG-110的热膨胀系数,分析不同压缩应力对IG-110热膨胀系数的影响.结果表明,压缩应力对IG-110的热膨胀系数影响显著.与未加载时相比,分别加载20、30、40 MPa压缩应力石墨试样平行加载方向的平均热膨胀系数由3.71×10~(-6) K~(-1)逐渐增大至4.20×10~(-6)、4.41×10~(-6)、4.78×10~(-6) K~(-1),分别提高约13.2%、18.9%和28.8%;而垂直加载方向的平均热膨胀系数则由4.03×10~(-6)K~(-1)逐渐减小至3.80×10~(-6)、3.79×10~(-6)、3.75×10~(-6)K~(-1),分别降低约5.7%、6.0%和6.9%.压缩应力状态下石墨热膨胀系数的变化可能与应力导致石墨内部微裂纹的张开和闭合有关.     

Investigation of oxidation behaviors of nuclear graphite being developed and IG-110 based on gas analysis

The oxidation behaviors of the nuclear graphite being developed were investigated using gas chromatograph at 873–1373 K. The oxidation experiments were carried out with the gas flow rate of 0.2 L/min and the oxygen concentrations of 7, 10 and 20 mol%. The oxidation reaction began at 973 K and was accelerated with the increase of temperature. At 1173–1273 K, the oxidation was limited by oxygen supplied to graphite and the reaction rate held steady. From 1273 to 1373 K, the oxidation rate increased obviously due to the significant reaction between CO₂ and graphite. At the low temperature regime (973–1073 K), the apparent activation energies with the oxygen mole fractions of 7%, 10% and 20% were 298, 324 and 321 kJ/mol, respectively. Scanning electron microscope was applied to reveal the pore development of the graphite oxidized at different temperatures. The effect of CO combustion at temperature below 1173 K was discussed based on the oxidation behaviors of the graphite being developed and IG-110. It was suggested that the ASTM D7542-15 standard should be adjusted to fit some popular graphite, such as graphite IG-110.     

Comparison of oxidation model predictions with gasification data of IG-110, IG-430 and NBG-25 nuclear graphite

A phenomenological oxidation kinetics model of graphite is presented and its results are compared with the reported experimental gasification data for nuclear graphite of IG-110, IG-430 and NBG-25. The model uses four elementary chemical kinetics reactions, employs Gaussian-like distributions of the specific activation energies for adsorption of oxygen and desorption of CO gas, and accounts for the changes in the effective surface areas of free active sites and stable oxide complexes with weight loss. The distributions of the specific activation energies for adsorption and desorption, the values of the pre-exponential rate coefficients for the four elementary chemical reactions and the surface area of free active sites are obtained from the reported measurements using a multi-parameter optimization algorithm. At high temperatures, when gasification is diffusion limited, the model calculates the diffusion velocity of oxygen in the boundary layer using a semi-empirical correlation developed for air flows at Reynolds numbers ranging from 0.001 to 100. The model also accounts for the changes in the external surface area, the oxygen pressure in the bulk gas mixture and the effective diffusion coefficient in the boundary layer with weight loss. The model results of the total gasification rate and weight loss with time in the experiments agree well with the reported measurements for the three types of nuclear graphite investigated, at temperatures from 723 to 1226 K and weight loss fractions up to ～0.86.     

Scaling analysis for the high temperature Gas Reactor Test Section (GRTS)

This paper presents an overview of a scaling analysis for a reduced scale Gas Reactor Test Section capable of modeling a variety of important phenomena in a Very High Temperature Gas Reactor. This research effort is being conducted at Oregon State University in support of an Idaho National Laboratory Lab Directed Research and Development project titled, Developing Core Flow Analysis Methods for the VHTR and GFR Designs. The INL point design for a prismatic core VHTR was selected for this scaling analysis, although the project maintains its secondary objective of co-generating Gas-Cooled Fast Reactor GFR-relevant thermal hydraulics data. The specific goal of the scaling analysis was to support the design of a test facility that can be used to produce benchmark data for depressurized conduction cool-down conditions. The scaling analysis determined that the GRTS will be capable of simulating core conduction and radiation heat transfer, vessel radiation heat transfer, core temperature profiles, air-ingress by lock-exchange, air-ingress by molecular diffusion, and single-phase air natural circulation. This paper shall focus on two aspects of the GRTS scaling analysis; air-ingress scaling analysis and scaling of the core heat transfer behavior for a DCC event.