Tritium Containment Research Articles

Containment of Tritiated Water Vapor by Concrete Walls of Fusion Reactor Building.

The diffusion of tritiated water vapor into concrete walls of a fusion reactor building is studied to evaluate the capability of the tritium containment of concrete materials. First, depth profiles of tritiated water in concrete are calculated to evaluate the capacity of the tritium containment by the sound concrete without cracks, and a 0.5-m-thick concrete wall is sufficient to prevent tritiated water releases to the environment in a normal operation of a fusion reactor over 50 yr. Second, simulations of the cleanup of tritiated water in the concrete reactor hall atmosphere taking into account the soaking are performed. Concrete porosity should be decreased to shorten the cleanup time of the reactor hall atmosphere. Surface coatings on the concrete, which apparently decrease the surface porosity, are effective measures to prevent the diffusion of tritiated water vapor into concrete during accidental releases.

Thermal, Fluid Flow, and Tritium Release Problems in Fusion Blankets

AbstractThe thermal, fluid flow, and tritium release problems represent a major part of the research and development issues for liquid-metal and solid breeder blankets. The issues are characterized, and recent progress in model development and experiments is described. The issues for liquid-metal blankets are dominated by magnetohydrodynamic (MHD) effects on fluid flow and heat transfer. New design solutions have been proposed, new facilities have been constructed, and models are being developed to address MHD effects. The problems of solid breeder blankets are dominated by tritium release, containment, and inventory. There has been remarkable progress in understanding tritium transport in solid breeders because of successful irradiation experiments of solid breeders in fusion reactors. New models on tritium transport have provided key tools for the analysis and interpretation of the experimental results, A new issue related to thermal control in solid breeder is the subject of active experimental and mod...

A Helium-Cooled Solid Breeder Concept for the Tritium-Producing Blanket of the International Thermonuclear Experimental Reactor

The usefulness of the tritium-producing blanket in the International Thermonuclear Experimentall Reactor (ITER) to the fusion research and development program can be maximized by selecting design parameters, features, and options that are reactor relevant without significantly increasing the risk in key areas such as device safety and operational reliability. For that reason, a helium-cooled solid breeder (SB) blanket is proposed since it combines the operation of the SB at high reactor-relevant temperatures with the operation of helium at moderate temperature and pressure to minimize risk. Results of the analysis done for this blanket concept indicate that it is very attractive. It can achieve a high tritium breeding ratio without breeding in the space-limited inboard region, It offers important safety features, including the use of inert gas with no chemical reaction or corrosion, low activation SB, and multiple containment of tritium. the concept provides great operational flexibility to accommodate changes in ITER operating parameters, such as power level, and to optimize the operating temperature of the structure. A novel and practical concept is proposed for the thermal resistance gap between the coolant and SB to allow their operating temperatures to be optimized.

Neutronics Analysis of the Laboratory Microfusion Facility

The radiological safety hazards of the experimental area (EA) for the proposed Inertial Confinement Fusion (ICF) Laboratory Microfusion Facility (LMF) have been examined. The EA includes those structures required to establish the proper pre-shot environment, point the beams, contain the pellet yield, and measure many different facets of the experiments. The radiation dose rates from neutron activation of representative target chamber materials, the laser beam tubes and the argon gas they contain, the air surrounding the chamber, and the concrete walls of the experimental area are given. Combining these results with the allowable dose rates for workers, we show how radiological considerations affect access to the inside of the target chamber and to the diagnostic platform area located outside the chamber. Waste disposal and tritium containment issues are summarized. Other neutronics issues, such as radiation damage to the final optics and neutron heating of materials placed close to the target, are also addressed. 16 refs., 2 figs., 1 tab.

Tritium-producing blanket for fusion engineering facility

The usefulness of ITER to fusion development can be substantially increased if the design parameters and features are selected so as to maximize reactor relevance without significantly increasing the risk. One critical area in which reactor relevance and risk are major issues is the tritium-producing blanket. A helium-cooled solid breeder blanket is found to offer the highest potential for a low-risk reactor-relevant blanket that can meet all the ITER requirements. It has many attractive features that include: (1) a large design margin to account for uncertainties in predicting the effects of the fusion environment; (2) an available data base that is rapidly expanding worldwide; (3) flexibility in accommodating changes in the operating parameters of ITER (e.g. fusion power and wall load); (4) solid breeder operation at reactor relevant temperatures and flexibility in setting the helium temperature to reduce risk and/or optimize the structure temperature while keeping the helium pressure at a modest level; and (5) safety features, including the use of inert gas with no chemical reaction or corrosion, a low activation solid breeder, and multiple containment of tritium.

The role of a blanket tritium system on the fusion fuel cycle

Abstract The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have an important impact on the operation of the fuel processing system. For some blankets, such an aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system in terms of component size and power consumption. The importance of the blanket interface to a fuel processing experiment, such as TSTA, has been identified. The initial work to define the blanket processing system, which is proposed to be added as part of TSTA, will be discussed here.

Tritium Proof-of-Principle Injector Experiment

The Tritium Proof-of-Principle (TPOP) pellet injector was designed and built by Oak Ridge National Laboratory (ORNL) to evaluate the production and acceleration of tritium pellets for fueling future fusion reactors. The injector uses the pipe-gun concept to form pellets directly in a short liquid-helium-cooled section of the barrel. Pellets are accelerated by using high-pressure hydrogen supplied from a fast solenoid valve. A versatile, tritium-compatible gas-handling system provides all of the functions needed to operate the gun, including feed gas pressure control and flow control, plus helium separation and preparation of mixtures. These systems are contained in a glovebox for secondary containment of tritium. Tritium experiments will be carried out at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL).

Apparatus for verification of beta heating driven layer uniformity in solid deuterium–tritium

A solid deuterium–tritium fuel layer within a spherical inertial confinement fusion target will be driven uniform by β‐decay energy if the target is held in an isothermal environment. The apparatus we constructed to verify this process provides an isothermal and radiation‐tight environment. Heat exchange between the target (≂20 K) and the environment (4.2 K) is regulated by controlling the helium exchange gas through a specially constructed manifold. A unique optical system maintained at low temperatures allows direct observation of the fuel layer uniformity. Tritium containment in the event of a target failure is assured by tungsten–inert gas welding of the stainless‐steel structure. This system conveniently fits a standard vendor supplied 100‐l Dewar, and is designed to minimize boiloff and cooldown losses by means of an efficient helium vapor counterflow system. We have also incorporated a vibration isolation system to permit holographic interferometry imaging and evaluation of the fuel layers.

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.

Cat-D fueled reversed-field pinch reactor assessment

A quantitative comparison of the technology requirements, environmental and cost issues of DD. Compact Reversed-Field Pinch Reactor (CRFPR) relative to a DT/CRFPR has been performed. The first wall/blanket energy recovery cycle for the DD reactor is simpler and more efficient than the DT reactor. In other technology areas (such as magnets and vacuum systems) DD requirements are not significantly different than the DT reactor. Tritium technology for processing the plasma exhaust is required to DD reactors, but no tritium containment around the blanket or heat transport system is needed. Safety analysis shows similar consequences for the release of activated corrosion products or activated first wall/blanket structure. Consequences of all postulated DD accidents for tritium releases are significantly smaller than those from the DT reactor. Cost studies have been performed for a series of DD reactors and compared with the DT reactor.

An Assessment of Problems Associated with Tritium Containment

A special task group was established as a part of the Blanket Comparison and Selection Study to investigate the problems associated with tritium containment in blanket systems. The goal of the task group was to review the existing experimental results and to define a set of ground rules that would form a common basis for each design group to resolve this particular problem. The work of the task group is summarized. Due to the scarcity of information, and often contradictory nature of the existing information, note that the conclusions are based on the best judgment of the task group. It is recommended that the conclusions be updated as more experimental results become available.

A Low-Cost Tritium Handling Facility

The tritium facility at KMS Fusion, Inc. supports our Inertial Confinement Fusion research program. The main function of the facility is to fill glass or polymer shells to a maximum pressure of 100 atm. with tritium (T/sub 2/) or deuterium-tritium (DT). The apparatus and procedures used in this process are described including provisions for tritium containment and monitoring. An estimate of equipment costs is given.

A Comparison of Li and 83Pb-17Li Primary Coolant Designs for the Pulse*Star ICF Reactor

Two different primary coolants, Li and 83Pb-17Li, have been examined for use in Pulse Star, a pooltype inertial confinement fusion reactor, and a balance-of-plant design has been generated for each coolant. The use of 83Pb-17Li eliminates concern about the large amount of stored chemical energy found in pure Li fusion reactors. A secondary loop was not included in the 83Pb-17Li coolant design because of the relative nonreactivity of lead-lithium. The design utilizing Li as a primary coolant includes a sodium secondary loop to prevent direct contact between irradiated Li and high-pressure water in the case of a steam generator leak. The secondary loop requires additional piping, pumps, heat exchanger area, and steam generator buildings. These additional costs are mitigated by the low pumping power requirement of Li compared with that of high-density 83Pb-17Li. A cost analysis revealed that the additional costs of the Li coolant design are only slightly greater ($13.5 million) than the cost savings due to the lower pumping power. Preliminary studies indicate that tritium containment will be more costly for the 83Pb-17Li coolant design than for the one involving pure Li because the insolubility of tritium in 83Pb-17Li creates large driving forces for tritium leakage into themore » surrounding plant. The tradeoff between the two safety concerns of chemical stability in the case of 83Pb-17Li and practicable tritium containment in the case of pure Li needs to be investigated.« less

Counter Current Extraction System for Tritium Recovery from 17Li-83Pb

A counter current extraction system is proposed here to recover tritium from 17Li-83Pb. The multiple stage extraction system is used to reduce the resistance to the molecular diffusion in the liquid phase, while a counter current helium purge gas is used to reduce the resistance in the gas phase. To reduce the purge gas requirement and alleviate the tritium containment problem, hydrogen is added to the purge gas. An isotope separation system is designed to separate the hydrogen from the tritium.

Modeling tritium containment in high temperature fusion reactors

Leakage of tritium resulting from permeation through surfaces such as those of the heat exchanger is a problem of special concern to fusion reactors. In this paper permeation rates are evaluated and their impact on the design of a fusion power system is assessed.

A Comparison of All-Remote and Personnel Access Maintenance Operations for INTOR

The INTOR reference design was developed to permit limited maintenance operations external to the reactor in a “hands-on” mode; all internal operations would be remotely accomplished. The design embodies those requirements for shielding, tritium containment and cleanup, and confinement of contaminated particulate matter to permit personnel access. The cost reflects these requirements, at least to first order. The impact of personnel access on the reactor design and its cost are cause to reexamine the maintenance approach on which much of the present configuration is based. The purpose of this study is to compare the benefits and costs associated with personnel access maintenence procedures to those associated with all-remote maintenance procedures and to identify modifications to the baseline design that could enhance maintenance operations.

Authors

Authors

A low cost tritium handling facility

The tritium facility at KMS Fusion, Inc. supports our Inertial Confinement Fusion research program. The main function of the facility is to fill glass or polymer shells to a maximum pressure of 100 atm. with tritium (T/sub 2/) or deuterium-tritium (DT). The apparatus and procedures used in this process are described including provisions for tritium containment and monitoring. An estimate of equipment costs is given.

Nuclear technologies?Their role and future impact

TRW briefly treats the unresolved nuclear feasibility issues, and then discusses the requirements and timing of a fusion nuclear technology program aimed at resolving these feasibility issues. A table summarizes what the author believes are the key nuclear technologies and their feasibility. The feasibility issues associated with tritium breeding, recovery, and containment are unique to fusion and their resolution will significantly impact the feasibility and attractiveness of fusion as an energy option. Options for doing low fluence integrated testing in the 2990's is focused upon. The Argonne National Laboratory discusses the ongoing activities in materials and blanket technology, and the fact that they have progressed well, both in focusing on the near-term needs, particularly in the plasma-wall and high heat flux area, and in the longer term by concentrating on major feasibility issues.

D-d Tokamak Reactor Assessment

A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases.