分析束强流中子发生器中子比产额及靶寿命

将在分析束(D_1~-)轰击下,大面积高速旋转靶半寿命、中子比产额等指标与非分析束作了比较。中子比产额从1.1×10~(11)s~(-1)·mA~(-1)提高到1.6×10~(11)s~(-1)·mA~(-1),氚钛靶半寿命提高一个数量级,靶寿期由1mA·h/cm~2提高到6.9 mA·h/cm~2。     

NSRL衍射和散射线站硬X射线参数测量

描述了NSRL新建X射线衍射和散射光束线站的主要指标测试及其结果,包括X射线单色光斑的CCD成像(4mm×0.5mm)、波长标定和光子通量(0.4×10~9—2.1×10~9 photons·s~(-1)·(100mA)~(-1))的测量;已在该线站上应用同步辐射X射线成功地收集了5组700余幅磷脂酶A2和神经毒蛋白晶体的衍射数据,处理结果良好。     

内蒙古自治区环境天然贯穿辐射水平调查研究

本文报道了内蒙古自治区天然环境陆地γ辐射水平调查方案、质量保证措施及调查结果。在全区110万余km~2的面积上以25×25km网格布设了1018个测点。全区天然环境原野γ辐射剂量率(离地1m高处的空气吸收剂量率,下同)范围为(0.96—18.62)×10~(-8)Gy·h~(-1),按测点、面积和人口加杈均值分别为5.45、5.32和5.72×10~(-8)Gy·h~(-1);全区道路γ辐射剂量率范围为(1.07—26.08)×10~(-8),平均值为5.92×10~(-8)Gy·h~(-1);全区建筑物室内γ辐射剂量率范围为3.82—18.94)×10~(-8)Gy·h~(-1),按点、按人口加杈均值分别为9.26和9.49×10~(-8)Gy·h~(-1);宇宙射线电离成分所致空气吸收剂量率室外变化范围为(3.12—5.55)×10~(-8)Gy·h~(-1),按点、按人口加杈均值分别为3.93和3.85×10~(-1),室内变化范围为(2.60—4.66)×10~(-8)Gy·h~(-1),按点、按人口加杈均值分别为3.52和3.40×10~(-8)Gy·h~(-1);天然贯穿辐射剂量率室外变化范围为(8.31—11.26)×10~(-8)Gyh~(-1),按点、按人口加杈均值分别为9.38和9.57×10~(-8)Gyh~(-1),室内变化范围为(11.20—14.67)×10~(-8)Gyh~(-1),按点、按人口加杈均值分别为12.78和12.89×10~(-8)Gy·h~(-1);天然贯穿辐射所致人均年有效剂量当量及集体年有效剂量当量分别为0.84mSv和1     

协同萃取体系的研究——Ⅱ.DMHMP和HEHEHP协同萃取UO_2(NO_3)_2的机理

本文研究了甲基膦酸二(1-甲庚)酯(DMHMP)和2-乙基已基膦酸单(2-乙基己基)酯(HEHEHP)协同萃取硝酸钠酰。用斜率法测得协萃络合物的组成为UO_2(NO_3)(HA_2)·DMHMP。在20℃时(离子强度为0.5)协萃反应平衡常数1gβ_(12)=4.95。并求得了热力学函数ΔH、ΔZ和ΔS分别为—2.50×10~4J·mol~(-1)、-2.75×10~4J·mol~(-1)和8.61J·mol~(-1)·K~(-1)。依据锥角堆积模型讨论了协萃反应机理。     

~(134)Cs经由菜豆叶系向其他部分转移的研究

本文测定了~(134)Cs通过菜豆叶系向其他部分转移的转移因子TF,研究了污染溶液酸度、叶面~(134)Cs放射性活度、菜豆的生长阶段、Cs载体及Cs的化学形式对TF的影响。实验结果表明,对开始结荚期(荚果长1cm)的菜豆在pH5.5、~(134)Cs表面活度为3.1×10~2Bq/m~2测得的TF(荚果)为7.3×10~(-2)Bq·kg~(-1)(鲜重)/Bq·m~(-2),高于其他生长阶段的TF(荚果);K~+的存在使~(134)Cs的TF提高;Cs的化学形式影响~(134)Cs的转移,~(134)CsI更易被菜豆吸收。     

二(2-乙基己基)亚砜从硝酸介质中萃取铀(Ⅵ)的动力学研究

采用恒界面池法研究了二(2-乙基己基)亚砜(DEHSO)从稀硝酸介质中萃取铀(Ⅵ)的动力学行为。结果表明,在该条件下,DEHSO萃取铀(Ⅵ)的过程为扩散控制。硝酸根浓度是萃取过程重要动力学因子。萃取动力学方程为:R=K_fA[UO_2~+][H~+]~(-0.33)[NO_3~-]~(2·16)[DEH-SO]~(0.63)。萃取速率常数K_f为7.90×10~(-2)m·s~(-1)·l~(2.5)·mol~(-2.5)萃取表观活化能为0.9KJ·mol~(-1)。     

金及其伴生元素的动力堆中子活化分析

苏联利用动力堆中子活化分析岩石和矿石中的金及其伴生元素含量。被辐照的样品置于电离室备用孔道。该孔道位于活化区外石墨反射层,直径50mm,此处最大热中子流密度为10~9cm~(-2)·s~(-1),温度80—90°C。样品能在这里长时间(几昼夜)被辐照。利用中子活化分析方法测定金基于放射性俘获的~(197)Au (n,γ)~(193)Au (T1/2=65h)反应和记录的41~2KeVγ辐射能。样品破碎到200目,其含金范围为2×(10~(-5)—10~(-3))％。为定量测量利用标准样P3C-1和P3C-2,其含金量分别为5.21×10~(-3)和9.4×10~(-5)％。被辐照的样品放入聚乙烯安瓿中(直径和高分别为40和20mm),样品质量为15—20g。不仅用全谱中子流还要用镉上中子进行辐照(在后一种情况下安瓿被厚度为O.5mm的镉屏围绕。)。辐照时间与金含量多     

HFETR十年运行对公众辐射危害分析

根据高通量工程试验堆(HFETR)十年运行期间放射性流出物的排放量、剂量学模式与参数、以及危害估算模式,对堆址80km 范围内公众的有效剂量和辐射危害进行了计算和分析。辐射危害的最大个人危险1.34×10~(-10)—2.24×10~(-7),为可忽略的危害水平,且低于我国不同产业非辐射死亡的个人危险(10~(-5)—10~(-3));而群体危险8.45×10~(-7)—2.88×10~(-3)人·a~(-1),与堆址 80km 范围非辐射致死性癌症死亡人数(3.44—3.78)×10~3人·a~(-1)相比,也是可以忽略的。     

中子四圆衍射仪

本文介绍了我国第一台四圆衍射仪的用途、结构和主要性能。这台谱仪是用于分析单晶和磁结构,中子的入射波长固定在1.69×10~(-10)m,谱仪的能量分辨率为4.7%,在样品位置上的最大中子注量为10~6n·cm~(-2)·s~(-1),最后与国外同类谱仪的分辨率进行了比较。     

在存在十六烷基三甲基氯化铵的条件下铬蓝G高灵敏度分光光度法测定铀

铀铬蓝G络合物和十六烷基三甲基氯化铵在水溶液中形成的三元络合物pH,值在4.0—6.2间波长在648 nm处具有最大吸光度。在0.2-2mg/l范围中的铀遵循比尔定律,mol吸光系数达1.32×10~5l·mol~(-1)·cm~(-1)。Sandell灵敏度为1.8×10~(-3) ug铀/cm~2。当使用钙-1,2-环己烷-二胺四乙酸作掩蔽剂时仅铝和铍有干扰。     

Attachment of Membrane Separator for Removal of Radon to Ionization Chamber Installed for Tritium Stack Monitor

The effect of background counts induced by α-particles mainly from Rn in air was removed by the attachment of a water-permselective membrane separator to a tritium stack monitor equipped previously. Water vapor, after permeating the membrane selectively, was carried by N₂ gas into an ionization chamber where the activity of tritium was measured. The consumption of N₂ gas for carrier was reduced by recycling the gas through dehumidification processes; (1) condensation by pressure, (2) condensation by refrigeration and (3) adsorption (by adsorbents), so that there is no added daily work arising from the attachment for maintenance of the tritium monitor.     

Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm³. Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10⁻¹³–1.8 × 10⁻¹³ m²/s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.     

用SSNTD法测量装修后室内氡浓度

采用固体核径迹探测器（SSNTD）,对装修后的房屋进行室内氡浓度调查.结果表明,用SSNTD测得的装修后居室内的氡浓度比以往用瞬时法测得的、未经装修的增加一倍.对有关影响因素和防氡措施进行了讨论.     

束流发射度的测量

在束流输运系统中,快速而精确地测定束流的发射度,对于提高束流传输效率、调节并保证束流达到预期的场所,是十分重要的。通常在测定束流发射度时采用移动缝隙法或电磁偏转方法。这些方法不仅需要附加一些较复杂的装置,而且测量时费时费力。特别是这些方法都必须采用缝隙装置,在测量时截留了大部分束流流强,因而必须强迫冷却,而在测量后,这些缝隙装置又带有很强的放射性。除了以上这些问题以外,由于束流在缝隙上的散射,又使测量结果的误差增大。Fermi实验室的E.R.Gray、CERN的G.Baribaud和C.Metzger曾用测量束流剖面的方法来测定束流的发射度。他们测量一段漂移空间内三点的     

基岩裂隙水渗流数值模拟研究综述

裂隙岩体中的渗流规律与传统多孔介质渗流规律存在本质的区别,裂隙岩体渗流具有不均匀性、各向异性和非连续性等特点,如何更好地刻画裂隙水渗流的这种特点是当前裂隙水数值模拟的热点与难点。介绍了裂隙岩体地下水渗流特性,对目前裂隙水渗流模型及其优缺点进行比较、分析,同时对主流数值计算方法和数值模拟软件进行概括介绍。     

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the ²¹⁰Po and ²⁰³Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.     

A Validation of ATR LOCA Thermal-hydraulic Code with a Statistical Approach

When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method.     

防氚高分子复合材料的制备及其性能

利用氯化聚乙烯、聚偏二氯乙烯(Polyvinylidene chloride,PVDC)、高密度聚乙烯等高分子材料复合防氚材料。采用渗透实验测量氚气渗透率;通过测量断裂强力和剥离力测量力学性能;通过测量浸泡化学试剂前后的溶失率、断裂强力下降率、剥离力下降率等表征材料的耐化学性能。结果表明,氚气在防氚高分子复合材料中的渗透率为3.44×10-10 m3/(m2?s);与PVC相比,氚气渗透率降低约70倍。PVDC薄膜复合后断裂强力提高约20倍,经纬向断裂强力为721 N。防氚高分子复合材料浸泡5种化学试剂后,最大溶失率为0.29%、断裂强力最大下降率为5.1%,其力学性能、耐化学性能均满足《防护服装酸碱类化学品防护服》(GB24540-2009)等相关标准的要求。     

Evaluation of Exposure due to Terrestrial Gamma-Radiation by Response Matrix Method

This paper describes an analytical method for determining terrestrial γ-radiation through conversion of the pulse-height distribution obtained with NaI(T1) scintillator into a true incident photon spectrum by means of a response matrix method. The resulting true spectrum also serves as bisis for estimating the exposure rate from the same radiation.

The value of exposure rate estimated by this method is compared with that measured with an ionization chamber, revealing good agreement between the results.     

Verification of Source Term Estimation Method against Measured Data for Decommissioning Waste from a CANDU Reactor

The method for the establishment of an equilibrium core model proposed in the previous paper and the source term calculation method proposed in this paper for the characterization of decommissioning waste were verified by comparing the nuclide inventory estimated by MCNP/ORIGEN2 simulations with the measured nuclide inventory according to a chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. At first, the time-average pseudoequilibrium full-core model of Wolsong Unit 1 was developed on the basis of the previously proposed modeling method for the activation of in-core and ex-core structural components. Then, the application level of the neutron flux and cross section in the radionuclide buildup calculation were compromised. Fourteen major actinides and fission products were considered to represent the irradiated fuel condition, and a geometry simplification was also introduced in the burned full-core model for MCNP simulation. The assumption of a constant neutron flux and capture cross section as a function of the irradiation time was applied in the radionuclide buildup calculation in ORIGEN2. As a result, the values estimated from the analysis system agreed with the measured data within a difference range of 30%. Therefore, it was found that the MCNP/ORIGEN system and source term characterization method proposed can be viable to estimate the source terms of the decommissioning waste from a CANDU reactor.