高通量工程试验堆堆内单晶硅旋转体温度测量

本文论述了高通量工程试验堆堆内单晶硅旋转体温度测量装置,介绍了在反应堆各种功率下单晶硅辐照样品温度的测量结果,并且根据温度测量值估算了单晶硅的 γ 发热率,进行了误差分析,同时作者还进行了单晶硅温度与轴流泵转速关系的试验.测量结果表明,在高通量堆首炉工况下单晶硅辐照装置内部冷却水不会沸腾.     

中子掺杂硅单晶的放射性强度

用反应堆辐照单晶硅进行中子掺杂,人们对辐照后的放射性水平极为关切。我们利用FELIXC-512计算机进行了数值计算。其结果与实验数据符合较好。我们将放射性强度与辐照中几个主要参量的关系列表给出,便于查找、制定辐照工艺及了解产品性能。单晶硅中子掺杂主要核过程为:     

智能中子通量积分仪研制报告

介绍了用于测量单晶硅在反应堆辐照孔道中接受中子照射的积分通量仪器——中子通 量积分仪。文中简要说明了研制的意义及技术指标;着重阐述了仪器的工作原理、构成及其特点。由于采用了单片机技术,实现了仪器的智能化。经实际使用表明,该仪器具有功能强、精度高、应用范围广、自动化程度高、操作简便、工作可靠等特点。从而提高了单晶硅辐照     

智能中子通量积分仪研制报告

介绍了用于测量单晶硅在反应堆辐照孔道中接受中子照射的积分通量仪器——中子通量积分仪。文中简要说明了研制的意义及技术指标;着重阐述了仪器的工作原理、构成及其特点。由于采用了单片机技术,实现了仪器的智能化。经实际使用表明,该仪器具有功能强、精度高、应用范围广、自动化程度高、操作简便、工作可靠等特点。从而提高了单晶硅辐照技术水平,也提高了生产效率。     

影响反应堆压力容器钢辐照脆性的因素及控制措施

辐照性能是影响反应堆压力容器使用寿命和运行安全的重要形式.本文阐述了影响反应堆压力容器辐照脆性的主要因素,并从反应堆压力容器钢的化学成分、生产工艺和辐照后退火等方面提出了控制反应堆压力容器钢辐照脆性的主要措施.     

基于P89C669的反应堆样品照射控制装置设计

介绍反应堆中子活化自动辐照检测系统的设计和实现过程.系统采用PHILIPS公司具有ISP功能的单片机P89C669作为控制核心,通过气动传输装置控制样品进入反应堆和离开反应堆.安装在反应堆堆口处的光电传感器检测管道内的样品,精确控制样品的辐照时间和计算样品的冷却时间,并通过串行接口实现与计算机的通讯,实时向计算机传输样品辐照的信息,单片机驱动液晶显示器完成辐照时间、冷却时间的动态数据显示.     

高通量堆NTD硅系统活化剂量计算

根据单晶硅及靶桶材料成分、测量的辐照孔道中子通量谱与辐照时间,采用点燃耗程序ORIGEN与蒙特卡罗程序MCNP耦合计算高通量堆中子嬗变掺杂(NTD)硅辐照系统活化后的外照射剂量当量率及各种活化产物放射性核素衰减变化情况,同时对各种活化核素剂量率贡献及相应衰减时间进行了分析。通过计算结果与堆厅γ电离室剂量率监测对比验证及堆厅屏蔽层厚度的保守估算,表明目前NTD硅系统转运过程屏蔽设计满足辐射防护要求,并提出有益建议。     

基于欧拉指数方法的活化计算程序开发及应用

材料活化产生的放射性不仅对反应堆系统安全产生重要的影响,还会使反应堆退役后存在大量核废料的后处理问题。本文基于欧拉指数方法,采用EAF数据库,自主开发了活化计算程序EuACT,对ZIRLO、Zr-4、M5、N18包壳材料的活化特性进行了计算与分析,并与欧洲活化程序FISPACT计算结果进行了对比。分别选取0.5、1.0和1.5a的辐照时间,计算3种情况下辐照停堆后不同包壳材料的放射性比活度以及衰变余热随停堆时间的变化,并对包壳材料活化特性进行初步分析。结果表明:EuACT与FISPACT的计算结果符合良好;仅从停堆后放射性比活度和衰变余热的角度分析,Zr-4相比其他3种材料具有一定优势。     

300#反应堆单晶硅中子嬗变掺杂计算

介绍了300#反应堆单晶硅中子嬗变掺杂辐照量计算的原理,给出了计算软件类结构及其主要类中的实现函数,描述了该程序在辐照计算,原始和出厂清单处理,以及各类原始数据处理方面的功能和特点,对相应的界面都给出了图示,并对软件的特性和应用前景作了分析。     

反冲释放对反应堆活化腐蚀产物源项的影响研究

中子与靶核碰撞时引起的靶核反冲释放，对于反应堆活化腐蚀产物源项分析有非常重要的影响。对于使用水冷方式的反应堆，在辐照区反冲释放可使活化腐蚀产物离开壁面进入到冷却剂中，并随冷却剂迁移到非辐照区，使非辐照区的设备也带有放射性。本文研究了反冲释放在反应堆内的作用方式，建立了反冲释放的计算模型和程序模块，并集成到活化腐蚀产物源项分析程序CATE中，利用改进后的CATE程序，计算分析了堆芯与蒸汽发生器中主要的活化腐蚀产物核素⁵⁸Co与⁶⁰Co在考虑反冲释放前后的数值，明确了反冲释放效应的影响程度。计算结果表明：考虑反冲释放前后堆芯处⁵⁸Co与⁶⁰Co活度的比值有所下降，而在蒸汽发生器中的比值则有所上升；反冲释放的总作用概率与腐蚀产物层厚度相关，会随着反应堆的运行而逐渐降低，反应堆运行初期作用概率的数量级在10^-1，对活化腐蚀产物的迁移有显著影响，100 d后作用概率的数量级下降到10^-3，对活化腐蚀产物源项的影响较小。     

低活化马氏体钢的微观结构与力学性能

介绍了作为聚变反应堆候选结构材料的低活化马氏体钢的基本设计思路,初步确定了材料的化学成分和热处理工艺,研究了材料的冶金特性、微观组织和力学性能.同时,对比了添加少量钇和硅对材料性能的影响,发现添加硅可以提高材料强度,同时能保证材料具有足够的塑性和韧性;钇的添加对改善材料的塑性很有帮助,但是会使材料强度降低.     

Advanced pebble bed high temperature reactor with central graphite column for future applications

Design evaluations of the advanced pebble bed high temperature reactor, AHTR, with central graphite column are given. This reactor, as a nuclear heat source, is suitable for coal refinement as well as for electricity generation with closed gas turbine primary helium circuit. With this design of the central graphite column, it is possible to limit the core temperatures under the required value of about 1600°C in case of accident conditions, even with higher thermal power and higher core inlet and outlet temperatures. The designs of core internals are described. The after heat removal system is integrated in the prestressed concrete reactor pressure vessel, which is based on the principals of natural convection.Research work is being carried out, whereby the spherical fuel elements are coated with a layer of silicon carbide, to improve the corrosion resistance as well as the effectiveness of the fission products barrier.     

Models test on dynamic structure–structure interaction of nuclear power plant buildings

A reactor building of an NPP (nuclear power plant) is generally constructed closely adjacent to a turbine building and other buildings such as the auxiliary building, and in increasing numbers of NPPs, multiple plants are being planned and constructed closely on a single site. In these situations, adjacent buildings are considered to influence each other through the soil during earthquakes and to exhibit dynamic behaviour different from that of separate buildings, because those buildings in NPP are generally heavy and massive. The dynamic interaction between buildings during earthquake through the soil is termed here as ‘dynamic cross interaction (DCI)’. In order to comprehend DCI appropriately, forced vibration tests and earthquake observation are needed using closely constructed building models. Standing on this background, Nuclear Power Engineering Corporation (NUPEC) had planned the project to investigate the DCI effect in 1993 after the preceding SSI (soil–structure interaction) investigation project, ‘Model Tests on Embedment Effect of Reactor Building’. The project consists of field and laboratory tests. The field test is being carried out using three different building construction conditions, e.g. a single reactor building to be used for the comparison purposes as for a reference, two same reactor buildings used to evaluate pure DCI effects, and two different buildings, reactor and turbine building models to evaluate DCI effects under the actual plant conditions. Forced vibration tests and earthquake observations are planned in the field test. The laboratory test is planned to evaluate basic characteristics of the DCI effects using simple soil model made of silicon rubber and structure models made of aluminum. In this test, forced vibration tests and shaking table tests are planned. The project was started in April 1994 and will be completed in March 2002. This paper describes an outline and the summary of the current status of this project.     

Comparison of radiation damage in silicon induced by proton andneutron irradiation

The subject of radiation damage to Si detectors induced by 24-GeV/c protons and nuclear reactor neutrons has been studied. Detectors fabricated on single-crystal silicon enriched with various impurities have been tested. Significant differences in electrically active defects have been found between the various types of material. The results of the study suggest for the first time that the widely used nonionizing energy loss (NIEL) factors are insufficient for normalization of the electrically active damage in case of oxygen- and carbon-enriched silicon detectors. It has been found that a deliberate introduction of impurities into the semiconductor can affect the radiation hardness of silicon detectors     

Study of VVER fuel with dopant additives

This is a report of a study of the effect of alloy additives on the properties of fuel under conditions typical of water cooled reactors. The behavior of uranium oxide fuel with added mixtures of the oxides of aluminum, silicon, niobium, and iron during reactor irradiation of experimental fuel elements is investigated in the MIR research reactor. The feasibility of using aluminum-silicate additives for improving the operating characteristics of fuel pellets under reactor irradiation conditions is demonstrated. Translated From Atomnaya énergiya, Vol. 105, No. 4, Pp. 205–210, October, 2008.     

Fabrication of SiC–SiC composites for fuel cladding in advanced reactor designs

Both advanced fission reactor concepts and fusion energy systems demand materials that can survive extremely harsh operating environments having persistent high temperature and high neutron flux conditions. Silicon carbide fiber/silicon carbide matrix (SiC–SiC) composites have shown promise for these applications, which include fuel cladding and reactor structural components. However, the composite fabrication process is time consuming and the fabrication of complicated geometries can be difficult.In this work, SiC–SiC and carbon fiber–SiC composite samples were fabricated using chemical vapor infiltration (CVI), and the mechanical and thermal properties of samples with a range of densities and total infiltration times were characterized and compared. Both sample density and the reinforcing fiber material were found to have a very significant influence on the composite mechanical and thermal material properties. In particular, internal porosity is found to have a significant effect on the mechanical response, as can be observed in the crack propagation in low density samples. In order to better understand the densification of the composites, a computer model is being developed to simulate the diffusion of reactants through the fiber preform, and SiC deposition on the fiber surfaces. Preliminary modeling has been correlated with experimental results and shows promising results.     

Computational estimate of the corrosion resistance of fuel microelements with silicon carbide cladding in a water medium at supercritical pressure

A computational estimate of the corrosion resistance of a silicon carbide coating on spherical fuel microelements in a supercritical-pressure water medium at 350–700°C is made for the nominal operating regime of the core of a light-water nuclear reactor. A model of the thermal dissociation of water and dissolution of a silicon oxide film is used to estimate the corrosion resistance of silicon carbide. An expression describing the dependence of the corrosion depth of a silicon carbide layer on the temperature and pressure of the medium (concentration of the products of dissociation of water) and the operating time of the fuel microelements is obtained on the basis of the model developed. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 168–173, March, 2007. An erratum to this article is available at .     

On the use of silicon as thermal neutron filter

A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 μeV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.     

Design concept of a small nuclear reactor for large-diameter NTD-Si using a conventional PWR full-length fuel assembly

A design concept for a small nuclear reactor dedicated to large-diameter neutron transmutation doping silicon (NTD-Si) is proposed. Conventional PWR (Pressurized Water Reactor) full-length fuel assembly is used to assure stable and reliable supply of fuel. Criticality, neutron transportation, and core burn-up calculations are performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the proposed reactor can be critical over 18 years, and excess reactivity can be suppressed by a combination of Gd₂O₃ burnable poison and soluble boron. Preliminary steady-state single-channel thermal hydraulic analysis showed that heat removal from core is possible under 1 atm operating pressure. Si ingots up to 30 cm in diameter can be irradiated in the reactor irradiation channels, and the uniform irradiation condition can be achieved for a large-diameter Si ingot.     

Second phase formation at microvoids and gas atom entrapment at fusion reactor first wall surfaces

Results of cascade production and annealing studies for iron are projected to the currently considered fusion reactor first wall materials. Collision cascades initiated by primary knock-on atoms (PKA) with energies above 50 keV, in iron, separate into distinct subregions. Most PKA produced during fission reactor neutron irradiation have energies below 50 keV, but the energies of PKA produced by 14.1 MeV neutron irradiation commonly lie in the 50–500 keV range in iron and vanadium alloys. Computer experiment simulation indicates that, when present, carbon, silicon, and nitrogen, freed by irradiation from bulk precipitates, should tend to re-precipitate on the facets of microvoids which form in cascade subregions during short-term annealing.