中子辐照后6Li-Al合金靶片中氚的测量方法研究

建立了流气式系统捕集和化学转移法测量^6LiAl合金靶片在中子辐照过程中的渗漏氚和辐照后靶片中总氚量的方法。渗漏氚的捕集方法是：在流气式系统中，用含少量氢的惰性载气将渗漏氚载带出来，经高温催化氧化后被乙二醇鼓泡器捕集；靶片中的氚转移法是用NaOH溶液将合金靶片溶解，气相中的氚采用渗漏氚的捕集方法捕集，液相中的氚则蒸馏到馏分中。最后用液体闪烁计数器分别测量乙二醇鼓泡器和馏分中的氚量。测量结果与理论氚产量基本相符。     

Hardening and Recovery of Neutron-Irradiated Fe-Cu-C Alloy

Internal friction measurements were performed on neutron-irradiated and annealed Fe-0.2w/0Cu-0.0066w/0C alloy, in an attempt to gain information on the role of interstitial C atoms in radiation effects. It was revealed that C atoms are trapped by defects produced by neutron-irradiation, and released upon annealing at about 400°C. The radiation hardening induced in the neutron-irradiated Fe-Cu-C alloy recovers by annealing in at least three steps of 150°–250°C, 350°–450°C and above 550°C. The second step corresponds to the release of C atoms from traps.     

Snoek Peak of Neutron-Irradiated Fe-3.48% Ni Alloy

The Snoek peak of a neutron-irradiated Fe-3.48% Ni alloy was measured to determine the resolution of interstitial impurity atoms.

The internal friction was measured after isothermal annealing on specimens irradiated at 75° ± 10°C to a dose of ～3.6 × 10¹⁹ nvt.

The change of the Snoek peak height, which corresponds to the amount of nitrogen in solution, was analyzed upon assuming the applicability of the rate equation.

As a result of calculation, the order of the reaction was found to be about 2.6, and the activation energy for resolution of nitrogen atoms about 2.9 eV.     

Preliminary Study on Production of Xenon-133 from Neutron-Irradiated Uranium Metal and Oxides by Oxidation

A method was developed for recovering the fission product ¹³³Xe from several kinds of reactor-irradiated U targets, including Al-U alloy, metallic U, and uranium oxides.

In order to observe the release of ¹³³Xe from U compounds at high temperatures, irradiated targets were heated at 500°~1,000°C in vacuum or under gas flow. The liberated ¹³³Xe was trapped on charcoal beds, and the release rate of ¹³³Xe from the compounds was determined by measuring the activity accumulating in the traps.

More than 90% of the ¹³³Xe was liberated from the alloy upon melting and from metal and uranium oxide upon oxidation.

The isolated Xe was purified by a system embodying cold traps and cryogenic distillation.

The final products were sealed in ampoules. They proved to possess radiochemical purity exceeding 99.9%, and less than 1 μ/ampoule (1 ml) of non-radioactive gaseous contaminants.     

利用伴随α粒子能谱分析固体氚靶中氚浓度深度分布

在目前的氘氚中子发生器源中子分析过程中,固体氚靶中氚浓度深度分布信息的缺失是普遍遇到的问题。为解决此问题,本文建立了利用伴随粒子能谱反演氚浓度深度分布的模型,采用来自氚钛靶的α实验能谱作为模型测试对象,通过该模型获得了氚钛靶中氚浓度深度分布的数据。结果表明,氚浓度随氚钛靶深度的增加呈双峰趋势,两峰之间的氚浓度波谷位于靶中0.94 μm处,该深度正是入射氘粒子的射程极限。所得的氚浓度深度分布趋势与其他实验方法测量结果相符,表明该模型能为氘氚中子发生器的源中子分析提供即时的氚浓度深度分布信息。     

Estimation of the Tritium Behavior in the Pebble Type Gas Cooled Reactor for Hydrogen Production

In Korea, a nuclear hydrogen program has been established to develop and demonstrate mass production system for hydrogen generation. The objective of this study is to establish the evaluation procedure for predicting the tritium behavior in the 300 MWth Pebble type gas cooled reactor which is the one of the candidate reactors for nuclear hydrogen development and demonstration plant. The tritium generated by the fission reaction can be leaked to the helium coolant from the coated ceramic particles and fuel elements. The annual total release rate of the tritium is estimated as 0.47% from the fuel kernel to the helium coolant by the numerical method. Tritium attributed by ⁶Li existing as impurities in the reflector can be released to the helium coolant by the diffusion process and the total annual release rate of the tritium is estimated as 5.3% through the reflector to the helium coolant. Based on the Siverts' law, tritium permeation from the primary coolant to the hydrogen production system is also evaluated and the result is calculated as 76?0.23 Bq/g-H₂ with respect to the PRF (Permeation Reduction Factor= 10?1000) in case of the normal operation of the 300 MWth Pebble type reactor.     

Temperature and Concentration Dependence of Tritium Release Behavior from Solidified LiF-PbF2 Pellet Irradiated by Thermal Neutron

Thin pellets of the LiF-PbF₂ system with no porosity were irradiated by thermal neutron. Then release experiment of tritium produced in the pellets was carried out at some constant temperatures. Thus diffusion coefficients of tritium in the LiF-PbF₂ system were estimated to obtain an activation energy of diffusion, which was found to be 0.51 eV at LiF mole fraction of 0.467. It was also observed that decrease of LiF concentration in the pellet brings increase of diffusion coefficient of tritium.     

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

CFETR水冷陶瓷增殖剂包层模块产氚率实验验证

水冷陶瓷增殖剂（WCCB）包层作为中国聚变工程试验堆（CFETR）候选包层之一,承担着氚增殖、核热提取、屏蔽等重要涉核功能,其中子学设计的可靠性直接影响CFETR氚自持目标的实现。为验证中子学设计工具,即MCNP和FNEDL3.0数据库,在WCCB包层中子学设计中的可靠性,基于研制出的WCCB包层模块,在DT中子环境下开展中子学实验,对以产氚率（TPR）为代表的中子学参数进行了模拟值（C）和实验值（E）对比分析。结果表明,模块中轴线位置处TPR的C/E为0.97?1.08,而模块边缘位置处TPR的C/E为0.65?0.82;模块钛酸锂层边缘区197Au（n,γ）198Au反应率的C/E为0.72?0.90,表明模块边缘区存在非期望的散射中子,导致该区TPR模拟值和实验值偏离较大。     

Application of a 6LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement in D-T Neutron Fields

⁶LiF small neutron detectors with an optical fiber have been used to measure ⁶Li(n,α)T reaction rate distributions at thermal research reactors and accelerator facilities. In the present study, we developed an experimental method for the measurement of tritium production rate (TPR) of ⁶Li using this small detector in deuterium-tritium (D-T) neutron fields. Reaction rate measurements with the detector were conducted in the D-T neutron fields at the Fusion Neutronics Source (FNS) facility. From the results, we determined that this detector can be used to measure the TPR distribution in soft neutron spectrum fields such as in a Be assembly. It is difficult to obtain ⁶Li(n,α)T reaction rate separately in hard neutron spectrum fields such as in a Li₂O assembly, because many kinds of charged particle production reactions need to be taken into consideration. However, a time-dependent reaction rate measurement method combined with the ⁶LiF detector and the ZnS detector is effective to separate the ⁶Li(n,α)T reaction from other reactions even in a hard spectrum field, and it can be applied to the measurement of the TPR distribution accurately.     

Application of Alpha-Track Method to Tritium Production Rate Measurements

The α-track method with a solid state nuclear track detector was used for the measurement of tritium production rates (TPR). Cellulose nitrate track detectors (Kodak LR-115) with Li-containing radiators were placed in a simulated fusion blanket of a Li₂O slab assembly with a Be layer and irradiated by D-T neutrons. After chemical etching, the etched-through tracks in the detectors were counted. Lithium metal was the most suitable radiator to distinguish tracks formed by α-particles of the LI(n, α)T reaction from the background. The experimental results agreed well with the calculations. This shows that the indirect α-track method is reliable in estimating TPR distribution.     

Separation of Plutonium-238 from Fission Products by Solvent Extraction Using HDEHP

A procedure for separating ²³⁸Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO₃ solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 10² for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5 mg of ²³⁸Pu was obtained with an efficiency exceeding 95%.     

Production of high-purity medical radio isotope 64Cu with accelerator-based neutrons generated with 9 and 12 MeV deuterons

We conducted a feasibility study for producing a high-purity medical radioisotope ⁶⁴Cu from natural zinc with accelerator-based neutrons. ⁶⁴Cu isotopes were produced via the ⁶⁴Zn(n,p) reaction. The accelerator-based neutrons were generated via the C(d,n) reaction using low-energy deuterons of 9 and 12 MeV on a 1-mm-thick carbon target. First, the production purity was estimated using the evaluated nuclear data library JENDL-4.0 and our previously measured thick target neutron yield. We found that even when natural zinc was used as the starting material, significantly high-purity ⁶⁴Cu could be obtained. Next, irradiation experiments for producing ⁶⁴Cu using natural zinc were conducted at Kyushu University Tandem Laboratory, with the amounts of ⁶⁴Cu isotopes and other gamma-emission nuclides measured by a high-purity germanium detector. As a result, high-purity ⁶⁴Cu isotopes of 1.11(49) × 10⁰ and 3.70 (17) × 10⁰ Bq/g/μC were produced with incident deuteron energies of 9 and 12 MeV, respectively.     

Preparation of Uranium-233 with Low Uranium-232 Content

Abstract

Thorium dioxide, previously freed of uranium and protactinium contamination by means of ion-exchange, was irradiated for 70.5 and 140hr respectively in the core and graphite thermal column positions of the KUR (Kyoto University Reactor). After being cooled for more than 1½yr, the ²³³U produced was chemically purified by ion exchange, and the resulting ²³³U specimens were electro-deposited on a stainless steel plate for submission to α-spectrometric measurement with a silicon surface barrier detector. From the α-spectra thus obtained, the ²³²U to ²³³U atomic ratios of the specimens were evaluated and compared with the calculated values. This revealed that the rate of formation of ²³²U depended largely on the ²³²Th(n, 2n)²³¹Th reaction by fast neutrons, and consequently, on the neutron energy distribution prevailing at the position of irradiation. The purest ²³³U sample obtained from the thorium target, irradiated with pure thermal neutrons in the graphite thermal column, yielded a ²³²U to ²³³U atomic ratio of 3.02×10^?9, which corresponded to 6.62×10^?6 in α-activity ratio.