Oxidized Tritium around a Research Reactor Site

The concentration and distribution of tritium in environmental samples obtained from sites near the Kyoto University Research Reactor were studied. About 5 GBq of tritium is discharged yearly from the KURR stack. The concentrations of tritium in the exhaust air, atmospheric moisture and precipitates were monitored to estimate not only any effects of tritium sources on the concentrations in the nearby environment but also the dilution factor of pollutants at the site boundary. The concentrations of tritium in surface water at the site were also monitored to identify the possibility of pollution in the water system. In both cases, there was slight contamination in samples near the site. The increased annual dose to an adult from tritium discharged in the atmosphere was estimated to be about five orders of magnitude lower than that from natural background radiation.     

中国HCCB-TBM氚增殖单元三维多物理场耦合氚输运模拟分析

中国氦冷固态实验包层模块(HCCB-TBM)将在国际热核聚变实验堆(ITER)上安装测试,以验证其氚增殖能力与核热移出能力。HCCB-TBM中的氚输运与流体的传热和传质、氢同位素交换、结构材料的SORET效应密切相关。考虑以上物理因素,基于商业软件COMSOL完成了HCCB-TBM氚增殖单元多物理场耦合的氢同位素输运模拟分析。分析结果表明:球床吹洗气体中含氢有助于抑制氚渗透损失;当吹洗气体含氢浓度为4.66×10^-2 mol/m³时,产生约13.2倍的氚渗透阻止效应。     

CFETR氚自持分析评估与验证策略

中国聚变工程实验堆(China Fusion Engineering Test Reactor,简称CFETR)的主要目标之一是实现氚自持。采用氚平衡法对CFETR不同运行工况下的氚自持条件进行了分析评估。结果表明:在500 MW运行阶段,CFETR实现氚自持所需的最小氚增殖比(TBR_r)为1.098,小于CFETR增殖包层可达到的氚增殖比(TBR_a),即在理论上满足氚自持条件。在此基础上,提出了CFETR未来通过定期的氚衡算来验证氚自持的基本策略。在基准输入参数和氚存量测量精度限制(1%)条件下,CFETR氚自持验证实验的运行周期需要大于22 d(氦冷包层)或87 d(水冷包层)。     

PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident: Part 2. Formation and destruction of iodine oxides compounds under irradiation - Experimental results modelling

In the case of a severe accident in a nuclear Light Water Reactor (LWR), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core would induce air radiolysis. The air radiolysis products (ARP) could, in turn, oxidise gaseous molecular iodine (I₂) into aerosol-borne iodine-oxygen-nitrogen compounds, abbreviated as iodine oxides (IO_x). These reactions involve the conversion of a gaseous iodine compound resulting in a change of the iodine depletion rate from the containment atmosphere. Kinetic data were produced within the first part of PARIS project on the air radiolysis products formation and destruction. The second part of the PARIS project as presented in this paper deals with the impact of the ARP on the conversion of I₂ into IO_x. The objective was to provide a database to develop new or to validate existing kinetic models of formation and destruction of iodine oxides.The iodine tests of the PARIS project, performed at very low, realistic iodine concentrations, constitute an important database to further develop or validate empirical and mechanistic models on radiolytic I₂ oxidation. In the presence of painted surface areas or silver aerosol surface areas, radiolytic I₂ oxidation is negligible compared to I₂ adsorption on these surfaces for the conditions examined. However, radiolytic I₂ oxidation remains very efficient if surface areas are small or if they are made of the relatively non-reactive stainless steel.     

用液闪直接测量核设施周围环境中的氚

某些核设施运行时会释放氚,从而引起周围环境中氚活度浓度水平的变化。对核设施周边区域空气、地下水、雨水和海水样品中的氚分别用内标准法(简称“内标法”)和外标准淬灭指示参数法(简称“外标法”)进行了液闪测量。两种标准方法测量数据的相对偏差在-4.0%～4.0%。根据内标法的探测效率与仪器给出的淬灭指示参数制作了4种环境水样的淬灭校正曲线。在环境样品测量中,内标法和外标法的探测效率最大差值约为1.6%,痕量¹⁴C和其它β放射性核素对³H的计数率影响可忽略。对探测效率为21.5%～24.5%无严重淬灭的样品,用液闪直接测量并根据外标法的淬灭校正曲线计算氚活度浓度,相对偏差在-6.35%～4.41%,基本可满足核设施氚常规监测的要求。     

氚增殖剂球床组件堆内辐照物理热工计算分析

为验证在中国先进研究堆(CARR)内进行国际热核聚变实验堆(ITER)氚增殖包层模块(TBM)辐照实验的可行性和安全性,进行了氚增殖剂球床组件堆内辐照物理及热工计算分析。氚增殖剂包层模块主要是固态氚增殖剂陶瓷球床。本文采用Monte Carlo粒子输运模拟程序对氚增殖剂球床进行堆内建模,计算球床的中子注量率、能量沉积和产额,得到不同功率下球床的中子注量率、发热功率和产氚速率以及球床组件引入反应堆的反应性。根据物理计算得到的组件各部件发热情况建立热工计算一维模型,通过更改反应堆功率得到满足实验要求的工况并采用三维程序进行验证。物理与热工计算分析的结果表明,在反应堆运行功率为20 MW的工况下球床组件各部件的温度均不超过限值。     

CFETR氦冷固态包层及其辅助系统动态氚输运分析

实现氚自持、建立完整的氚循环系统并保证氚安全是中国聚变工程实验堆(CFETR)的主要目标之一。在CFETR氦冷固态包层及其辅助系统设计过程中,需对系统级氚输运行为进行详细分析,包括氚滞留量、释放量、浓度的动态变化等。基于已建立的动态氚分析程序TriSim-Dynamic,在此基础上进行修改完善,利用该程序对CFETR氦冷固态包层及其辅助系统氚动态输运进行分析模拟,得到了冷却剂及提氚吹扫气中氚浓度、氚分压,管壁及结构材料中氚盘存量,氚通过包层结构材料和辅助系统管壁向真空室、水冷系统及建筑的渗透通量动态变化,并将其稳态值与已进行基准校核的稳态氚分析程序TriSim-SA及理论解析解进行比较,以初步验证分析结果的准确性,数据结果也对CFETR氚安全分析提供了一定的参考。     

核电站氚的排放量及浓度限值比较分析

氚是核电站运行过程中向环境中排放较大的放射性核素之一,控制核设施中氚的产生和排放量越来越引起人们的重视。本文通过分析核电站产生氚的主要途径,结合国际上的运行经验参数,对比分析了不同国家、不同堆型核电站氚的排放量和浓度限值。分析结果表明:三十年间,全球核电站流出物中气态氚的排放量显著高于液态氚,重水堆是各堆型核电站中氚排放的主要贡献者,也是氚排放所致公众剂量的主要来源。为了更加有效的控制氚的排放,法国等国家核安全监管机构根据电站的装机容量、排放工艺、堆型等制定了各自国家核电站氚的年排放总量限值;加拿大等国的监管机构根据剂量限值制定了导出排放限值,该值的优点是便于审查核电站正常运行时氚的排放量;其它核电国家则是以剂量限值的形式提出了氚的排放限值。     

The Significance of Tritium Releases to the Environment

Tritium is produced naturally and was present in low concentrations in precipitation and natural bodies of water before atmospheric testing of nuclear weapons. Other sources of tritium are now present from which tritium is released to the environment. Nuclear reactor tritium production, according to recent estimates, will equal natural tritium production before the year 2000. Predicted increases of tritium in the environment will take place first on a local ecological level and then appear on a biospheric level. Tritium introduced into the environment as THO will move through ecological systems in the same manner as stable water. Tritium will enter the hydrologic cycle either via evapo-transpiration or the surface bodies of water. Ecological experiments have been conducted to determine the movement of tritium in the environment. Field-grown plants were exposed to liquid and vapor THO for periods of one-half and one hours. Tritium concentrations were determined in leaf samples collected after exposure for periods of time up to 45 days. Tritium decays rapidly in the plant species studied and exhibited a three component half-life when plants were exposed to THO vapor. The length of exposure, and sources of THO in the soil affect the half-time of tritium in the plant tissues. Data produced in ecological experiments on tritium movement are used in a theoretical consideration of acute and chronic vapor releases of tritium in an agricultural environment.     

Fusion tritium research facilities in KAERI

Korea Domestic Agency (KODA) is developing a nuclear fusion fuel storage and delivery system (SDS) as one of the Korean procurement packages. Korea Atomic Energy Research Institute (KAERI) is operating the following basic scientific research laboratories for an SDS and tritium supply study: a metal hydride bed preparation laboratory, hydrogen isotope recovery and delivery performance test rig, in-bed calorimetry performance test rig, and tritium shipping container integrity test facility. Furthermore, the development of a test blanket module (TBM) is required to test and validate the design concept of tritium breeding blankets relevant to fusion power plants. KAERI is also operating the following laboratories for the TBM research, such as a tritium extraction performance test rig, High-flux Advanced Neutron Application Reactor (HANARO), and Experimental Loop for Liquid Breeder (ELLI).     

Neutronic performance of new coolants in a fusion–fission (hybrid) reactor

Selection of coolant used in the fuel zone of a fusion–fission (hybrid) reactor affects the neutronic performance of the blanket much. Recently, two coolants namely, Flinabe and Li₂₀Sn₈₀ have been investigated to use in fusion reactors as tritium breeder and energy carrier due to their advantages of low melting point, low vapor pressure. In this study, neutronic performance of these coolants in a hybrid reactor using Canada Deuterium Uranium Reactor (CANDU) spent fuel was investigated for an operation period of 48 months. And also that of natural lithium and Flibe was also examined for comparison. Neutron transport calculations were conducted on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation.     

Source Term and Mobilization Assessment in NSSR-1 of ITER

We describe the radioactive sources in the International Thermonuclear Experimental Reactor (ITER). The most important sources are co-deposited tritium, tritiated water, tokamak dust, and corrosion products. The co-deposited tritium is limited to 1 kg-T; the total on-site tritium inventory in the Basic Performance Phase (BPP) is 4 kg-T. Tritiated water concentrations are kept below 0.2 g-T/m³ in the divertor; other coolant loops have lower tritium concentrations. The in-vessel dust inventory is up to 100 kg-W, 100 kg-Be, and 200 kg-C. The activated corrosion product inventory is kept below 10 kg per loop.     

Korea's progress on the ITER tritium systems

Korea has been operating four units of CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant has been removed by Wolsong TRF (tritium removal facility) since 2007. Korea has developed a 500-kCi BU-type tritium transport container. Furthermore, strictly controlled tritium release to the environment from the CANDU nuclear power plant in Korea will also be a helpful experience for ITER. The BU-type tritium transport container will be unpacked and the quantity of tritium in the metal tritides of the primary vessel will be measured accurately before the tritium is liberated by heating the metal tritides. In the ITER fuel cycle plant, Korea is responsible for the development and supply of the SDS (fuel storage and delivery system). We present the on-going R&D progress to use a non-nuclear material ZrCo for the storage and delivery of tritium. Especially we present the delivery characteristics of the ZrCo hydride bed by considering the fueling requirement and maintainability of the ZrCo hydride bed. We also present various test results for the experimental ZrCo beds and the test programs for the leak tight large scroll pump with a magnetic coupling-drive.     

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.     

Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM

Tritium behavior in the reactor such as production, diffusion and release are accompanied by their adsorption and desorption in graphite materials, which are essential to the safety of high temperature gas cooled reactor (HTGR). In order to study this important issue, hydrogen instead of tritium is experimentally used in this work and justified viable by theory. By performing multiple sets of comparative experiments, the features of hydrogen adsorption and desorption behavior changing by adsorption temperature and time in typical graphites used in HTR-PM (High Temperature Gas Cooled Reactor – Pebble Bed Module), i.e. reflective layer, fuel element and boron carbon bricks, have been observed and analyzed. Furthermore, the adsorption rates of hydrogen in the three materials as above at different conditions are also given. Based on the experimental results, tritium behavior in the HTR-PM was inferred and estimated, which is significant for the further study on the mechanism of tritium transport.     

The TFTR Lithium Blanket Module program

The Lithium Blanket Module (LBM) is an approximately 80×80×80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li₂O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li₂O pellets with satisfactory reproducibility were developed using purified Li₂O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1–1 nCi/g).     

Progress on ruthenium release and transport under air ingress conditions

A particular concern in the event of a hypothetical severe accident is the potential release of highly radiotoxic fission product (FP) isotopes of ruthenium. The highest risk for a large quantity of these isotopes to reach the containment arises from air ingress following vessel melt-through. One work package (WP) of the source term topic of the EU 6th Framework Network of Excellence project SARNET is producing and synthesizing information on ruthenium release and transport with the aim of validating or improving the corresponding modelling in the European ASTEC severe accident analysis code. The WP includes reactor scenario studies that can be used to define conditions for new experiments.The experimental database currently being reviewed includes the following programmes:

•

AECL experiments conducted on fission product release in air; results are relevant to CANDU loss of end-fitting accidents;

•

VERCORS tests on FP release and transport conducted by CEA in collaboration with IRSN and EDF; additional tests may potentially be conducted in more oxidizing conditions in the VERDON facility;

•

RUSET tests by AEKI investigating ruthenium transport with and without other FP simulants;

•

Experiments by VTT on ruthenium transport and speciation in highly oxidizing conditions.

In addition to the above, at IRSN and at ENEA modelling of fission product release and of fuel oxidation is being pursued, the latter being an essential boundary condition influencing ruthenium release.Reactor scenario studies have been carried out at INR, EDF and IRSN: calculations of air ingress scenarios with respectively ICARE/CATHARE V2; SATURNE-MAAP; and ASTEC codes provided first insights of thermal-hydraulic conditions that the fuel may experience after lower head vessel failure.This paper summarizes the status of this work and plans for the future.     

Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.     

Irradiation testing of design models for the GCFR fuel pressure equalization (vent) system

The major consideration in the design of the pressure equalization system for the gas-cooled fast breeder reactor is the release and venting of gaseous and volatile fission products. Single vented rods have been irradiated in the thermal flux of the Oak Ridge Research Reactor (ORR) at GCFR operating conditions of 12–15 kW/ft and 565–685°C cladding outside temperature to determine the fission product release and to verify the design concept. Results obtained to date from measurements of fission gas release and transport have been compared with predictions based on design assumptions to verify analytical models and have established a degree of conservatism of design assumptions.The release of radioactive gases from the fuel matrix was measured directly at 12 kW/ft in an operating fuel rod and was found to be significantly less than the design assumption of 100% instantaneous release and less than predictions using the diffusion model with Findlay's coefficients. Although solid state diffusion was found to be the dominant process delaying the venting of fission gases in the experimental irradiation, fission gas interdiffusion in helium will be the dominant venting transport process for the reactor design. Delay of fission gases by adsorption on charcoal was verified at trap operating temperatures for burn-ups up to 54 000 MWd/t. Volatile fission products (cesium and iodine) did not migrate beyond the fuel-blanket interface. The feasibility of the vented-fuel-rod design concept has been established.     

An Overview of INEL Fusion Safety R&D Facilities

The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.