Tritium Control Research Articles

Tritium sorption by cement and subsequent release

Abstract In a fusion reactor or tritium-handling facilities, contamination of concrete by tritium and subsequent release from it to the reator or experimental room is a matter of problem for safe control of tritium and management of operational environment. In order to evaluate this tritium behavior, interaction of tritiated water with concrete or cement should be clarified. In the present study, HTO sorption and subsequent release from cement were experimentally studied. 1. (1) Sorption experiments were conducted using columns packed with cement particles of different sizes. From the analysis of the breakthrough curve, tritium diffusivity in macropores and microparticles were evaluated. 2. (2) From the short-term tritium release experiments, effective desorption rate constants were evaluated and the effects of temperature and moisture were studied. 3. (3) In the long-term tritium release experiments to 6000 h, the tritium release mechanism was found to be composed of three kinds of water: initially from capillary water, and in the second stage from gel water and from the water in the cement crystal. 4. (4) Tritium release behavior by heat treatment to 800°C was studied. A high temperature above 600°C was required for the tritium trapped in the crystal water to be released.

National Ignition Facility Environmental Protection Systems

The conceptual design of Environmental Protection Systems (EPS) for the National Ignition Facility (NIF) is described. These systems encompass tritium and activated debris handling, chamber, debris shield and general decontamination, neutron and gamma monitoring, and radioactive, hazardous and mixed waste handling. Key performance specifications met by EPS designs include limiting the tritium inventory to 300 Ci and total tritium release from NIF facilities to less than 10 Ci/yr. Total radiation doses attributable to NIF shall remain below 10 mrem/yr for any member of the general public and 500 mrem/yr for NIF staff. ALARA-based design features and operational procedures will, in most cases, result in much lower measured exposures. Waste minimization, improved cycle time and reduced exposures all result from the proposed CO2 robotic arm cleaning and decontamination system, while effective tritium control is achieved through a modern system design based on double containment and the proven detritiation technology.

The control of tritium in ETHEL

The operation of the European Tritium Handling Experimental Laboratory (ETHEL) will require the implementation of means and procedures for allowing tritium control within the facility. For that purpose, account must be taken of the particular characteristics of tritium, such as its high mobility, capacity to dissolve in materials, often limited precision when performing inventory measurements. This paper estimates the influence of these effects on the overall tritium balance in ETHEL. By employing available models for predicting tritium hold-up, it is estimated that three to four grams of tritium may potentially remain irreversibly fixed in various plant items of the standard laboratory infrastructure (exclusive of experimental circuits). On the other hand, the highest overall precision that may be attained with the present plant regarding inventory measurements is estimated to be of the order of few percent. On the basis of the above estimates, the allowable limits for the Material Unaccounted For (MUF) are discussed.

The Receipt, Storage and Distribution of Tritium in ETHEL

ETHEL, a new research facility located at the JRC-Ispra, is envisaged to be licensed for 37 PBq (1 MCi or 100 g) of tritium. A central service of laboratory operations is the overall control of tritium within the facility. The intended management of tritium has been developed on the basis of ETHEL’s general safety criteria and the requirements of typical experiments with regard to quantity, quality and frequency of tritium distributions. The present paper describes the method of supplying tritium to the laboratory, its accountancy and long-term storage, its distribution to individual experiments and, finally, its retrieval from experiments or the laboratory’s tritium recovery system.

Crucial issues on liquid metal blanket design

Typical design concepts of liquid metal breeder blankets for power reactors are explained and characterized. The major problems of these concepts are described for both water-cooled blankets and self-cooled blankets. Three crucial issues of liquid metal breeder blankets are investigated. They are in the fields of magnetohydrodynamics, tritium control and safety. The influence of the magnetic field on liquid metal flow is of special interest for self-cooled blankets. The main problems in this field and the status of the related R&D-work are described. Tritium permeation losses to the cooling water is a crucial issue for water-cooled blankets. Methodes for its reduction are discussed. An inherent problem of all liquid breeder blankets is the potential release of activated products in the case of chemical reactions between the breeder material and water or reactive gases. The most important issues in this field are described.

Some considerations on tritium control for the European ceramic BIT DEMO blanket

One of the objectives assigned to a DEMO blanket will be to limit tritium leakage to the steam circuit by permeation through the steam generator tubing of the tritium unavoidably present in the coolant. The control of these losses will require, on one hand, to minimize the amount of tritium reaching the coolant under both nominal and accidental conditions, and, on the other hand, to have an efficient tritium recovery process and chemistry control for maintaining at a low level the concentration in reduced tritiated species in the coolant.

Iter/tiber helium-cooled solid and liquid breeder test module designs and evaluations

This paper reports the scoping phase designs of the helium-cooled solid breeder and liquid breeder test modules proposed to be used in the International Thermonuclear Experimental Reactor (ITER). It also includes detailed test module development plans from which test tasks that are required for ITER were identified. Many of the identified tasks can be addressed by pre-ITER experiments, thus minimizing the capital and operating costs of ITER. Tasks that need to be addressed by ITER are in the areas of tritium self-sufficiency and control, structure and solid breeder thermomechanical behavior under neutron fluence, and system reliability.

Liquid Metal MHD Energy Conversion in Fusion Reactors

The purpose of the present work is to assess the viability of a novel approach to high-temperature fusion blanket power conversion - liquid-metal magnetohydrodynamics (LMMHD). The following power-conversion LMMHD PCSs and lithium-cooled blanket LMMHD PCSs. LMMHD PCSs might enable converting high-temperature blanket energy into electricity at efficiencies of the order of 60%. Being free of moving parts, at least in the high-temperature range, the LMMHD PCS technology appears to be better suited than PCS technology based on turbomachinery. Indeed, certain LMMHD PCSs could be designed to be completely static and practically hermetically sealed. This provides a simple system, with expected high reliability, facilitated maintenance, and tritium control. Moreover, in none of the LMMHD PCSs considered is there a need for lithium-to-water heat exchange, thus avoiding safety and tritium handling difficulties.

Conceptual Design of A Self-Cooled FLiBe Blanket

There are a number of attractive features for Flibe, such as: (1) low-pressure operation, (2) relatively inert to air and water, (3) low activation, (4) high-temperature stability, and (5) low electrical conductivity. If these features can be incorporated into a self-cooled blanket, a simple, high-performance, and reliable blanket may be possible. Detailed analysis has uncovered some attractive features and some potential problems. A highly compact and efficient primary loop can be developed with a conversion efficiency of 46%. The blanket is designed to be inherently safe for at least a 4.7 MW/m/sup 2/ neutron wall loading. Since no reflector is needed, the mass utilization is very high. A similar blanket for MINIMARS has a mass utilization figure of 138 kW(e)/tonne, the highest among the four blanket candidates. In addition, a very simple blanket can be designed. There is, however, concern about the lithium doping idea for tritium control. The issues cannot be resolved until experimental verification is available. However, if a vanadium alloy is used for the structural material, a large fraction of tritium will permeate back through the first wall to the plasma. Permeation through the first wall will result in a tritium partial pressure of < 10/sup -4/more » Torr. If this is the case, the primary blanket tritium recovery system can be combined with the plasma exhaust system. A highly efficient tritium cleanup system will still be required to reduce the tritium permeation problem. A countercurrent extraction system can be used for tritium cleanup.« less

Tritium Control in NET: Preliminary Design Considerations

Tritium Control in NET: Preliminary Design Considerations

Tritium Control in a Helium-Cooled Ceramic Blanket for a Fusion Reactor

The Karlsruhe Nuclear Center is studying the conceptual design of a helium-cooled poloidal ceramic blanket with beryllium multiplier for the Next European Torus. To maintain the tritium losses from the plant below 10 Ci/day, a separate helium purge flow is required. In the case of a high-pressure purge flow, tritium containment requires an oxidizing atmosphere in the purge flow region. The consequences of the resulting T/sub 2/O partial pressure on Li/sub 2/ O are assessed. A purge flow at 1 bar does not require an oxidizing atmosphere.

Tritium Control in Helium-Cooled Blankets

Tritium Control in Helium-Cooled Blankets

Tritium Monitoring — Canadian Experience and Instrumentation

The control and monitoring of tritium are important aspects of the Canadian heavy water based nuclear power program. The development of tritium monitoring in Canada is discussed with particular attention being paid to experience gained at the Chalk River Nuclear Laboratories and at Ontario Hydro CANDU nuclear generating stations. Presently a coordinated tritium monitoring development program is underway to provide improved tritium monitoring instrumentation to meet the identified needs of fission reactors, tritium handling facilities, and fusion reactors. This program is outlined with some particular requirements of fusion reactors being discussed in detail.

Experimental Design to Study Tritium Oxidation as a Means of Tritium Control

The design of an experiment for determining the kinetic rate of the tritium oxidation reaction when tritium atoms are present is described. Determination of the reaction rate will be done by analyzing the ratio of T2 and T2O product as a function of time, temperature, tritium concentration, and oxygen concentrations. A model of a proposed surface reaction is given.

Tritium control in helium-cooled blankets

As a part of the Blanket Comparison and Selection Study (BCSS), GA Technologies was responsible for the design of helium-cooled, solid- and liquid-metal breeder blankets. Conceptual blanket designs were developed, including the consideration of the generation, transport, and extraction of tritium. Evaluations were made of the inventory and leakage of tritium for helium-cooled Li/sub 2/O and LiAlO/sub 2/ and liquid lithium breeder blankets for tokamak and tandem mirror reactors. To facilitate the evaluation, a solid breeder tritium code TRIT4 was developed. The results from this study indicate that tritium inventories and leakages are acceptable for the proposed helium-cooled blankets. An assumption made in the tritium leakage calculations was that tritium is released to the helium purge and coolant streams as T/sub 2/ and remains in that form. If oxidation to T/sub 2/O is possible, significant reduction in the tritium leakage will be possible. We conclude that more experimental data on breeder material properties and tritium permeation behavior are needed. However, we are certain that an adequate number of different techniques are available to control the breeder tritium inventory and leakage to an acceptable level in helium-cooled solid- and lithium-breeder blankets.

Tritium monitoring - Canadian experience and instrumentation

The control and monitoring of tritium are important aspects of the Canadian heavy water based nuclear power program. The development of tritium monitoring in Canada is discussed with particular attention being paid to experience gained at the Chalk River Nuclear Laboratories and at Ontario Hydro CANDU nuclear generating stations. Presently a coordinated tritium monitoring development program is underway to provide improved tritium monitoring instrumentation to meet the identified needs of fission reactors, tritium handling facilities, and fusion reactors. This program is outlined with some particular requirements of fusion reactors being discussed in detail.

An examination of tritium oxidation as a means of tritium control

An examination of tritium oxidation as a means of tritium control

Tritium Technology Review

The methods used in recovering, containing and controlling tritium in fusion reactor systems are the subject of the paper. Safety, cost, and breeding needs for tritium control are outlined.

Tritium inventory and permeation in TFTR

The control of tritium in TFTR has both important safety and environmental implications. Current modelling techniques allow realistic predictions of the tritium inventory in and permeation through in-torus components. The source of the tritium was computed using a three-dimensional description of the neutral particle flux on the TFTR vacuum vessel wall from the neutral transport code DEGAS, under plasma conditions modelled with the one-dimensional transport code BALDUR. The movable limiter (graphite) was modelled using an extension of the Local Mixing Model for hydrogen retention and isotope exchange. In particular, the calculations consider the inventory and recycling for the movable limiters (which are expected to experience transient temperatures up to 2000°C) as well as the bumper limiters and protective plates which will remain somewhat cooler. Transient effects in the bulk graphite limiter and in the stainless steel wall were modelled using the DIFFUSE code with materials parameters taken from the most recent literature. It appears that for the expected material properties, specifically the hydrogen recombination constant measured in a clean TFTR environment, and operating scenario, there will be essentially no tritium permeation through the stainless steel walls or bellows and less than 0.1 kCi of tritium inventory in the stainless steel first wall. The limiters may contain as much as 5 kCi of tritium.

Oxidation of tritium gas in glovebox atmosphere using CuO column.

A multiple barrier containment concept for tritium control is proposed to provide minimal release to the environment. The rationale for this design is the prevention of the dilution of tritium within each barrier so that the tritium can be recovered before it penetrates to the outer barrier. Oxidation performances of CuO sponge column in the inert atmosphere of glovebox system are discussed and following results are obtained. The oxidation column efficiency or KFaVcat/Q plays an important role in the tritium level decrease on the glovebox atmosphere. The recommended value of the column efficiency is 20–100. The isotope effect in chemical oxidation of H2, D2 and HT is 1, 1/6 and 1/12, respectively. Use of CuO sponge for the atmosphere in the presence of oxygen is not recommended because of the retardation of chemical oxidation rate.