环境大气氚的测定技术研究

文章分析环境大气中氚的来源，并描述环境大气氚浓度测量系统的组成与测量原理。研究了累积氚取样装置中催化炉的催化温度和气体流速对催化效率的影响、冷阱冷却温度对氚回收率的影响，刻度了氚电解浓集装置的回收率，在此基础上测定了某核设施周围环境大气氚浓度。结果表明：在催化温度为350℃、催化效率为98%、电解浓集装置平均回收率为54%的实验条件下，本测试方法可高灵敏度地测定特定范围内外环境低水平大气氚浓度变化，探测下限可达0.1Bq/m³。     

充氚不锈钢微观组织及断裂特征

采用力学拉伸实验测定充氚不锈钢的断裂强度值,采用拉伸断口进行SEM观察和正电子湮灭(PAT)分析,采用TEM动态拉伸实验观察和记录材料在微观断裂过程中的行为,通过对比分析氚对不锈钢断裂过程的影响。结果表明,高温充氚后,室温存放2a,样品中氚衰变产生的氦累积已达约30ppm;氚、氦使样品断裂强度降低,内部缺陷增多,正电子寿命变长。TEM观察未发现明显的氦泡组织;动态拉伸实验表明,充氚促进裂纹尖端位错的发射和增殖;HR-1、HR-2不锈钢微观断裂过程相似,可表述为氚致微裂纹的形核-形成微空洞-微空洞长大-空洞连接(断裂)。氚、氦使无位错区减小甚至消失。     

充氚不锈钢中氦泡长大行为的加热效应

以室温贮存经历的充氚不锈钢试样为研究对象,计算了充氚不锈钢中氚、氦浓度的深度分布,利用透射电镜观察了充氚不锈钢在加热过程中氦泡的演化行为。结果表明:在氚压0.131MPa、780℃充氚8h后,不锈钢中氚在深度方向分布均匀,平均浓度为110μL/L;在空气室温环境下放置6a后,不锈钢中氚衰变的氦浓度在深度方向分布均匀,平均浓度为60μL/L;对充氚不锈钢加热处理后,在550℃/1h时效即可观察到氦泡;在950℃/1h和1050℃/1h时效时,氦泡明显长大,大的可达100nm,小的可达30nm,在晶界、晶内和位错处均可见氦泡。     

A thermal desorption study of the surface interaction between tritium and type 316 stainless steel

The thermal desorption of tritium from a type 316 stainless steel exposed to gaseous tritium (HT-gas) was studied. In the desorption spectra of tritium, four distinct peaks appeared at about 430 (peak HT-1), 540 (peak HTO), 750 (peak HT-II) and 970 K (peak HT-III). The activation energies for the peaks HT-I, HTO and HT-II were determined as 20 ± 4, 27 ±7 and 55 ±9 kJ/mol, respectively. It is presumed that the tritium leading to the peak HTO forms tritoxyl ion (OT^?) by combining with surface O^2? ion or by the isotopic exchange with the protium of surface OH^? ion on the sorption process. The OT ion combines with the OH^? ion or the dissolved protium into HTO at the desorption process. The tritium leading to both peaks HT-II and HT-III is sorbed and dissociates into atomic species which diffuse into the bulk, while that leading to the peak HT-I is very weakly held on the topmost surface of stainless steel in the form of a molecule or HT⁺ ion.     

不锈钢和铜吸氚及其干法去污

为了对不锈钢和无氧铜吸氚后氚在其内部的分布情况及除氚去污方法进行研究,对模拟吸氚及加热去污后的样品进行了酸蚀刻以考察氚在金属层中的分布情况;单独加热或加热结合通入空气、O3和紫外线(UV)进行去污,考察不同去污方式的去污效果。结果表明:金属在表层1μm内吸附了大量的氚,约占总量的42%;加热到500℃及联合去污不锈钢的最佳去污因子达到286,铜为150,通入气体在中温条件下对金属去污最有效,加热是金属去污最有效方式;氚热解吸形态分析表明氚污染不锈钢有4种吸附态。     

Permeation Analysis of Tritium Through the Vessel Wall of a Titanium Hydride Storage Container

Abstract

A preliminary design for a stainless steel vessel for the long-term storage of hydrogen isotopes has been proposed. The immobilised hydrogen, as a titanium hydride, could be stored in a stainless steel vessel for this application. The vessel, as a primary package, is designed to form titanium hydride and to contain the hydrogen isotopes and helium-3 produced from the decay of tritium. In order to predict the possibility of contamination and the deterioration of the mechanical properties, a numerical diffusion analysis calculation of the hydrogen isotopes and helium inside the stainless steel vessel was carried out. Numerical results showed that a negligible amount of tritium would be released by permeation through a 0.7 cm thick vessel wall at normal conditions over the entire period of the storage. When the vessel is heated up to a temperature of 600^°C for the routine conditions of activation or exothermic hydriding, tritium loss or contamination would be of little concern. However, if the vessel were exposed to fire conditions with a temperature of 800^°C, permeation of hydrogen through the vessel wall would result in a serious increase in the amount of tritium escaping, in a very short time.     

金属样品残余氚测量方法

为定量评价氚在结构材料和老龄贮氢材料内部的滞留量,用化学蚀刻法测定不锈钢内部氚浓度大小、分布情况及同位素交换后老龄贮氚铀床的氚滞留量。结果表明,贮氚13年的不锈钢样品中氚主要存在于样品内表面由表及里的120μm范围内,样品蚀刻深度110.6μm范围内,不锈钢的平均氚滞留量～9.37×10-4mmol/g,贮氚铀粉平均氚滞留量～4.16×10-5mmol/g。该方法对测量金属中微量氚有较高灵敏度,可检测金属中残余氚的滞留量。     

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100-600 °C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300-600 °C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150 °C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.     

Impact of nuclear irradiation on helium bubble nucleation at interfaces in liquid metals coupled to permeation through stainless steels

The impact of nucleating gas bubbles in the form of a dispersed gas phase on hydrogen isotope permeation at interfaces between liquid metals, like LLE, and structural materials, like stainless steel, has been studied. Liquid metal to structural material interfaces involving surfaces, may lower the nucleation barrier promoting bubble nucleation at active sites. Hence, hydrogen isotope absorption into gas bubbles modelling and control at interfaces may have a capital importance regarding design, operation and safety.He bubbles as a permeation barrier principle is analysed showing a significant impact on hydrogen isotope permeation, which may have a significant effect on liquid metal systems, e.g., tritium extraction systems. Liquid metals like LLE under nuclear irradiation in, e.g., breeding blankets of a nuclear fusion reactor would generate tritium which is to be extracted and recirculated as fuel. At the same time that tritium is bred, helium is also generated and may precipitate in the form of nano bubbles.Phenomena modelling is exposed and implemented in openFROM^® CFD tool for 0D to 3D simulations. Results for a 1D case show the impact of a He dispersed phase of nano bubbles on hydrogen isotopes permeation at an interface. In addition, a simple permeator simulation, consisting in a straight 3D pipe is exposed showing the effect of a He dispersed gas phase on hydrogen isotope permeation through different stainless steels. Results show the permeation reduction as a function of the interface area covered by He bubbles.Our work highlights the effect of gas bubble nucleation at interfaces and the importance of controlling these phenomena in nuclear technology applications.     

Isotope exchange reaction between tritiated water and hydrogen on SiC

SiC has been considered as a primary candidate material for a first wall component in future fusion reactor because it has been claimed that SiC has excellent high-temperature properties, good chemical stability and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, tritium trapping capacity on the surface of SiC was experimentally obtained at the temperature range of 25-800 °C in consideration of tritium trapping to the experimental system. The capacity, which was independent of the water vapor pressure in the gas phase and the temperature, was determined as about 10⁶ Bq/cm². The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface was quantified at the temperature of 25, 500 and 700 °C in consideration of the behavior of tritium trapping at change of experimental condition by the numerical curve fitting method applying the serial reactor model. The reaction rate was observed to be constant as 3.48 × 10⁻⁵ m/s. Additionally tritium release behavior from the surface of SiC in water vapor atmosphere was predicted and compared with that for graphite and stainless steel.     

氚污染部件干法去污技术

针对核退役工程中存在的大量含氚废物处理问题,利用设计组装的干法去污处理装置对氚污染的金属进行了加热、紫外线、臭氧去污研究.结果表明,升高温度可明显提高去污效果;220 ℃用365 nm紫外线照射3 h对不锈钢的表面去污效率可达99%;臭氧与加热联合作用更有利于提高去污效率,220 ℃时去污3 h,臭氧对不锈钢、铝、黄铜的去污效率可达95%以上;而去污完毕经放置后,金属的氚表面活度会有所增加.     

Surface effects on tritium diffusion in niobium,zirconium and stainless steel

Tritium diffusion in niobium, Zircaloy-2 and stainless steel has been studied by measurement of both tritium concentration profiles and surface tritium release rates. Concentration profiles show buildup of tritium in the surface layers of a specimen with classical diffusion behavior at depths greater than about 5 μm from the surface.Application of a two-region diffusion model to the experimental data gives tritium diffusion coefficients in the surface films which are lower than the bulk diffusion coefficients by two orders of magnitude in stainless steel and eight to ten orders of magnitude in niobium and Zircaloy over a temperature range of interest for fission and fusion reactor systems. The surface effect appears to be a consequence of oxide film formation and is not due to the helium injected into specimens along with the tritium.     

氚在真空贮存容器材料中的吸附和渗透行为

研究了贮存氚靶约4 a和20 a的两个316 L不锈钢真空贮存容器(以下简称贮存容器)及其垫片材料对氚的吸附行为,并对氚在贮存容器材料中的渗透速率进行了测量和分析。结果表明,贮存容器外表面氚污染为几十Bq/cm2,不锈钢与陶瓷中吸附的氚活度均为106Bq/g;热解吸至1 273 K过程中,材料中99%的氚释放出来;在解吸出的氚中,陶瓷中的HTO比例高于不锈钢;贮存温度对氚靶贮存容器的渗氚速率有较大影响,夏季约为冬季的4倍。上述结果提示,氚在贮存容器材料内表面吸附后,一部分会向晶格扩散并滞留下来;另一部分则透过材料向外环境渗透,其中温度是影响氚向外环境渗透的主要因素之一。     

Neutronics Calculations in Pellets and Blankets of Laser Fusion Reactor Concept SENRI-I

The neutronic properties of SENRI-I, a reference design of laser fusion reactor proposed by Institute of Engineering, Osaka University, are discussed on the basis of the one-dimensional neutron transport calculations in burning DT plasmas and blankets. The softening of the fusion neutron energy spectrum, the neutron heating and the neutron multiplication are studied and discussed for the compressed DT pellets with various thickness of fuel plasmas and lead or lead-polyethylene tampers.

The neutronic and thermal features in the blanket of the SENRI-I design are also examined. The tritium breeding ratio is high enough (~1.6), depending on the neutron energy spectrum from a pellet. The maximum temperature increase per 1,000 MJ DT fusion reactions is ~3°C in the inner liquid Li layer and ~1.5°C in the stainless steel first wall. A parametric study is also presented on the effect of varying the thickness of the inner Li blanket ΔRⁱ to examine the thickness required for the enough tritium breeding ratio and energy deposition.     

Release behavior of bred tritium from LiAlO2

The release behavior of bred tritium to the blanket purge gas is mainly controlled by such bulk phenomena as tritium forming reaction, diffusion of tritium in grain, interaction of tritium with irradiation defects, and absorption together with such surface phenomena as adsorption, isotope exchange reaction between molecular form hydrogen in purge gas and tritium on grain surface (isotope exchange reaction 1), isotope exchange reaction between water vapor and tritium on grain surface (isotope exchange reaction 2), and water formation reaction at addition of hydrogen. Following the observation of the present authors that the isotope exchange reaction 2 is much faster than the isotope exchange reaction 1, the release curve of bred tritium obtained at purge with humidified gas was used for estimation of the effective diffusivity of bred tritium in LiAlO₂. Then, the effective diffusivity of tritium in grain of LiAlO₂ is obtained as D_T = 2.5 × 10⁻⁷exp(−110 [kJ]/RT) [m²/s]. This equation gives the larger diffusivity than any other diffusivity presented so far because the mass transfer resistance at the grain surface is expected to be eliminated in the estimation procedure of this study.     

Design of a multipurpose laboratory scale apparatus for the investigation of hydrogen isotopes in PbLi and permeation technologies

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain.A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550 °C and it will be operated with hydrogen partial pressure in the range 10²–10⁴ Pa. The facility can work with desorption and absorption technique.Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.     

Sorption of gaseous tritium on the surface of type 316 stainless steel

The sorption of gaseous tritium on the type 316 stainless steel was studied. The stainless steel was first contacted with gaseous tritium, and then the remaining tritium was evacuated. During a gradual etching from the surface by an acid solution, the tritium was released as HTO with a fraction of HT. They were radioassayed separately. The HTO mostly originates from the tritium present on the outer-most surface and about 90% of it could be released easily into water. However, the rest is sorbed tightly and remained in the surface layer. A fraction of the sorbed-tritium will diffuse atomically through the surface layer into the bulk of stainless steel and is released as HT by etching. The activation energy of the diffusion was determined as 32.8 kJ/mol.     

A tritium permeation model for conceptual fusion reactor designs

A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10^?3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×10⁴ s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.     

Determination of Tritium Concentration in Fuel Reprocessing Liquid Waste by Vacuum Freeze Distillation

Abstract

The vacuum freeze distillation method was successfully applied to the preparation of radiochemically pure water samples used for determining the tritium concentration in highly radioactive liquid waste produced in nuclear fuel reprocessing.

This method yielded a decontamination factor exceeding 10⁵ for radioruthenium and other long-lived fission products. The isotope effect on the tritium concentration was corrected by applying an enrichment factor determined from experiment.

A small portion of the high level liquid waste solution was taken from the Reprocessing Plant of the Japan Atomic Energy Research Institute, neutralized with 2 N sodium hydroxide, and subjected to the vacuum freeze distillation. The distillate was subsequently analyzed for tritium by liquid scintillation counting. The tritium concentration in the waste solution was found to be approximately 0.13 μCi/ml, which corresponded to about 70% of the total tritium contained in the original spent fuel rods.     

Tritium dose criteria and radiological impact of a tritium plant

Radiation safetry criteria adopted in Russia (in the former USSR) distinguish five classes of tritium compounds. The lowest permissible tritium concentration in the air is set for insoluble tritium compounds (3.10⁵ times lower than that for HT). Russia's criteria for tritiated radioactive waste are outlined. It is explained why the tritium weighting factor of two is used as a basis for the tritium dose criteria development in this country. The ecological situation nearby a large tritium processing plant is considered. Amounts of tritiated waste produced at the plant, sources of tritium effluents, tritium content in the air, water, snow, soil and vegetation as well as HTO sorption parameters of various food products are reported. On the basis of HTO near-surface concentrations in the air and public doses measured 3 km away from the plant stack, the tritium dose factor was calculated.