Detritiation Factor Research Articles

JET exhaust detritiation system replacement—design, commissioning, and operation

The exhaust detritiation system (EDS) of the Joint European Torus (JET) active gas handling system removes tritium from exhaust gases and contaminated air prior to discharge to stack. The system was replaced prior to JET’s tritium experimental campaign due to corrosion damage. In this paper, the following topics are presented: system function and description, corrosion and failure of the previous EDS, enhancements for the replacement EDS including testing of acid passivation column, commissioning experience, operating experience and performance. Acids including hydrofluoric were generated in catalytic recombiners following inadvertent halogen ingress. Pitting corrosion was identified with breakthrough failure of condensate pipework. To mitigate a recurrence, a novel acid passivation sub-system was designed. The acid passivation testing demonstrated successful performance of the design and calcium-hydroxide passivation packing tested, with acids below the limit of detectability at the test column outlet. Issues encountered during commissioning are described and key performance metrics such as detritiation factor reported. Finally, operational issues and system performance is presented.

Performance of exhaust detritiation system for a fusion test device in the initial phase of the operation

Abstract The exhaust detritiation system (EDS) has been operating for the deuterium plasma experiment in large fusion test device since 2016. The EDS consists of two systems: the molecular sieves (MS) type for plasma exhaust gas, the polymer membrane (PM) type for vacuum vessel purge gas during the maintenance activity. The tritium removal performance of the EDS was evaluated for four years. As the operation results, the maximum detritiation factor (DF) of the MS system was achieved to be more than 103 even though the average tritium concentration in the process gas was less than 104 Bq/m3. On the other hand, the DF of the PM system was less than 103, because the average tritium concentration during the vacuum vessel maintenance activity was at the order of less than 103 Bq/m3. Throughout the initial phase of the deuterium plasma experiment, the tritium concentration at the outlet of EDS was maintained to be less than the order of 102 Bq/m3. Also, a failure analysis was performed to evaluate the reliability of the EDS in the initial phase of the operation. Over the four years of operation, the overall failure rate of the MS system was on the order of 10−5 events/hour.

Initial operation results of exhaust detritiation system using a polymer membrane

An exhaust detritiation system using commercially available hollow fiber-type polyimide membrane modules, a PM system, was installed and applied for tritium recovery in the vacuum vessel purge gas of a large fusion test device. The PM system is operated annually and there were no serious malfunctions after starting the operation for approximately 2.5 y. The continuous recovery of tritiated water without the heating operation and the switching of valves as well as the regeneration operation of molecular sieves, was demonstrated. The tritium in the process gas was successfully recovered by the PM system even though the average tritium concentration at the inlet of the PM system was less than the order of 10−3 Bq/cm3. The detritiation factor and tritium recovery ratio were greater than 103 and 0.97, respectively. The tritium concentration at the outlet of the PM system was maintained in the order of 10−6 Bq/cm3 irrespective of the tritium concentration at the inlet of the PM system. Thus, the operating data indicates that the PM system has a potentially higher detritiation factor than DF = 103.

Analysis of the Transient Regimes of a Detritiation Facility Operation

The National Research and Development Institute for Cryogenics and Isotopic Technologies (ICSI Rm. Valcea) was established in 1970 as a research entity with the main goal of developing a heavy water production technology. One of the departments of the institute, entitled ICSI Nuclear, is oriented toward the development of a heavy water detritiation technology with the main beneficiary being the Cernavoda nuclear power plant. Thus, ICSI Nuclear has completed the technical project of the Cernavoda Tritium Removal Facility (CTRF). This paper focuses on the CTRF front end (water detritiation system) and will analyze achievement of the nominal regime at start-up and analysis of the transient regimes that may occur during normal operation and their impact on the water detritiation factor. Analysis results can become input for improvement of the CTRF control and instrumentation system, for subsequent operating facility procedures, and for verification of the isotopic exchange software models corresponding to these modes of operation.

Cernavoda tritium removal facility: A key tritium supplier for future fusion facilities

National R&D for Cryogenics and Isotopes Separation (ICSI) completed the conceptual design of the Cernavoda Tritium Removal Facility (CTRF) in 2015. CTRF is located at CNE11Centrala Nucleara Electrica - NPP Cernavoda and is sized to process 40 kg/h heavy water from 2 CANDU reactors, with a detritiation factor of 100, over 40 years design life. CTRF removes tritium using liquid phase catalytic exchange (LPCE) paired with cryogenic distillation (CD). The design of CTRF uses expertise from ICSI (Pilot Plant for Tritium and Deuterium Separation) and from Canada (Kinectrics), together with experience from Wolsung TRF project, research tritium laboratories and industry. The CTRF design uses the most advanced TRF technology, including recent safety requirements for a tritium industrial facility as specified by the Romanian Regulator.Construction of CTRF is planned to start in 2020 and the detritiation of heavy water from first reactor is scheduled to begin in 2026. After this date tritium stored will become available to be used for fusion research and industrial facilities. First phase of detritiation process is to reduce the moderator tritium content from 65 Ci/kg to 10 Ci/kg, second phase to maintain the tritium concentration up to 10 Ci/kg and third phase to use CTRF to reduce tritium content as low as possible before decommissioning of the site.This paper will present the technologies used, provide a prediction of tritium production, and considerations for possible use of CTRF for supplying He-3, as a by-product of the detritiation process.

Methodology to identify appropriate options to manage tritiated waste

Fusion facilities will use tritium as a part of fuel. As a consequence, most of waste is expected to be contaminated with tritium and specific waste management strategies need to be developed taking into account physical and chemical properties of tritium. In France, the reference strategy for waste exceeding acceptance limits for disposal due to its tritium inventory is an interim storage period up to 50 years, corresponding approximately to detritiation factor of 16 due to radioactive decay.This article is mainly focused on the most tritiated categories of waste that would require an interim storage period longer than 50 years. Alternative strategies or additional processing options have been sought for these categories and the article defines methodology of identification of the Best Available Technique (BAT) for reduction of tritium inventory in the waste. Furthermore, the methodology is demonstrated on the case of purely tritiated metallic waste made from stainless steel. The methodology has been divided into three main steps: 1) Identification of the most relevant detritiation techniques; 2) Comparative analysis; and 3) Economical assessment. In order to simplify the first step, a software tool has been developed by CEA and its main features are described in this article. The results given by this tool are then analysed in the second step, which takes into account several non-economic criteria. Results of the comparative analysis show that detritiation of purely tritiated metallic waste made from stainless steel appears to be very attractive, allowing a significant reduction of interim storage duration before disposal in a surface repository. The last step, briefly described in this article, is ongoing and will economically assess the results from the previous steps.

Enhanced configuration of a water detritiation system; impact on ITER Isotope Separation System based cryogenic distillation

Tritiated water is generated in the ITER systems by various sources and may contain deuterium and tritium at various concentrations. The reference process for the ITER Water Detritiation System is based on Combined Electrolysis Catalytic Exchange (CECE) configuration. During long time operation of the CECE process, the accumulation of deuterium in the electrolysis unit and consequently along the Liquid Phase Catalytic Exchange (LPCE) column is unavoidable with consequences on the overall detritiation factor of the system. Beside the deuterium issue in the process, the large amount of the tritiated water with tritium activity up to 500Ci/kg in the electrolysis cells is a concern from the safety aspect of the plant.The enhanced configuration of a system for processing tritiated water allows mitigation of the effects due to deuterium accumulation and also reduction of tritium inventory within the electrolysis system. In addition the benefits concerning to the interface between the water detritiation system and tritium recovery based cryogenic distillation are also presented.

Effect of Residual Tritiated Water on Air Detritiation Dryer Packed with Silica Gel

For a practical air detritiation system, the effect of the residual tritiated water left on the silica was demonstrated. The residual tritiated water significantly reduced the detritiation factor of the air detritiation dryer. To maintain the detritiation factor more than 350, the space velocity lower than 140 h-1 should be selected for the design and the operation of an air detritiation dryer packed with silica gel. The HTO/H2O separation factor evaluated using a dryer with residual tritiated water was 1.17 on the average. Silica gel is superior to commercial molecular sieves in dehydration ratio at room temperature. Dehydration from silica gel was dependent on the accumulated amount of purge gas, independent on flow rate. Dehydration from silica gel showed a slight pressure dependency. Consequently, silica gel is a promising desiccant agent for air detritiation dryer.

Development of tritiated vapor absorbent applicable to the atmospheric detritiation system in a nuclear facility

The combination of hydrogen oxidation reactor packed noble metal catalysts and water vapor absorber has been applied to the atmospheric detritiation system of the tritium handling facility. Commercial synthetic zeolite such as molecular sieve 5A has been used as an adsorbent of ADS absorber. In the case of application of molecular sieve 5A to the ADS absorber of a large-scale tritium handling facility such as a future fusion plant, an absorber becomes huge due mainly to the difficulty in dehydration from molecular sieve 5A. Hence, application of CaY Faujasite-type zeolite with a high framework silica-to-alumina ratio to the adsorbent for atmospheric detritiation system was investigated. It was clear that the dehydration behavior at room temperature was significantly improved using CaY zeolite. In contrast, detritiation factor for CaY zeolite with a high framework silica-to-alumina ratio depended strongly on the space velocity through the absorber. To apply CaY zeolite with a high framework SiO(2)/Al(2)O(3) ratio to the ADS absorbent, the space velocity less than 250h(-1) was recommended to maintain the detritiation factor more than 1000. The steep increase in water adsorption at the relative pressure lower than 0.05 is a feature of synthetic zeolite with calcium cation. However, such an increase was not observed in water adsorption on CaY zeolite with a framework SiO(2)/Al(2)O(3) ratio more than 7.0. Consequently, the CaY zeolite with the framework SiO(2)/Al(2)O(3) ratio of 5.0 is a promising candidate as absorbent of ADS absorber.

Optimisation of operating parameters for plasma facing components detritiation

Within the framework of management of waste produced by fusion facility, the presence of high tritium content in large pieces is very specific. This question is very sensitive in the case of maintenance of plasma facing components (PFC). In order to reach several goals (tritium recovery, limitation of tritium content in the waste), several treatment methods have been analysed and thermal detritiation is selected as a reference process. In the case of ITER facility a tritium recovery station based on this technique is foreseen. In order to better understand which parameters have important influence on the efficiency of the treatment (detritiation factor), a recall of the process principle and of the main data concerning tritium behaviour in different materials is done. For example, low tritium diffusion coefficient in Be which is the material of the first layer on some components can limit drastically the process performance. In this paper a recall of the different process parameters is presented. Some important parameters are selected in order to lead a sensitivity study on adequate variation domains. It allows studying several scenarios of treatment and analysing the consequences of them. The main parameters of this study are the level of temperature, the duration of treatment, the renewal rate of the furnace atmosphere, the presence of Be or not. The main response to this sensitivity analysis is the detritiation factor, however some other consequences such as tritiated water flow rate and also tritium concentration in water are evaluated. Then the consequences of change of operating parameters are compared in order to identify what could be the most appropriate operating domain.

Heavy water purification from tritium by CECE process

The experimental industrial plant for hydrogen isotope separation on the basis of the Combined Electrolysis and Catalytic Exchange (CECE) process is operated trouble free over a period of more than 10 years in Petersburg Nuclear Physics Institute. In parallel with carrying out the research of CECE process the plant is used for reprocessing heavy water wastes contaminated by tritium; several tons of reactor quality heavy water have been conditioned. The plant was updated to provide high factor of heavy water detritiation. Detritiation factor of about 2000 was achieved and heavy water with reduced contents of tritium less than 10 5 Bq/kg was withdrawn at the rate of 7–9 kg/day as a top product. The experience gained during the plant operation shows the high efficiency of hydrogen isotope separation with CECE process. This process can be used for tritium and protium removal from reactor heavy water.

Improvement of PNPI Experimental Industrial Plant Based on CECE Process for Heavy Water Detritiation

An updated experimental industrial plant of PNPI for the development of CECE technology is described. Experimental results for heavy water detritiation in different operating modes are shown. The effect of pressure, temperatures and gas flow rate on the detritiation factor for the CECE process is presented.

Techniques for Tritium Recovery from Carbon Flakes and Dust at the JET Active Gas Handling System

Detritiation of highly tritium contaminated carbon and metal material used as first wall armour is a key issue for fusion machines like JET and ITER. Re-deposited carbon and hydrogen in the form of flakes and dust can lead to a build-up of the tritium inventory and therefore this material must be removed and processed. The high tritium concentration of the flake and dust material collected from the JET vacuum vessel makes it unsuitable for direct waste disposal without detritiation. A dedicated facility to process the tritiated carbon flake material and recover the tritium has been designed and built. In several test runs active material was successfully processed and detritiated in the new facility. Samples containing only carbon and hydrogen isotopes have been completely oxidized without any residue. Samples containing metallic impurities, e.g. beryllium, require longer processing times, adjusted processing parameters and yield an oxide residue. The detritiation factor was 2 x 104. In order to simulate in-vessel and ex-vessel detritiation techniques, the detritiation of a carbon flake sample by isotopic exchange in a hydrogen atmosphere was investigated. 2.8 % of tritium was recovered by this means.

Studies on the behaviour of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

Confinement and the removal of tritium are key subjects for the safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released as vapour and is in contact with the inner walls. The inner walls of the ITER tritium plant building will also be exposed to tritium in an accident. The tritium released in the buildings is removed by the atmosphere detritiation systems (ADS), where the tritium is oxidized by catalysts and is removed as water. A special gas of SF6 is used in ITER and is expected to be released in an accident such as a fire. Although the SF6 gas has potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed as yet. Tritiated water is produced in the regeneration process of ADS and is subsequently processed by the ITER water detritiation system (WDS). One of the key components of the WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R&D task: (1) tritium behaviour in concrete; (2) the effect of SF6 on the performance of ADS and (3) tritium durability of the electrolysis cell of the ITER-WDS. (1) The tritiated water vapour penetrated up to 50 mm into the concrete from the surface in six months' exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of the cement paste plays an important role in tritium trapping and penetration into concrete materials when concrete is exposed to tritiated water vapour. It is required to evaluate the effect of coating on the penetration rate quantitatively from the actual tritium tests. (2) SF6 gas decreased the detritiation factor of ADS. Since the effect of SF6 depends closely on its concentration, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of arrangement of components in the buildings. (3) It was expected that the electrolysis cell of the ITER-WDS could endure 3 years' operation under the ITER design conditions. Measuring the concentration of the fluorine ions could be a promising technique for monitoring the damage to the electrolysis cell.

Recent Activities on Tritium Technologies for ITER and Fusion Reactors at JAEA

The design studies of Atmosphere Detirtiation System (ADS) have been carried out in Japan Atomic Energy Agency (JAEA) as a contribution of Japan to ITER. The performance of ADS has also been investigated under accidental conditions such as fire and co-existing of a poison gas for catalyst like SF6. There is no degradation of Detritiation Factor (DF) under co-existing of CO or CO2 up to 20% as a simulated fire condition. However, only 0.1% of SF6 degrades the DF from more than 1000 to 50, following reduction of water by SF4 etc. (decomposition products of SF6) at 773K of catalyst bed.For the tritium processing technologies, our efforts have been focused on the R & D of the tritium recovery system of breeding blanket. In case of ITER Test Blanket Module, a cryogenic molecular sieve bed system was designed and demonstrated. Furthermore, electro-chemical pumping system using a proton conductor is also investigated to design more effective system. The durability of electrolysis cell for Water Detritiation System (WDS) has been investigated and it is expected that the cell can endure more than 3 years’ operation under the ITER WDS design condition.A series of fundamental studies on tritium safety technologies has been carried out as another major activity of JAEA for ITER and future fusion demo reactors. Tritium behavior in various confinement materials, tritium monitoring & accountancy, and detritiation were studied under collaboration programs with universities, using Caisson Assembly for Tritium Safety study.

Oxidation Performance Test of Detritiation System Under Existence of SF6

To obtain performance data of atmosphere detritiation system at the off normal events such as SF6 release for the safety of ITER, the detritiation experiment was planned and performed at Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency (JAEA) using a small scaled detritiation system for the oxidation performance test which can process gas flow rate of 0.06 m3/hr in once through. The detritiation system consists of two oxidation catalyst beds (473K and 773K) for converting hydrogen isotopes and tritiated methane in compounds to water vapor and a bubbler for removing water vapor. SF6 gas is used as an electric insulation gas of Neutral Beam Injection system (NBI) in ITER, and is expected to be released in an accident such as fire. In this time, the performance of oxidation catalyst bed of the detritiation system for hydrogen under existence of SF6 which are released from NBI was investigated.The SF6 gas was notably decomposed in the case of the catalyst bed temperature higher than 623K. In addition, when 0.05% or more of SF6 was introduced with 1% of hydrogen, a part of the water produced by the 473K catalyst bed was reduced to hydrogen due to the reaction with the decomposed gas in SF6. Consequently, the detritiation factor (D.F.) of the detritiation system was decreased to less than 50 from > 10000 of its initial value. Since the effect of SF6 depends on its concentration closely, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of the arrangement of components using SF6 in the buildings.

Operational analysis of a liquid phase catalytic exchange column for a detritiation of heavy water

An operational analysis for a liquid phase catalytic exchange column was carried out by the model equations, composed of a material balance and equilibrium relationships for a scrubbing and catalyst bed. The solutions were determined by using the Newton–Raphson method and the parametric investigations were simulated for the effect of the operating variables on the detritiation factor. It was observed that the maximum detritiation performance could be obtained at certain temperature conditions by considering the flow ratio of the water vapor to the given hydrogen stream in the catalyst bed and the water vapor–liquid equilibrium in the scrubbing bed. The optimum operating temperature should be chosen at the larger detritiation conditions providing a reasonable processing flow rate.

Heavy water detritiation by combined electrolysis catalytic exchange at the experimental industrial plant

The experimental industrial plant has been in operation in Petersburg Nuclear Physics Institute to develop the combined electrolysis and catalytic exchange (CECE) process for hydrogen isotope separation since 1995. The plant is also used for processing tritiated heavy water waste; several tons of reactor quality heavy water have been obtained. Last year, the plant was updated to provide a means for heavy water detritiation with high detritiation factor. This paper describes the experimental results for heavy water detritiation. During heavy water purification from tritium detritiation factor about 10 3 was achieved. Heavy water with reduced contents of tritium less than 10 5 Bq/kg was withdrawn at the rate of 4.5 kg per day as a top product. The sample points located along the separation column allowed obtaining tritium concentration profiles within the column. The experimental data are compared with the ones predicted by computer simulation. CECE process is attractive for use in the heavy water hi-flux PIK reactor detritiation plant. This process can find an application at the ITER (International Thermonuclear Experimental Reactor) fusion facility.

Design and Operational Experience with a Pilot-Scale CECE Detritiation Process

ABSTRACTA 5 Mg/annum Combined Electrolysis Catalytic Exchange (CECE) Facility was designed, constructed and operated to demonstrate the CECE process for heavy water detritiation. In this demonstration facility, a liquid-phase catalytic exchange (LPCE) column, using AECL’s wetproofed catalyst, separated tritium from deuterium and a specially designed, low-inventory electrolytic cell provided tritium-enriched deuterium to the LPCE column. An overhead recombiner, also using wetproofed catalyst, produced detritiated heavy water. Tritium was removed from the electrolysis cell as tritiated deuterium gas and packaged as a titanium deuteride. The design detritiation factor of 100 was readily achieved using a 370 GBq/kg heavy water feed. Design features, operational experience and results from the 4-month, 2 000-h operation are described.

A Systematic Approach to the Design of a Large-Scale Detritiation System for Controlled Thermonuclear Fusion Experiments

The most common basic process of air detritiation, which employs oxidation of tritiated gases in a catalytic recombiner and subsequent collection of HTO on molecular sieve dryers, can also be used for a large-scale detritiation system for the next-step deuterium-tritium fusion device. Performance, economy, and reliability can be improved by modifying the design of basic elements, i.e., the recombiners and molecular sieve dryers, and by rearranging them in a system permitting multiple process path choices for optimum performance depending on demand. These improvements should result in a system that is (a) free of secondary tritium release by permeation; (b) economical, with <1 kW power required in a ready-to-operate [open quotes]hot standby[close quotes] condition; (c) capable of reducing inlet humidity of the order of 10 000 ppm (volume) to 0.01 ppm at the outlet by using two adsorber stages in series; and (d) capable of providing the best starting condition for water processing: little or no dilution by H[sub 2]O from isotopic swamping due to the use of two adsorber stages. The system detritiation factor is defined and discussed, and the overriding importance of high water retention efficiency is demonstrated. 4 refs., 5 figs.