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The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain ...

The tritium exhaust behavior from the Large Helical Device (LHD) was observed in the first deuterium plasma experimental campaign. Tritium in the exhaust gas was monitored at the conducted by use of the ionization chamber and water bubbler system with the discrimination of chemical forms. The observation results indicated that (i) tritium on the surface of the first wall and divertor tiles as plasma facing components was released by the hydrogen isotope exchange reaction of the glow discharge cleaning operation and the diffusion-limited process was suggested in the tritium release behavior from the bulk, (ii) the amount of tritium release from the LHD vacuum vessel was about one-third of the produced tritium and the mostly produced tritium was still retained at the end of plasma experimental campaign, (iii) the ratio of exhausted tritium from the LHD vacuum vessel was larger than that in the case of JT-60U based on the carbon materials as the plasma-facing components. It indicated that the tritium inventory would be reduced and controlled by the kind of plasma-facing materials.

The WWR-K is 6-MW(thermal) light-water, tank-type reactor with thermal neutron spectrum. It is the exclusive multipurpose research reactor in the Republic of Kazakhstan. The WWR-K is owned by the I...

Tritium release behavior in a tungsten sample after exposing to tritium ions with energy about 200 eV created by glow discharge has been studied by both β-ray–induced X-ray spectrometry (BIXS) and ...

The Chinese (CN) Helium Cooled Ceramic Breeding (HCCB) Test Blanket Module (TBM) (CN HCCB TBM) set with its ancillary systems will demonstrate the feasibility of in-pile tritium production/breeding in ITER for fuel self-sufficiency and high-grade fusion energy conversion to heat and extraction for a future magnetic confined fusion reactor. Tritium release in some major components of the recently designed TBM systems through permeation and natural leakage was estimated with simple diffusion/permeation and leak rate calculation models. Results showed that because of the tritium permeation barrier coating for tritium confinement in some tritium containments, total tritium release to the environment by permeation in the CN HCCB TBM and ancillary systems will be kept well below 2 Ci/full-power day. However, tritium release through natural leakage from components can be neglected compared with permeation. Equipped with ITER tritium safety guarantee facilities like the tritium monitoring and detritiation...

The high contamination potential of the release of radioactive tritium facilitates the demand for and development of a stringent and comprehensive approach to operational maintenance of tritium systems. Prompt and efficient maintenance is necessary to ensure the accepted operational safety basis is adhered to and a continued safe state of operation is achieved. This will help to mitigate and avoid potential hazards that result from a tritium release to the public and facility personnel. Because of the hazards associated with a release of tritium contamination the process systems are in large kept within a series of inerted glovebox environments that must be maintained to keep structural integrity. The nature of a tritium release from a glovebox could have significant consequences for the general public as well as for personnel. As such, the maintenance philosophy is developed to help facilitate operations in the adherence to the facility’s safety code of conduct.To effectively facilitate the safe ...

Tritium Process Laboratory (TPL) in Japan is operated by the Japan Atomic Energy Agency (JAEA) and was established as the only facility to handle over one gram of tritium in Japan in 1985. Since March 1988, the TPL has been operated safely with tritium, and no tritium release accidents have been occurred. The maximum tritium concentration of a three-month average in a stream from a stack of TPL to environment was 350 Bq/m3, and is about 14 times smaller than that of the legal release limit in Japan. The failure data have been analyzed for several main components of the safety systems such as pumps and monitors. The tritium waste data has also been accumulated as liquid and solid waste from TPL. Through this operating experience, a significant database for the safety systems of the TPL has been accumulated. This data can provide a source of reliability information for a future fusion facilities.

The Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is to be installed in one half of equatorial port #2 for testing in ITER machine. Liquid Pb-Li and solid Li2TiO3 are the tri...

The impact of water vapor and temperature on the release of tritium from stainless steel was studied. Degreased stainless steel samples loaded with tritium at room temperature following a 24-h dega...

Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in...

The main purpose of the Tritium Recovery System (TRS) is to reduce to a maximum possible extent the release of tritium from the facility following a tritium release in confinement boundaries and also to have provisions to recover both elemental and vapors tritium from the purging gases during maintenance and components replacement from various systems processing tritium. This work/paper proposes a configuration of Tritium Recovery System wherein elemental tritium and water vapors are recovered in a separated, parallel manner. The proposed TRS configuration is a combination of permeators, a platinum microreactor (MR) and a trickle bed reactor (TBR) and consists of two branches: one branch for elemental tritium recovery from tritiated deuterium gas and the second one for tritium recovery from streams containing a significant amount of water vapours but a low amount, below 5%, of tritiated gas. The two branches shall work in a complementary manner in such a way that the bleed stream from the permeators shall be further processed in the MR and TBR in view of achieving the required decontamination level. A preliminary evaluation of the proposed TRS in comparison with state of the art tritium recovery system from tritium processing facilities is also discussed. (authors)

The assessment of the environmental impact of tritium release from nuclear facilities is a topic of interest in many countries. In the IAEA's Environmental Modelling for Radiation Safety (EMRAS I) programme, progresses for routine releases were done and in the EMRAS II programme a dedicated working group (WG 7 - Tritium Accidents) focused on the potential accidental releases (liquid and atmospheric pathways). The progresses achieved in WG 7 were included in a complex report - a technical document of IAEA covering both liquid and atmospheric accidental release consequences. A brief description of the progresses achieved in the frame of EMRAS II WG 7 is presented. Important results have been obtained concerning washout rate, the deposition on the soil of HTO and HT, the HTO uptake by leaves and the subsequent conversion to OBT (organically bound tritium) during daylight. Further needs of the processes understanding and the experimental efforts are emphasised.

The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.

Tritium release behavior for thermal neutron-irradiated Li0.17Pb0.83 eutectic alloy was studied. Main tritium release peak was observed in the temperature just a little higher than melting point in...

The Rokkasho Reprocessing Plant, located in northeastern Aomori prefecture, is currently undergoing the final commissioning test using actual spent nuclear fuels (the Active Test). Tritium had been discharged from the reprocessing plant and some effects to the environment were observed since the Active Test had been started at the reprocessing plant on March 31, 2006. The purpose of this study is to predict the environmental impact of the reprocessing plant once it becomes operational. An atmospheric dispersion simulation system, developed by the authors, has been utilized to predict the expected tritium dispersion during the plant future operations. In this study, a simulation of tritium dispersion was carried out using wind velocities and wind directions as prediction factors for future HTO activity. The simulation results were compared with actual HTO activity measurements taken during the autumn of 2006 and the spring and autumn of 2007. The results of the simulation appear to accurately refle...

The assessment of environmental impact of tritium release from nuclear facilities is a topic of interest in many countries. In the IAEA’s Environmental Modelling for Radiation Safety (EMRAS I) prog...

Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO.In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so fa...

The present authors have developed the tritium release model to represent the tritium release behavior from solid breeder materials (Li2ZrO3, Li2TiO3, Li4SiO4, LiAlO2 and Li2O). It has been found that water is released from solid breeder materials into the purge gas due to desorption of physically and chemically adsorbed water and water generation reaction and that this water affects the tritium release behavior. In this study, the amount of adsorbed water and its desorption rate for Li2ZrO3 were quantified. It was found in this experiment that Li2ZrO3 has the largest adsorption amount among the solid breeder candidates. It was also observed that Li2ZrO3 has the largest water generation capacity among the solid breeder candidates. A unique reaction at around 550°C which made up approximately 80% of the capacity of water generation was also observed. It is considered that the phase change of ZrO2 at around 550°C supplies oxygen to promote water generation reaction. Tritium release behavior from Li2...

A tritium release model has been developed by the present authors. The tritium release curves estimated by this tritium model give good agreement with experimental curves for L1 4 SiO 4 , Li 2 TiO 3 , Li 2 ZrO 3 or LiAlO 2 under various purge gas conditions in our out-of-pile bred tritium release. The characteristics of tritium release behavior from various solid breeder materials carried out by us and in EXOTIC experiments at Petten are discussed in this study.

To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 ∼ 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D.

Within the IAEA program EMRAS, an exercise has been performed by 7 countries to evaluate the consequences of an acute atmospheric release of tritium (10g). This study aimed at giving practical technical information to decision-makers. Three scenarios with different meteorological conditions were modeled. The objective of this paper is to give the main information about transfer and impact, evaluate uncertainties in models/assessments and so help to set countermeasures. From the results of this exercise, reference activity values for countermeasures can be discussed. All pathways and chemical forms (HTO-HT-OBT) are considered.

Thermal desorption behavior of tritium has been investigated for SS316 and F82H irradiated by 580MeV proton (SINQ-target3) up to 5.0 ∼5.9 dpa and 6.3∼9.1 dpa, respectively, in order to understand tritium transport in the irradiated materials. While the tritium release has only one peak at 670 K from irradiated SS316, that has two peaks at 510 K and 670 K from irradiated F82H. Those results indicate that only one kind of trap site exists in the SS316, and at least two kinds of trap site exist in F82H. As the results of tritium transport analysis of tritium release behavior, it was found that the trap site at 670 K for SS316 and F82H could be controlled by the same trap mechanism. As to the chemical form of tritium released from the steels, 1/2 and 1/3 of tritium was release as water vapor form from SS316 and F82H, respectively. It could be attributed to the growth of surface oxide on the metal surfaces during the TDS.

One of the main concerns related to licensing of ITER is the amount of potentially tritium release into the environment and the qualification of the barriers against tritium release. The final barrier of tritium release from fuel cycle is the Water Detritiation System (WDS) which will be operated in combination with the Isotope Separation System (ISS). To investigate the performances of various components of these systems, an experimental facility based on Combined Electrolysis Catalytic Exchange (CECE) process with a Cryogenic Distillation (CD) process was built at Tritium Laboratory Karlsruhe. The investigations are focused on two main issues: to quantify the separation performances of deuterium and tritium within the Liquid Phase Catalytic Exchange (LPCE) and CD processes in steady state and in dynamic mode of operation and to develop an integrated control system to be used in ITER ISS, in order to minimize the tritium inventory and to reduce at maximum extent the tritium releases. At TLK the two systems, CECE and CD have been commissioned and the experimental program and preliminary functionality tests of the main components are presented.

The behaviour of tritium released into the International Thermonuclear Experimental Reactor (ITER) working rooms in the form of HTO or T2 as a result of some postulated reference accidents is analysed. The dynamics of air and of surface detritiation are studied for different rooms with specific wall coatings at a design ventilation throughput. Adsorption in the walls and on the surface of equipment is shown to be an important factor that may determine the level of air contamination and the time needed to detritiate the contaminated rooms. The role of different factors in the detritiation process is discussed. Measures are proposed that can be adopted at the design stage to mitigate the consequences of tritium release into the working rooms of the ITER and to reduce the time needed to decontaminate room air to a permissible level.

CAP88-PC, Version 2.0, and NORMTRI, which are based on the Gaussian plume model, were used to estimate the chronic dose due to a continuous, ground-level, atmospheric release of tritium as tritiated water. For the same conditions the predictions of CAP88-PC were found to be higher by a factor of 3 or less than those of NORMTRI. The major differences are due to the use of higher dose coefficients in CAP88-PC andNORMTRI's method of calculating the tritium content of food.

The computer program TriStat (Tritium dose assessment for stationary release) was used to estimate the human dose under stationary release and to obtain a conservative estimate of the dose after an accidental release as well. The atmospheric behavior of tritium is described by a Gaussian dispersion model. The tritium concentration in the atmosphere, soil, vegetables and cereals were estimated on the basis of tritium inventory of the facility and the release rate of tritium. In the model description, the specific tritium concentrations for the free water component and the organic component are essential. The food chain for humans was modeled by assuming a forage compartment, a plant compartment and an animal compartment. In the model, a virtual plant and a virtual animal were defined. The calculation revealed that the exchange of HTO between atmosphere and plant leaves has a critical role for increasing the human dose both for stationary and accidental release of tritium.

At the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI), the three-dimensional 'TBEHAVIOR' code has been developed and improved to understand initial tritium behavior and tritium confinement performance in a ventilated room of a fusion reactor in case of tritium leak event. The purpose of this study was mainly focused to; 1) investigate the effect of atmospheric exchange time per hour on the tritium confinement performance in an actual scaled tritium handling room after off-normal tritium release; 2) investigate the effect of atmospheric exchange time per hour on the time necessary for detecting tritium release; 3) investigate the suitable location of exhaust ducts and alarm monitors. The simulated room used in the present analysis is approximately 3000 m{sup 3} of tritium handling room (12.00 m{sup W}, 29.00 m{sup D} and 8.50 m{sup H}) with six supply ducts and six exhaust ducts. Atmospheric exchange time per hour is changed as a parameter from 0.67 to 3.33 times per hour.

A model to explain tritium release behavior from irradiated Li4SiO4, in the model reported so far by the present authors, it is required to use so small reaction rates for the surface reactions as one several thousandth of the observed values reported in the previous papers to get the good fitting. In this study the mass transfer resistance between grain surface and surface water is newly introduced because it is preferable to use the same reaction rate as that reported previously. The estimated values using the new model give good agreement with the observed tritium release curves and also with the release curves estimated using the model so far.

In Korea, Wolsong Tritium Removal Facility (WTRF) is scheduled to begin operation in 2005 to reduce the amount of tritium generated in the moderator and coolant. The objective of this study is to evaluate the environmental impact of tritium released from WTRF in the postulated accident. In order to achieve this, a computer code was developed at KAIST (Korea Advanced Institute of Science and Technology). This code can be used to evaluate the individual and public dose with the source term. This source term can represent not only the concentration of tritium that will be stored at the long term tritium storage vault located in the underground of WTRF building but also may be released to the environment from the WTRF online system by variously postulated accidents. To validate this code, calculated results were compared with the previous reference under the same assumption. Even if the most severe postulated accident that the tritium may be released through the fracture of the storage vault was occurred, the result of individual dose at the exclusion area boundary is turned out to be within the radiation dose limit.

This paper proposes a model to explain tritium release behavior of an irradiated Li4SiO4 sample made by Forschungszentrum Karlsruhe. The release curves were obtained in a series of experiments carr...

The atmosphere detritiation system (ADS) for ITER, which is a system provided to remove tritium in building and confinement atmospheres to secure the tritium safety for workers and public in both normal and off-normal conditions based on the ALARA principle, has been designed. In normal condition including maintenance activities, the ADS removes tritium in effluent gas and secondary confinement atmosphere. In the event of off-normal tritium release, the ADS, as a whole, maintains the tritium safety by mitigating the effect of the tritium release according to the operation scenario established based on the present design.

ETDOSE is a simple computer code for calculating distribution patterns of tritium in the environment for acute and chronic releases of HT and HTO. It calculates tritium concentrations in air, soil, plant free water and plant organic material, and estimates dose impacts from inhalation of air and ingestion of food. ETDOSE includes two different models for calculating HTO re-emission from the soil surface for a chronic tritium release. These two models were tested in the IAEA's model validation program BIOMASS (BIOsphere Modeling and ASSessment methods) 1 .

In order to demonstrate tritium confinement performance in DT (deuterium-tritium) fusion reactors at any accident, TPL/JAERI has carried out intentional tritium release experiments using Caisson Assembly for Tritium Safety study (CATS). Some data have been accumulated such as actual tritium behavior in the normal ventilated area and response of tritium monitors and emergency isolation valves. The experimental results of confinement of released tritium of about 3 GBq in 12m 3 of the Caisson under ventilation of 50 m 3 /h were consistent with those of simulation by a three dimensional eddy flow calculation code modified from FLOW-3D. Using this simulation code, the tritium confinement performance in large rooms can be demonstrated for actual fusion reactors.

At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate tritium behavior in the case where a tritium leak event should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak event should happen in a ventilated room. The RNG model was found to be valid for eddy flow calculation in the 50m3/h ventilated Caisson with acceptable engineering precision. The calculated initial and removal tritium concentration histories after intended tritium release were consistent with the experimental observations in the 50 m3/h ventilated Caisson. It is found that the flow near a wall plays an important role for the tritium transport in the ventilated room. On the other hand, tritium behavior intentionally ...

The paper presents the expediency of developing the system for tritium removal from large air volumes in working rooms. The system is intended both to prevent ejection of air contaminated by tritium in the atmosphere in emergency tritium release from processing equipment of the experimental closed loop modelling the ITER vacuum-fuel system, and to purify air to sanitary standards. Schematic layout and a design of the system elements are presented. Operation principle consists in timely detection of tritium leakage by control pickups, in locking and switching off the ventilation, as well as in air purification using gas purifying facility (AGPF). In the AGPF gaseous tritium oxidizes on alumopalladium catalyst to oxide, and produced ultraheavy water is absorbed by zeolites. The facility is composed of air heater, converter (column with catalyst), water cooler, two parallel adsorbers (columns with zeolite) and gas blower.

Since lithium-containing ceramics (e.g. Li 2 O, LiAlO 2 , Li 4 SiO 4 , Li 2 ZrO 3 , Li 2 TiO 3 ) are considered as breeding materials in the blanket of the next generation fusion reactors, several studies are in progress to evaluate their behaviour under irradiation in both operating and accidental conditions. Based on safety and economic considerations tritium inventory and release are the most critical issues for blanket concept. Investigation of tritium transport processes by using comprehensive physical-mathematical models is one of the current activities in this area. Although some analytical models and numerical methods dealing with tritium transport and release in fine-grained ceramic were already developed and applied to interpret results from in-situ and/or post-irradiation annealing experiments, it is necessary that presently available computer codes enlarge their range of applicability to be able to predict, with increased accuracy, the tritium release response for a wider range of experimental conditions and material characteristics. This paper reviews the tritium modelling activity and summarizes the existing transport models and computer codes highlighting models development and focusing on major changes and evolutionary improvements. 1 Validation of models by comparison of calculated results with experimental ones is also reported and discussed. Areas of future applications are identified and emphasis is placed upon the growing need of developing more accurate computer codes with the aim to improve the accuracy of blanket tritium inventory estimations.

The Replacement Tritium Facility (RTF) at the Savannah River Site in the United States was designed and built to handle kilogram levels of tritium. The RTF was started up in January 1994. All the design objectives were achieved. To minimize tritium release to the environment, the tritium handling process is installed inside nitrogen-atmosphere gloveboxes. Any tritium that might leak from the process to the gloveboxes is recovered by stripper systems. The tritium concentration in the gloveboxes is normally maintained at below 0.1 Ci/m{sup 3}. During a large tritium leak from the process to the glovebox, the stripper system lowered the tritium concentration in the glovebox from about 8,000 Ci/m{sup 3} to about 100 Ci/m{sup 3} in one hour. After that the tritium concentration decreased very slowly. It required 5 days of stripping before the concentration was down to about 10 Ci/m{sup 3}.

An analysis of the heat transport and water detritiation systems of ITER has been performed in order to determine major pathways for tritium loss and estimate releases during normal operation (operational tritium release). Heavy water escape and tritium release estimates compiled on the basis of operating experiences of typical CANDU PWR and the Darlington Tritium Removal Facility (DTRF) have been appropriately scaled on the basis of water and tritium inventories and tritium concentrations to fit ITER design and operating conditions. The paper estimates the chronic and acute tritium releases to the environment in elemental and oxide forms, via waterborne and airborne pathways of the ITER water systems. The results of the analysis will be used to demonstrate that the ITER design will meet the dose limits for occupational and accidental tritium releases. 5 refs., 5 tabs.

Regrowth in tritium desorption from type 316 stainless steel, copper and borosilicate glass was studied. It was found that the tritium, which was penetrated into materials by long term contact, could not be easily desorbed by a stream of nitrogen gas ( dry, wet or 10 % H 2 ) at room temperature. The ratio of the tritium amount desorbed from surface by each purging to the tritium concentration in the gaseous phase under the sorption /desorption equilibrium on the surface was found to be constant through the repeated desorption. The amount of tritium desorbed by each gas purging was found to decrease by repeated desorption.

The design of the ITER tritium systems must ensure that the public is protected from operational and accidental tritium release and that worker exposures are minimized. The ITER public exposure limits are consistent with international recommendations and the As Low As Reasonably Achievable (ALARA) concept is being implemented. We describe design approaches that will allow us to achieve these objectives. Although the design is still in its early stages, we have estimated the tritium flows and inventories and have started the safety analysis. Operational and accidental releases from ITER have not yet been determined, but we have used experience from other facilities to estimate ITER releases.

A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences frommore » short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.« less

This paper reviews the basic mechanisms governing tritium release rates from a spill of tritiated water and develops a mathematical model which can be used in these types of calculations. The model...

This paper reports that it is important in the design of future fusion reactors and associated facilities that incorporate passive safety to take account of the possible environmental impact of accidental tritium release. Reliable information on dose consequences can be obtained by evaluating urine samples from persons exposed to tritium. Translating the results of the environmental HT experiment performed in France in 1986 into worst-case exposure conditions, the effective dose equivalent to an individual with highest exposure at a distance of 800 m (typical for site boundaries) is {approximately}1 {times} 10{sup {minus}4} Sv per gram of tritium emitted as HT when inhalation and skin absorption are considered. From this value, maximum permissible amounts of accidentally released HT can be derived on the basis of regulatory or anticipated dose limits.

Tritium release from 1.5 mm diameter Li2ZrO3 spheres has been examined in post-irradiation annealing tests, to support the Canadian sphere-pac breeder-blanket concept. Rapid tritium release was obs...

Modeling of the tritium reaction at the surface of Li2O was performed using data obtained in in-situ and out-of-pile tritium release experiments. In this model the effects of H2O and H2 in the swee...

In the design of facilities which handle tritium, radiation safety assessments for various operational parameters and conditions are required. A computer-based system, entitled RAPTIAH, has been developed which performs the analyses and provides the fundamental information required for such assessment. RAPTIAH starts by estimating fluid leak rates from individual components. From the leak rates it derives the rate of tritium release in each area of the facility. Next, the airborne tritium concentrations in areas of the facility and the release of tritium to the environment are obtained. Finally, using component reliability data and the pattern of maintenance, the potential tritium exposure to occupational individuals is estimated. RAPTIAH may be used to identify design measures which would enhance radiation protection in tritium handling facilities.

In-situ tritium release behavior from Li/sub 2/O powder has been investigated under neutron irradiation at high temperatures. It was found that the chemical form of the released tritium depended strongly on the experimental conditions such as pre-treatment (dehumidification) of the sample and the H/sub 2/ addition to the sweep gas of helium. Water-soluble component was observed to be a main species from the sample pre-dried in N/sub 2/ stream at 800/sup 0/C for 3 days, while waterinsoluble component had fairly large proportion on the sample pre-dried for 5 days. In the case of He-4%H/sub 2/ sweep gas, more than 90% of released tritium was water-insoluble even for the sample pre-dried for 3 days. Tritium recovery rate was also accelerated by H/sub 2/ addition. These experimental results showed that the oxygen activity of the experimental system essentially affected to tritium release behavior from Li/sub 2/ sample.