新型自补偿γ量热计的刻度及测量

用电加热方式对新型自补偿γ量热计算计进行了刻度，测量了低功率反应堆的Ｋ１１孔道和７号孔道内不锈钢的γ释热率，在相同γ释热条件下，使γ释热率的测量误差由没刻度时的±８．０％减小到±１．０％。     

蒙特卡罗方法用于HFETR堆芯γ释热的可行性研究

采用蒙特卡罗方法对高通量工程试验堆(HFETR)堆芯内的γ释热进行了计算,并将计算结果与实测值进行了比较.结果表明:用蒙特卡罗方法计算HFETR堆的γ释热率是可行的,具有满意的计算精度.因此,在实际工程中可采用蒙特卡罗方法来计算HFETR及堆芯内任意位置的γ释热.     

瞬发和缓发γ射线对堆内构件释热率影响的研究

为提高核电设计中反应堆堆内构件释热率计算的准确性,本文在原来MCNP外中子源模型计算方法的基础上,计算分析瞬发裂变γ对堆内构件释热率的贡献。计算结果显示,考虑瞬发裂变γ使得堆内构件的释热率增加9%～38%,离堆芯越近的堆内构件的增加值越大。另外,分析认为缓发γ对堆内构件释热率的贡献与瞬发裂变γ相当。因而反应堆堆内构件释热率计算中除了考虑中子及中子俘获所生γ的贡献,还应该考虑瞬发裂变γ和缓发γ的贡献。     

HFETR堆芯及φ63辐照孔道γ释热研究

采用蒙特卡罗方法对高通量工程试验堆堆芯内的γ释热分布进行了详细的计算分析和研究,并对堆芯内φ63辐照孔道在不同状态下的γ释热分布进行了详细准确的研究。计算结果表明：堆芯内γ释热功率为3．29MW,燃料元件功率62．8MW,分别占堆芯总功率（70MW）的4．7％和89．7％;3个φ63辐照孔道内单位质量介质γ释热率分别为：G7孔道为3．016W／g,P12孔道为3．733W／g,P15孔道为3．627W／g。本研究为HFETR堆芯及各组件的热工安全分析提供了必要的数据,保证了反应堆的安全运行,节约了反应堆的运行成本,提高了反应堆的经济性。     

反应堆中心孔道辐照材料的中子与γ释热研究

反应堆辐照材料上中子与γ的释热率是该材料在堆中热工计算的重要输入参数.本文基于蒙特卡罗粒子输运程序(MCNP),计算了某堆首炉高热中子堆芯布置下,L12中心孔道中不同材料(水、T6061铝、单晶硅、不锈钢、锆合金)轴向的中子、γ释热率分布.计算结果表明,活性区轴向高度为0～1000 mm,中子与γ在材料上的最大释热率点...     

HFETR堆芯γ释热率径向分布的简易求法

本文以实验为基础，讨论了ＨＦＥＴＲ堆芯材料γ释热径向分，并且描述了材料γ释热率与反应堆热功率，热中子注量率之间的关系。同时，给出了它们之间的关系表达式。     

研究堆内材料释热率测量装置研制

为了探究材料释热率在研究堆孔道内的轴向分布规律,以高通量工程试验堆（HFETR）G7孔道为例,设计一种材料释热率测量装置。通过数值模拟方法得到释热率测量装置及试验段在载荷作用下的应变分布云图,采用物理计算得到量热计校对桥和测量桥的温度参数,并利用本装置在G7孔道开展释热率测量试验。结果表明,该装置整体结构满足强度要求,试验段量热计之间需加装保护管;计算得出样品、校对桥和测量桥的温度低于材料熔点,装置满足热工要求;试验测得的释热率值随堆功率变化规律性强,且不同材料在不同能量等级的γ射线环境下,对γ的吸收性是有区别的。因此,本装置可以作为HFETR释热率测量工具,为确定不同材料在堆内释热率分布情况提供保障。      

单棒式自补偿γ量热计及其应用

本文介绍了一种新型单棒式自补偿γ量热计(以下简称新型γ量热计)的设计及其在游泳池堆(SPR-300)和高通量工程试验反应堆(HFETR)上的应用。     

HFETR实验孔道内γ释热率的测定

本文介绍了测定HFETR实验孔道内γ释热率的两种方法——测温法和真空升温法,测量误差为±8.4％。     

γ刻度室的散射测量

本文介绍了铅准直器 γ 刻度室的散射实验研究。用几种方法测量了不同使用距离上的相对散射系数,其结果小于5%。测量了 γ 散射能量,并探讨了准直器的散射,从而了解了辐射场的性质。文中还提出了值得注意的几个问题。     

Gamma-ray attenuation technique for measuring void fraction in horizontal gas-liquid two-phase flow

The measurement of void fraction is of importance to the oil industry and chemical industry. In this article, the principle and mathematical method of determining the void fraction of horizontal gas-liquid flow by using a single-energy γ-ray system is described. The γ-ray source is the radioactive isotope of 241Am with γ-ray energy of 59.5 keV. The time-averaged value of the void fraction in a 50.0-mm i.d. transparent horizontal pipeline is measured under various combinations of the liquid flow and gas flow. It is found that increasing the gas flow rate at a fixed liquid flow rate would increase the void fraction. Test data are compared with the predictions of the correlations and a good agreement is found. The result shows that the designed γ-ray system can be used for measuring the void fraction in a horizontal gas-liquid two-phase flow with high accuracy.     

一种因参考光源强度变化导致的热释光剂量测量结果偏差的修正方法

在热释光剂量测量中，参考光源强度是影响测量结果的重要因素之一。对参考光源强度的调整往往通过调整测量仪器的高压来实现，但调高高压参数会导致测量仪器暗电流增加，进而导致测量值偏差增大。频繁的调整高压参数，也可能会使仪器的稳定性变差。针对参考光源强度变化导致的测量偏差，提出利用测量时的参考光源强度值与校准测量时参考光源强度值的变化求出相应的修正系数，进行测量结果修正。该方法经标准值样片测量检验，其测量值更准确、误差更小，与标准辐照值的偏差小于5%，表明该方法可行，可在实际工作中应用。     

Current status of radiation therapy for prostate cancer

Radiotherapy for the treatment of prostate cancer has been extensively explored in the past. Along with the comprehensive understanding of the biology of prostate cancer and rapid advances in terms of technology, the out- come of treatment for the patients with prostate cancer has improved. The authors review radiotherapy as the primary treatment for the disease, with particular emphasis on the technological advances from both the radiobiological and radiophysics aspects. Nonconventional fractionated irradiation like hyper- or hypo-fractionation has been imple- mented in the clinic, the final results still need to be confirmed in the future. Technological advances like IMRT, IGRT, in the last two decades have significantly improved the delivery of external radiotherapy to the prostate. This has re- sulted in an overall increase in the total dose that can be safely delivered to the prostate, which has led to modest im- provements in the biochemical outcome. However, establishing the standard therapy for prostate cancer remains con- troversial. It is hoped that the next decades will bring continued advances in the development of biologicals that will further improve current clinical outcomes.     

Thermoelastic analyses of cladding on beam transient of lead–bismuth cooled accelerator-driven system

The thermoelastic analyses of cladding for lead–bismuth cooled accelerator-driven system (ADS) are conducted for the beam transients. The beam transients are considered to be caused by the abnormal behavior of the accelerator and are peculiar to ADS. The program of the thermoelastic analyses is developed for the evaluation of the stresses of the cladding. This program is intended to analyze a fuel pin of a cylindrical model, and solves the thermoelastic problem by the use of the finite-element-method. The beam transients are analyzed by employing the ADS dynamic calculation code and the program of the thermoelastic analyses for the ADS designed by Japan Atomic Energy Agency. As a result, the transformation of the beam shape does not cause the cladding failure. However, the ductile failure is caused by the beam incident position change in several seconds. These results are also compared with those of the creep analyses conducted in the previous study, and both the creep and the ductile failure are revealed to be caused by the beam incident position change. Consequently, the beam incident position change is concluded to have a high risk of cladding failure.     

Adsorption of iodine on graphite in High Temperature Gas-Cooled Reactor systems: A review

Adsorption of iodine on graphite is of great interest for operation and safety of high temperature nuclear reactors. Graphite can adsorb significant amounts of iodine and retain it for a long period of time. Significant amount of work on this subject has been done in the past. Various types of adsorption apparatus have been designed and data were collected. The types of graphite used in past studies are not available anymore, and as a consequence the data are not applicable for the new type of commercial nuclear grade graphites. However, the past experimental systems, data, and their analysis are useful to design a better experimental system, collect more accurate data, and, provide better understanding of the adsorption process and data. In addition the existing data can be used to generate a framework to understand the types of adsorption processes taking place. In this work, we have conducted an exhaustive literature review and further analyzed the data. Four adsorption isotherms; the Langmuir, the Freundlich, and the two isotherms proposed in the International Atomic Energy Agency (IAEA) Tecdoc-978 were used to correlate the available equilibrium adsorption data. For most of the data, the simple Langmuir and the Freundlich isotherms provided a reasonable fit of the data. The Polyani's potential theory was also used to check the consistency of the data and as indicated by the theory, most of the data set provided a single characteristic curve. The isosteric heats of adsorption calculated using the literature data suggested that iodine-adsorption on graphite could be a chemisorption process.     

Gamma radiation levels in the ambient environment of the QNPP Base

Monitoring results of gamma dose rate level in 1992～2004 in the ambient environment of the Qinshan Nuclear Power Plants(QNPP)Base,the northeast of Zhejiang Province,are reported in this paper.It is shown that the gamma dose rate of five monitoring sites of 2.5 km to QNPP Base is 84～113 nGy/h,with an average of 96 nGy/h in the 13 years.The average value is close to the background level of 93 nGy/h prior to operation of the QNPP Base,and is lower than the monitoring result of 101 nGy/h at the reference site in Hangzhou City.Within 50 km from the QNPP Base,the cumulative dose rate of the thermoluminescent dosimeter(TLD)is 90 nGy/h,which is lower than the back- ground level of 111 nGy/h.     

Radiation synthesis and characterization of polyacrylic acid hydrogels

The pH-sensitive polyacrylic acid （PAA） hydrogels were synthesized by gamma-ray irradiation at an ambient temperature. The influences of dose, monomer concentration, cross-linking agent content, pH, and ionic strength on the swelling ratio （SR） of the PAA hydrogels were investigated in detail. The results show that the SR of the hydrogel decreases with an increase in the dose, monomer concentration, and cross-linking agent content. In alkaline solution, the SR of the hydrogels is much higher than that in acid solution. Also, the ionic strength can influence the SR of the hydrogels. The more the concentration, the lower the SR.     

Comparative Studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 Data Libraries on the Monte Carlo Continuous Energy Modeling of the Gas Turbine-Modular Helium Reactor Operating with Thorium Fuels

One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile ²³²Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of ²³⁹Pu, ²³³U and ²³⁵U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B.     

Study on Risk-Informed In-Service Inspection for BWR Piping

Risk-informed in-service inspection for piping was studied for a BWR plant. Piping segment rupture probabilities were determined by Bayesian transform from piping failure events in the database of the OECD-NEA Piping Failure Data Exchange project. Based on the methodology of the Westinghouse Owners Group, core damage frequency induced by each segment rupture was determined by the use of a surrogate component in the PSA model. Nondestructive examinations were added to leak examinations for segments of the resultant high risk significance. The changes from current examinations gave around 29% reduction of segments subject to both leak and nondestructive examinations within the total segments. Deterministic insights and engineering judgments on top of risk significance should be applied to obtain the final decision of inspection methods. An extent-of-examination was studied by the adoption of the Perdue-Abramson model in the Westinghouse Owners Group methodology. The necessary leak frequency of a crack in a segment was calculated by the probabilistic fracture mechanics code PRAISE. Two segments of high risk significance showed lower or slightly higher extents-of-examination, respectively, than the current extent-of-examination. To contribute to the enhancement of the scientific rationality of piping inspections, technical knowledge was accumulated.     

Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea

Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.