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Results for “Tritium Purification”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m{sup 3}Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

SHINE Vacuum Pump Test Verification

Normetex pumps used world-wide for tritium service are no longer available. DOE and other researchers worldwide have spent significant funds characterizing this pump. Identification of alternate pumps is required for performance and compatibility with tritium gas. Many of the pumps that could be used to meet the functional performance requirements (e.g. pressure and flow conditions) of the Normetex pump have features that include the use of polymers or oils and greases that are not directly compatible with tritium service. This study assembles a test system to determine the flow characteristics for candidate alternate pumps. These tests are critical to the movement of tritium through the SHINE Tritium Purification System (TPS).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Logical Approach to Tritium Vacuum Pump Selection

Introduction Vacuum pumps are essential to the operation of tritium purification and recovery processes. However, not all vacuum pumps sold meet the requirements for tritium service. Manufacturers routinely do not fully appreciate the hazards associated with tritium, or the complex materials interactions. Therefore, a decision tree was established detailing which pump types, and certain manufacturers, are either fully compatible, limited use, or to avoid at all costs to establish a more structured and uniform approach to tritium vacuum pumps.

ANGELETTE, LUCAS↗
DOE OSTI2016

Performance testing of a prototype Pd-Ag diffuser

The fusion fuel cycle has gained significant attention over the last decade as interest in fusion programs has increased. One of the critical components of the fusion process is the tritium fuel cycle. The tritium fuel cycle is designed to supply and recycle process tritium at a specific throughput rate. One of the most important processes within the tritium fuel cycle is the clean-up of the of the process tritium. This step will initially separate the hydrogen isotopes (H2, D2, and T2) from the rest of the process gas using Pd-Ag diffusers or permeators. The Pd-Ag diffuser is an integral component for any tritium purification system; whether part of the United States’ defense mission or fusion programs. Domestic manufacturers of Pd-Ag diffusers are extremely limited and only a few manufacturers exist. Johnson-Matthey (JM) Pd-Ag diffusers (permeators) have previously been evaluated for the separation of hydrogen isotopes from non-hydrogen gas species in the process. JM is no longer manufacturing Pd-Ag diffusers and a replacement vendor needs to be identified to support future needs. A prototype Pd-Ag diffuser has been manufactured by Power and Energy, and is considered a potential replacement for the JM diffuser for tritium service. New diffuser designs for a tritium facility for any fusion energy applications must be characterized by evaluating their operating envelope prior to installation in a tritium processing facility. The prototype Pd-Ag diffuser was characterized to determine the overall performance as a function of the permeation of hydrogen through the membrane. The tests described in this report consider the effects of feed gas compositions, feed flow rates, pump configuration and internal tube pressure on the permeation of H2 through the Pd-Ag tubes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Evaluation of getters for methane and ammonia decomposition

Getters are used in many tritium purification systems for impurity removal. One getter material, SAES® ST909, can be used to decompose various impurities (methane, carbon oxides, ammonia and others), and release the hydrogen bound in these compounds for further processing. Because, ST909 absorbs eight to nine percent of its initial weight in nitrogen, alternate materials may provide greater operational flexibility to accommodate variable operating carrier gases. The gettering capability of ST909 is related to the tri-metallic alloy (zirconium, manganese, and iron) composition. However, simply using a material with the same composition may not provide similar methane and ammonia decomposition conversion efficiency. For this testing, alternative materials to ST909 are compared to track methane and ammonia decomposition performance. Bench scale tests were performed at elevated temperature using various gas mixtures. Changes in methane and ammonia decomposition performance were tracked using RGA analysis during bench scale testing. Analytical comparisons of chemically equivalent materials to ST909 were also performed and correlated to bench scale test results. Characterization testing performed to evaluate chemically equivalent materials included: X-ray fluorescence, X-ray diffraction, and inductively coupled plasma-mass spectrometry (ICP-MS). Results are expounded upon and differences in gettering capability of alternate materials are highlighted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Development of tritium technologies at KAERI

Korea has been operating a CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant is removed by the Wolsong TRF (Tritium Removal Facility) and measurement campaigns of tritium near the power plant have shown the efficiency of the TRF system. The HANARO reactor uses heavy water as both reflector and moderator. In HANARO the tritiated water removal system consists of compressors, condensers, and adsorption beds. A tritium behavior analysis code (TRIBAC) for a Very High Temperature Gas-Cooled Reactor (VHTR) is under development at KAERI. The TRIBAC computer software has been equipped with models for tritium production, purification, and leakage, as well as chemisorption and tritium behavior, in the hydrogen production system. Korea takes part into the ITER program and is responsible for the supply of an SDS (Tritium Storage and Delivery System). Within this program Korea has launched an experimental program to study the physico-chemical properties of metal and their hydrides in which hydrogen isotope gases can be stored and removed safely.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Application of computational fluid dynamics for the simulation of cryogenic molecular sieve bed absorber of hydrogen isotopes recovery system for Indian LLCB-TBM

One of the most challenging tasks in the design of the fuel cycle system lies in the effective design of Tritium Extraction System (TES) which involves proper extraction and purification of tritium in the fuel cycle of the fusion reactor. Indian Lead Lithium cooled Ceramic Breeder Test Blanket Module (LLCB-TBM) would extract hydrogen isotopes through Cryogenic Molecular Sieve Bed (CMSB) adsorber system. A prototype Hydrogen Isotopes Recovery System (HIRS) is being developed to validate the concepts for tritium extraction by adsorption mass transfer mechanism. In this study, a design model has been developed and analyzed to simulate the adsorption mass transfer kinetics in a fixed bed adsorption column. The simulation leads primarily to effective design of HIRS, which is a state-of-the-art technology. The paper describes the process simulation approach and the results of Computational Fluid Dynamics (CFD) analysis. The effects of different operating conditions are studied to investigate their influence on the hydrogen isotopes adsorption capacity. The results of the present simulation study would be used to understand the best optimized transport phenomenon before realizing the TES as a system for LLCB-TBM. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Evaluation of a cleaning method for the tritium environment

Solvents can be used to clean process equipment and piping that will come in contact with tritium containing gas streams; but, the ways in which solvents interact with catalysts is also of interest. This method of cleaning does not account for impurities within the bulk of a material that could be released from catalyst or containment barriers during high temperature operation. One solvent of interest which can be used because of its ability to both degrease and be analyzed for carbon content is Vertrel™ MCA. Tests were performed to determine carbon, sulfur, and chlorine concentrations for two candidate materials which were subjected to Vertrel MCA cleaning. Comparisons with high temperature hydrogen reduction are provided. Solvent analysis, LECO carbon/low sulfur determination testing, and neutron activation testing were performed to provide comparisons. Results show potential risks of using solvent cleaning and analysis as the only method of determining impurity concentration of materials which will be exposed to high temperature testing.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Separation of Hydrogen Using Pd/Ag Membranes: Experimental and Modeling Results with Potential Application to Direct Internal Recycle

Implementation of fusion energy requires processing the deuterium-tritium (D-T) mixture used to fuel the reaction, and separation of hydrogen isotopes from other gases is imperative. Specifically, the separation of hydrogen isotopes from helium is a matter of importance to the fusion fuel cycle community. Initial testing with a palladium-silver (Pd-Ag) membrane indicates that even moderate vacuum (~100 torr permeate pressure) can provide a high degree of separation (>90%) at a high ratio of H 2 to He. Given the presence of He in many fusion systems, a high technology readiness level (TRL) for Q 2 /He (where Q represents any isotope of hydrogen) separations is needed. This study demonstrates the efficacy of H 2 removal from He via permeation and potential applications for direct internal recycle. Modeling will accompany the experimental campaign to generate a predictive capability and quantify the separation performance. Modeling from previous hydrogen permeation studies has demonstrated that the typical Sieverts’ law fails to predict the measured permeation rates at high hydrogen fluxes. Existing models are being refined to integrate the effects of surface phenomena into permeation predictions, which have been expanded to account for mixtures with large ranges of Q 2 concentrations. These data will improve the TRL of permeators as a separation technology for the fusion fuel cycle.

08 HYDROGEN↗
DOE OSTI2019

CNL Technologies for Heavy Water Detritiation - 19437

Canadian Nuclear Laboratories (CNL) is responsible for a large amount of heavy water owned by the Government of Canada. A significant portion of this heavy water has been downgraded and is chemically and/or radiologically contaminated (besides tritium). Storage, monitoring and disposal of this water incurs significant costs. Upgrading following chemical purification (including removal of radionuclides other than tritium) of this water is feasible with commercial processes. However, removal of the tritium to the level that would allow the heavy water to be available for non-nuclear applications requires very specialized processes. CNL has developed and demonstrated the Combined Electrolysis and Catalytic Exchange (CECE) detritiation technology to reduce the tritium to the level at which the heavy water can be classified as non-radioactive. The CECE process consists of a Liquid Phase Catalytic Exchange (LPCE) bed, where the transfer of hydrogen isotopes between water and gaseous hydrogen occurs, and an electrolysis cell (E-cell) that produces hydrogen and separates the isotopes of hydrogen, adding to the enrichment achieved in the LPCE catalyst bed. The application of the CECE technology to the tritium-contaminated heavy water would restore its full market value. As part of a project supported by CNL Decommissioning and Waste Management, a business case is being developed to build and operate a facility that will purify and detritiate heavy water to render it commercially attractive. The main goal of the current study is to evaluate the net benefits of detritiation and product sale versus the cost of permanent storage or disposal. Such a facility would not only eliminate the need to store the heavy water as a radioactive waste material and provide revenue, but would also serve as a demonstration of CNL's technology for similar applications elsewhere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Investigations into the Gas Phase Reactions of Zirconium Tetrachloride (ZrCl) from UNF Cladding

Zirconium-based cladding materials that surround used nuclear fuel (UNF) rods represent a significant fraction of the total amount of high level waste that must be disposed of in long term geological repositories. The decontamination, recovery and reuse of these materials would significantly reduce the high level waste associated with light water reactors and could provide a new source of hafnium free zirconium for the fabrication of new nuclear components. A step in a strategy to recycle zirconium from used cladding materials is the high yield chlorination of Zircaloys to produce zirconium tetrachloride (ZrCl{sub 4}). The ZrCl, is contaminated with other metal chloride species that originate from the alloying metals as well as contaminants from oxide layers on the cladding, trace fission products, fuel residues, activation byproducts, and tritium. The properties of these complex mixtures must be better understood before purification protocols can be developed that will allow zirconium chloride to be recycled further. First order thermodynamic modeling of high temperature mixtures of ZrCl, and other metal chlorides in the gas phase may not accurately predict their properties. Computational modeling studies indicate that mixed metal adducts may form under the conditions of sublimation-based purification protocols. In light of these predictions, alternate strategies may also be needed to obtain pure ZrCl, from such mixtures. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Isotope exchange of ND 3 on Pt catalyst-loaded 13X molecular sieve

During D-T fusion operations the capture, purification, and recycling of unburned tritium will be crucial, as the formation of tritium containing molecules require additional processing. Here, removing the tritium can require costly processing to be unbound from the tritium-containing molecules and improvements to these processes will be necessary moving forward. Existing techniques using sorbent material beds to remove tritium-containing molecules from process gas streams undergo repeated high-heat cycling which leads to diminished bed lifespans, necessitating replacement and associated downtime. This work demonstrates the capture and isotopic exchange technique of deuterated ammonia (ND3), used as a surrogate for tritium, at ambient temperature using a Pt catalyst-loaded 13X molecular sieve. Unmodified 13X molecular sieve is capable of adsorbing and retaining the ND3 however, incorporation of a catalyst facilitates the isotopic exchange of the hydrogen isotopes. The effluent gas streams were analyzed in conjunction with desorbed ammonia isotopologues post-exchange to verify these results. Isotopically exchanging and removing heavier hydrogen isotopes using this technique provides an alternative to traditional removal methods.

08 HYDROGEN↗
DOE OSTI2015

Management of tritium European Spallation Source

The European Spallation Source (ESS) will produce tritium via spallation and activation processes during operational activities. Within the location of ESS facility in Lund, Sweden site it is mandatory to demonstrate that the management strategy of the produced tritium ensures the compliance with the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment were also analyzed. Calculations show that the annual release of tritium during the normal operations represents a small fraction from the estimated total dose. However, more refined calculations of migration of activated-groundwater should be performed for higher hydraulic conductivities, with the availability of the results on soil examinations. With the assumption of 100% release of tritium to the atmosphere during the occurring of the extreme accidents, it was found as well that the total dose complies with the constraint. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Combustion water purification techniques influence on OBT analysing using liquid scintillation counting method

In order to determine organically bound tritium (OBT) from environmental samples, these must be converted into water, measurable by liquid scintillation counting (LSC). For this purpose we conducted some experiments to determine OBT level of a grass sample collected from an uncontaminated area. The studied grass sample was combusted in a Parr bomb. However usual interfering phenomena were identified: color or chemical quench, chemiluminescence, overlap over tritium spectrum because of other radionuclides presence as impurities ({sup 14}C from organically compounds, {sup 36}Cl as chloride and free chlorine, {sup 40}K as potassium cations) and emulsion separation. So the purification of the combustion water before scintillation counting appeared to be essential. 5 purification methods were tested: distillation with chemical treatment (Na{sub 2}O{sub 2} and KMnO{sub 4}), lyophilization, chemical treatment (Na{sub 2}O{sub 2} and KMnO{sub 4}) followed by lyophilization, azeotropic distillation with toluene and treatment with a volcanic tuff followed by lyophilization. After the purification step each sample was measured and the OBT measured concentration, together with physico-chemical analysis of the water analyzed, revealed that the most efficient method applied for purification of the combustion water was the method using chemical treatment followed by lyophilization.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Local Gas Purification System in Spent Nitride Fuel Oxidation

A local gas-purification system of the voloxidation operation is examined within the framework of the technology of regeneration of spent mixed uranium-plutonium nitride fuel from fast reactors. The system includes a metal-fabric filter, HEPA filter, ruthenium capture unit, an adsorber for capturing iodine, a unit for additional oxidation and trapping of hydrogen (tritium), and a unit for capturing {sup 14}CO{sub 2}. A description is given of each unit, the overall scheme, and the possibilities of the gas-purification system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Results of Speciation Determination Measurements of Insulating Sulfur Hexafluoride Gas

Under the direction of the NNSA NA-231 Mo-99 Program, SRNL has provided technical assistance to U.S. companies under cooperative agreements. One such area of technical assistance has been for the recommendation for contamination management methods in the eventuality that tritium contaminates the insulating gas of a high-voltage ion source used in the accelerator-based process developed by SHINE Medical Technologies to produce Mo-99. The SHINE accelerator process uses tritium as the target of a deuterium ion beam, where the ion beam is produced within a high-voltage ion source insulated with sulfur hexafluoride (SF 6 ), and is separated from the tritium target by an extensive pump train. Normal operating conditions preclude the conditions necessary for contamination of the insulating gas, however certain atypical conditions could result in a tritium contamination event. Accelerator based processes for medical isotope production represent an anomalous tritium contamination challenge due to the presence of SF 6 . SF 6 decomposes through normal use as an insulator gas, and the exact speciation of the decomposition byproducts depends on a variety of factors. This creates a scenario where the anticipation of species present in the insulating gas mixture is difficult. In turn, the fate of tritium within such a chemical system is unclear. This document discusses the results of measurements made in-situ of the SF 6 insulating gas in use within a SHINE Medical Technology’s high-voltage ion source. This was done in an effort to determine the chemical speciation of the gas, allowing for the recommendation of management strategies in the event that tritium contaminates the insulating gas. The results of the measurements reported in this work suggest that tritium contamination in the insulating gas may result in two chemical forms of tritium in the system, tritiated water and molecular tritium. These two forms of tritium would be captured by two methods. Tritiated water may be captured and removed from the SF 6 by an adsorbent canister already present within the pressure vessel that houses the insulating gas while in use. It may also be separated from the SF 6 during routine purification of the SF 6 , which is performed by SHINE personnel routinely. The purification cart will perform all non-embedded systems cleanup necessary to remediate tritiated species in the system. The purification cart passes the insulating gas through several filters, followed by the liquefication of the SF 6 . The filters will separate tritiated water from the SF 6 at this stage. It is expected that molecular tritium will accumulate in the gas-phase during the liquification step, which is separated and stored. Tritium may be recovered from this mixture through more conventional hydrogen purification methods such as a palladium diffuser.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Decommissioning of the NPP A-1 heavy water management system - 15121

The paper deals with experience and techniques in the application of decontamination technology and remotely controlled robotic devices for the decontamination and dismantling of the heavy water management system during undergoing decommissioning process of the A-1 NPP as well as treatment of arising liquid waste All of these activities are characterized by high level of radioactivity and contamination. During operation of the A-1 NPP the heavy water management system fulfilled several functions: (1) cooling of the moderator, (2) topping up of the moderator in the active zone, (3) drainage, feed and storage of heavy water, (4) capture of leaked heavy water. Closely associated with heavy water management were systems for separation and isotopic purification of heavy water (distilling station, removing organic impurities station, isotopic purification station). The NPP heavy water system is located inside the power block building in several rooms where the inner surface contamination is up to the level of 10 E4 Bq.cm{sup -2}, dose rate up to 15,0 mGy.h{sup -1} and the feeding pipeline contained liquid RAW with high tritium content. Pre-dismantling decontamination of the heavy water management system and associated systems had to be performed with the intention of improving the radiation situation, i.e. reducing contamination of internal surface and dose rates in the vicinity of the equipment. Various sophisticated procedures such as retrieval of sludge using remotely operated devices and loop chemical decontamination of internal surfaces were applied. The generated radioactive waste such as spent decontamination solutions and sludge were immediately solidified on-site into a geo-polymer matrix using a mobile conditioning facility. For dismantling and fragmentation of some devices, mobile robotic systems MT80 and MT15 were applied which had been developed, designed and constructed as general-purpose decommissioning equipment. Special tooling was developed for application with the robots, such as hydraulic shears, circular saw, reciprocating saw, circular pipe cutter and a system for quick tool-change without direct intervention of the operators. Most of the operations are remotely controlled on basis of visual information from four cameras, with consistent radiation protection of the operators. The implementation of partial decommissioning tasks in such hard and unfavorable radiation conditions (moreover accompanied by generation of specific high-content alpha radwaste) has introduced in practice several new unique decontamination and dismantling techniques and new procedures for the effective on-site processing of generated secondary radioactive waste, including sludge and sorbents. These advantageous techniques, equipment and experience will be exploited in the continuation of the A-1 NPP decommissioning process which has been organized according to distinctively licensed phases and are also being applied in the decommissioning of other NPP in Slovakia as well as abroad. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Evaluation of Getters for Helium Purification

• Helium-3 is a vital commodity in both medical and defense applications • Ultra-low cryogenic coolant (MRI imaging, etc.) • Neutron detectors • Being investigated as possible fusion fuel • Managed and distributed by DOE-OS • New PDRD project exploring helium isotope separation • Limited supply • Only available from tritium decay • Prohibitively difficult to separate from helium-4 • SRS Tritium Facility provides a steady source

Angelette, Lucas M.↗