Fusion safety and environmental R&D tasks in Japan

This paper aims at listing and evaluating the status of all the research and development (R&D) tasks necessary for the construction of a safe and environmentally benign fusion experimental reactor. At this time, it is not possible to define precisely the R&D tasks necessary for the licensing approval and those that are useful in improving safety but not necessarily required for licensing because the licensing procedure itself is still being discussed. Among the R&D tasks, the most important are considered to be those related to tritium safety, namely, those effective in reducing the uncertainty in tritium inventory in the plasma facing components and blanket, uncertainty in tritium permeation and leakage, and those to clarify tritium behavior in the containment and in the environment. The R&D tasks with second priority are judged to be those related to mobilization of the activation products such as activated erosion dust or the corrosion products. The volatilization of structural metal caused by the oxidation at high temperature seems to be highly unlikely but some experiments are needed to assure that this is the case.     

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.     

Effect of Tritium and Dissolved Oxygen on Anodic Polarization of SUS304 Stainless Steel in Sulfuric Acid Solution

Since exotic corrosion of stainless steels in tritiated water can be expected, the anodic polarization of a SUS304 stainless steel sample in approximately 5 wt% sulfuric acid solution was performed at various concentrations of tritium and dissolved oxygen (hereafter DO) in the electrolyte. The inhibitory effect of tritium on the passivation could be observed with DO even at a tritium concentration in the electrolyte of as low as 2.2 kBq cm^?3. This effect became more pronounced as the tritium concentration increased. It was suggested that the inhibitory reaction depending on tritium concentration would compete with the self-passivation depending on the DO concentration (hereafter [DO]), since it was found that there is a threshold [DO] for self-passivation at each tritium concentration.     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

Tritium absorption of co-deposited carbon films

Co-deposited carbon film with different deuterium concentration, D/C, were exposed to tritium gas at the temperature of 423 K, and then the atomic ratio of absorbed tritium to carbon, T/C, was evaluated. The obtained data were discussed with crystal structure of the carbon film. The T/C increased with decreasing D/C of carbon film. The carbon film with low D/C had more defective structure. The reduction of D/C by the heating before tritium exposure led to the increase of absorption amount. These results suggest that carbon film with more defective structure and low D/C film could absorb large amount of tritium. The hydrogen isotope concentration in the present experiment was saturated below the orders of 10^?4, which was 3–4 orders of magnitude smaller than that of co-deposited carbon film with hydrogen isotope.     

Tritium implantation and depth profiling in fusion related materials

Tritium behaviour in solids and particularly its permeation and inventory in the first wall, limiters, breeding blanket materials and in other structural elements of fusion reactors is a subject of great concern in all projects aiming at D + T fusion. In the present work elastic recoil detection (ERD) under ⁴He bombardment and the T(d, α)n nuclear reaction analysis (NRA) in the forward detection geometry were applied to the depth profiling of tritium at submicron distances below the surface of selected fusion related materials. Experimental results obtained for tritium implanted in titanium, graphite and lithium aluminate LiAlO₂ are presented as the examples.     

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.     

Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 °C) and the beam-off (20 °C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 × 10⁴ MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environment.     

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.D.T.I. Dr. Trippe Ing. GmbH, Karlsruhe.     

Isotopic Effects of Hydrogen During the Decomposition of Water in Electrolysis with a Solid Polymer Electrolyte

The partition coefficients of hydrogen isotopes in the decomposition of water in electrolyzers with a solid polymer electrolyte are measured in a wide range of deuterium content in the initial water (from the natural content up to close to 100%), including for water containing trace concentrations of tritium. Equations describing the temperature dependence of the partition coefficients (temperature range 298–333 K) and the deuterium content dependence of the tritium partition coefficient are obtained. Partition coefficients are presented for binary isotopic mixtures protium–deuterium, protium–tritium, deuterium–tritium. 1 figure, 3 tables, 17 references.     

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.     

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.     

Adsorption Characteristics of Water Vapor on Gamma-Lithium Aluminate

Lithium aluminate (LiAl0₂) is one of the probable candidates for tritium breeding material because of its potential to give a fair tritium breeding ratio, stability at high temperature and stability in atmosphere containing water vapor. Clarification of the adsorption performances of water on LiAl0₂ is important for optimization of the way to recover bred tritium from LiAl0₂ and for estimation of the tritium inventory in the breeding blanket and for quantification of the hydrogen isotope exchange reaction. The amount of water captured on LiAl0₂ was studied in the temperature range of 373~1,100K using the breakthrough curve method, and adsorption isobar and isotherm of water on LiAl0₂ were proposed based on the data obtained. The water capture phenomena of LiAl0₂ was attributed to the dissociative chemisorption, and the apparent activation energy was determined to be 32.2 kJ/mol.K. The tritium inventory by sorption for LiAl0₂ was compared with that for Li₂0.     

Assessment of tritium levels in rivers and precipitation in north-western Greece before the ITER operation

The project ITER aims to demonstrate that fusion is the energy source of the future. The prototype Tokamak machine is intended to start operation at about 2019 and tritium is one of the major contaminants that can be accidentally released in the environment. Nowadays environmental tritium levels are of natural origin except in the vicinity of nuclear facilities. The evaluation of background tritium levels is essential in the context of a future possibility of accidental tritium releases. For this purpose and also because of the lack of relevant information, an extended programme of river and rain water sampling was implemented in north-western Greece. Water samples from six major rivers in this area and rain water samples were analysed for tritium content. The rivers under investigation were Aliakmonas River, Pinios River, Arachthos River, Kalamas River, Aoos River and Louros River, which originate from the central Greek mountain range Pindos, and flow to Aegean and Ionian Sea.The tritium concentrations were determined by the Liquid Scintillator Analyser Tri-Carb 3170TR/SL. The statistical analysis of data revealed that there is a seasonal variation of tritium concentration in rain samples and a less pronounced seasonal variation in river samples. The weighted mean tritium concentration for rain samples was determined equal to 0.90 ± 0.08 Bq L^?1 (7.6 ± 0.7 TU) and the respective mean value for river samples was 0.94 ± 0.04 Bq L^?1 (7.9 ± 0.3 TU). Further analysis has demonstrated that river waters tend to show lower tritium concentrations than the concurrently measured tritium concentrations in rain samples, during spring and summer (at 47% and 71% of the sampling stations, respectively), while this observation is reversed during autumn and winter (at 44% and 35% of the sampling stations, respectively). This may be attributed to rain water remaining underground for a long period allowing tritium to decay and when it reappears as river water, the tritium concentration values are lower when compared to the rain water concentrations. Rough estimates of the residence time of underground waters in the study area provided values, which ranged from 0.5 to 11.7 years, with a mean value of 5.2 ± 0.9 years.     

Isotope Exchange Capacity on Li4SiO4 and Comparison of Tritium Inventory in Various Solid Breeder Blankets

It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of a breeder material, absorption of water vapor into bulk of the grain, adsorption of water on surface of the grain, and exchange capacity of tritium to be trapped to surface of the grain together with two types of isotope exchange reactions for evaluation of the tritium inventory in a solid breeder blanket under various conditions. The isotope exchange capacity on the Li₄SiO₄ surface is experimentally obtained in this study. Most of the properties required for evaluation of the tritium inventory for various blanket materials have been already quantified by the present authors. Then it has become possible to compare the tritium inventory in solid breeder blankets packed with either Li₂O, LiAlO₂, Li₂ZrO₃, Li₂TiO₃ or Li₄SiO₄ using the calculation model previously presented by the present authors.     

Safety Analysis of ITER EDA Design by GEMSAFE

General Methodology of Safety Analysis and Evaluation for Fusion Systems (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) design in the stage of Engineering Design Activities (EDA) to identify Design Basis Events (DBEs) and the related safety features, which were compared with those of the ITER design in the stage of Conceptual Design Activities (CDA). As a result, 18 DBEs for the EDA design were selected in comparison with 25 DBEs for the CDA design. DBEs related to the fuel area were categorized in higher event category than those of the CDA design due to the increase of the mobile tritium contained in some components. It was necessary to reduce the inventory of the tritium absorbed in the tokamak dust in the EDA design as well as in the CDA design. Some measures were recommended to reduce mobile tritium dissolved in the coolant in the single cooling loop due to the increase of this estimated inventory.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

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T riSim-Dynamic,在此基础上进行修改完善,利用该程序对CFETR...     

Effect of sweep gas species on tritium release behavior from lithium titanate packed bed during 14 MeV neutron irradiation

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li₂TiO₃) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.     

Overview of tritium safety technology at the Tritium Process Laboratory of JAERI

The Tritium Process Laboratory of the Japan Atomic Energy Research Institute is the only laboratory in Japan where grams of tritium can be handled to carry out R&D on tritium processing and tritium safety handling technologies for fusion reactors. The tritium inventory is approximately 13 grams. Since 1988, basic research has been performed using gram-level tritium quantities. During the past 5 years, approximately 1 kilogram of tritium has been handled in experimental apparatus. The total amount of tritium released through the stack of TPL was controlled to less than 1 Ci without any accidents. In order to establish more complete tritium safety for DT fusion reactors, main R&D areas on tritium safety technology at TPL were focused on a new compact tritium confinement system, reliable tritium accounting and inventory control, new tritium waste treatments, and tritium release behavior into a room.