Catalytic membrane reactor for tritium extraction system from He purge

International Nuclear Information System (INIS)

Santucci, Alessia; Incelli, Marco; Sansovini, Mirko; Tosti, Silvano

2016-01-01

Highlights: • In the HCBB blanket, the produced tritium is recovered by purging with helium; membrane technologies are able to separate tritium from helium. • The paper presents the results of two experimental campaigns. • In the first, a Pd–Ag diffuser for hydrogen separation is tested at several operating conditions. • In the second, the ability of a Pd–Ag membrane reactor for water decontamination is assessed by performing isotopic swamping and water gas shift reactions. - Abstract: In the Helium Cooled Pebble Bed (HCPB) blanket concept, the produced tritium is recovered purging the breeder with helium at low pressure, thus a tritium extraction system (TES) is foreseen to separate the produced tritium (which contains impurities like water) from the helium gas purge. Several R&D activities are running in parallel to experimentally identify most promising TES technologies: particularly, Pd-based membrane reactors (MR) are under investigation because of their large hydrogen selectivity, continuous operation capability, reliability and compactness. The construction and operation under DEMO relevant conditions (that presently foresee a He purge flow rate of about 10,000 Nm 3 /h and a H 2 /He ratio of 0.1%) of a medium scale MR is scheduled for next year, while presently preliminary experiments on a small scale reactor are performed to identify most suitable operative conditions and catalyst materials. This work presents the results of an experimental campaign carried out on a Pd-based membrane aimed at measuring the capability of this device in separating hydrogen from the helium. Many operative conditions have been investigated by considering different He/H 2 feed flow ratios, several lumen pressures and reactor temperatures. Moreover, the performances of a membrane reactor (composed of a Pd–Ag tube having a wall thickness of about 113 μm, length 500 mm and diameter 10 mm) in processing the water contained in the purge gas have been measured by using

Catalytic membrane reactor for tritium extraction system from He purge

Energy Technology Data Exchange (ETDEWEB)

Santucci, Alessia, E-mail: alessia.santucci@enea.it [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Incelli, Marco [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); DEIM, University of Tuscia, Via del Paradiso 47, 01100 Viterbo (Italy); Sansovini, Mirko; Tosti, Silvano [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy)

2016-11-01

Highlights: • In the HCBB blanket, the produced tritium is recovered by purging with helium; membrane technologies are able to separate tritium from helium. • The paper presents the results of two experimental campaigns. • In the first, a Pd–Ag diffuser for hydrogen separation is tested at several operating conditions. • In the second, the ability of a Pd–Ag membrane reactor for water decontamination is assessed by performing isotopic swamping and water gas shift reactions. - Abstract: In the Helium Cooled Pebble Bed (HCPB) blanket concept, the produced tritium is recovered purging the breeder with helium at low pressure, thus a tritium extraction system (TES) is foreseen to separate the produced tritium (which contains impurities like water) from the helium gas purge. Several R&D activities are running in parallel to experimentally identify most promising TES technologies: particularly, Pd-based membrane reactors (MR) are under investigation because of their large hydrogen selectivity, continuous operation capability, reliability and compactness. The construction and operation under DEMO relevant conditions (that presently foresee a He purge flow rate of about 10,000 Nm{sup 3}/h and a H{sub 2}/He ratio of 0.1%) of a medium scale MR is scheduled for next year, while presently preliminary experiments on a small scale reactor are performed to identify most suitable operative conditions and catalyst materials. This work presents the results of an experimental campaign carried out on a Pd-based membrane aimed at measuring the capability of this device in separating hydrogen from the helium. Many operative conditions have been investigated by considering different He/H{sub 2} feed flow ratios, several lumen pressures and reactor temperatures. Moreover, the performances of a membrane reactor (composed of a Pd–Ag tube having a wall thickness of about 113 μm, length 500 mm and diameter 10 mm) in processing the water contained in the purge gas have been

Tritium extraction from Pb-17Li by bubble columns

International Nuclear Information System (INIS)

Malara, C.

1995-01-01

Tritium extraction from the Pb-17Li liquid breeder of a fusion reactor can be efficiently carried out by bubble columns. To this aim, a mathematical model describing the complex fluid-dynamics of a bubble extractor is here presented. The model equations are made dimensionless and, together with the proper boundary conditions, numerically solved by the orthogonal collocation technique. Moreover, in order to better understand the role played by the different parameters in determining the performance of a bubble column, a closed solution of the model is obtained by introducing suitable hypotheses. A parametric analysis of the extraction efficiency of a bubble column as a function of the process parameters is carried out and, on this basis, the design of a tritium extraction system from the Pb-17Li breeder of a DEMO-type fusion reactor is proposed. 17 refs., 3 figs., 2 tabs

Assessment of the cryogenic distillation system in Cernavoda tritium removal facility

International Nuclear Information System (INIS)

Pasca, Gheorghe; Draghia, Mirela; Porcariu, Florina; Ana, George

2010-01-01

Full text: This paper aims at presenting an assessment of the Cryogenic Distillation system (CD) in the Cernavoda Tritium Removal Facility (CTRF). The cryogenic distillation system is one of the key components of the CTRF which comprises other systems as: the liquid phase catalytic exchange system, designed to transfer tritium from heavy water to a deuterium stream to be fed into the CD system; the atmosphere detritiation system; the tritium recovery system; the tritium/hydrogen monitoring system; the central interlocking system; the tritium extraction and storage system. Thus, the need to build a tritium separation and recovery system results from economic opportunities offered both by heavy water reuse and tritium production, but, at the same time, it offers an alternative for the storage of tritiated heavy water as radioactive waste. (authors)

Optimizing tritium extraction from a Permeator Against Vacuum (PAV) by dimensional design using different tritium transport modeling tools

Energy Technology Data Exchange (ETDEWEB)

Martinez, P., E-mail: pablomiguel.martinez@ciemat.es [CIEMAT-LNF (Laboratorio Nacional de Fusion), Madrid (Spain); Moreno, C. [CIEMAT-LNF (Laboratorio Nacional de Fusion), Madrid (Spain); Martinez, I. [SENER Ingenieria y Sistemas, Provenca 392, 4a 08025 Barcelona (Spain); Sedano, L. [CIEMAT-LNF (Laboratorio Nacional de Fusion), Madrid (Spain)

2012-08-15

The Permeator Against Vacuum (PAV) has been conceived as the simplest, cost effective and reliable technology system dedicated to tritium extraction from breeding liquid metals. An optimal design of a PAV requires a detailed hydraulic design optimization for established operational ranges (HCLL at low velocities of {approx}1 mm/s or DCLL in the ranges of tens of cm/s). The present work analyses the PAV extraction efficiency dependency on the design parameters as optimum on-line Tritium Extraction System (TES). Three different models have been built for that purpose: one through physically refined 1D tritium transport computation using TMAP7 (unique simulation tool with QA for ITER); and two further detailed models on 2D/3D FEM tool (COMSOL Multi-physics 4.0). The geometry used in this work is a simplification of Fuskite{sup Registered-Sign} conceptual design developed at CIEMAT, consisting of a set of cylindrical and concentric {alpha}-Fe double membranes enclosing a vacuumed space and in contact with in-pipe flowing LiPb eutectic. The aim of this paper is to give the first steps to establish the optimal design parameters of a PAV and evaluate the state-of-the-art of these models.

Method and equipment to extract and recover tritium from light and heavy water

International Nuclear Information System (INIS)

Butler, J.P.; Hammerli, M.M.

1979-01-01

A combined electrolysis-catalysis exchange process is proposed to extract tritium, with the aim of being able to reduce the tritium value in the heavy water moderation and cooling system of nuclear power plants. The tritium-contained water is brought into contact in a catalyst column in countercurrent with electrolytically produced hydrogen in this process, the hydrogen is then fed to a catalyst-containing separation column whilst the tritium-contained water flows to the electrolysis cell which produces the hydrogen gas. The process is described in detail. (UWI) [de

Irradiation of lithium aluminate and tritium extraction

International Nuclear Information System (INIS)

Roth, E.; Abassin, J.J.; Botter, F.; Briec, M.; Chenebault, P.; Masson, M.; Rasneur, B.; Roux, N.

1984-12-01

After preselection of the preparation procedures, following short irradiations, γ LiAl0 2 samples submitted to 2.10 19 fast neutrons cm -2 and 1.5 10 20 thermal neutrons cm -2 fluences experienced no apparent damage. Post-irradiation tritium extraction from samples irradiated to 2.10 17 neutrons/cm 2 in quartz ampoules produced mostly tritiated water. When in-pile experiments are performed the sample container material influences greatly the measured ratio of tritium gas to tritiated water - Stainless steel capsules yield more T 2 gas than quartz capsules probably because of a reduction process. Difficulties in interpretation arise from adsoption of tritiated water on the measuring lines. Both experiments showed that much faster extraction rates are obtained from small grain size samples than from large ones at the same open porosity. If diffusion in the grains controls the extraction rates, apparent D values vary from 10 -16 to 1.5 10 -15 cm 2 S -1 in the temperature range explored. Around 500 0 C small grain samples reached equilibrium tritium concentration of a few mCi in 4 hours. Such values are suitable for a blanket concept

Oxidative Tritium Decontamination System

International Nuclear Information System (INIS)

Gentile, Charles A.; Parker, John J.; Guttadora, Gregory L.; Ciebiera, Lloyd P.

2002-01-01

The Princeton Plasma Physics Laboratory, Tritium Systems Group has developed and fabricated an Oxidative Tritium Decontamination System (OTDS), which is designed to reduce tritium surface contamination on various components and items. The system is configured to introduce gaseous ozone into a reaction chamber containing tritiated items that require a reduction in tritium surface contamination. Tritium surface contamination (on components and items in the reaction chamber) is removed by chemically reacting elemental tritium to tritium oxide via oxidation, while purging the reaction chamber effluent to a gas holding tank or negative pressure HVAC system. Implementing specific concentrations of ozone along with catalytic parameters, the system is able to significantly reduce surface tritium contamination on an assortment of expendable and non-expendable items. This paper will present the results of various experimentation involving employment of this system

Tritium Assay and Dispensing of KEPRI Tritium Lab

International Nuclear Information System (INIS)

Sohn, S. H.; Song, K. M.; Lee, S. K.; Lee, K.W.; Ko, B. W.

2009-01-01

The Wolsong Tritium Removal Facility(WTRF) has been constructed to reduce tritium levels in the heavy water systems and environmental emissions at the site. The WTRF was designed to process 100 kg/h of heavy water with the overall tritium extraction efficiency of 97% per single pass and to produce ∼700 g of tritium as T2 per year at the feed concentration of 0.37 TBq/kg. The high purity tritium greater than 99% is immobilized as a metal hydride to secure its long term storage. The recovered tritium will be made available for industrial uses and some research applications in the future. Then KEPRI is constructing the tritium lab. to build-up infrastructure to support tritium research activities and to support tritium control and accountability systems for tritium export. This paper describes the initial phases of the tritium application program including the laboratory infrastructure to support the tritium related R and D activities and the tritium controls in Korea

Confinement and Tritium Stripping Systems for APT Tritium Processing

Energy Technology Data Exchange (ETDEWEB)

Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Heung, L.K.

1997-10-20

This report identifies functions and requirements for the tritium process confinement and clean-up system (PCCS) and provides supporting technical information for the selection and design of tritium confinement, clean-up (stripping) and recovery technologies for new tritium processing facilities in the Accelerator for the Production of Tritium (APT). The results of a survey of tritium confinement and clean-up systems for large-scale tritium handling facilities and recommendations for the APT are also presented.

Confinement and Tritium Stripping Systems for APT Tritium Processing

International Nuclear Information System (INIS)

Hsu, R.H.; Heung, L.K.

1997-01-01

This report identifies functions and requirements for the tritium process confinement and clean-up system (PCCS) and provides supporting technical information for the selection and design of tritium confinement, clean-up (stripping) and recovery technologies for new tritium processing facilities in the Accelerator for the Production of Tritium (APT). The results of a survey of tritium confinement and clean-up systems for large-scale tritium handling facilities and recommendations for the APT are also presented

Measurement of Tritium Activity in Plants by Ice Extraction Method

International Nuclear Information System (INIS)

Pelled, O.; Ovad, S.; Tubul, Y.; Tsroya, S.; Gonen, R.; Abraham, A.; Weinstein, M.; German, U.

2014-01-01

cell and causing its death. This process continues until the cells are almost totally dehydrated. In the temperature range of -20° to -60° C the intra-cellular water freeze forms 'sharp' ice crystals that cause the death of the cells. Water (H2O) and tritiated water (HTO) behave nearly identically in both liquid and vapour phases. The freeze-drying method, although relatively simple, requires the use of dedicated systems and is time consuming. When a plant is frozen in a closed bag, ice is accumulated on the exterior surface of the plant and in the plastic bag that contained the sample, producing a 'self-freeze drying' effect. This ice may be directly used for tritium evaluation if the tritium measurement results are compatible with the generally accepted freeze-drying (lyophilization) method. The present work presents a comparison of this simple Ice Extraction Method (IEM) for tritiated water analysis with the standard lyophilization method

Process for the extraction of tritium from heavy water

International Nuclear Information System (INIS)

Dombra, A.H.

1984-01-01

The object of the invention is achieved by a process for the extraction of tritium from a liquid heavy water stream comprising: contacting the heavy water with a countercurrent gaseous deuterium stream in a column packed with a water-repellent catalyst such that tritium is transferred by isotopic exchange from the liquid heavy water stream to the gaseous deuterium stream

Design and test about de tritium system to filling tritium glove box

International Nuclear Information System (INIS)

Lei, Jiarong; Du, Yang; Yang, Yong

2008-01-01

In order to deal tritium permeated from inflating tritium system at the scene of inflating tritium, dealing waste tritium gas system was designed according to demand and action of dealing waste tritium gas from inflating tritium, and the data of character and volume about appliance of catalyst reaction and drying agent was calculated. Through the test at the scene of inflating tritium, it is result that dealing waste tritium gas system's efficiency reaches above 85% average in circulatory system, so that it can be used in practice extensively. (author)

An assembly of tritium production experiment

International Nuclear Information System (INIS)

Abe, Toshihiko

1981-01-01

An assembly for tritium production experiment, i.e. Tritium Extraction System (TREX) constructed as a small scale test facility for tritium production, and Tritium Removal System (TRS) attached to TREX, and the preliminary results of the experiments with them are described. The radiological safety of the process and operation is also an important consideration. Lithium-aluminum alloy was selected as the most promising target material. The following matters are involved in the scope of production technology: the selection of a target material and target preparation, reactor irradiation, the construction of a facility for the extraction of tritium from the irradiated target, the establishment of the optimum conditions of extraction, the purification, collection and storage of tritium, and the inspection of the product. The tritium production experiment at JAERI is yet on the initial stage; the development is to be continued with the stepwise increase of the scale of tritium production. (J.P.N.)

Tritium systems concepts for the next European torus (NET)

International Nuclear Information System (INIS)

Sood, S.K.; Bagli, K.S.; Busigin, A.; Kveton, O.K.; Dombra, A.H.; Miller, A.I.

1986-09-01

The study deals with the design of the various tritium processing facilities that will be required for the Next European Torus (NET) design. The reference data for the design of the NET Tritium Systems was provided by the NET team. Significant achievements of this study were: (a) Identification of new ways of handling some problems for example: 1) Recovery of tritium from the helium purge of the lithium-ceramic blanket using a novel Adsoprtion and Catalytic Exchange Process, 2) A new way of combining fuel component separation and coolant water detritiation using cryogenic distillation, 3) The use of parasitic refrigeration for the cryogenic isotope separation, 4) Tritium extraction from effluent gas streams at their respective sources, 5) Attempt to eliminate the need for Air Cleanup Systems. (b) Identification of uncertainties, for example: composition of plasma exhaust, required helium purge rate of Li-Pb for tritium recovery, uncertainty in requirements for decontaminating blanket sectors, etc. (c) Review of ways to limit tritium permeation into steam by swamping with hydrogen and to provide quantitative estimates for this permeation

Development of tritium technology at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Anderson, J.L.; Bartlit, J.R.

1982-01-01

The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for large scale fusion reactor systems starting with the Fusion Engineering Device (FED) or the International Tokamak Reactor (INTOR). This paper briefly describes the fuel cycle and safety systems at TSTA including the Vacuum Facility, Fuel Cleanup, Isotope Separation, Transfer Pumping, Emergency Tritium Cleanup, Tritium Waste Treatment, Tritium Monitoring, Data Acquisition and Control, Emergency Power and Gas Analysis systems. Discussed in further detail is the experimental program proposed for the startup and testing of these systems

Recent progress of China HCCB TBM tritium system

Energy Technology Data Exchange (ETDEWEB)

Luo, Deli, E-mail: luodeli2005@hotmail.com; Huang, Guoqiang; Huang, Zhiyong; Qin, Cheng; Song, Jiangfeng; He, Kanghao; Chen, Chang’an; Zhang, Guikai; Fu, Jun; Yao, Yong; An, Yongtao

2016-11-01

Highlights: • Comparing with our previous design, improvements have been made according to the up-to-date experiments and simulations: (1) The palladium alloy tube in the previous design is now removed in the upgraded one and the cryogenic molecular sieve bed is replaced by the getter bed to reduce tritium inventory; (2) Hot metal reduction bed is relocated from T-Plant to Port Cell; (3) TAS is now integrated into TES. • The proposed coolant purification is based on catalytic oxidation and molecular sieve bed adsorption for tritium removal, as well as hot metal adsorption for the elimination of non-tritium gaseous impurities. Some operation parameters and functional components are improved. The interface with the high pressure HCS and other plant systems was incorporated taking into account of the requirement from the ITER port management group meetings. - Abstract: China tritium system including Tritium Extraction System (TES) with Tritium Accountancy System (TAS) integrated in and Coolant Purification System (CPS), which is subordinate to Helium Coolant System (HCS), is of great importance for China Helium Cooled Ceramic Breeder Test Blanket Module (CN HCCB TBM). The purge gas (99.9% He + 0.1% H{sub 2}) carrying Q{sub 2}O (Q = H, D, T) and Q{sub 2} from Li{sub 4}SiO{sub 4} ceramic breeder flows through the reduction bed where Q{sub 2}O is reduced into Q{sub 2} and then absorbed by the getter bed. The HT/HTO ratio and the total tritium are determined by TAS. Catalytic oxidation combines with molecular sieve absorption and hot metal purification are applied to remove tritium and other impurities in helium coolant. A loop including depressurization, helium-sweeping assisted thermal desorption, and cold trapping for the regeneration of saturated molecular sieve bed until the concentration of the desorbed Q{sub 2}O is reduced to an acceptable level. This paper introduces the recent progress of China tritium system including updated conceptual designs of TES and

Technical solutions for tritium removal from Cernavoda NPP heavy water systems

International Nuclear Information System (INIS)

Barariu, Gheorghe; Panait, Adrian

2002-01-01

In CANDU nuclear plants 2400 KCi/GW(e) - year tritium is generated. At a CANDU - 600 reactor similar to Cernavoda NPP Unit 1, 1500 KCi/year of tritium is generated 95% being in the D 2 O moderator, which can achieve a radioactivity level of 80 - 100 Ci/kg. Tritium in heavy water contributes with 30 - 50% to the doses received by operation personnel and with 20% to the radioactivity released to the environment. The extraction of tritium heavy water at CANDU reactors implies the following possibilities: - the radioactivity level reduction in the operation area; - the maintenance and repair cost reduction due to reduction of personnel protection measures and increased labor productivity; - the increase of NPP utilization factor by shutdown time reduction for maintenance and repair; - tritium concentration reduction from technological systems, ensuring thus the possibility of redesigning the systems in order to lower the cost of investment; - profitable use of extracted tritium. Technical measures provided by AECL project for CANDU 600 at Cernavoda make possible to satisfy the current standards concerning tritium concentration in the operation area atmosphere of 5 x 10 -6 Ci/m 3 . The regulations recommend that the radioactivity level should be maintained as low as possible in conformity with ALARA principles. Also, it is possible that norms will become more restrictive in the future, so the tritium removal technology is a good preventive measure which may become very necessary. The methods, which currently reached the industrial or pilot stages, are based on catalyzed chemical exchange, the heavy water electrolysis, and deuterium distillation. They are known as: VPCE - Vapour Phase Catalytic Exchange; LPCE - Liquid Phase Catalytic Exchange; DE - Direct Electrolysis; CD - Cryogenic Distillation. As transfer processes the catalyzed chemical exchange and heavy water electrolysis are used while concentration of tritium gas is done by cryogenic distillation. At present the

Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

Energy Technology Data Exchange (ETDEWEB)

Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

1997-11-30

The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

International Nuclear Information System (INIS)

Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R.; Carson, S.D.; Peterson, P.K.

1997-01-01

The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term

The ITER tritium systems

International Nuclear Information System (INIS)

Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

2007-01-01

ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

HYLIFE-II tritium management system

International Nuclear Information System (INIS)

Longhurst, G.R.; Dolan, T.J.

1993-06-01

The tritium management system performs seven functions: (1) tritium gas removal from the blast chamber, (2) tritium removal from the Flibe, (3) tritium removal from helium sweep gas, (4) tritium removal from room air, (5) hydrogen isotope separation, (6) release of non-hazardous gases through the stack, (7) fixation and disposal of hazardous effluents. About 2 TBq/s (5 MCi/day) of tritium is bred in the Flibe (Li 2 BeF 4 ) molten salt coolant by neutron absorption. Tritium removal is accomplished by a two-stage vacuum disengager in each of three steam generator loops. Each stage consists of a spray of 0.4 mm diameter, hot Flibe droplets into a vacuum chamber 4 m in diameter and 7 m tall. As droplets fall downward into the vacuum, most of the tritium diffuses out and is pumped away. A fraction Φ∼10 -5 of the tritium remains in the Flibe as it leaves the second stage of the vacuum disengager, and about 24% of the remaining tritium penetrates through the steam generator tubes, per pass, so the net leakage into the steam system is about 4.7 MBq/s (11 Ci/day). The required Flibe pumping power for the vacuum disengager system is 6.6 MW. With Flibe primary coolant and a vacuum disengager, an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate vacuum disengager operation with Flibe. A secondary containment shell with helium sweep gas captures the tritium permeating out of the Flibe ducts, limiting leaks there to about 1 Ci/day. The tritium inventory in the reactor is about 190 g, residing mostly in the large Flibe recirculation duct walls. The total cost of the tritium management system is 92 M$, of which the vacuum disengagers cost = 56%, the blast chamber vacuum system = 15%, the cryogenic plant = 9%, the emergency air cleanup and waste treatment systems each = 6%, the protium removal system = 3%, and the fuel storage system and inert gas system each = 2%

Development of a tritium recovery system from CANDU tritium removal facility

International Nuclear Information System (INIS)

Draghia, M.; Pasca, G.; Porcariu, F.

2015-01-01

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)

Development of a tritium recovery system from CANDU tritium removal facility

Energy Technology Data Exchange (ETDEWEB)

Draghia, M.; Pasca, G.; Porcariu, F. [SC.IS.TECH SRL, Timisoara (Romania)

2015-03-15

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)

Overview of tritium processing development at the tritium systems test assembly

International Nuclear Information System (INIS)

Anderson, J.L.

1986-01-01

The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 10 8 Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered

Tritium monitor and collection system

Science.gov (United States)

Bourne, G.L.; Meikrantz, D.H.; Ely, W.E.; Tuggle, D.G.; Grafwallner, E.G.; Wickham, K.L.; Maltrud, H.R.; Baker, J.D.

1992-01-14

This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter. 7 figs.

ARIES-I tritium system

International Nuclear Information System (INIS)

Sze, D.K.; Tam, S.W.; Billone, M.C.; Hassanein, A.M.; Martin, R.

1990-09-01

A key safety concern in a D-T fusion reactor is the tritium inventory. There are three components in a fusion reactor with potentially large inventories, i.e., the blanket, the fuel processing system and the plasma facing components. The ARIES team selected the material combinations, decided the operating conditions and refined the processing systems, with the aiming of minimizing the tritium inventories and leakage. The total tritium inventory for the ARIES-I reactor is only 700 g. This paper discussed the calculations and assumptions we made for the low tritium inventory. We also addressed the uncertainties about the tritium inventory. 13 refs., 2 figs., 3 tabs

Construction and operation of a tritium extraction facility at the Savannah River Site. Final environmental impact statement

International Nuclear Information System (INIS)

1999-03-01

DOE proposes to construct and operate a Tritium Extraction Facility (TEF) at H Area on the Savannah River Site (SRS) to provide the capability to extract tritium from commercial light water reactor (CLWR) targets and from targets of similar design. The proposed action is also DOE's preferred alternative. An action alternative is to construct and operate TEF at the Allied General Nuclear Services facility, which is adjacent to the eastern side of the SRS. Under the no-action alternative DOE could incorporate tritium extraction capabilities in the accelerator for production of tritium. This EIS is linked to the Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling, from which DOE determined that it would produce tritium either in an accelerator or in a commercial light water reactor. The purpose of the proposed action and alternatives evaluated in this EIS is to provide tritium extraction capability to support either tritium production technology. The EIS assesses the environmental impacts from the proposed action and the alternatives, including the no action alternative

RAMI modeling of plant systems for proposed tritium production and extraction facilities

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

The control of life-cycle cost is a primary concern during the development, construction, operation, and decommissioning of DOE systems and facilities. An effective tool that can be used to control these costs, beginning with the design stage, is called a reliability, availability, maintainability, and inspectability analysis or, simply, RAMI for short. In 1997, RAMI technology was introduced to the Savannah River Site with applications at the conceptual design stage beginning with the Accelerator Production of Tritium (APT) Project and later extended to the Commercial Light Water Reactor (CLWR) Tritium Extraction Facility (TEF) Project. More recently it has been applied to the as-build Water Treatment Facilities designed for ground water environmental restoration. This new technology and database was applied to the assessment of balance-of-plant systems for the APT Conceptual Design Report. Initial results from the Heat Removal System Assessment revealed that the system conceptual design would cause the APT to fall short of its annual production goal. Using RAM technology to immediately assess this situation, it was demonstrated that the product loss could be gained back by upgrading the system's chiller unit capacity at a cost of less than $1.3 million. The reclaimed production is worth approximately $100 million. The RAM technology has now been extended to assess the conceptual design for the CLWR-TEF Project. More specifically, this technology and database is being used to translate high level availability goals into lower level system design requirements that will ensure the TEF meets its production goal. Results, from the limited number of system assessments performed to date, have already been used to modify the conceptual design for a remote handling system, improving its availability to the point that a redundant system, with its associated costs of installation and operation may no longer be required. RAMI results were also used to justify the elimination

Studies on chemical phenomena of high concentration tritium water and organic compounds of tritium from viewpoint of the tritium confinement

International Nuclear Information System (INIS)

Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori; Sugiyama, Takahiko; Okuno, Kenji

2009-01-01

As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated two research programs on chemical phenomena of high concentration tritium water and organic compounds of tritium from view point of the tritium confinement have been conducted by the C01 team. The results are summarized as follows: (1) Chemical effects of the high concentration tritium water on stainless steels as structural materials of fusion reactors were investigated. Basic data on tritium behaviors at the metal-water interface and corrosion of metal in tritium water were obtained. (2) Development of the tritium confinement and extraction system for the circulating cooling water in the fusion reactor was studied. Improvement was obtained in the performance of a chemical exchange column and catalysts as major components of the water processing system. (J.P.N.)

PERMCAT experiments with tritium at high helium flow rates relevant for the tritium extraction systems using the CAPER facility at TLK

Energy Technology Data Exchange (ETDEWEB)

Bükki-Deme, András, E-mail: andras.buekki-deme@kit.edu; Demange, David; Le, Thanh-Long; Fanghänel, Eleonore; Simon, Karl-Heinz

2016-11-01

Highlights: • We examined PERMCAT reactor efficiency processing tritiated water at high Helium carrier flow rates. • We have found that – as expected from previous studies – that the swamping ratio (ratio between the impurity and purge side flow rates) has a key effect on the decontamination factors. • On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high impurity flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions. - Abstract: Experiments are still necessary to consolidate the processes retained for the Tritium Extraction Systems of the European ITER Test Blanket Modules (TBM). A PERMCAT reactor combines a catalyst promoting isotope exchange reactions and a Pd/Ag membrane allowing tritium recovery from complex gaseous mixtures containing tritium in different chemical forms. Originally developed for the Tokamak Exhaust Processing, the PERMCAT process is also candidate to detritiate the water arising from an adsorption column installed in the TBM ancillary systems. We discuss the results of an extensive experimental campaign using a PERMCAT reactor to process Q{sub 2}O containing impurity gas mixtures at high flow rates. Two different experimental configurations were studied, namely PERMCAT stand-alone, and PERMCAT in combination with a zeolite molecular sieve bed (MSB, previously loaded with Q{sub 2}O) under regeneration. On the one hand, many expected behaviors were observed, such as the key influence of the swamping ratio (ratio between the impurity and purge side flow rates) on the decontamination factors. On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions.

Experience in handling concentrated tritium

International Nuclear Information System (INIS)

Holtslander, W.J.

1985-12-01

The notes describe the experience in handling concentrated tritium in the hydrogen form accumulated in the Chalk River Nuclear Laboratories Tritium Laboratory. The techniques of box operation, pumping systems, hydriding and dehydriding operations, and analysis of tritium are discussed. Information on the Chalk River Tritium Extraction Plant is included as a collection of reprints of papers presented at the Dayton Meeting on Tritium Technology, 1985 April 30 - May 2

Tritium Systems Test Facility. Volume I

International Nuclear Information System (INIS)

Anderson, G.W.; Battleson, K.W.; Bauer, W.

1976-10-01

Sandia Laboratories proposes to build and operate a Tritium Systems Test Facility (TSTF) in its newly completed Tritium Research Laboratory at Livermore, California (see frontispiece). The facility will demonstrate at a scale factor of 1:200 the tritium fuel cycle systems for an Experimental Power Reactor (EPR). This scale for each of the TSTF subsystems--torus, pumping system, fuel purifier, isotope separator, and tritium store--will allow confident extrapolation to EPR dimensions. Coolant loop and reactor hall cleanup facilities are also reproduced, but to different scales. It is believed that all critical details of an EPR tritium system will be simulated correctly in the facility. Tritium systems necessary for interim devices such as the Ignition Test Reactor (ITR) or The Next Step (TNS) can also be simulated in TSTF at other scale values. The active tritium system will be completely enclosed in an inert atmosphere glove box which will be connected to the existing Gas Purification System (GPS) of the Tritium Research Laboratory. In effect, the GPS will become the scaled environmental control system which otherwise would have to be built especially for the TSTF

Tritium effluent removal system

International Nuclear Information System (INIS)

Lamberger, P.H.; Gibbs, G.E.

1978-01-01

An air detritiation system has been developed and is in routine use for removing tritium and tritiated compounds from glovebox effluent streams before they are released to the atmosphere. The system is also used, in combination with temporary enclosures, to contain and decontaminate airborne releases resulting from the opening of tritium containment systems during maintenance and repair operations. This detritiation system, which services all the tritium handling areas at Mound Facility, has played an important role in reducing effluents and maintaining them at 2 percent of the level of 8 y ago. The system has a capacity of 1.7 m 3 /min and has operated around the clock for several years. A refrigerated in-line filtration system removes water, mercury, or pump oil and other organics from gaseous waste streams. The filtered waste stream is then heated and passed through two different types of oxidizing beds; the resulting tritiated water is collected on molecular sieve dryer beds. Liquids obtained from regenerating the dryers and from the refrigerated filtration system are collected and transferred to a waste solidification and packaging station. Component redundancy and by-pass capabilities ensure uninterrupted system operation during maintenance. When processing capacity is exceeded, an evacuated storage tank of 45 m 3 is automatically opened to the inlet side of the system. The gaseous effluent from the system is monitored for tritium content and recycled or released directly to the stack. The average release is less than 1 Ci/day. The tritium effluent can be reduced by isotopically swamping the tritium; this is accomplished by adding hydrogen prior to the oxidizer beds, or by adding water to the stream between the two final dryer beds

Overview of the tritium system of Ignitor

International Nuclear Information System (INIS)

Rizzello, C.; Tosti, S.

2008-01-01

Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium-tritium mixtures and recovering the plasma exhaust. In fact, the tritium system of Ignitor provides for injecting deuterium-tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability. In this work an analysis of the designed tritium system of Ignitor is summarized

Mixing and mass transfer in bubble extractor: its application to tritium extraction from 17Li83Pb

International Nuclear Information System (INIS)

Baratti, R.; Polcaro, A.M.; Ricci, P.F.; Viola, A.; Pierini, G.

1986-01-01

The tritium extraction from the liquid alloy 17Li83Pb has been examined taking into consideration the equations related to the design of bubble extractors in order to verify which are the highest tritium recovery efficiencies which can be realized so as to minimize the tritium permeation into the water of the cooling system. For the tritium desorption from alloy, flowing countercurrent to a helium stream in a bubble extractor, the axial dispersion in liquid and gaseous phase and the effect of gas phase expansion caused by reduced hydrostatic head in the extractor are taken into account. Taking into consideration some NET technical specifications, such as the alloy volume in the blanket of 65 m 3 and the tritium generation rate of 78 g.d -1 , the results are presented and discussed

Tritium extraction from neutron-irradiated lithium aluminate.; Extraccion del tritio generado por irradiacion neutronica de aluminato de litio.

Energy Technology Data Exchange (ETDEWEB)

Garcia H, F

1995-10-01

Lithium aluminate is being strongly considered as a breeder material because of its thermophysical, chemical and mechanical stability at high temperatures and its favorable irradiation behavior. Furthermore, it is compatible with other blanket and structural materials. In this work, the effects of calcination temperature during preparation, extraction temperature and sweep gas composition were observed. Lithium aluminate prepared by four different methods, was neutron irradiated for 30 minutes at a flux of 10{sup 12} -10{sup 13} n/cm{sup 2} s in the TRIGA Mark III reactor at Salazar, Mexico; and the tritium extraction rate was measured. Calcination temperature do not affect the tritium extraction rate. However, using high calcination temperature, gamma lithium aluminate was formed. The tritium extraction at 600 Centigrade degrees was lower than at 800 Centigrade degrees and the tritium amount extracted by distillation of the solid sample was higher. The sweep gas composition showed that tritium extraction was less with Ar plus 0.5 % H{sub 2} that with Ar plus 0.1 % H{sub 2}. This result was contrary to expected, where the tritium extraction rate could be higher when hydrogen is added to the sweep gas. Probably this effect could be attributed to the gas purity. (Author).

Development of a tritium monitor combined with an electrochemical tritium pump using a proton conducting oxide

Energy Technology Data Exchange (ETDEWEB)

Tanaka, M. [National Institute for Fusion Science, Toki, Gifu (Japan); Sugiyama, T. [Nagoya University, Fro-cho, Chikusa-ku, Nagoya (Japan)

2015-03-15

The detection of low level tritium is one of the key issues for tritium management in tritium handling facilities. Such a detection can be performed by tritium monitors based on proton conducting oxide technique. We tested a tritium monitoring system composed of a commercial proportional counter combined with an electrochemical hydrogen pump equipped with CaZr{sub 0.9}In{sub 0.1}O{sub 3-α} as proton conducting oxide. The hydrogen pump operated at 973 K under electrolysis conditions using tritiated water vapor (HTO). The proton conducting oxide extracts tritium molecules (HT) from HTO and tritium concentration is measured by the proportional counter. The advantage of the proposed tritium monitoring system is that it is able to convert HTO into molecular hydrogen.

The MOZART in-pile tritium extraction experiment

International Nuclear Information System (INIS)

Briec, M.

1990-01-01

In-pile tritium extraction behavior of various ceramics was compared in the MOZART experiment. The influence of temperature and purge gas composition was studied. The experimental results are analyzed by taking into account the processes of diffusion in the grain and desorption at grain surface. This analysis confirms that a better knowledge of the desorption process is necessary for a satisfactory explanation of the experimental data

Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

Science.gov (United States)

Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

2017-11-01

The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

Tritium extraction mechanisms from lithium aluminates during in pile irradiation experiments

International Nuclear Information System (INIS)

Briec, M.; Roth, E.

1987-04-01

The principal aim was to determine ranges of parameters governing tritium release from γ lithium aluminates within which acceptable rates for their contemplated usage as tritium breeder material in a fusion reactor blanket could be obtained. in the first place values of every quantity involved should be known as well as possible. Reproducible results should be a criterium of validity of the selected parameters. It is shown from a description of a series of experiments that processes limiting tritium release rates are not the same in different temperature ranges. By varying the composition of purge gases used for tritium extraction, the level of irradiation fluxes, and by studying simultaneously samples of different textures, results were obtained and an assignment of the respective role of defect formation, texture, surface effect is attempted to interpret them

Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

International Nuclear Information System (INIS)

Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

1995-01-01

The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

The LLNL portable tritium processing system

International Nuclear Information System (INIS)

Anon.

1995-01-01

The end of the Cold War significantly reduced the need for facilities to handle radioactive materials for the US nuclear weapons program. The LLNL Tritium Facility was among those slated for decommissioning. The plans for the facility have since been reversed, and it remains open. Nevertheless, in the early 1990s, the cleanup (the Tritium Inventory Removal Project) was undertaken. However, removing the inventory of tritium within the facility and cleaning up any pockets of high-level residual contamination required that we design a system adequate to the task and meeting today's stringent standards of worker and environmental protection. In collaboration with Sandia National Laboratory and EG ampersand G Mound Applied Technologies, we fabricated a three-module Portable Tritium Processing System (PTPS) that meets current glovebox standards, is operated from a portable console, and is movable from laboratory to laboratory for performing the basic tritium processing operations: pumping and gas transfer, gas analysis, and gas-phase tritium scrubbing. The Tritium Inventory Removal Project is now in its final year, and the portable system continues to be the workhorse. To meet a strong demand for tritium services, the LLNL Tritium Facility will be reconfigured to provide state-of-the-art tritium and radioactive decontamination research and development. The PTPS will play a key role in this new facility

The tritium operations experience on TFTR

International Nuclear Information System (INIS)

Halle, A. von; Anderson, J.L.; Gentile, C.; Grisham, L.; Hosea, J.; Kamperschroer, J.; LaMarche, P.; Oldaker, M.; Nagy, A.; Raftopoulos, S.; Stevenson, T.

1995-01-01

The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grams of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the U.S. Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described. (orig.)

The tritium operations experience on TFTR

International Nuclear Information System (INIS)

von Halle, A.; Gentile, C.

1994-01-01

The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grains of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes- the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the US Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described

Tritium Systems Test Facility

International Nuclear Information System (INIS)

Cafasso, F.A.; Maroni, V.A.; Smith, W.H.; Wilkes, W.R.; Wittenberg, L.J.

1978-01-01

This TSTF proposal has two principal objectives. The first objective is to provide by mid-FY 1981 a demonstration of the fuel cycle and tritium containment systems which could be used in a Tokamak Experimental Power Reactor for operation in the mid-1980's. The second objective is to provide a capability for further optimization of tritium fuel cycle and environmental control systems beyond that which is required for the EPR. The scale and flow rates in TSTF are close to those which have been projected for a prototype experimental power reactor (PEPR/ITR) and will permit reliable extrapolation to the conditions found in an EPR. The fuel concentrations will be the same as in an EPR. Demonstrations of individual components of the deuterium-tritium fuel cycle and of monitoring, accountability and containment systems and of a maintenance methodology will be achieved at various times in the FY 1979-80 time span. Subsequent to the individual component demonstrations--which will proceed from tests with hydrogen (and/or deuterium) through tracer levels of tritium to full operational concentrations--a complete test and demonstration of the integrated fuel processing and tritium containment facility will be performed. This will occur near the middle of FY 1981. Two options were considered for the TSTF: (1) The modification of an existing building and (2) the construction of a new facility

Tritium module for ITER/Tiber system code

International Nuclear Information System (INIS)

Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

1988-01-01

A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

Tritium behavior pattern in some soil-plant systems in a tropical environment

International Nuclear Information System (INIS)

Soman, S.D.; Iyengar, T.S.; Sadarangani, S.H.; Vaze, P.K.

1975-01-01

A study of the distribution pattern of tritium in the soil/plant environment gives valuable ecological information on the natural water balance. The results of such a study for the conditions obtaining in India are given in this paper. Field studies are carried out by injection of tritium into some soil/plant systems and following the transfer pathways. The method of extraction for tissue-free-water-tritium (TFWT) is based on the vacuum freeze-drying technique while the tissue-bound-tritium (TBT) is estimated by a modified version of the Shoniger method. The determination of residence time of tritium in aqueous and organic phase in a number of tropical trees has been carried out both for stem-injection as well as intake from the soil. From the results of this study the tree biomass and transpiration rates have been determined. The tritium profile over time, for an acute exposure in certain trees such as Morinda Tinetoria, Achras Sapota etc. shows significantly different patterns compared to the normal pattern shown by Mangifera Indica, Terminalia Catappa, Ficus Glomerata etc. The period of investigation in each case varied from 400 to 1000 h. In most of the cases, the TBT fractions were very low compared to TFWT fractions in the initial stages. The tritium behavior in the tree reflects significant characteristics of the tritium behavior in the soil system. The authors have found that the leaf sampling can be used as an indicator of total environmental tritium behavior. (author)

Tritium contaminated waste management at the tritium systems test assembly

International Nuclear Information System (INIS)

Jalbert, R.A.; Carlson, R.V.

1987-01-01

The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to move toward full operation of an integrated, full-sized, computer-controlled fusion fuel processing loop. Concurrent nonloop experiments further the development of advanced tritium technologies and handling methods. Since tritium operations began in June 1984, tritium contaminated wastes have been produced at TSTA that are roughly typical in kind and amount of those to be produced by tritium fueling operations at fusion reactors. Methods of managing these wastes are described, including information on some methods of decontamination so that equipment can be reused. Data are given on the kinds and amounts of wastes and the general level of contamination. Also included are data on environmental emissions and doses to personnel that have resulted from TSTA operations. Particular problems in waste managements are discussed

TFTR tritium inventory accountability system

International Nuclear Information System (INIS)

Saville, C.; Ascione, G.; Elwood, S.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Stencel, J.; Voorhees, D.; Tilson, C.

1995-01-01

This paper discusses the program, PPPL (Princeton Plasma Physics Laboratory) Material Control and Accountability Plan, that has been implemented to track US Department of Energy's tritium and all other accountable source material. Specifically, this paper details the methods used to measure tritium in various systems at the Tokamak Fusion Test Reactor; resolve inventory differences; perform inventory by difference inside the Tokamak; process and measure plasma exhaust and other effluent gas streams; process, measure and ship scrap or waste tritium on molecular sieve beds; and detail organizational structure of the Material Control and Accountability group. In addition, this paper describes a Unix-based computerized software system developed at PPPL to account for all tritium movements throughout the facility. 5 refs., 2 figs

Tritium system design studies of fusion experimental breeder

International Nuclear Information System (INIS)

Deng Baiquan; Huang Jinhua

2003-01-01

A summary of the tritium system design studies for the engineering outline design of a fusion experimental breeder (FEB-E) is presented. This paper is divided into three sections. In first section, the geometry, loading features and tritium concentrations in liquid lithium of tritium breeding zones of blanket are described. The tritium flow chart corresponding to the tritium fuel cycle system has been constructed, and the inventories in ten subsystems are calculated using SWITRIM code in section 2. Results show that the necessary initial tritium storage to start up FEB-E with fusion power of 143 MW is about 319 g. In final section, the tritium leakage issues under different operation circumstances have been analyzed. It was found that the potential danger of tritium leakage could be resulted from the exhausted gas of the diverter system. It is important to elevate the tritium burnup fraction and reduce the tritium throughput. (authors)

Five years of tritium handling experience at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Carlson, R.V.

1989-01-01

The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility designed to develop and demonstrate, in full scale, technologies necessary for safe and efficient operation of tritium systems required for tokamak fusion reactors. TSTA currently consists of systems for evacuating reactor exhaust gas with compound cryopumps; for removing impurities from plasma exhaust gas and recovering the chemically-combined tritium; for separating the isotopes of hydrogen; for transfer pumping; or storage of hydrogen isotopes; for gas analysis; and for assuring safety by the necessary control, monitoring, and tritium removal from effluent streams. TSTA also has several small scale experiments to develop and test new equipment and processes necessary for fusion reactors. In this paper, data on component reliability, failure types and rates, and waste quantities are presented. TSTA has developed a Quality Assurance program for preparing and controlling the documentation of the procedures required for the design, purchase, and operation of the tritium systems. Operational experience under normal, abnormal, and emergency conditions is presented. One unique aspect of operations at TSTA is that the design personnel for the TSTA systems are also part of the operating personnel. This has allowed for the relatively smooth transition from design to operations. TSTA has been operated initially as a research facility. As the system is better defined, operations are proceeding toward production modes. The DOE requirements for the operation of a tritium facility like TSTA include personnel training, emergency preparedness, radiation protection, safety analysis, and preoperational appraisals. The integration of these requirements into TSTA operations is discussed. 4 refs., 3 figs., 3 tabs

The application of integrated safety management principles to the Tritium Extraction Facility project

International Nuclear Information System (INIS)

Hickman, M.O.; Viviano, R.R.

2000-01-01

The DOE has developed a program that is accomplishing a heightened safety posture across the complex. The Integrated Safety Management (ISM) System (ISMS) program utilizes five core functions and seven guiding principles as the basis for implementation. The core functions define the work scope, analyze the hazards, develop and implement hazard controls, perform the work, and provide feedback for improvement. The guiding principles include line management responsibility, clear roles and responsibilities, competence per responsibilities, identification of safety standards/requirements, tailored hazard control, balanced priorities, and operations authorization. There exists an unspecified eighth principle, that is, worker involvement. A program requiring the direct involvement of the employees who are actually performing the work has been shown to be quite an effective method of communicating safety requirements, controlling work in a safe manner, and reducing safety violations and injuries. The Tritium Extraction Facility (TEF) projects, a component of the DOE's Commercial Light Water Reactor Tritium Production program, has taken the ISM principles and core functions and applied them to the project's design. The task of the design team is to design a facility and systems that will meet the production requirements of the DOE tritium mission as well as a design that minimizes the workers' exposure to adverse safety situations and hazards/hazardous materials. During the development of the preliminary design for the TEF, design teams consisted of not only designers but also personnel who had operational experience in the existing tritium and personnel who had operational experience in the existing tritium and personnel who had specialized experience from across the DOE complex. This design team reviewed multiple documents associated with the TEF operation in order to identify and document the hazards associated with the tritium process. These documents include hazards

Design of a permeator against vacuum for tritium extraction from eutectic lithium-lead in a DCLL DEMO

Energy Technology Data Exchange (ETDEWEB)

Garcinuño, Belit, E-mail: belit.garcinuno@ciemat.es [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Rapisarda, David [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Fernández, Iván [Fundación & Departamento de Ingeniería Energética, UNED, Madrid (Spain); CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Moreno, Carlos; Palermo, Iole; Ibarra, Ángel [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain)

2017-04-15

Highlights: • A conceptual design of a Permeator Against Vacuum is presented. • The efficiency is dependent on geometry and Tritium transport. • The use of different membrane materials is discussed. • A squared PAV with alternated PbLi flowing and vacuum flat ducts is designed. • 80% efficiency of Tritium extraction is accomplished under DCLL-BB requirements. - Abstract: One of the most important issues in future fusion power plants is the extraction of tritium generated in the breeders in order to achieve self-sufficiency. When the breeder is a liquid metal one of the most promising techniques is the Permeation Against Vacuum, whose principle is based on tritium diffusion through a permeable membrane in contact with the liquid metal carrier and its further extraction by a vacuum pump. A conceptual design of permeator has been developed, taking into account the features of a DEMO reactor with a Dual Coolant Lithium Lead (DCLL) breeder blanket. The study is based on the analysis of different membranes and geometries aiming at the overall efficiency (extraction capability) of the device, as well as its compatibility with the breeder material. The permeator is based on a rectangular section multi-channel distribution where the liquid metal channels and vacuum channels are alternated in order to maximize the contact area and therefore to promote tritium transport from the bulk to the walls. The resulting permeator design has an excellent estimated extraction efficiency, of 80%, in a relatively compact device.

Tritium accountancy in fusion systems

Energy Technology Data Exchange (ETDEWEB)

Klein, J.E.; Clark, E.A.; Harvel, C.D.; Farmer, D.A.; Tovo, L.L.; Poore, A.S. [Savannah River National Laboratory, Aiken, SC (United States); Moore, M.L. [Savannah River Nuclear Solutions, Aiken, SC (United States)

2015-03-15

The US Department of Energy (DOE) has clearly defined requirements for nuclear material control and accountability (MCA) of tritium whereas the International Atomic Energy Agency (IAEA) does not since tritium is not a fissile material. MCA requirements are expected for tritium fusion machines and will be dictated by the host country or regulatory body where the machine is operated. Material Balance Areas (MBA) are defined to aid in the tracking and reporting of nuclear material movements and inventories. Material sub-accounts (MSA) are established along with key measurement points (KMP) to further subdivide a MBA to localize and minimize uncertainties in the inventory difference (ID) calculations for tritium accountancy. Fusion systems try to minimize tritium inventory which may require continuous movement of material through the MSA. The ability of making meaningful measurements of these material transfers is described in terms of establishing the MSA structure to perform and reconcile ID calculations. For fusion machines, changes to the traditional ID equation will be discussed which includes breeding, burn-up, and retention of tritium in the fusion device. The concept of 'net' tritium quantities consumed or lost in fusion devices is described in terms of inventory taking strategies and how it is used to track the accumulation of tritium in components or fusion machines. (authors)

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

International Nuclear Information System (INIS)

Malara, C.; Casini, G.; Viola, A.

1995-01-01

The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.)

A novel atmospheric tritium sampling system

Science.gov (United States)

Qin, Lailai; Xia, Zhenghai; Gu, Shaozhong; Zhang, Dongxun; Bao, Guangliang; Han, Xingbo; Ma, Yuhua; Deng, Ke; Liu, Jiayu; Zhang, Qin; Ma, Zhaowei; Yang, Guo; Liu, Wei; Liu, Guimin

2018-06-01

The health hazard of tritium is related to its chemical form. Sampling different chemical forms of tritium simultaneously becomes significant. Here a novel atmospheric tritium sampling system (TS-212) was developed to collect the tritiated water (HTO), tritiated hydrogen (HT) and tritiated methane (CH3T) simultaneously. It consisted of an air inlet system, three parallel connected sampling channels, a hydrogen supply module, a methane supply module and a remote control system. It worked at air flow rate of 1 L/min to 5 L/min, with temperature of catalyst furnace at 200 °C for HT sampling and 400 °C for CH3T sampling. Conversion rates of both HT and CH3T to HTO were larger than 99%. The collecting efficiency of the two-stage trap sets for HTO was larger than 96% in 12 h working-time without being blocked. Therefore, the collected efficiencies of TS-212 are larger than 95% for tritium with different chemical forms in environment. Besides, the remote control system made sampling more intelligent, reducing the operator's work intensity. Based on the performance parameters described above, the TS-212 can be used to sample atmospheric tritium in different chemical forms.

Ontario Hydro diversifies into tritium

International Nuclear Information System (INIS)

Anon.

1983-01-01

A report is given on a plant which is to be built at the Darlington Candu reactor site in Canada for the extraction of tritium from heavy water. As tritium is used as a fuel in fusion research the market for it is expected to grow. The design of the system is outlined with the help of a flow diagram. (U.K.)

Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

Energy Technology Data Exchange (ETDEWEB)

Mertz, G.

1999-12-16

This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements.

Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

International Nuclear Information System (INIS)

Mertz, G.

1999-01-01

This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements

Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Konishi, S.; Yoshida, H.; Naruse, Y.; Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L.

1990-01-01

Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, ''JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing

Operation of the TSTA (Tritium Systems Test Assembly) with 100 gram tritium

International Nuclear Information System (INIS)

Sherman, R.H.; Bartlit, J.R.

1988-01-01

In March of 1988 full operation of the 4-column isotope separation system (ISS) was realized in runs that approximated the design load of tritium. Previous operations had been fraught with operating difficulties principally due to external systems. This report will examine the recent highly successful 6-day period of operation. During this time the system was cooled from room temperature, loaded with hydrogen isotopes including 109 grams of tritium, integrated with the transfer pumping, impurity injection, and impurity removal systems, as well as the remote computer control system. At the end of the operation 12 grams of tritium having a measured purity of 99.987% (remainder deuterium) were offloaded from the system. Observed profiles in the columns in general agree with computer models. A Height Equivalent to a Theoretical Plate (HETP) of 5.0 cm is confirmed. 3 refs., 5 figs

Technology and component development for a closed tritium cycle

International Nuclear Information System (INIS)

Penzhorn, R.D.; Haange, R.; Hircq, B.; Meikle, A.; Naruse, Y.

1991-01-01

A brief summary on recent advances in the field of tritium technology concerning the most important subsystems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation, the tritium storage and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevant amounts of tritium that constitute the basis of proposed integral process concepts are described. 48 refs., 2 tabs

Technology and component development for a closed tritium cycle

International Nuclear Information System (INIS)

Hircq, B.; Penzhorn, R.D.; Haange, R.; Naruse, Y.

1991-01-01

A brief summary on recent advances in the field of tritium technology concerning the most important subsystems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation, the tritium storage and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevants amounts of tritium that constitute the basis of proposed integral process concepts are described. 48 refs

Review of the Tritium Extraction Facility design

International Nuclear Information System (INIS)

Barton, R.W.; Bamdad, F.; Blackman, J.

2000-01-01

The Defense Nuclear Facilities Safety Board (DNFSB) is an independent executive branch agency responsible for technical safety oversight of the US Department of Energy's (DOE's) defense nuclear facilities. One of DNFSB's responsibilities is the review of design and construction projects for DOE's defense nuclear facilities to ensure that adequate health and safety requirements are identified and implemented. These reviews are performed with the expectation that facility designs are being developed within the framework of a site's Integrated Safety Management (ISM) program. This paper describes the application of ISM principles in DNFSB's ongoing review of the Tritium Extraction Facility (TEF) design/construction project

Review of the Tritium Extraction Facility Design

International Nuclear Information System (INIS)

Ronald W. Barton; Farid Bamdad; Joel Blackman

2000-01-01

The Defense Nuclear Facilities Safety Board (DNFSB) is an independent executive branch agency responsible for technical safety oversight of the U.S. Department of Energy's (DOE's) defense nuclear facilities. One of DNFSB's responsibilities is the review of design and construction projects for DOE's defense nuclear facilities to ensure that adequate health and safety requirements are identified and implemented. These reviews are performed with the expectation that facility designs are being developed within the framework of a site's Integrated Safety Management (ISM) program. This paper describes the application of ISM principles in DNFSB's ongoing review of the Tritium Extraction Facility (TEF) design/construction project

Overview of tritium systems for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Bartlit, J.R.; Gruetzmacher, K.M.; Fleming, R.B.

1987-01-01

The Compact Ignition Tokamak (CIT) is being designed at several laboratories to produce and study fully ignited plasma discharges. The tritium systems which will be needed for CIT include fueling systems and radiation monitoring and safety systems. Design of the tritium systems is the responsibility of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. Major new tritium systems for CIT include a pellet injector, an air detritiation system and a glovebox atmosphere detritiation system. The pellet injector is being developed at Oak Ridge National Laboratory. 7 refs., 2 figs

Process and system for removing tritium

International Nuclear Information System (INIS)

Ridgely, J.N.

1976-01-01

A process and system for removing tritium, particularly from high temperature gas cooled atomic reactors (HTGR), is disclosed. Portions of the reactor coolant, which is permeated with the pervasive tritium atom, are processed to remove the tritium. Under conditions of elevated temperature and pressure, the reactor coolant is combined with gaseous oxygen, resulting in the formation of tritiated water vapor from the tritium in the reactor coolant and the gaseous oxygen. The tritiated water vapor and the remaining gaseous oxygen are then successively removed by fractional liquefaction steps. The reactor coolant is then recirculated to the reactor

Parameters governing tritium extraction rates from lithiated ceramics. The case of lithium aluminate

International Nuclear Information System (INIS)

Roth, E.; Botter, F.; Briec, M.; Rasneur, B.; Roux, N.

1986-10-01

Significant discrepancies between results of authors comparing tritium extraction rates from different lithiated ceramics are found in the literature. Recent results obtained at C.E.A., principally on lithium aluminates, show that, for a given ceramic, parameters other than textural (grain size, porosity, etc...) may play a predominant role. Enhancements of extraction rates have been induced by adding MgO to the solid or H 2 and CO to the sweep gas, but other factors, probably related to the surface condition of samples, may produce even greater effects. Results of investigations of the influence of exposure to air at given partial pressures of water vapor or of CO 2 show that strict preirradiation procedures must be adopted for preparation, storage and handling of ceramic tritium breeders

User's manual for the ARMLID (Argonne metallic lithium/isotopic dilution) tritium assay system

International Nuclear Information System (INIS)

Porges, K.G.; Bretscher, M.M.; Bennett, E.F.; DiIorio, G.; Mattas, R.F.; Lewandowski, E.F.

1992-08-01

The Argonne Metallic Lithium - Isotopic Dilution (ARMLID) system described in this report, originally developed at ANL for other purposes, was recently redeployed to measure the tritium production rate (TPR) in a series of US/Japanese collaborative fusion blanket integral experiments, involving large assemblies of fusion breeder blanket materials that were irradiated with a fusion neutron source at FNS/JAERI, Japan. Whereas previous uses of the ARMUD scheme involved just a few samples, its application infusion blanket TPR mapping called for large sample numbers per experiment, implying a commensurate scale of sample fabrication and encapsulation, on one hand, and tritium extraction and counting on the other hand. To shorten the time required for these various tasks, yet still yield reliable and accurate results, both the sample fabrication - encapsulation facility and the tritium extraction system had to be extensively revised from original versions that were designed for accuracy, but not necessarily for speed. The present report describes overall revisions in sufficient detail to serve as a User's Manual for this facility, and/or suggest how a new system might be put together. Either possibility may develop in the near future, in support of ITER design studies. Preliminary and partial descriptions of various aspects and features of the system were presented orally, in the course of annual ANL/JAERI/UCLA ''workshops'', over the last 34 years, as well as elsewhere

Tritium system test assembly control system cost estimate

International Nuclear Information System (INIS)

Stutz, R.A.

1979-01-01

The principal objectives of the Tritium Systems Test Assembly (TSTA), which includes the development, demonstration and interfacing of technologies related to the deuterium--tritium fuel cycle for fusion reactor systems, are concisely stated. The various integrated subsystems comprising TSTA and their functions are discussed. Each of the four major subdivisions of TSTA, including the main process system, the environmental and safety systems, supporting systems and the physical plant are briefly discussed. An overview of the Master Data Acquisition and Control System, which will control all functional operation of TSTA, is provided

Selection of fluids for tritium pumping systems

International Nuclear Information System (INIS)

Chastagner, P.

1984-02-01

The degradation characteristics of three types of vacuum pump fluids, polyphenyl ethers, perfluoropolyethers and hydrocarbon oils were reviewed. Fluid selection proved to be a critical factor in the long-term performance of tritium pumping systems and subsequent tritium recovery operations. Thermal degradation and tritium radiolysis of pump fluids produce contaminants which can damage equipment and interfere with tritium recovery operations. General characteristics of these fluids are as follows: polyphenyl ether has outstanding radiation resistance, is very stable under normal diffusion pump conditions, but breaks down in the presence of oxygen at anticipated operating temperatures. Perfluoropolyether fluids are very stable and do not react chemically with most gases. Thermal and mechanical degradation products are inert, but the radiolysis products are very corrosive. Most of the degradation products of hydrogen oils are volatile and the principal radiolysis product is methane. Our studies show that polyphenyl ethers and hydrocarbon oils are the preferred fluids for use in tritium pumping systems. No corrosive materials are formed and most of the degradation products can be removed with suitable filter systems

Operating experience with TFTR's Tritium Storage and Delivery System

International Nuclear Information System (INIS)

Voorhees, D.R.

1995-01-01

The Tritium Storage and Delivery System (TSDS) at TFTR was fabricated at Monsanto Mound Lab in the late 1970's and delivered to PPPL in the early 1980's. Commissioning progressed slowly and was finally completed in 1992 following a series of Preoperational tests and Integrated Systems tests. Those tests included thorough leak testing of glove boxes and process piping, electrical interlocks and controls, instrumentation calibrations, volume determinations and verification of uranium bed capacity. The system accepted tritium in dilute form in May of 1993 and began serious usage of pure tritium in November 1993. As the throughput of high purity tritium increased, shortcomings of the system became evident and extensive repairs were implemented. System leakage and material compatibility were the primary causes of the problems. To date, the system has received, stored and delivered over 500 kCi of tritium and is performing very well. The dedicated quadrupole mass spectrometer and beta scintillator system has been analyzing tritium bearing and pure gas streams for over 3 years with minimal downtime

Development of tritium cleanup system for LHD

International Nuclear Information System (INIS)

Sakuma, Yoichi; Kawano, Takao; Shibuya, Mamoru; Kabutomori, Toshiki

2000-01-01

Energy is vital for humans and we have been consuming a large amount of fossil fuel especially from the beginning of the industrial revolution. Nowadays its huge consumption has however come to threaten our life and we have to prepare nonfossil fuels, for instance solar energy, biomass energy, nuclear energy and so on. Fusion energy is an unlimited resource and one of the strongest candidates of the future energy source. At the National Institute for Fusion Science (referred to as 'NIFS' hereafter), we have constructed a new fusion experimental device called large helical device (referred to as 'LHD' hereafter) in 1998. The device will generate a small amount of tritium, as a fusion product. In order to remove it from the exhaust gas, we have designed a tritium cleanup system based on a new concept. This system is mainly composed of a palladium permeater, a decomposer and hydrogen absorbing alloys. It may perfectly recover the tritium from exhaust gas without oxidizing it. This system is applicable for the future needs at fusion power plants. In order to remove tritium discharged from fusion experimental facilities, it is usual to employ a system by which tritiated constituents, in various chemical forms, are entirely converted to a form of water vapor by catalytic oxidation. The water vapor containing tritiated form is then absorbed by molecular sieve (referred to as 'wet system' hereafter). However, in the case of LHD, it is not rational to deliberately convert the discharged tritium into the water vapor, because the tritium discharged from LHD is almost in a form of hydrogen molecules. Moreover, the tritium in the form of water vapor affects the human body 18000 times stronger than that of hydrogen molecules. In accordance with these view points, we have developed another type of tritium cleanup system based on a new concept, in which hydrogen molecules including tritiated ones (HT, DT and T 2 ) found in the exhaust gas of LHD are directly fixed to hydrogen

Safety analysis of tritium processing system based on PHA

International Nuclear Information System (INIS)

Fu Wanfa; Luo Deli; Tang Tao

2012-01-01

Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)

Separation of tritium from other hydrogen isotopes

International Nuclear Information System (INIS)

Roth, E.

1988-01-01

The paper describes a plant that has been operated at Marcoule for tritium production and used thermal diffusion enrichment, a facility that was built in Saclay to enrich hydrogen in tritium for low level measurements, and the Laue Langevin Institute tritium extraction plant. Details are given on the project under construction for the tritium separation facility at JET using Gas Chromatography, and on proposals for circuits for NET. Studies on catalysers for liquid phase catalytic exchange, on electrolysers, or different gas chromatography arrangements, are described. Systems designed for reprocessing plants, for detritiation of heavy water by distillation are briefly accounted for

A system dynamics model for tritium cycle of pulsed fusion reactor

International Nuclear Information System (INIS)

Zhu, Zuolong; Nie, Baojie; Chen, Dehong

2017-01-01

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

A system dynamics model for tritium cycle of pulsed fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: dehong.chen@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2017-05-15

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

Is Tritium an Issue for High Temperature Reactors?

International Nuclear Information System (INIS)

Fütterer, Michael A.; D'Agata, Elio; Raepsaet, Xavier

2014-01-01

In a High Temperature Reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy. (author)

Is tritium an issue for high temperature reactors?

Energy Technology Data Exchange (ETDEWEB)

Fütterer, Michael A., E-mail: michael.fuetterer@ec.europa.eu [European Commission – Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); D’Agata, Elio [European Commission – Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Raepsaet, Xavier [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DM2S, 91191 Gif-sur-Yvette Cedex (France)

2016-09-15

In a high temperature reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy.

Development of wet-proofed catalyst and catalytic exchange process for tritium extraction

Energy Technology Data Exchange (ETDEWEB)

Song, Myung Jae; Son, Soon Hwan; Chung, Yang Gun; Lee, Gab Bock [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1996-12-31

To apply a liquid phase catalytic exchange(LPCE) process for the tritium extraction from tritiated heavy water, the wet proofed catalyst to allow the hydrogen isotopic exchange reaction between liquid water and hydrogen gas was developed. A styrene divinyl benzene copolymer was selected as am effective catalyst support and prepared by suspension copolymerization. After post-treatment, final catalyst supports were dipped in chloroplatinic acid solution. The catalyst support had a good physical properties at a particular preparation condition. The catalytic performance was successfully verified through hydrogen isotopic exchange reaction in the exchange column. A mathematical model for the tritium removal process consisted of LPCE front-ended process and cryogenic distillation process was established using the NTU-HTU method for LPCE column and the FUG method for cryogenic distillation column, respectively. A computer program was developed using the model and then used to investigate optimum design variables which affect the size of columns and tritium inventory (author). 84 refs., 113 figs.

Tritium systems test assembly stabilization

International Nuclear Information System (INIS)

Jasen, William G.; Michelotti, Roy A.; Anast, Kurt R.; Tesch, Charles

2004-01-01

The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium technology Research and Development (R and D) primarily for future fusion power reactors. The facility was conceived in mid 1970's, operations commenced in early 1980's, stabilization and deactivation began in 2000 and were completed in 2003. The facility will remain in a Surveillance and Maintenance (S and M) mode until the Department of Energy (DOE) funds demolition of the facility, tentatively in 2009. A safe and stable end state was achieved by the TSTA Facility Stabilization Project (TFSP) in anticipation of long term S and M. At the start of the stabilization project, with an inventory of approximately 140 grams of tritium, the facility was designated a Hazard Category (HC) 2 Non-Reactor Nuclear facility as defined by US Department of Energy standard DOE-STD-1027-92 (1997). The TSTA facility comprises a laboratory area, supporting rooms, offices and associated laboratory space that included more than 20 major tritium handling systems. The project's focus was to reduce the tritium inventory by removing bulk tritium, tritiated water wastes, and tritium-contaminated high-inventory components. Any equipment that remained in the facility was stabilized in place. All of the gloveboxes and piping were rendered inoperative and vented to atmosphere. All equipment, and inventoried tritium contamination, remaining in the facility was left in a safe-and-stable state. The project used the End Points process as defined by the DOE Office of Environmental Management (web page http://www.em.doe.- gov/deact/epman.htmtlo) document and define the end state required for the stabilization of TSTA Facility. The End Points process added structure that was beneficial through virtually all phases of the project. At completion of the facility stabilization project the residual tritium inventory was approximately 3,000 curies, considerably less than the 1.6-gram threshold for a HC 3 facility. TSTA is now

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

2013-10-15

Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

Design and operational experience with a portable tritium cleanup system

International Nuclear Information System (INIS)

Maienschein, J.L.; Wilson, S.W.; Garcia, F.

1991-06-01

We built a portable tritium cleanup system to scavenge tritium from contaminated gases in any tritium-containing system in the LLNL Tritium Facility. The cleanup system uses standard catalytic oxidation of tritium to water followed by water removal with a molecular sieve dryer. The cleanup unit, complete with instrumentation, is contained in a portable cart that is rolled into place and connected to the apparatus to be cleaned. The cleanup systems is effective, low-tech, simple, and reliable. The nominal flow rate of the system is 30 liters/minute, and the decontamination factor is > 1000. In this paper we will show design information on our portable cleanup system, and will discuss our operational experience with it over the past several years

Compartmental model for tritium persistence in the soil-plant system

International Nuclear Information System (INIS)

Iyengar, T.S.; Sadarangani, S.H.; Vaze, P.K.; Soman, S.D.

1977-01-01

A three-component computer model for tritium persistence in the soil-plant system, on the basis of an exponential polynomial is attempted. A series of field experiments with four species of trees, viz. Cardia sebastina, Terminalia catappa, Aracaria bidwilli and Mangifera indica, were carried out to generate data for testing the model. It is observed that there are two short-term components and one long-term component for tritium mean residence time, corresponding to the three phases of tritium in the system, viz. Tissue-Free-Water-Tritium, labile component of Tissue-Bound-Tritium and non-labile component of Tissue-Bound-Tritium. (author)

Tritium permeation through iron

International Nuclear Information System (INIS)

Hagi, Hideki; Hayashi, Yasunori

1989-01-01

An experimental method for measuring diffusion coefficients and permeation rates of tritium in metals around room temperature has been established, and their values in iron have been obtained by using the method. Permeation rates of tritium and hydrogen through iron were measured by the electrochemical method in which a tritiated aqueous solution was used as a cathodic electrolyte. Tritium and hydrogen were introduced from one side of a membrane specimen by cathodic polarization, while at the other side of the specimen the permeating tritium and hydrogen were extracted by potentiostatical ionization. The amount of permeated hydrogen was obtained by integrating the anodic current, and that of tritium was determined by measuring the radioactivity of the electrolyte sampled from the extraction side. Diffusion coefficients of tritium (D T ) and hydrogen (D H ) were determined from the time lag of tritium and hydrogen permeation. D T =9x10 -10 m 2 /s and D H =4x10 -9 m 2 /s at 286 K for annealed iron specimens. These values of D T and D H were compared with the previous data of the diffusion coefficients of hydrogen and deuterium, and the isotope effect in diffusion was discussed. (orig.)

Results of tritium tests performed on Sandia Laboratories decontamination system

International Nuclear Information System (INIS)

Gildea, P.D.; Wall, W.R.; Gede, V.P.

1978-05-01

The Tritium Research Laboratory (TRL), a facility for performing experiments using gram amounts of tritium, became operational on October 1, 1977. As secondary containment, the TRL employs sealed glove boxes connected on demand to two central decontamination systems, the Gas Purification System and the Vacuum Effluent Recovery System. Performance tests on these systems show the tritium removal systems can achieve concentration reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass at inlet concentrations of 1 part per million or less for both tritium and tritiated methane

The movement of tritium in ecological systems

International Nuclear Information System (INIS)

Polevoy, Y; Laichter, Y.

1988-11-01

This literature survey summarizes the interaction of tritium gas and tritiated water with various components of the ecological system. The intake of tritium gas and tritiated water in plants and soil is described as well as the location of the highest measurable concentration. This information may serve as a basis for risk assessment from tritium to man through the food chain and enables effective tracing of its concentration in the environment. (author)

Tritium-containment systems: a tradeoff study

International Nuclear Information System (INIS)

Folkers, C.L.; Cena, R.J.

1978-01-01

Various design parameters are evaluated that affect the performance of tritium-containment systems for fusion reactors. Our study included a review of such parameters as tritium forms, impurities, catalysts, adsorbents, getters, and as low as reasonably achievable principles. We organized these schemes, which can be considered for treating either air or inert atmospheres, so one could easily make orderly choices and tradeoffs for optimum performance. The relationships examined involved purification-system decontamination factors, flow rates, recycling and leakage, and environmental losses

Operational experience with two tritium-effluent-monitoring systems

International Nuclear Information System (INIS)

Haynie, J.S.; Gutierrez, J.A.

1982-01-01

Two new tritium stack monitoring systems were designed and built. The operational experience of a wide-range detector with a useful range of a few μCi/m 3 to 10 8 μCi/m 3 , and a second monitoring system using an improved Kanne chamber and a new electrometer, called a Model 39 Electrometer-Chargemeter are discussed. Both tritium chambers have been designed to have a reduced sensitivity to tritium contamination, a fast response, and an integrating chargemeter with digital readout for easy conversion to microcuries. The calibration of these monitors and advantages of using these chambers over conventional systems are discussed

Tritium contamination of concrete walls and floors in tritium-handling laboratory

International Nuclear Information System (INIS)

Kawano, T.; Kuroyanagi, M.; Tabei, T.

2006-01-01

A tritium handling laboratory was constructed at the National Institute for Fusion Science about twenty years ago and it was recently closed down. We completed the necessary work that is legally required in Japan at the laboratory, when the use of radioisotopes is discontinued, involving measurements of radioactive contamination. We mainly used smear and direct-immersion methods for the measurements. In applying the smear method, we used a piece of filter paper to wipe up the tritium staining the surfaces. The filter paper containing the tritium was placed directly into a dedicated vial, a scintillation cocktail was then poured over it, and the tritium was measured with a liquid scintillation counter. With the direct-immersion method, a piece of concrete was placed directly into a vial containing a scintillation cocktail, and the tritium in the concrete was measured with a liquid scintillation counter. As well as these measurements, we investigated water-extraction and heating-cooling methods for measuring tritium contamination in concrete. With the former, a piece of concrete was placed into water in a tube to extract the tritium, the water containing the extracted tritium was then poured into a dedicated vial containing a scintillation cocktail, and the tritium contamination was measured. With the latter, a piece of concrete was placed into a furnace and heated to 800 degrees centigrade to vaporize the tritiated water into flowing dry air. The flowing air was then cooled to collect the vaporized tritiated water in a tube. The collected water was placed in a vial for scintillation counting. To evaluate the direct-immersion method, ratios were determined by dividing the contamination measured with the heating-cooling method by that measured with the direct-immersion method. The average ratio was about 2.5, meaning a conversion factor from contamination obtained with the direct-immersion method to that with the heating-cooling method. We also investigated the

Analysis of the organically bound tritium

International Nuclear Information System (INIS)

Baglan, N.; Alanic, G.

2011-01-01

In environmental samples, tritium is very often combined with the fraction of bulk water accumulated in the sample but also in the form of organically bound tritium. When the tritium is organically bound, 2 forms can coexist: the exchangeable fraction and the non-exchangeable fraction. The analysis of the different forms of tritium present in the sample is necessary to assess the sanitary hazards due to tritium. The total tritium is obtained from the analysis of the water released when the fresh sample is burnt while the organically bound tritium is obtained from the analysis of the water released when the dry extract of the sample is burnt. The measurement of the exchangeable fraction and the non-exchangeable fraction requires an additional stage of labile exchange. The exchangeable fraction is determined from the analysis of the water released during the labile exchange and the non-exchangeable fraction is determined from the water released during the combustion of the dry extract of the labile exchange

The tritium systems test assembly: Overview and recent results

International Nuclear Information System (INIS)

Bartlit, J.R.; Anderson, J.L.

1988-01-01

The fusion technology development program for tritium in the US is centered around the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The TSTA is a full-scale system of reactor exhaust gas reprocessing for an ITER-sized machine. That is, TSTA has the capacity to process tritium in a closed loop mode at the rate of 1 kg per day, requiring a tritium inventory of about 100 g. The TSTA program also interacts with all other tritium-related fusion technology programs in the US and all major programs abroad. This report summarizes the current status, results and interactions of the TSTA. Special emphasis is given to operations in May/June using large compound cryopumps that completed the fuel loop integration of all TSTA subsystems for the first time. 6 refs., 2 figs

Separation of tritium from gaseous and aqueous effluent systems

International Nuclear Information System (INIS)

Kobisk, E.H.

1977-01-01

Removal or reduction of tritium content in a wide variety of effluent streams has been extensively studied in the United States. This paper specifically reviews three processes involving tritium separation in the gaseous phase and the aqueous phase. Diffusion through a selective Pd-25Ag alloy membrane at temperatures up to 600 0 C and at pressures up to 700 kg/cm 2 has resulted in successful separation of hydrogen-deuterium mixtures with an associated separation factor of 1.65 (and gives a calculated separation factor for hydrogen-tritium mixtures of 2.0). Use of a single palladium bipolar membrane in an electrolysis system has been found to yield a hydrogen-deuterium separation factor of 4 and a hydrogen-tritium factor of 6 to 11 without the production of gaseous hydrogen. Finally, countercurrent catalytic exchange between tritium-containing hydrogen gas and water has yielded a separation factor of 6.3. The specific advantages of each of these systems will be discussed in terms of their potential applications. In all cases, further investigations are necessary to scale the systems to handle large quantities of feed material in a continuous mode and to minimize energy requirements. Such separative systems must necessarily be cascaded to yield gaseous or aqueous product streams suitable for recycling to the tritium producing systems, for storage or for discharge to the environment. (orig./HP) [de

Tritium system for a tokamak reactor with a self-pumped limiter

International Nuclear Information System (INIS)

Hassanein, A.M.; Sze, D.K.

1986-01-01

Benefits of the self-pumping system are the elimination of vacuum ducts, pumps, and penetration shielding (except for a very small startup system), and the reduction of tritium recycle and refueling. In addition, a self-pumped system may perform better and last longer than alternative systems such as a pumped limiter. The reference case here is a self-cooled lithium/vanadium blanket with a first wall/limiter. This concept combines the functions of first wall and limiter into a single first-wall structure. The wall is shaped in accordance with the outermost plasma flux surface. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. The entire wall area is used for helium trapping. The tritium inventory, tritium permeation rate, and plasma protium concentration for the vanadium wall as a function of the number of years of operation are calculated. The tritium inventory is acceptable, the protium concentration in the plasma is acceptably small, and the tritium permeation rate is moderate. At the start of operation, it is equal to about five times the tritium burnup rate. This tritium will enter the coolant and the cost of the blanket tritium recovery system will be higher

Study on the tritium behaviors in the VHTR system. Part 2: Analyses on the tritium behaviors in the VHTR/HTSE system

Energy Technology Data Exchange (ETDEWEB)

Kim, Eung S. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States); Oh, Chang H., E-mail: Chang.Oh@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States); Patterson, Mike [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States)

2010-07-15

Tritium behaviors in the very high temperature gas reactor (VHTR)/high temperature steam electrolysis (HTSE) system have been analyzed by the TPAC developed by Idaho National Laboratory (INL). The reference system design and conditions were based on the indirect parallel configuration between a VHTR and a HTSE. The analyses were based on the SOBOL method, a modern uncertainty and sensitivity analyses method using variance decomposition and Monte Carlo method. A total of 14 parameters have been taken into account associated with tritium sources, heat exchangers, purification systems, and temperatures. Two sensitivity indices (first order index and total index) were considered, and 15,360 samples were totally used for solution convergence. As a result, important parameters that affect tritium concentration in the hydrogen product have been identified and quantified with the rankings. Several guidelines and recommendations for reducing modeling uncertainties have been also provided throughout the discussions along with some useful ideas for mitigating tritium contaminations in the hydrogen product.

Conceptual design of an emergency tritium clean-up system

International Nuclear Information System (INIS)

Muller, M.E.

1978-01-01

The Los Alamos Scientific Laboratory (LASL) has been selected by the Department of Energy (DOE) to design, build, and operate a facility to demonstrate the operability of the tritium-related subsystems that would be required to successfully develop fusion reactor systems. An emergency tritium clean-up subsystem (ETC) for this facility will be designed to remove tritium from the cell atmosphere if an accident causes the primary and secondary tritium containment to be breached. Conceptually, the ETC will process cell air at the rate of 0.65 actual m 3 /s and will achieve an overall decontamination factor of 10 6 per tritium oxide (T 2 O). Following the maximum credible release of 100 g of tritium, the ETC will restore the cell to opertional status within 24 h without a significant release of tritium to the environment

Japanese university program on tritium radiobiology and environmental tritium

International Nuclear Information System (INIS)

Okada, Shigefumi

1989-01-01

The university program of the tritium study in the Special Research Project of Nuclear Fusion (1980-1989) is now on its 9th year. The study's aim is to assess tritium risk on man and environment for development of Japanese Nuclear Fusion Program. The tritium study begun by establishing various tritium safe-handling devices and methods to protect scientists from tritium contamination. Then, the tritium studies were initiated in three areas: The first was the studies on biological effects of tritiated water, where their RBE values, their modifying factors and mechanisms were investigated. Also, several human monitoring systems for detection of tritium-induced damage were developed. The second was the metabolic studies of tritium, including a daily tritium monitoring system, methods to enhance excretion of tritiated water from body and means to prevent oxidation of tritium gas in the body. The third was the study of environmental tritium. Tritium levels in environmental waters of various types were estimated all-over in Japan and their seasonal or regional variation were analyzed. Last two years, the studies were extended to estimate tritium activities of plants, foods and man in Japan. (author)

Isotopic fractionation of tritium in biological systems.

Science.gov (United States)

Le Goff, Pierre; Fromm, Michel; Vichot, Laurent; Badot, Pierre-Marie; Guétat, Philippe

2014-04-01

Isotopic fractionation of tritium is a highly relevant issue in radiation protection and requires certain radioecological considerations. Sound evaluation of this factor is indeed necessary to determine whether environmental compartments are enriched/depleted in tritium or if tritium is, on the contrary, isotopically well-distributed in a given system. The ubiquity of tritium and the standard analytical methods used to assay it may induce biases in both the measurement and the signification that is accorded to the so-called fractionation: based on an exhaustive review of the literature, we show how, sometimes large deviations may appear. It is shown that when comparing the non-exchangeable fraction of organically bound tritium (neOBT) to another fraction of tritium (e.g. tritiated water) the preparation of samples and the measurement of neOBT reported frequently led to underestimation of the ratio of tritium to hydrogen (T/H) in the non-exchangeable compartment by a factor of 5% to 50%. In the present study, corrections are proposed for most of the biological matrices studied so far. Nevertheless, the values of isotopic fractionation reported in the literature remain difficult to compare with each other, especially since the physical quantities and units often vary between authors. Some improvements are proposed to better define what should encompass the concepts of exchangeable and non-exchangeable fractions. Copyright © 2014 Elsevier Ltd. All rights reserved.

Tritium inventories and tritium safety design principles for the fuel cycle of ITER

International Nuclear Information System (INIS)

Cristescu, I.R.; Cristescu, I.; Doerr, L.; Glugla, M.; Murdoch, D.

2007-01-01

Within the tritium plant of ITER a total inventory of about 2-3 kg will be necessary to operate the machine in the DT phase. During plasma operation, tritium will be distributed in the different sub-systems of the fuel cycle. A tool for tritium inventory evaluation within each sub-system of the fuel cycle is important with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems; however, tritium accounting may be achieved by modelling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the sub-systems. To get reliable results, an accurate dynamic modelling of the tritium content in each sub-system is necessary. A dynamic model (TRIMO) for tritium inventory calculation reflecting the design of each fuel cycle sub-systems was developed. The amount of tritium needed for ITER operation has a direct impact on the tritium inventories within the fuel cycle sub-systems. As ITER will function in pulses, the main characteristics that influence the rapid tritium recovery from the fuel cycle as necessary for refuelling are discussed. The confinement of tritium within the respective sub-systems of the fuel cycle is one of the most important safety objectives. The design of the deuterium/tritium fuel cycle of ITER includes a multiple barrier concept for the confinement of tritium. The buildings are equipped with a vent detritiation system and re-circulation type room atmosphere detritiation systems, required for tritium confinement barrier during possible tritium spillage events. Complementarily to the atmosphere detritiation systems, in ITER a water detritiation system for tritium recovery from various sources will also be operated

Improved iodine and tritium control in reprocessing plants

International Nuclear Information System (INIS)

Henrich, E.; Schmieder, H.; Roesch, W.; Weirich, F.

1981-01-01

During spent fuel processing, iodine and tritium are distributed in many aqueous, organic and gaseous process streams, which complicates their control. Small modifications of conventional purex flow sheets, compatible with processing in the headend and the first extraction cycle are necessary to confine the iodine and the tritium to smaller plant areas. The plant area connected to the dissolver off-gas (DOG) system is suited to confine the iodine and the plant area connected to the first aqueous cycle is suited to confine the tritium. A more clear and convenient iodine and tritium control will be achieved. Relevant process steps have been studied on a lab or a pilot plant scale using I-123 and H-3 tracer

Tritium isotope fractionation in biological systems and in analytical procedures

International Nuclear Information System (INIS)

Kim, M.A.; Baumgaertner, Franz

1989-01-01

The organically bound tritium (OBT) is evaluated in biological systems by determining the tritium distribution ratio (R-value), i.e. tritium concentrations in organic substance to cell water. The determination of the R-value always involves isotope fractionation is applied analytical procedures and hence the evaluation of the true OBT -value in a given biological system appears more complicated than hitherto known in the literature. The present work concentrates on the tritium isotope fractionation in the cell water separation and on the resulting effects on the R-value. The analytical procedures examined are vacuum freeze drying under equilibrium and non-equilibrium conditions and azeotropic distillation. The vaporization isotope effects are determined separately in the phase transition of solid or liquid to gas in pure tritium water systems as well as in real biological systems, e.g. corn plant. The results are systematically analyzed and the influence of isotope effects on the R-value is rigorously quantified

Development of compact tritium confinement system using gas separation membrane

International Nuclear Information System (INIS)

Hayashi, Takumi; Okuno, Kenji

1994-01-01

In order to develop more compact and cost-effective tritium confinement system for fusion reactor, a new system using gas separation membranes has been studied at the Tritium Process Laboratory in the Japan Atomic Energy Research Institute. The preliminary result showed that the gas separation membrane system could reduce processing volume of tritium contaminated gas to more than one order of magnitude compared with the conventional system, and that most of tritiated water vapor (humidity) could be directly recovered by water condenser before passing through dryer such as molecular sieves. More detail investigations of gas separation characteristics of membrane were started to design ITER Atmospheric Detritiation System (ADS). Furthermore, a scaled polyimide membrane module (hollow-filament type) loop was just installed to investigate the actual tritium confinement performance under various ITER-ADS conditions. (author)

Tritium interactions of potential importance to fusion reactor systems: technology requirements

International Nuclear Information System (INIS)

Wilkes, W.R.

1976-01-01

The tritium technology requirements created by the controlled thermonuclear research program to develop a demonstration fusion power reactor by the year 2000 are reviewed. It is found that the majority of the technological advances which are needed to ensure adequate tritium containment in a tritium breeding power reactor need to be demonstrated on a pilot scale by approximately 1983, so that they may be incorporated into EPR-II, the second of two planned experimental power reactors. The most important advances include development of containment materials with permeabilities to tritium well below measured values for stainless steel; large scale, low inventory deuterium-tritium separation systems; and improved monitoring and assay systems. There are less critical requirements for information about the effects of tritium and helium on the mechanical properties of materials, the effects of tritium on biological systems, and data on physical and chemical properties of tritium. Substantial progress needs to be made on these problems early enough to permit possible solutions to be tested on EPR-I. In addition, major improvements in tritium handling equipment are required for EPR-I. Those technological problems for which solutions have not yet been demonstrated by EPR-II must be solved by 1989 if they are to be assured successful application in the demonstration reactor

Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

International Nuclear Information System (INIS)

Pearson, G.G.; Meixler, L.D.; Sirsingh, R.A.P.

1992-01-01

This paper reports on the TFTR Tritium Stack and Area Monitoring Systems which have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. Area monitors provide surveillance of the air in various rooms at TFTR. Stack monitors monitor the air at the TFTR test site that is exhausted through the HVAC systems, from the room exhaust stacks and the tritium systems process vents. The philosophies for the implementation of the Stack and Area Tritium Monitoring Systems at TFTR are to use hardwired controls wherever personnel protection is involved, and to take advantage of modern intelligent controllers to provide a distributed system to support the functions of tracking, displaying, and archiving concentration levels of tritium for all of the monitored areas and stacks

Chemical form of tritium released from solid breeder materials and the influences of it on a bred tritium recovery systems

International Nuclear Information System (INIS)

Furukubo, Y.; Nishikawa, M.; Nishida, Y.; Kinjyo, T.; Tanifuji, Takaaki; Kawamura, Yoshinori; Enoeda, Mikio

2004-01-01

The ratio of HTO in total tritium was measured at release of the bred tritium to the purge gas with hydrogen using the thermal release after irradiation method, where neutron irradiation was performed at JRR-3 reactor in JAERI or KUR reactor in Kyoto University. It is experimentally confirmed in this study that not a small portion of bred tritium is released to the purge gas in the form of HTO form ceramic breeder materials even when hydrogen is added to the purge gas. The chemical composition is to be decided by the competitive reaction at the grain surface of a ceramic breeder material where desorption reaction, isotope exchange reaction 1, isotope exchange reaction 2 and water formation reaction are considered to take part. Observation in this study implies that it is necessary to have a bred tritium recovery system applicable for both HT and HTO form to recover whole bred tritium. The chemical composition also decides the amount of tritium transferable to the cooling water of the electricity generation system through the structural material in the blanket system. Permeation behavior of tritium through some structural materials at various conditions are also discussed. (author)

Tritium fractionation in biological systems and in analytical procedures

International Nuclear Information System (INIS)

Kim, M.A.; Baumgaertner, F.

1991-01-01

The organically bound tritium (OBT) is evaluated in biological systems by measuring the tritium distribution ratio (R-value), i.e. tritium concentrations in organic substance to tissue water. The determination of the R-value is found to involve always isotope fractionation in applied analytical procedures and hence the evaluation of the true OBT-value in a given biological system appears more complicated than hitherto known in the literature. The present work concentrates on the tritium isotope fraction in the tissue water separation and on the resulting effects on the R-value. The analytical procedures examined are vacuum freeze drying under equilibrium and non-equilibrium conditions and azeotropic distillation. The vaporization isotope effects are determined separately in the phase transition of solid or liquid to gas in pure water systems as well as in real biological systems, e.g. maize plant. The results are systematically analysed and the influence of isotope effects on the R-value is rigorously quantified. (orig.)

Tritium problems in fusion reactor systems

International Nuclear Information System (INIS)

Hickman, R.G.

1975-01-01

A brief introduction is given to the role tritium will play in the development of fusion power. The biological and worldwide environmental behavior of tritium is reviewed. The tritium problems expected in fusion power reactors are outlined. A few thoughts on tritium permeation and recent results for tritium cleanup and CT 4 accumulation are presented. Problems involving the recovery of tritium from the breeding blanket in fusion power reactors are also considered, including the possible effect of impurities in lithium blankets and the use of lithium as a regenerable getter pump. (auth)

Fuel cleanup system for the tritium systems test assembly: design and experiments

International Nuclear Information System (INIS)

Kerr, E.C.; Bartlit, J.R.; Sherman, R.H.

1980-01-01

A major subsystem of the Tritium Systems Test Assembly is the Fuel Cleanup System (FCU) whose functons are to: (1) remove impurities in the form of argon and tritiated methane, water, and ammonia from the reactor exhaust stream and (2) recover tritium for reuse from the tritiated impurities. To do this, a hybrid cleanup system has been designed which utilizes and will test concurrently two differing technologies - one based on disposable, hot metal (U and Ti) getter beds and a second based on regenerable cryogenic asdorption beds followed by catalytic oxidation of impurities to DTO and stackable gases and freezout of the resultant DTO to recover essentially all tritium for reuse

A tritium target system for μCF

International Nuclear Information System (INIS)

Zmeskal, J.; Ackerbauer, P.; Durham, W.B.; Heard, H.C.; Neumann, W.; Bossy, H.

1990-12-01

An apparatus has been constructed for the safe handling of tritium as part of a series of muon-catalyzed fusion experiments. The equipment was designed to handle 100 kCi of tritium. The main parts of this system are the oil-free high vacuum and transfer system, and the quadrupole mass analyzer for a direct determination of the target content. The system was used successfully for five continuous periods of operation of over one month each. A new target system was constructed at Lawrence Livermore National Laboratory (LLNL) for ultimate use at Paul Scherrer Institute (PSI) to investigate the high temperature and high pressure region. 9 refs., 4 figs

The Tritium Systems Test Assembly applicability to ITER

International Nuclear Information System (INIS)

Anderson, J.L.

1988-01-01

The Tritium Systems Test Assembly (TSTA), is operated by the Los Alamos National Laboratory (LANL) under the sponsorship of the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The objectives of the TSTA project are to develop, demonstrate, and evaluate the exhaust gas processing and tritium related safety systems for the magnetic fusion energy program. The applicability of these processes for the ITER Tokamak is discussed

STAR facility tritium accountancy

International Nuclear Information System (INIS)

Pawelko, R. J.; Sharpe, J. P.; Denny, B. J.

2008-01-01

The Safety and Tritium Applied Research (STAR) facility has been established to provide a laboratory infrastructure for the fusion community to study tritium science associated with the development of safe fusion energy and other technologies. STAR is a radiological facility with an administrative total tritium inventory limit of 1.5 g (14,429 Ci) [1]. Research studies with moderate tritium quantities and various radionuclides are performed in STAR. Successful operation of the STAR facility requires the ability to receive, inventory, store, dispense tritium to experiments, and to dispose of tritiated waste while accurately monitoring the tritium inventory in the facility. This paper describes tritium accountancy in the STAR facility. A primary accountancy instrument is the tritium Storage and Assay System (SAS): a system designed to receive, assay, store, and dispense tritium to experiments. Presented are the methods used to calibrate and operate the SAS. Accountancy processes utilizing the Tritium Cleanup System (TCS), and the Stack Tritium Monitoring System (STMS) are also discussed. Also presented are the equations used to quantify the amount of tritium being received into the facility, transferred to experiments, and removed from the facility. Finally, the STAR tritium accountability database is discussed. (authors)

Development of a compact tritium activity monitor and first tritium measurements

Energy Technology Data Exchange (ETDEWEB)

Röllig, M., E-mail: marco.roellig@kit.edu; Ebenhöch, S.; Niemes, S.; Priester, F.; Sturm, M.

2015-11-15

Highlights: • We report about experimental results of a new tritium activity monitoring system using the BIXS method. • The system is compact and easy to implement. It has a small dead volume of about 28 cm{sup 3} and can be used in a flow-through mode. • Gold coated surfaces are used to improve significantly count rate stability of the system and to reduce stored inventory. - Abstract: To develop a convenient tool for in-line tritium gas monitoring, the TRitium Activity Chamber Experiment (TRACE) was built and commissioned at the Tritium Laboratory Karlsruhe (TLK). The detection system is based on beta-induced X-ray spectrometry (BIXS), which observes the bremsstrahlung X-rays generated by tritium decay electrons in a gold layer. The setup features a measuring chamber with a gold-coated beryllium window and a silicon drift detector. Such a detection system can be used for accountancy and process control in tritium processing facilities like the Karlsruhe Tritium Neutrino Experiment (KATRIN). First characterization measurements with tritium were performed. The system demonstrates a linear response between tritium partial pressure and the integral count rate in a pressure range of 1 Pa up to 60 Pa. Within 100 s measurement time the lower detection limit for tritium is (143.63 ± 5.06) · 10{sup 4} Bq. The system stability of TRACE is limited by a linear decrease of integral count rate of 0.041 %/h. This decrease is most probably due to exchange interactions between tritium and the stainless steel walls. By reducing the interaction surface with stainless steel, the decrease of the integral count rate was reduced to 0.008 %/h. Based on the first results shown in this paper it can be concluded that TRACE is a promising complement to existing tritium monitoring tools.

Lithium test module on ITER: engineering design of the tritium recovery system

International Nuclear Information System (INIS)

Finn, P.A.

1988-01-01

The design presented is an overview of the tritium recovery system for a lithium module on an ITER type reactor. The design of a tritium recovery system for larger blanket units, sectors, etc. could use the information developed in this report. A goal of this design was to ensure that a reliable, integrated performance of the tritium recovery system could be demonstrated. An equally important goal was to measure and account for the tritium in the liquid lithium blanket module and its recovery system in order to validate the operation of the blanket module

Conceptual design of an emergency tritium clean-up system

International Nuclear Information System (INIS)

Muller, M.E.

1978-01-01

The Los Alamos Scientific Laboratory (LASL) has been selected to design, build, and operate a facility to demonstrate the operability of the tritium-related subsystems that would be required to successfully develop fusion reactor systems. Basically, these subsystems would consist of the deuterium-tritium fuel cycle and associated environmental control systems. An emergency tritium clean-up subsystem (ETC) for this facility will be designed to remove tritium from the cell atmosphere if an accident causes the primary and secondary tritium containment to be breached. Conceptually, the ETC will process cell air at the rate of 0.65 actual m 3 /s (1385 ACFM) and will achieve an overall decontamination factor of 10 6 for tritium oxide (T 2 O). Following the maximum credible release of 100 g of tritium, the ETC will restore the cell to operational status within 24 h without a significant release of tritium to the environment. The basic process will include compression of the air to 0.35 MPa (3.5 atm) in a reciprocating compressor followed by oxidation of the tritium to T 2 O in a catalytic reactor. The air will be cooled to 275 K (350 0 F) to remove most of the moisture, including T 2 O, as a condensate. The remaining moisture will be removed by molecular sieve dryer beds that incorporate a water-swamping step between beds, allowing greater T 2 O removal. A portion of the detritiated air will be recirculated to the cell; the remainder will be exhausted to the building ventilation stack to maintain a slight negative pressure in the cell. The ETC will be designed for maximum flexibility so that studies can be performed that involve various aspects of room air detritiation

Current operations and experiments at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Bartlit, J.R.; Anderson, J.L.

1985-01-01

The Tritium Systems Test Assembly (TSTA) has continued to move toward operation of a fully-integrated, full-sized, computer-controlled fusion fuel processing loop. Concurrent, nonloop experiments have answered important questions on new components and issues such as palladium diffusion membranes, ceramic electrolysis cells, regenerable tritium getters, laser Raman spectroscopy, unregenerable tritium inventory on molecular sieves, tritium contamination problems and decontamination methods, and operating data on reliability, emissions, doses, and wastes generated. 4 refs., 2 figs

Tritium systems for the TITAN reversed-field pinch fusion reactor design

International Nuclear Information System (INIS)

Martin, R.C.; Sze, D.K.; Bartlit, J.R.; Gierszewski, P.J.

1987-01-01

Tritium systems for the TITAN reversed-field pinch (RFP) fusion reactor study have been designed for two blanket concepts. The TITAN-1 design uses a self-cooled liquid-lithium blanket. The TITAN-2 design uses a self-cooled aqueous-solution blanket, with lithium nitrate dissolved in the water for tritium breeding. Tritium inventory, release, and safety margins are within regulatory limits, at acceptable costs. Major issues for TITAN-1 are plasma-driven permeation, the need for a secondary coolant loop, tritium storage requirements, redundancy in the plasma exhaust system, and minimal isotopic distillation of the exhaust. TITAN-1 fuel cleanup, reprocessing, and air detritiation systems are described in detail

Unclassified information on tritium extraction and purification technology: attachment 1

International Nuclear Information System (INIS)

McNorrill, P.L.

1976-01-01

Several tritium recovery and purification techniques developed at non-production sites are described in the unclassified and declassified literature. Heating of irradiated Li-Al alloy under vacuum to release tritium is described in declassified reports of Argonne National Laboratory. Use of palladium membranes to separate hydrogen isotopes from other gases is described by Argonne, KAPL, and others. Declassified KAPL reports describe tritium sorption on palladium beds and suggest fractional absorption as a means of isotope separation. A thermal diffusion column for tritium enrichment is described in a Canadian report. Mound Laboratory reports describe theoretical and experimental studies of thermal diffusion columns. Oak Ridge reports tabulate ''shape factors'' for thermal diffusion columns. Unclassified journals contain many articles on thermal diffusion theory, experiments, and separation of gas mixtures by thermal diffusion columns; much of these data can be readily extended to the separation of hydrogen-tritium mixtures. Cryogenic distillation for tritium recovery is described in the Mound Laboratory reports. Process equipment such as pumps, valves, Hopcalite beds, and uranium beds are described in reports by ANL, KAPL, and MLM, and in WASH-1269, Tritium Control Technology

The data collection system for failure/maintenance at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Casey, M.A.; Gruetzmacher, K.M.; Bartlit, J.R.; Cadwallader, L.C.

1988-01-01

A data collection system for obtaining information which can be used to help determine the reliability and vailability of future fusion power plants has been installed at the Los Alamos National Laboratory's Tritium Systems Test Assembly (TSTA). Failure and maintenance data on components of TSTA's tritium systems have been collected since 1984. The focus of the data collection has been TSTA's Tritium Waste Tratment System (TWT), which has maintained high availability since it became operation in 1982. Data collection is still in progress and a total of 291 failure reports are in the data collection system at this time, 47 of which are from the TWT. 6 refs., 2 figs., 2 tabs

Radiological safety system based on real-time tritium-in-air monitoring indoor and in effluents

International Nuclear Information System (INIS)

Bidica, Nicolae; Sofalca, Nicolae; Balteanu, Ovidiu-Ioan; Stefan, Ioana-Iuliana

2007-01-01

In this paper we describe an improved real-time tritium monitoring system designed for Heavy Water Detritiation Pilot Plant of National Institute for Cryogenics and Isotopes Separation, Rm. Valcea, Romania. The system consists of three fixed tritium-in-air monitors which measure continuously tritium-in-air concentration (in both species: vapour and gas) in working areas and gaseous effluents. Portable tritium monitors with ionization chamber, and tritium-in-air collector combined with liquid scintillation counter method are also used to supplement fixed instrument measurements. The main functions of tritium monitoring system are: to measure tritium-in air concentration in working areas and gaseous effluents; to alarm the personnel if tritium concentration thresholds are exceeded; to integrate tritium activity released to the environment during a week and to cut off normal ventilation when the activity threshold is exceeded and start the air cleaning system. Now, several especial functions have been added, so that by using appropriate conversion factors, the tritium monitoring system is able to estimate the effective dose rate before starting an activity into the monitored area, during this activity, or soon as the activity was finished. Another new function has been added by coupling tritium-in-air monitoring system with control access system. This is very useful for quick estimation of tritium doses. For routine dosimetric survey, one the internal dose for individuals by measuring tritium in urine is estimated. With all these features our tritium monitoring system is really a safety system for personnel and for environment. (authors)

Tritium permeation and recovery

International Nuclear Information System (INIS)

Bond, R.A.; Hamilton, A.M.

1987-01-01

The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. In this appendix, tritium transport in the DEMO breeding blanket is considered with emphasis on the permeation rate from the lithium-lead breeder into the coolant. A computational model used to calculate the tritium transport in the breeder blanket is described. Results are reported for the tritium transport in the NET/INTOR type blanket as well as the DEMO blanket in order to provide a comparison. In addition, results are presented for the helium coolant tritium extraction analysis. (U.K.)

Performances of an atmospheric tritium sampler and its application

International Nuclear Information System (INIS)

Inoue, Yoshikazu; Kahn, B.; Carter, M.W.

1983-01-01

A sampling system for atmospheric tritium in the form of water vapor, hydrogen and hydrocarbons was designed and built. The air was passed first through molecular sieve which adsorbed water vapor, then over palladium catalyst which oxidized hydrogen and adsorbed resulting water in situ, and finally over hot Hopcalite catalyst, which oxidized hydrocarbons and the resulting water was adsorbed on a following molecular sieve column. Three water samples were extracted from adsorbers and their tritium contents were measured by liquid scintillation counting. Performances of this sampler were examined for retrieval of tritiated water from molecular sieve, oxidation of hydrogen on palladium catalyst and oxidation of methane on Hopcalite. The portable sampler was applied to analyze tritium in a duct air of a heavy water moderated research reactor. More than 99% of total tritium was in vapor form. Trace amounts of tritiated hydrogen and hydrocarbon were also detected. This tritium sampler is applicable to detect all of atmospheric tritium as high as ten times of ambient levels. (author)

Advancement in tritium transport simulations for solid breeding blanket system

Energy Technology Data Exchange (ETDEWEB)

Ying, Alice, E-mail: ying@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Zhang, Hongjie [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Merrill, Brad J. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-11-01

In this paper, advancement on tritium transport simulations was demonstrated for a solid breeder blanket HCCR TBS, where multi-physics and detailed engineering descriptions are considered using a commercial simulation code. The physics involved includes compressible purge gas fluid flow, heat transfer, chemical reaction, isotope swamping effect, and tritium isotopes mass transport. The strategy adopted here is to develop numerical procedures and techniques that allow critical details of material, geometric and operational heterogeneity in a most complete engineering description of the TBS being incorporated into the simulation. Our application focuses on the transient assessment in view of ITER being pulsed operations. An immediate advantage is a more realistic predictive and design analysis tool accounting pulsed operations induced temperature variations which impact helium purge gas flow as well as Q{sub 2} composition concentration time and space evolutions in the breeding regions. This affords a more accurate prediction of tritium permeation into the He coolant by accounting correct temperature and partial pressure effects and realistic diffusion paths. The analysis also shows that by introducing by-pass line to accommodate ITER pulsed operations in the TES loop allows tritium extraction design being more cost effective.

Operating experience and procedures at the tritium systems test assembly

International Nuclear Information System (INIS)

Carlson, R.V.; Binning, K.E.; Cole, S.P.; Jenkins, E.M.; Wilhelm, R.C.; Cole, S.P.

1988-01-01

Operating procedures are important for the safe and efficient operation of the Tritium Systems Test Assembly (TSTA). TSTA has been operating for four years with tritium in a safe and efficient manner. The inventory of tritium in the process loop is 100 grams and several milestone runs have been completed. This paper describes the methods used to operate TSTA. 3 refs., 1 fig

Tritium Systems Test Assembly: design for major device fabrication review

International Nuclear Information System (INIS)

Anderson, J.L.; Sherman, R.H.

1977-06-01

This document has been prepared for the Major Device Fabrication Review for the Tritium Systems Test Assembly (TSTA). The TSTA is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for fusion reactor systems. The principal objectives for TSTA are: (a) demonstrate the fuel cycle for fusion reactor systems; (b) develop test and qualify equipment for tritium service in the fusion program; (c) develop and test environmental and personnel protective systems; (d) evaluate long-term reliability of components; (e) demonstrate long-term safe handling of tritium with no major releases or incidents; and (f) investigate and evaluate the response of the fuel cycle and environmental packages to normal, off-normal, and emergency situations. This document presents the current status of a conceptual design and cost estimate for TSTA. The total cost to design, construct, and operate TSTA through FY-1981 is estimated to be approximately $12.2 M

Monitoring system of the Tritium Research Laboratory, Sandia Laboratories, Livermore, CA

International Nuclear Information System (INIS)

Wall, W.R.; Hafner, R.S.; Westfall, D.L.; Ristau, R.D.

1978-11-01

Automated tritium monitoring is now in use at the Tritium Research Laboratory (TRL). Betatec 100 tritium monitors, along with several Sandia-designed accessories, have been combined with a PDP 11/40 computer to automatically read and record tritium concentrations of room air, containment, and cleanup systems. Each individual monitoring system, in addition to a local display in the area of interest, has a visible/audible display in the control room. Each system is then channeled into the PDP 11/40 computer, providing immediate assessment of the status of the entire laboratory from a central location. Measurement capability ranges from μCi/m 3 levels for room air monitoring to kCi/m 3 levels for glove box and cleanup systems monitoring. In this report the overall monitoring system and its capabilities are discussed, with detailed descriptions given of monitors and their components

Environmental effects of normal and off-normal releases of tritium from CTR systems

International Nuclear Information System (INIS)

McKone, T.E.

1978-08-01

Near term fusion technology will utilize the deuterium-tritium reaction. To quantify the magnitude of the hazard presented by major tritium release mechanisms, a method is presented for determining doses to the public from releases of tritium as tritiated water vapor or tritiated lithium compounds. Inclusion of this method in a computer model is described. This model uses the Gaussian dispersion method to predict distribution of tritium species in the downwind environment. Movement of tritium into biological systems is determined by treating these systems as a series of interacting water compartments. Dispersion and uptake calculations are applied to two sample sites in order to predict health effects. These effects are compared to the long range effect of introducing tritium into the world water system

TFTR tritium handling concepts

International Nuclear Information System (INIS)

Garber, H.J.

1976-01-01

The Tokamak Fusion Test Reactor, to be located on the Princeton Forrestal Campus, is expected to operate with 1 to 2.5 MA tritium--deuterium plasmas, with the pulses involving injection of 50 to 150 Ci (5 to 16 mg) of tritium. Attainment of fusion conditions is based on generation of an approximately 1 keV tritium plasma by ohmic heating and conversion to a moderately hot tritium--deuterium ion plasma by injection of a ''preheating'' deuterium neutral beam (40 to 80 keV), followed by injection of a ''reacting'' beam of high energy neutral deuterium (120 to 150 keV). Additionally, compressions accompany the beam injections. Environmental, safety and cost considerations led to the decision to limit the amount of tritium gas on-site to that required for an experiment, maintaining all other tritium in ''solidified'' form. The form of the tritium supply is as uranium tritide, while the spent tritium and other hydrogen isotopes are getter-trapped by zirconium--aluminum alloy. The issues treated include: (1) design concepts for the tritium generator and its purification, dispensing, replenishment, containment, and containment--cleanup systems; (2) features of the spent plasma trapping system, particularly the regenerable absorption cartridges, their integration into the vacuum system, and the handling of non-getterables; (3) tritium permeation through the equipment and the anticipated releases to the environment; (4) overview of the tritium related ventilation systems; and (5) design bases for the facility's tritium clean-up systems

A Next Generation Digital Counting System For Low-Level Tritium Studies (Project Report)

International Nuclear Information System (INIS)

Bowman, P.

2016-01-01

Since the early seventies, SRNL has pioneered low-level tritium analysis using various nuclear counting technologies and techniques. Since 1999, SRNL has successfully performed routine low-level tritium analyses with counting systems based on digital signal processor (DSP) modules developed in the late 1990s. Each of these counting systems are complex, unique to SRNL, and fully dedicated to performing routine tritium analyses of low-level environmental samples. It is time to modernize these systems due to a variety of issues including (1) age, (2) lack of direct replacement electronics modules and (3) advances in digital signal processing and computer technology. There has been considerable development in many areas associated with the enterprise of performing low-level tritium analyses. The objective of this LDRD project was to design, build, and demonstrate a Next Generation Tritium Counting System (NGTCS), while not disrupting the routine low-level tritium analyses underway in the facility on the legacy counting systems. The work involved (1) developing a test bed for building and testing new counting system hardware that does not interfere with our routine analyses, (2) testing a new counting system based on a modern state of the art DSP module, and (3) evolving the low-level tritium counter design to reflect the state of the science.

Modification and testing of the Sandia Laboratories Livermore tritium decontamination systems

International Nuclear Information System (INIS)

Gildea, P.D.; Birnbaum, H.G.; Wall, W.R.

1978-08-01

Sandia Laboratories, Livermore, has put into operation a new facility, the Tritium Research Laboratory. The laboratory incorporates containment and cleanup facilities such that any tritium accidentally released is captured rather than vented to the atmosphere. This containment is achieved with hermetically sealed glove boxes that are connected on demand by manifolds to two central decontamination systems called the Gas Purification System (GPS) and the Vacuum Effluent Recovery System (VERS). The primary function of the GPS is to remove tritium and tritiated water vapor from the glove box atmosphere. The primary function of the VERS is to decontaminate the gas exhausted from the glove box pressure control systems and vacuum pumps in the building before venting the gas to the stack. Both of these systems are designed to remove tritium to the few parts per billion range. Acceptance tests at the manufacturer's plant and preoperational testing at Livermore demonstrated that the systems met their design specifications. After preoperational testing the Gas Purification System was modified to enhance the safety of maintanance operations. Both the Gas Purification System and the Vacuum Effluent Recovery System were performance tested with tritium. Results show that concentraion reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass have been achieved for both systems at inlet concentrations of 1 ppM or less

Modification and testing of the Sandia Laboratories Livermore tritium decontamination systems

International Nuclear Information System (INIS)

Gildea, P.D.; Birnbaum, H.G.; Wall, W.R.

1979-01-01

Sandia Laboratories, Livermore, has put into operation a new facility, the Tritium Research Laboratory. The laboratory incorporates containment and cleanup facilities such that any tritium accidentally released is captured rather than vented to the atmosphere. This containment is achieved with hermetically sealed glove boxes that are connected on demand by manifolds to two central decontamination systems called the Gas Purification System (GPS) and the Vacuum Effluent Recovery System (VERS). The primary function of the GPS is to remove tritium and tritiated water vapor from the glove box atmosphere. The primary function of the VERS is to decontaminate the gas exhausted from the glove box pressure control systems and vacuum pumps in the building before venting the gas to the stack. Both of these systems are designed to remove tritium to the few parts per billion range. Acceptance tests at the manufacturer's plant and preoperational testing at Livermore demonstrated that the systems met their design specifications. After preoperational testing the Gas Purification System was modified to enhance the safety of maintanance operations. Both the Gas Purification System and the Vacuum Effluent Recovery System were performance tested with tritium. Results show that concentration reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass have been achieved for both systems at inlet concentrations of 1 ppM or less

Physical inventory by use of modeling for the tritium aqueous waste recovery system

International Nuclear Information System (INIS)

Sienkiewicz, C.J.; Lentz, J.E.; Wiggins, D.V.

1988-01-01

Physical inventory requirements for the Tritium Aqueous Waste Recovery System (TAWRS) presented constraints that required unique solutions. Available analytical techniques for which sound measurement control practices existed could not be readily adapted to the system without significant modifications and expense. Based on the assumption that would accurately estimate total system inventory given a few key measurements, a model was developed for TAWRS. Tritium concentrations in two streams, the tritiated feed stream to the process and the tritiated hydrogen stream generated by the electrolysis cells, provided the key values to the model. The proposed mathematical model relates the tritium concentration throughout the system to the tritium concentration in these two streams. Testing of the system using low-level tritiated feed water was conducted to characterize tritium distribution in the system and to relate key values to total inventory. 4 refs., 2 figs.,

Consideration for a tritium removal facility at the Cernavoda Nuclear Power Station

International Nuclear Information System (INIS)

2006-01-01

Full text: A pre-feasibility study considering process options for a Tritium Removal Facility at the Cernavoda Nuclear Power Station has been completed by ICIT and AECL. Three different process options were considered. These three options differ in the front-end process used to transfer tritium from heavy water to deuterium gas. All three options use cryogenic distillation (CD) as a back end process to extract tritium from the deuterium gas stream and concentrate it into a small volume stream of pure DT or T 2 that can be immobilized on a titanium sponge. The first option for the front-end process is Liquid Phase Catalytic Exchange (LPCE). The LPCE column is used to transfer the tritium from the heavy water to a recirculating stream of deuterium gas. The separation of hydrogen isotopes takes place in the cryogenic distillation column. Tritium-depleted deuterium gas from the CD system is fed back to the LPCE column. The cryogenic distillation system concentrates the tritium into a small volume of elemental tritium for storage. Tritiated heavy water that has been purified to remove catalyst poisons is fed to the top of the LPCE column. The heavy water leaving the column is depleted in deuterium. Both existing detritiation plants built to detritiate CANDU reactors (the Darlington TRF in Canada and the Wolsung TRF in Korea) use variations of the LPCE-CD process. The second option uses electrolysis to convert tritiated heavy water into oxygen and tritiated deuterium gas. The deuterium gas is sent to the Cryogenic Distillation system to extract and concentrate the tritium. The tritium depleted deuterium gas is recombined with the electrolytic oxygen to give a tritium-depleted heavy water product. The third option uses a Combined Electrolysis and Catalytic Exchange (CECE) front end. A CECE process concentrates the tritium in the water and, using water electrolysis, converts the concentrated tritium into deuterium gas. An overhead catalytic recombiner converts the

Investigations of titamium and zirconium hydrides to determine suitability of recoverable tritium immobilization for the Pickering tritium removal system

International Nuclear Information System (INIS)

Noga, J.O.

1981-11-01

A tritium removal system will be constructed at Pickering Nuclear Generating station to reduce the adverse effects of this radioactive hydrogen isotope. This report summarizes various properties of titanium and zirconium sponge hydrides which have been selected as suitable candidates for tritium product immobilization. Equilibrium pressure-composition-temperature data indicates that both materials behave suitably to provide a safe, solid form of tritium storage. Titanium tritide is recommended as the best choice due to higher dissociation pressures which can be achieved at equivalent temperatures when compared to zirconium tritide. Higher dissociation pressures would result in faster and more efficient recovery of tritium gas from the immobilized state. It is evident from the stability of these compounds that their utilization as tritides will greatly enhance the integrity of tritium storage

Tritium fuel cycle modeling and tritium breeding analysis for CFETR

Energy Technology Data Exchange (ETDEWEB)

Chen, Hongli; Pan, Lei; Lv, Zhongliang; Li, Wei; Zeng, Qin, E-mail: zengqin@ustc.edu.cn

2016-05-15

Highlights: • A modified tritium fuel cycle model with more detailed subsystems was developed. • The mean residence time method applied to tritium fuel cycle calculation was updated. • Tritium fuel cycle analysis for CFETR was carried out. - Abstract: Attaining tritium self-sufficiency is a critical goal for fusion reactor operated on the D–T fuel cycle. The tritium fuel cycle models were developed to describe the characteristic parameters of the various elements of the tritium cycle as a tool for evaluating the tritium breeding requirements. In this paper, a modified tritium fuel cycle model with more detailed subsystems and an updated mean residence time calculation method was developed based on ITER tritium model. The tritium inventory in fueling system and in plasma, supposed to be important for part of the initial startup tritium inventory, was considered in the updated mean residence time method. Based on the model, the tritium fuel cycle analysis of CFETR (Chinese Fusion Engineering Testing Reactor) was carried out. The most important two parameters, the minimum initial startup tritium inventory (I{sub m}) and the minimum tritium breeding ratio (TBR{sub req}) were calculated. The tritium inventories in steady state and tritium release of subsystems were obtained.

A programmable autosampler for a field deployable tritium analysis system

International Nuclear Information System (INIS)

Hofstetter, K.J.; Cable, P.R.; Beals, D.M.; Jones, J.

1996-01-01

Researchers in the Environmental Technology Section of the Savannah River Technology Center, in cooperation with Sampling Systems, Inc. are developing a fully programmable, remotely operated, fixed volume, automatic sampler for use with the field deployable tritium analysis system currently under development at U. of GA's Center for Applied Isotope Studies. The sampler will collect a limited-volume sample and perform on-line sample purification for tritium analyses from multiple collection sites. Pneumatically operated stainless steel samplers operate satisfactorily upon remote activation. The one-step purification system removes all impurities with interfere with tritium analysis by liquid scintillation. Field testing has confirmed system operation. The autosampler may act as a stand-alone device and is enclosed in a rugged, field-portable case with wheels. The system weighs about 40 lbs

Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

Tritium monitoring at the Sandia Tritium Research Laboratory

International Nuclear Information System (INIS)

Devlin, T.K.

1978-10-01

Sandia Laboratories at Livermore, California, is presently beginning operation of a Tritium Research Laboratory (TRL). The laboratory incorporates containment and cleanup facilities such that any unscheduled tritium release is captured rather than vented to the atmosphere. A sophisticated tritium monitoring system is in use at the TRL to protect operating personnel and the environment, as well as ensure the safe and effective operation of the TRL decontamination systems. Each monitoring system has, in addition to a local display, a display in a centralized control room which, when coupled room which, when coupled with the TRL control computer, automatically provides an immediate assessment of the status of the entire facility. The computer controls a complex alarm array status of the entire facility. The computer controls a complex alarm array and integrates and records all operational and unscheduled tritium releases

Tritium supply assessment for ITER and DEMOnstration power plant

International Nuclear Information System (INIS)

Ni, Muyi; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

2013-01-01

Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource

Tritium supply assessment for ITER and DEMOnstration power plant

Energy Technology Data Exchange (ETDEWEB)

Ni, Muyi, E-mail: muyi.ni@fds.org.cn; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

2013-10-15

Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

Tritium confinement in a new tritium processing facility at the Savannah River Site

International Nuclear Information System (INIS)

Heung, L.K.; Owen, J.H.; Hsu, R.H.; Hashinger, R.F.; Ward, D.E.; Bandola, P.E.

1991-01-01

A new tritium processing facility, named the Replacement Tritium Facility (RTF), has been completed and is being prepared for startup at the Savannah River Site (SRS). The RTF has the capability to recover, purify and separate hydrogen isotopes from recycled gas containers. A multilayered confinement system is designed to reduce tritium losses to the environment. This confinement system is expected to confine and recover any tritium that might escape the process equipment, and to maintain the tritium concentration in the nitrogen glovebox atmosphere to less than 10 -2 μCi/cc tritium

JAERI Fuel Cleanup System (J-FCU) stand-alone tritium test at the TSTA

International Nuclear Information System (INIS)

Konishi, Satoshi; Hayashi, Takumi; Inoue, Masahiko

1993-03-01

JAERI designed, fabricated, and installed the JAERI Fuel Cleanup System (J-FCU) as a subsystem of simulated fusion fuel loop at the TSTA. The main function of the J-FCU is to purify and to recover hydrogen isotopes from simulated plasma exhaust while exhausting tritium free impurities. After a lot of deuterium tests, a first tritium test of the J-FCU was performed with one gram of tritium at the TSTA on June 1991. Main purpose of this test was to evaluate the total integrity and function of the J-FCU system with a DT mixture. Through this test, the J-FCU was operated well and its function with tritium was demonstrated. This report describes the detail test results of the J-FCU first tritium test and discuss its functions by stand-alone mode. Residual tritium inventory of the J-FCU system was also discussed. (author)

Pre-Conceptual Design for Northstar ⁹⁹Mo Process Tritium Removal System

Energy Technology Data Exchange (ETDEWEB)

Nobile, Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichert, Heidi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, William Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taylor, Craig Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gordon, John Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-01-12

In this report we describe a preliminary concept for a Tritium Removal System (TRS) to remove tritium that is generated in the ⁹⁹Mo production process. Preliminary calculations have been performed to evaluate an approximate size for the system. The concept described utilizes well-established detritiation technology based on catalytic oxidation of tritium and tritiated hydrocarbons to water in a high temperature (400 °C) reactor and capture of water in a molecular sieve bed. The TRS concept involves use of a single system that would cycle through each of the seven online target systems and remove tritium that has been accumulated after one week’s run time. The TRS would perform cleanup operations on each target system for a period of approximately 24 hours. This would occur while the system is still online and just prior to target replacement, so tritium levels would at their minimum values for target replacement. In the concept, during normal operation a small fraction (1%) of the helium recirculating in the system would be diverted through the TRS and returned to the flow loop. With this approach sufficient levels of detritiation can be accomplished in a 24 hour period. In the study it was found that because of the need to maintain low oxygen levels in the system (<100 ppm) this increases the size of the catalytic reactor. As a result of this finding, consideration should be given to other methods for removing tritium from the system. Other methods such as catalytic exchange of tritium with an unsaturated organic compound and subsequent trapping on activated carbon or molecular sieve could offer advantages of reducing reactor size and operation at lower reactor temperature. However the most significant advantage of such an approach would be the ability to operate in very low oxygen environments, which would eliminate any concerns for oxidation of the target.

Radiological safety system based on real-time tritium-in-air monitoring indoors and in effluents

International Nuclear Information System (INIS)

Bidica, N.; Sofalca, N; Balteanu, O.; Srefan, I.

2006-01-01

Exposure to tritium is an important health hazard in any tritium processing facility so that implementing a real-time tritium monitoring system is necessary for its operation in safety conditions. The tritium processing facility operators need to be informed at any time about the in-air tritium concentration indoors or in the stack effluents, in order to detect immediately any leaks in tritium containments, or any releases inside the buildings or to the environment. This information is very important for adopting if necessary protection measures and correcting actions as quickly as possible. In this paper we describe an improved real-time tritium monitoring system designed for the Heavy Water Detritiation Pilot Plant of National Institute for Cryogenics and Isotopes Separation, Rm. Valcea, Romania. The design of the Radiological Safety System implemented for the ICIT Water Detritiation Pilot Plant is intended to provide the maximum safety level based on the ALARA concept. The main functions of tritium monitoring system are: - monitoring the working areas and gaseous effluents by determination of the tritium-in-air activity concentration; - local and remote data display; - assessing of environment dose equivalent rates and dose equivalents in the working environment (for personnel exposure control and work planning); - assessing the total tritium activity released to the environment through ventilation exhaust stack; - safety functions, i.e., local and remote, locking/unlocking personnel access, process shut-down in emergency conditions and start of the air cleaning systems. With all these features our tritium monitoring system is really a safety system adequate for personnel and environmental protection. (authors)

Tritium inventory tracking and management

International Nuclear Information System (INIS)

Eichenberg, T.W.; Klein, A.C.

1990-01-01

This investigation has identified a number of useful applications of the analysis of the tracking and management of the tritium inventory in the various subsystems and components in a DT fusion reactor system. Due to the large amounts of tritium that will need to be circulated within such a plant, and the hazards of dealing with the tritium an electricity generating utility may not wish to also be in the tritium production and supply business on a full time basis. Possible scenarios for system operation have been presented, including options with zero net increase in tritium inventory, annual maintenance and blanket replacement, rapid increases in tritium creation for the production of additional tritium supplies for new plant startup, and failures in certain system components. It has been found that the value of the tritium breeding ratio required to stabilize the storage inventory depends strongly on the value and nature of other system characteristics. The real operation of a DT fusion reactor power plant will include maintenance and blanket replacement shutdowns which will affect the operation of the tritium handling system. It was also found that only modest increases in the tritium breeding ratio are needed in order to produce sufficient extra tritium for the startup of new reactors in less than two years. Thus, the continuous operation of a reactor system with a high tritium breeding ratio in order to have sufficient supplies for other plants is not necessary. Lastly, the overall operation and reliability of the power plant is greatly affected by failures in the fuel cleanup and plasma exhaust systems

Separation of tritium from gaseous and aqueous effluent systems

International Nuclear Information System (INIS)

Kobisk, E.H.

1977-01-01

Three processes are discussed for separating tritium from gaseous and aqueous effluent systems: separation in the gas phase using Pd-25 wt percent Ag alloy diffusion membranes; electrolytic separation in the aqueous phase using ''bipolar'' electrodes; and the countercurrent exchange of tritium-containing hydrogen gas with water on catalytic surfaces combined with separation by direct electrolysis

Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

Energy Technology Data Exchange (ETDEWEB)

Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Oji, L.N.

1997-11-14

Under the Tritium Facility Modernization {ampersand} Consolidation (TFM{ampersand}C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM{ampersand}C Project also provides for a new replacement R&D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H.

Development of tritium plant system for fusion reactors. Achievements in the 14-year US-Japan collaboration

International Nuclear Information System (INIS)

Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

2003-01-01

Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safe-operation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore, distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work. (author)

Development of Tritium Plant System for Fusion Reactors - Achievements in the 14-year US-Japan Collaboration -

Science.gov (United States)

Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safeoperation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore,distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work.

Tritium emissions reduction facility (TERF)

International Nuclear Information System (INIS)

Lamberger, P.H.; Hedley, W.H.

1993-01-01

Tritium handling operations at Mound include production of tritium-containing devices, evaluation of the stability of tritium devices, tritium recovery and enrichment, tritium process development, and research. In doing this work, gaseous process effluents containing 400,000 to 1,000,000 curies per year of tritium are generated. These gases must be decontaminated before they can be discharged to the atmosphere. They contain tritium as elemental hydrogen, as tritium oxide, and as tritium-containing organic compounds at low concentrations (typically near one ppm). The rate at which these gases is generated is highly variable. Some tritium-containing gas is generated at all times. The systems used at Mound for capturing tritium from process effluents have always been based on the open-quotes oxidize and dryclose quotes concept. They have had the ability to remove tritium, regardless of the form it was in. The current system, with a capacity of 1.0 cubic meter of gas per minute, can effectively remove tritium down to part-per-billion levels

Automation system for tritium contaminated surface monitoring

International Nuclear Information System (INIS)

Culcer, Mihai; Iliescu, Mariana; Curuia, Marian; Raceanu, Mircea; Enache, Adrian; Stefanescu, Ioan; Ducu, Catalin; Malinovschi, Viorel

2005-01-01

The low energy of betas makes tritium difficult to detect. However, there are several methods used in tritium detection, such as liquid scintillation and ionization chambers. Tritium on or near a surface can be also detected using proportional counter and, recently, solid state devices. The paper presents our results in the design and achievement of a surface tritium monitor using a PIN photodiode as a solid state charged particle detector to count betas emitted from the surface. That method allows continuous, real-time and non-destructively measuring of tritium. (authors)

Secondary containment system for a high tritium research cryostat

International Nuclear Information System (INIS)

Tsugawa, R.T.; Fearon, D.; Souers, P.C.; Hickman, R.G.; Roberts, P.E.

1976-01-01

A 4.2- to 300-K liquid helium cryostat has been constructed for cryogenic samples of D--T containing up to 4 x 10 14 dis/s (10,000 Ci) of tritium radioactivity. The cryostat is enclosed in a secondary box, which acts as the ultimate container in case of a tritium release. Dry argon is flushed through the box, and the box atmosphere is monitored for tritium, oxygen, and water vapor. A rupture disk and abort tank protect the box atmosphere in case the sample cell breaks. If tritium breaks into the box, a powdered uranium getter trap reduces the 4 x 10 14 dis/s (10,000 Ci) to 4 x 10 9 dis/s (0.1 Ci) in 24 h. A backup palladium--zeolite getter system goes into operation if an overabundance of oxygen contaminates the uranium getter

Process for recovering tritium from molten lithium metal

Science.gov (United States)

Maroni, Victor A.

1976-01-01

Lithium tritide (LiT) is extracted from molten lithium metal that has been exposed to neutron irradiation for breeding tritium within a thermonuclear or fission reactor. The extraction is performed by intimately contacting the molten lithium metal with a molten lithium salt, for instance, lithium chloride - potassium chloride eutectic to distribute LiT between the salt and metal phases. The extracted tritium is recovered in gaseous form from the molten salt phase by a subsequent electrolytic or oxidation step.

Tritium accounting for PHWR plants

International Nuclear Information System (INIS)

Nair, P.S.; Duraisamy, S.

2012-01-01

Tritium, the radioactive isotope of hydrogen, is produced as a byproduct of the nuclear reactions in the nuclear power plants. In a Pressurized Heavy Water Reactor (PHWR) tritium activity is produced in the Heat Transport and Moderator systems due to neutron activation of deuterium in heavy water used in these systems. Tritium activity build up occurs in some of the water systems in the PHWR plants through pick up from the plant atmosphere, inadvertent D 2 O ingress from other systems or transfer during processes. The tritium, produced by the neutron induced reactions in different systems in the reactor undergoes multiple processes such as escape through leaks, storage, transfer to external locations, decay, evaporation and diffusion and discharge though waste streams. Change of location of tritium inventory takes place during intentional transfer of heavy water, both reactor grade and downgraded, from one system to another. Tritium accounting is the application of accounting techniques to maintain knowledge of the tritium inventory present in different systems of a facility and to construct activity balances to detect any discrepancy in the physical inventories. It involves identification of all the tritium hold ups, transfers and storages as well as measurement of tritium inventories in various compartments, decay corrections, environmental release estimations and evaluation of activity generation during the accounting period. This paper describes a methodology for creating tritium inventory balance based on periodic physical inventory taking, tritium build up, decay and release estimations. Tritium accounting in the PHWR plants can prove to be an effective regulatory tool to monitor its loss as well as unaccounted release to the environment. (author)

Tritium in HTR systems

International Nuclear Information System (INIS)

Steinwarz, W.

1987-07-01

Starting from the basis of the radiological properties of tritium, the provisions of present-day radiation protection legislation are discussed in the context of the handling of this radionuclide in HTR plants. Tritium transportation is then followed through from the place of its creation up until the sink, i.e. disposal and/or environmental route, and empirical values obtained in experiments and in plant operation translated into guidelines for plant design and planning. The use of the example of modular HTR plants permits indication that environmental contamination via the 'classical' routes of air and water emissions, and contamination of products, and resulting consumer exposure, are extremely low even on the assumption of extreme conditions. This leads finally to a requirement that the expenditure for implementation of measures for further reduction of tritium activity rates be measured against low radiological effect. (orig.) [de

A Gas Target with a Tritium Gas Handling System

Energy Technology Data Exchange (ETDEWEB)

Holmqvist, B; Wiedling, T

1963-12-15

A detailed description is given of a simple tritium gas target and its tritium gas filling system, and how to put it into operation. By using the T (p,n) He reaction the gas target has been employed for production of monoenergetic fast neutrons of well defined energy and high intensity. The target has been operated successfully for a long time.

Electrochemical measurement of tritium and hydrogen permeation through iron membranes

International Nuclear Information System (INIS)

Hagi, Hideki; Hayashi, Yasunori

1988-01-01

Permeation rates of tritium and hydrogen through iron were measured by the electrochemical method in which an aqueous solution containing 3.7 x 10 12 Bq/m 3 tritium was used as a cathodic electrolyte. Tritium and hydrogen were introduced from one side of a specimen by cathodic polarization with a constant current density, while at the other side of the specimen the permeated tritium and hydrogen were extracted by potentiostatical ionization. Nearly all of the potentiostatic current of the extraction side stands for the ionization of hydrogen, because the concentration of tritium in the cathodic electrolyte is very small. The amount of permeated hydrogen was obtained by integrating the anodic current, and that of tritium was determined by measuring the radioactivity of the electrolyte sampled from the extraction side. The separation factor for permeation obtained under steady state conditions (the ratio of permeation rates of hydrogen to tritium divided by the ratio of the concentration of hydrogen to tritium in the charging electrolyte) is 12 at 288 K. This value is independent of cathodic current density. Diffusion coefficients of tritium (D T ) and hydrogen (D H ) in iron were determined from the time lag of tritium and hydrogen permeation. For annealed specimens at 286 K, D T = 9 x 10 -10 m 2 /s and D H = 4 x 10 -9 m 2 /s, and for 9% cold-worked specimens at 284 K, D T = 3 x 10 -10 m 2 /s and D H = 4 x 10 -10 m 2 /s. (author)

A system dynamics model for stock and flow of tritium in fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kwon, Saerom [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Sakamoto, Yoshiteru; Yamanishi, Toshihiko; Tobita, Kenji [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori-ken 039-3212 (Japan)

2015-10-15

Highlights: • System dynamics model of tritium fuel cycle was developed for analyzing stock and flow of tritium in fusion power plants. • Sensitivity of tritium build-up to breeding ratio parameters has been assessed to two plant concepts having 3 GW and 1.5 GW fusion power. • D-D start-up absolutely without initial loading of tritium is possible for both of the 3 GW and 1.5 GW fusion power plant concepts. • Excess stock of tritium is generated by the steady state operation with the value of tritium breeding ratio over unity. - Abstract: In order to analyze self-efficiency of tritium fuel cycle (TFC) and share the systems thinking of TFC among researchers and engineers in the vast area of fusion reactor technology, we develop a system dynamics (SD) TFC model using a commercial software STELLA. The SD-TFC model is illustrated as a pipe diagram which consists of tritium stocks, such as plasma, fuel clean up, isotope separation, fueling with storage and blanket, and pipes connecting among them. By using this model, we survey a possibility of D-D start-up without initial loading of tritium on two kinds of fusion plant having different plasma parameters. The D-D start-up scenario can reduce the necessity of initial loading of tritium through the production in plasma by D-D reaction and in breeding blanket by D-D neutron. The model is also used for considering operation scenario to avoid excess stock of tritium which must be produced at tritium breeding ratio over unity.

Tritium transport and control in the FED

International Nuclear Information System (INIS)

Rogers, M.L.

1981-01-01

The tritium systems for the FED have three primary purposes. The first is to provide tritium and deuterium fuel for the reactor. This fuel can be new tritium or deuterium delivered to the plant site, or recycled DT from the reactor that must be processed before it can be recycled. The second purpose of the FED tritium systems is to provide state-of-the-art tritium handling to limit worker radiation exposure and to minimize tritium losses to the environment. The final major objective of the FED tritium systems is to provide an integrated system test of the tritium handling technology necessary to support the fusion reactor program. Every effort is being made to incorporate available information from the Tritium System Test Assembly (TSTA) at Los Alamos National Laboratory, the Tokamak Fusion Test Reactor (TFTR) tritium systems, and the tritium handling information generated within DOE for the past 20 years

Pre-operational HTO/HT surveys in the vicinity of the Chalk River Laboratories tritium extraction plant

International Nuclear Information System (INIS)

Workman, W.J.G.; Brown, R.M.

1993-08-01

Surveys of the concentrations of HT and HTO in the atmosphere downwind of the Chalk River Laboratories reactor facilities were carried out in 1986 November, and in 1989 March, April and September under different conditions of air temperature, wind direction, and snow or vegetative cover. HT usually amounted to 1-5% of total tritium, but values up to 20% were observed, probably resulting from preferential removal of HTO. In all of the surveys, the greater persistence in the atmosphere of HT than of HTO was evident. The existing levels of HT are such that they will not be augmented significantly by chronic releases from the Tritium Extraction Plant (TEP) when it comes into operation. Hence, operation of the TEP will not facilitate studies of the environmental behaviour of chronically released HT. However, longer term studies of the distribution of HT from the existing facilities would be worthwhile. Soil and vegetation HTO levels in the study area are reported. Further studies of the distribution of tritium between the air, soil and vegetation in areas subjected to chronic exposure would be valuable

Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

International Nuclear Information System (INIS)

Hsu, R.H.; Oji, L.N.

1997-01-01

Under the Tritium Facility Modernization ampersand Consolidation (TFM ampersand C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM ampersand C Project also provides for a new replacement R ampersand D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H

A new combination of membranes and membrane reactors for improved tritium management in breeder blanket of fusion machines

International Nuclear Information System (INIS)

Demange, D.; Staemmler, S.; Kind, M.

2011-01-01

Tritium used as fuel in future fusion machines will be produced within the breeder blanket. The tritium extraction system recovers the tritium to be routed into the inner-fuel cycle of the machine. Accurate and precise tritium accountancy between both systems is mandatory to ensure a reliable operation. Handling in the blanket huge helium flow rates containing tritium as traces in molecular and oxide forms is challenging both for the process and the accountancy. Alternative tritium processes based on combinations of membranes and membrane reactors are proposed to facilitate the tritium management. The PERMCAT process is based on counter-current isotope swamping in a palladium membrane reactor. It allows recovering tritium efficiently from any chemical species. It produces a pure hydrogen stream enriched in tritium of advantage for integration upstream of the accountancy stage. A pre-separation and pre-concentration stage using new zeolite membranes has been studied to optimize the whole process. Such a combination could improve the tritium processes and facilitate accountancy in DEMO.

Development of a low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic

International Nuclear Information System (INIS)

Pawelko, R.; Shimada, M.; Katayama, K.; Fukada, S.; Terai, T.

2014-01-01

A new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology is operational at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The system is similar to a hydrogen/deuterium permeation measurement system developed at Kyushu University and also incorporates lessons learned from previous tritium permeation experiments conducted at the STAR facility. This paper describes the experimental system that is configured specifically to measure tritium mass transfer properties at low tritium partial pressures. We present preliminary tritium permeation results for α-Fe and α-Fe/LLE samples at 600degC and at tritium partial pressures between 1.0E-3 and 2.4 Pain helium. The preliminary results are compared with literature data. (author)

ITER fuel cycle systems layout

International Nuclear Information System (INIS)

Kveton, O.K.

1990-10-01

The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

International Nuclear Information System (INIS)

Guo Chunqiu; Xie Jiachun; Liu Xingmin

2013-01-01

In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

Vacuum sieve tray for tritium extraction from liquid Pb-17Li

Energy Technology Data Exchange (ETDEWEB)

Okino, Fumito, E-mail: fumito.okino@iae.kyoto-u.ac.jp [Graduate School of Energy Science, Kyoto University, Uji, Kyoto 611-0011 (Japan); Noborio, Kazuyuki [Institute of Sustainability Science, Kyoto University, Uji, Kyoto 611-0011 (Japan); Yamamoto, Yasushi [Faculty of Engineering Science, Kansai University, 3-3-35 Suita-shi, Osaka 564-8680 (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan)

2012-08-15

Formation of droplet of liquid Li-17Pb released from a nozzle into vacuum was studied for the evaluation of the feasibility as a tritium extraction process. Size of droplets formed from the nozzles was estimated by theoretical and experimental methods. For the theoretical estimation, the non-dimensional comparison of the physical bulk property of liquid Pb-17Li with water (H{sub 2}O) at ambient temperature was applied. It was found to be reasonable to apply the Plateau-Rayleigh-Instability theory for the droplet size formula of the fluid Pb-17Li for the nozzle diameter 0.4 mm-1.0 mm, temperature 400 Degree-Sign C-500 Degree-Sign C, at initial velocity of 3 m/s. The experimental results of the droplet size showed good agreement with the theory. This device was used for the parametric study of extraction of deuterium during their free fall in vacuum. The scaling of the device suggests the engineering feasibility of the process.

Tritium dosimetry and standardization

International Nuclear Information System (INIS)

Balonov, M.I.

1983-01-01

Actual problem of radiation hygiene such as an evaluation of human irradiation hazard due to a contact with tritium compounds both in industrial and public spheres is under discussion. Sources of tritium release to environment are characterized. Methods of tritium radiation monitoring are discussed. Methods of dosimetry of internal human exposure resulted from tritium compounds are developed on the base of modern representations on metbolism and tritium radiobiological effect. A system of standardization of permissible intake of tritium compounds for personnel and persons of population is grounded. Some protection measures are proposed as applied to tritium overdosage

The effective cost of tritium for tokamak fusion power reactors with reduced tritium production systems

International Nuclear Information System (INIS)

Gilligan, J.G.; Evans, K.

1983-01-01

If sufficient tritium cannot be produced and processed in tokamak blankets then at least two alternatives are possible. Tritium can be purchased; or reactors with reduced tritium (RT) content in the plasma can be designed. The latter choice may require development of magnet technology etc., but the authors show that the impact on the cost-of-electricity may be mild. Cost tradeoffs are compared to the market value of tritium. Adequate tritium production in fusion blankets is preferred, but the authors show there is some flexibility in the deployment of fusion if this is not possible

Tritium Systems Test Assembly operator training program

International Nuclear Information System (INIS)

Kerstiens, F.L.

1985-01-01

Proper operator training is needed to help ensure the safe operation of fusion facilities by personnel who are qualified to carry out their assigned responsibilities. Operators control and monitor the Tritium Systems Test Assembly (TSTA) during normal, emergency, and maintenance phases. Their performance is critical both to operational safety, assuring no release of tritium to the atmosphere, and to the successful simulation of the fusion reaction progress. Through proper training we are helping assure that TSTA facility operators perform their assignments in a safe and efficient manner and that the operators maintain high levels of operational proficiency through continuing training, retraining, requalification, and recertification

Trace hydrogen extraction from liquid lithium tin alloy

International Nuclear Information System (INIS)

Xie Bo; Hu Rui; Xie Shuxian; Weng Kuiping

2010-01-01

In order to finish the design of tritium extraction system (TES) of fusion fission hybrid reactor (FFHR) tritium blanket, involving the dynamic mathematical model of liquid metal in contact with a gaseous atmosphere, approximate mathematical equation of tritium in lithium tin alloy was deduced. Moreover, carrying process used for trace hydrogen extraction from liquid lithium tin alloy was investigated with hydrogen being used to simulate tritium in the study. The study results indicate that carrying process is effective way for hydrogen extraction from liquid lithium tin alloy, and the best flow velocity of carrier gas is about 4 L/min under 1 kg alloy temperatures and carrying numbers are the main influencing factors of hydrogen number. Hydrogen extraction efficiency can reach 85% while the alloy sample is treated 6 times at 823 K. (authors)

Analysis on tritium permeation in tritium storage bed with gas flowing calorimetry

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hirofumi; Hayashi, Takumi; Suzuki, Takumi; Nishi, Masataka [Japan Atomic Energy Research Inst., Naka Fusion Research Establishment, Department of Fusion Engineering Research, Naka, Ibaraki (Japan); Yoshida, Hiroshi [Japan Atomic Energy Research Inst., Naka Fusion Research Establishment, ITER-Joint Centeral Team, Naka, Ibaraki (Japan)

2000-10-01

Tritium permeation amount in a tritium storage bed with gas flowing calorimetric was evaluated under a condition of new operation mode for International Thermonuclear Experimental Reactor (ITER). As a result, tritium permeation under the new operation mode was estimated to be about twice of that under the practical operation mode. This result show that it would be regardless in a view point of material control of tritium, however, it was suggested to be required additional tritium removal or evacuate system in a view points of safety control or performance of accountability or thermal insulating of the tritium storage bed. (author)

Computer control of the TFTR tritium storage and delivery system

International Nuclear Information System (INIS)

Youssef, N.; Phillips, H.; Yemin, L.; Dong, J.; Pierce, C.

1980-01-01

The Tritium Storage and Delivery System (TSDS) will deliver to the torus the required tritium gas in precisely controlled injection profiles. This system will utilize advanced Central Instrumentation, Control and Data Acquisition (CICADA) computer-control techniques, in normal and malfunction-recovery modes of operation. The control scheme of the TSDS is built of three main control scenarios. An operating mode defines the permissives, sequence and path of a process during each scenario. The computerized control of the TSDS has four distinct advantages: (1) versatile control with fast response times both for tritium gas generation and for gas injection into the torus; (2) ease of selecting the proper operating modes of a control scenario, (3) ease of operation without disturbing the multiple levels of containment, and (4) simple fast trouble shooting of system malfunction utilizing programmed procedures and on-line diagnosis. The TSDS has both remote nd local control capability

Evaluation of tritium release behavior from Li{sub 2}TiO{sub 3} during DT neutron irradiation by use of an improved tritium collection method

Energy Technology Data Exchange (ETDEWEB)

Edao, Yuki, E-mail: edao.yuki@jaea.go.jp [Tritium Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kawamura, Yoshinori [Blanket Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Hoshino, Tsuyoshi [Breeding Functional Materials Development Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Ochiai, Kentaro [BA Project Coordination Group, Department of Fusion Power Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

2016-11-15

Highlights: • Behavior of tritium released from Li{sub 2}TiO{sub 3} under neutron irradiation was measured. • Tritium collection method with hydrophobic catalyst was demonstrated successfully. • Temperature of Li{sub 2}TiO{sub 3} was dominant to control the chemical form of tritium release. - Abstract: The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li{sub 2}TiO{sub 3} during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li{sub 2}TiO{sub 3}. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li{sub 2}TiO{sub 3} as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li{sub 2}TiO{sub 3} by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li{sub 2}TiO{sub 3} irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li{sub 2}TiO{sub 3} was the dominant parameter to control the chemical form of tritium released from the Li{sub 2}TiO{sub 3}.

Tritium conference days; Journees tritium

Energy Technology Data Exchange (ETDEWEB)

Garnier-Laplace, J.; Lebaron-Jacobs, L.; Sene, M.; Devin, P.; Chretien, V.; Le Guen, B.; Guetat, Ph.; Baglan, N.; Ansoborlo, E.; Boyer, C.; Masson, M.; Bailly-Du-Bois, P.; Jenkinson, St.; Wakeford, R.; Saintigny, Y.; Romeo, P.H.; Thompson, P.; Leterq, D.; Chastagner, F.; Cortes, P.; Philippe, M.; Paquet, F.; Fournier, M.

2009-07-01

This document gathers the slides of the available presentations given during this conference day. Twenty presentations out of 21 are assembled in the document and deal with: 1 - tritium in the environment (J. Garnier-Laplace); 2 - status of knowledge about tritium impact on health (L. Lebaron-Jacobs); 3 - tritium, discrete but present everywhere (M. Sene); 4 - management of tritium effluents from Areva NC La Hague site - related impact and monitoring (P. Devin); 5 - tritium effluents and impact in the vicinity of EDF's power plants (V. Chretien and B. Le Guen); 6 - contribution of CEA-Valduc centre monitoring to the knowledge of atmospheric tritiated water transfers to the different compartments of the environment (P. Guetat); 7 - tritium analysis in environment samples: constraints and means (N. Baglan); 8 - organically-linked tritium: the analyst view (E. Ansoborlo); 9 - study of tritium transfers to plants via OBT/HTO{sub air} and OBT/HTO{sub free} (C. Boyer); 10 - tritium in the British Channel (M. Masson and P. Bailly-Du-Bois); 11 - tritium in British coastal waters (S. Jenkinson); 12 - recent results from epidemiology (R. Wakeford); 13 - effects of tritiated thymidine on hematopoietic stem cells (P.H. Romeo); 14 - tritium management issue in Canada: the point of view from authorities (P. Thompson); 15 - experience feedback of the detritiation process of Valduc centre (D. Leterq); 16 - difficulties linked with tritiated wastes confinement (F. Chastagner); 17 - optimisation of tritium management in the ITER project (P. Cortes); 18 - elements of thought about the management of tritium generated by nuclear facilities (M. Philippe); 19 - CIPR's position about the calculation of doses and risks linked with tritium exposure (F. Paquet); 20 - tritium think tanks (M. Fournier). (J.S.)

A novel portable system for detecting and measuring tritium

International Nuclear Information System (INIS)

Barillari, Domenico

2007-01-01

A novel tritium detector configuration is described based on the anthracene scintillation method. Tritium-bearing samples are applied to a plate-bearing finely sublimed anthracene crystals and viewed in a field-able PMT-based reader against a standard plate. A microprocessor-based control and signal analysis system delivers a reading with a sensitivity of better than 5 nCi 3 H in approximately 3 min, and 2.3 nCi in 10 min of counting

Tritium monitoring techniques

International Nuclear Information System (INIS)

DeVore, J.R.; Buckner, M.A.

1996-05-01

As part of their operations, the U.S. Navy is required to store or maintain operational nuclear weapons on ships and at shore facilities. Since these weapons contain tritium, there are safety implications relevant to the exposure of personnel to tritium. This is particularly important for shipboard operations since these types of environments can make low-level tritium detection difficult. Some of these ships have closed systems, which can result in exposure to tritium at levels that are below normally acceptable levels but could still cause radiation doses that are higher than necessary or could hamper ship operations. This report describes the state of the art in commercial tritium detection and monitoring and recommends approaches for low-level tritium monitoring in these environments

Tritium containment of controlled thermonuclear fusion reactor

International Nuclear Information System (INIS)

Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

1979-01-01

It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

Environmental monitoring for tritium in tritium separation facility

International Nuclear Information System (INIS)

Varlam, Carmen; Stefanescu, Ioan; Steflea, Dumitru; Lazar, Roxana Elena

2001-01-01

The Cryogenic Pilot is an experimental project in the nuclear energy national research program, which has the aim of developing technologies for tritium and deuterium separation by cryogenic distillation. The experimental installation is located 15 km near the highest city of the area and 1 km near Olt River. An important chemical activity is developed in the area and chemical plants make up almost entire neighborhood of the Experimental Cryogenic Pilot. It is necessary to emphasize this aspect because the hall sewage system of the pilot is connected with the one of other three chemical plants from vicinity. This is the reason why we progressively established elements of an environmental monitoring program well in advance of tritium operation in order to determine baseline levels. The first step was the tritium level monitoring in environmental water and sewage from neighboring industrial activity. In this work, a low background liquid scintillation was used to determine tritium activity concentration according to ISO 9698/1998 standard. We measured drinking water, precipitation, river water, underground water and wastewater. The tritium level was between 10 TU and 27 TU what indicates that there is no source of tritium contamination in the neighborhood of Cryogenic Pilot. In order to determine baseline levels we decided to monitor monthly each location. In this paper it is presented a standard method used for tritium determination in water samples, the precautions needed to achieve reliable results and the evolution of tritium level in different location near the Experimental Pilot for Tritium and Deuterium Cryogenic Separation. (authors)

Environmental monitoring for tritium at tritium separation facility

International Nuclear Information System (INIS)

Varlam, C.; Stefanescu, I.; Steflea, D.; Lazar, R.E.

2001-01-01

The Cryogenic Pilot is an experimental project in the nuclear energy national research program, which has the aim of developing technologies for tritium and deuterium separation by cryogenic distillation. The experimental installation is located 15 km near the highest city of the area and 1 km near Olt River. An important chemical activity is developed in the area and the Experimental Cryogenic Pilot's, almost the entire neighborhood are chemical plants. It is necessary to emphasize this aspect because the sewerage system is connected with the other three chemical plants from the neighborhood. This is the reason that we progressively established elements of an environmental monitoring program well in advance of tritium operation in order to determine baseline levels. The first step was the tritium level monitoring in environmental water and waste water of industrial activity from neighborhood. In this work, a low background liquid scintillation is used to determine tritium activity concentration according to ISO 9698/1998. We measured drinking water, precipitation, river water, underground water and waste water. The tritium level was between 10 TU and 27 TU that indicates there is no source of tritium contamination in the neighborhood of Cryogenic Pilot. In order to determine baseline levels we decide to monitories monthly each location. In this paper a standard method is presented which it is used for tritium determination in water sample, the precautions needed in order to achieve reliable results, and the evolution of tritium level in different location near the Experimental Pilot Tritium and Deuterium Cryogenic Separation.(author)

Tritium migration in nuclear desalination plants

International Nuclear Information System (INIS)

Muralev, E.D.

2003-01-01

Tritium transport, as one of important items of radiation safety assessment, should be taken into consideration before construction of a Nuclear Desalination Plant (NDP). The influence of tritium internal exposition to the human body is very dangerous because of 3 H associations with water molecules. The problem of tritium in nuclear engineering is connected to its high penetration ability (through fuel element cans and other construction materials of a reactor), with the difficulty of extracting tritium from process liquids and gases. Sources of tritium generation in NDP are: nuclear fuel, boron in control rods, and deuterium in heat carrier. Tritium passes easily through the walls of a reactor vessel, intermediate heat exchangers, steam generators and other technological equipment, through the walls of heat carrier pipelines. The release of tritium and its transport could be assessed, using mathematical models, based on the assumption that steady state equilibrium has been attained between the sources of tritium, produced water and release to the environment. Analysis of the model shows the tritium concentration dependence in potable water on design features of NDP. The calculations obtained and analysis results for NDP with BN-350 reactor give good convergence. According to the available data, tritium concentration in potable water is less than the statutory maximum concentration limit. The design of a NDP requires elaboration of technical solutions, capable of minimising the release of tritium to potable water produced. (author)

A study of the tritium behavior in coolant and moderator system of heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

1993-12-15

The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

Energy Technology Data Exchange (ETDEWEB)

Giancarli, L.M., E-mail: luciano.giancarli@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Iseli, M.; Lepetit, L.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Livingston, D. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom); Nevière, J.C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ricapito, I. [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wyse, S. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom)

2014-10-15

Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.

System for deuterium-tritium mixture filling the working chamber of a dense plasma focus device

International Nuclear Information System (INIS)

Bondar', A.I.; Vyskubov, V.P.; Gerasimov, S.A.

1981-01-01

A gas-vacuum system designed for filling the gas-discharge chamber of a plasma focus device with equal-coaponent deuterium-tritium mixture is described. The system consists of a unit for gaseous mixture prepa ration and a unit for mixture absorption and device evacuation. The system provides the gaseous mixture purification of O 2 and N 2 impurities. Final tritium content in the gas-discharge chamber after tritium removal is not greater than 2x10 8 Bq/l. Tritium content in a sealed box in which the device is placed does not exceed 30 Bq/l that is less than limiting safe value. The conclusion is made that the described system design gives an opportunity to begin experimental studies at plasma focus devices with deuterium-tritium mixture [ru

Handling of tritium at TFTR

International Nuclear Information System (INIS)

Pierce, C.W.; Howe, H.J.; Yemin, L.; Lind, K.

1977-01-01

Some of the engineering approaches taken at TFTR for the tritium control systems are discussed as the requirements being placed on the tritium systems by the operating scenarios of the Tokamak. The tritium control systems presently being designed for TFTR will limit the annual release to the environment to less than 100 curies

Design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

International Nuclear Information System (INIS)

Finn, P.A.; Reedy, G.T.; Homa, M.I.; Clemmer, R.G.; Pappas, G.; Slawecki, M.A.; Graczyk, D.G.; Bowers, D.L.; Clemmer, E.D.

1983-01-01

The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

Evaluation of tritium transport in the biomass-fusion hybrid system and its environmental impact

Energy Technology Data Exchange (ETDEWEB)

Namba, Kyosuke [Graduate School of Energy Science, Kyoto University, Kyoto (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Kyoto (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Kyoto (Japan); Yamamoto, Yasushi [Faculty of Engineering Science, Kansai University, Osaka (Japan)

2015-10-15

Highlights: • We assumed that tritium migrates from biomass hybrid fusion system to fuel cell vehicles. • We developed a seven-compartment model to describe the water flow and tritium in an urban area Osaka. • Tritium concentration of surface soil water run by 4 Bq/L level after 60 years later. • The tritium does not deserve health hazard but easily detectable in the environment. - Abstract: The behavior of tritium contained in the biofuel produced by the fusion energy is analyzed. Hydrogen product is contaminated with tritium from breeding blanket of fusion plant within the regulation limit and released to atmosphere when used for fuel cell vehicles. In the model city, Osaka, seven-compartment model describes the behavior of exhausted tritium by adapting the environment water flow and its migration was analyzed with STELLA system dynamics code. Tritium (HTO) with a concentration of 5000 Bq//m{sup 3} exhausted from the running vehicle increases decades and reaches steady state after about 50 years, at around 40 Bq/m{sup 3} in atmosphere and 4 Bq/L in surface soil water that does not deserve health hazard, however causes contamination of large populated area.

Upgrade to the Tritium Remote Control and Monitoring System for TFTR D and D

International Nuclear Information System (INIS)

Sichta, P.; Oliaro, G.; Sengupta, S.

2002-01-01

Since 1988, the Tritium Remote Control and Monitoring System (TRECAMS) has performed crucial functions in support of D-T [deuterium-tritium] operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). Although plasma operations on TFTR were completed in 1997, the need for TRECAMS continued. During this period TRECAMS supported the TFTR tritium systems, the TFTR's Shutdown and Safing phase, and the TFTR Decontamination and Decommissioning (D and D) project. The most critical function of the TRECAMS in the post-TFTR era has been to provide a real-time indication of the airborne tritium levels in the tritium areas and the (HVAC) stacks. TRECAMS is a critical tool in conducting safe TFTR D and D tritium-line breaks and other tritium-related work activities. Beginning in 1998, the failure rate of the system's hardware sharply increased. Furthermore, the specialized knowledge required to maintain the original software and hardware was diminishing. It soon became apparent that a failure of the TRECAMS could significantly impact the TFTR D and D project's cost and schedule. To preclude this, the TRECAMS hardware and software was upgraded in the year 2000 to use modern components. This paper will describe that successful upgrade, including a review of the engineering processes and our operating experiences with the upgraded system

The monitoring system of the Tritium Research Laboratory, Sandia Laboratories, Livermore, California

International Nuclear Information System (INIS)

Hafner, R.S.; Westfall, D.L.; Ristau, R.D.

1978-01-01

Computerized tritium monitoring is now in use at the Tritium Research Laboratory (TRL). Betatec 100 tritium monitors, along with several Sandia designed accessories, have been combined with a PDP 11/40 computer to provide maximum personnel and environmental protection. Each individual monitoring system, in addition to a local display in the area of interest, has a visual/audible display in the control room. Each system is then channeled into the PDP 11/40 computer, providing immediate assessment of the status of the entire laboratory from a central location. Measurement capability ranges from uCi/m 3 levels for room air monitoring to KCi/m 3 levels for glove box and process system monitoring. The overall monitoring system and its capabilities will be presented

New arrangement for the air cleanup system to recover tritium

International Nuclear Information System (INIS)

Nishikawa, Masabumi; Takahashi, Kohsaku; Munakata, Kenzo; Fukada, Satoshi; Kotoh, Kenji; Takeishi, Toshiharu

1997-01-01

At present, the standard arrangement of the air cleanup system responsible for emergency tritium recovery from room air is a catalytic oxidation bed with a heater followed by an adsorption bed with a cooler. One disadvantage of this arrangement is that trouble with the heater or the cooler could result in a loss of capacity to recover tritium. Another disadvantage of the catalyst-adsorption-bed arrangement is that tritiated water must be recovered with a high decontamination factor after dilution with a large amount of water vapor in the working atmosphere. The performance of a new arrangement for the air cleanup system, which consists of a precious metal catalyst bed preceded by an adsorption bed without heating equipment, is discussed. According to calculations, most of the tritium released to the room air is recovered in the catalyst bed through oxidation, adsorption, and isotope exchange reaction when the new arrangement is applied. The adsorption bed placed before the catalyst bed dehumidifies the process gas to such a degree that the oxidation reaction of tritium in the catalyst bed is not hindered by water vapor. 15 refs., 6 figs., 6 tabs

Experimental Tritium Cleanup System availability analysis from 1984 to 1992

International Nuclear Information System (INIS)

Cadwallader, L.C.; Taylor, G.L.

1993-05-01

This report gives the availability percentage of the Experimental Tritium Cleanup System (ETC) at the Tritium Systems Test Assembly (TSTA), which is a fusion research and technology facility at the Los Alamos National Laboratory. The component failure reports, the numbers of components, and operating times or demands are all given in this report. Sample calculations of the failure rates obtained from these data are given in the appendices. While future fusion experiments might use different or more advanced means to detritiate room air, the analysis of this system gives a data point for an actual detritiation system. Such a data point can be extrapolated for comparison with fault tree results on system designs, or can be used in a Bayesian failure rate analysis for estimating reliability of a new type of system. The nine years of testing operations on TSTA's ETC result in a reasonable average availability value of 92% for the maximal tritium release event. The failure rates for new systems are expected to be lower than for the TSTA ETC, since improvements will be made in the design of the room air detritiation system based on the TSTA system experiences. Nonetheless, these TSTA data should be useful for future fusion reactor design work and safety assessment tasks

Tritium production in fusion reactors

International Nuclear Information System (INIS)

Roth, E.

1981-08-01

The present analyses on the possibilities of extracting tritium from the liquid and solid fusion reactor blankets show up many problems. A consistent ensemble of materials and devices for extracting the heat and the tritium has not yet been integrated in a fusion reactor blanket project. The dimensioning of the many pipes required for shifting the tritium can only be done very approximately and the volume taken up by the blanket is difficult to evaluate, etc. The utilization of present data leads to over-dimensioning the installations by prudence and perhaps rejecting the best solutions. In order to measure the parameters of the most promising materials, work must be carried out on well defined samples and not only determine the base physical-chemical coefficients, such as thermal conductivity, scattering coefficients, Sievert parameters, but also the kinetic parameters conventional in chemical engineering, such as the hourly space rates of degassing. It is also necessary to perform long duration experiments under radiation and at operating temperatures, or above, in order to study the ageing of the bodies employed [fr

Migration of tritium from a nuclear waste burial site

International Nuclear Information System (INIS)

Hawkins, R.H.

1975-09-01

The Savannah River Plant (SRP) has routinely and continuously monitored the local environment (land, water, air, flora, and fauna) since 1951. As part of this intensive program, a three-part study was made to assess the tritium migration from an onsite burial ground for solid nuclear wastes and the resulting dose-to-man. A major source of tritium is buried, massive, Li-Al residues (referred to as melts) from the thermal extraction step in the SRP tritium production process. A melt with its extraction crucible and lid were immersed in water to measure the amounts of tritium released as HTO and HT to the water and to air. The result was a rapid release of 23 curies, of which approximately 99 percent was HTO that remained in the immersion water, and 1 percent was HT that passed into the air. (auth)

Photoproduction of tritium

International Nuclear Information System (INIS)

Becker, J.A.; Anderson, J.D.; Weiss, M.S.

1995-01-01

3 H (Tritium) is required for maintenance of nuclear weapons in the stockpile. The National Defense need for 3 H was historically met by the Savannah River Facility. This facility is no longer safe for operation. 3 H decays with a mean lifetime τ = 17.8 y, and therefore new methods of 3 H production are required to meet US military requirements. Irradiation of 7 Li by low-energy photons produces tritium ( 3 H) via the photodisintegration process. Waste heat from the 7 Li target can be extracted and used for the direct generation of electricity. Other advantages include: negligible residual radioactivity, simple target technology, small low-energy electron accelerators for bremsstrahlung production (the photon source), developed liquid metal technology, modularity, simple extraction of 3 H from a recirculating 7 Li target, abundant supply of 7 Li, and straightforward target-accelerator-bremsstrahlung converter interface. A schematic plant characterized by very low risk is described, and a figure-of-merit is obtained

An estimation of tritium inventory limits for the E-Area vaults

International Nuclear Information System (INIS)

Yu, A.D.; Cook, J.R.

1991-01-01

At the request of Waste Management, Interim Waste Technology has conducted a modeling study to estimate the tritium inventory limits for the E-Area vaults. These inventory limits are based on the groundwater impact of the planned waste disposal. The tritium inventory limit for an Intermediate Level Tritium Vault (ILTV) is estimated to be 400,000 Curies with a 100 year storage period. During this period, it is assumed that the ILTV will be vented, any leachate will be extracted, and its performance will be carefully monitored. The tritium inventory limits for a Low Activity Waste Vault (LAWV) and an Intermediate Level Non-Tritium Vault (ILNTV) are estimated to be 15,000 and 11,000 Curies, respectively. Venting and leachate extraction were not assumed necessary. These operational alternatives would further enhance the performance of these vaults. These limits are significantly higher than the forecasted maximum tritium inventories for the vaults. Details of the modeling study are described in the attached report

Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

International Nuclear Information System (INIS)

Lewis, M.R.

2000-01-01

A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

Tritium resources available for fusion reactors

Science.gov (United States)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

Design and construction of thermal desorption measurement system for tritium contained materials

International Nuclear Information System (INIS)

Hara, M.; Hatano, Y.; Calderoni, P.; Shimada, M.

2014-01-01

The dual-mode thermal desorption analysis system was designed and built in Idaho National Laboratory (INL) to examine the evolution of the hydrogen isotope gas from materials. The system is equipped with a mass spectrometer for stable hydrogen isotopes and an ionization chamber for tritium components. The performance of the system built was tested with using tritium contained materials. The evolution of tritiated gas species from contaminated materials was measured successfully by using the system. (author)

Preparation of pyronaridine labelled with tritium

International Nuclear Information System (INIS)

Jiang Shangen; Zhang Liufang; Zheng Dongzhu; Feng Zheng; Wu Zufan

1987-01-01

Pyronaridine is a high efficient and low toxic new antimalarial drug. 3 H-pyronaridine was prepared by catalytic isotopic exchange in solution with tritium gas using PdO/BaSO 4 as catalyst. That crude product was purified by extraction. 3 H-NMR spectra of pyronaridine showed that tritium was labelled at the 6-position. Specific activity of 3 H-pyronaridine was 5.5 Ci/mmol and radiochemical purity over 95%

Review of tritium behavior in HTGR systems

International Nuclear Information System (INIS)

Gainey, B.W.

1976-01-01

The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

The design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

International Nuclear Information System (INIS)

Finn, P.A.; Bowers, D.L.; Clemmer, E.D.; Clemmer, R.G.; Graczyk, D.G.; Homa, M.I.; Pappas, G.; Reedy, G.T.; Slawecki, M.A.

1983-01-01

The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time; unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

A proposed model for the transfer of environmental tritium to man and tritium metabolism in model animals

International Nuclear Information System (INIS)

Saito, Masahiro; Ishida, M.R.

1987-01-01

To evaluate the accumulated dose in human bodies due to the environmental tritium, it is of required to establish an adequate model for the tritium transfer from the environment to man and to obtain enough information on the metabolic behaviour of tritium in animal bodies using model animal system. In this report, first we describe about a proposed model for the transfer of environmental tritium to man and secondly mention briefly about the recent works on the tritium metabolism in newborn animals which have been treated as a model system of tritium intake through food chain. (author)

Electrochemical measurement of tritium and hydrogen permeation through iron membranes

International Nuclear Information System (INIS)

Hagi, Hideki; Hayashi, Yasunori

1987-01-01

Permeation rates of tritium and hydrogen through iron were measured by the electro-chemical method in which an aqueous solution containing 3.7 x 10 12 Bq/m 3 tritium was used as a cathodic electrolyte. Tritium and hydrogen were introduced from one side of a specimen by cathodic polarization with a constant current density, while at the other side of the specimen the permeated tritium and hydrogen were extracted by potentiostatical ionization. Nearly all of the potentiostatic current on the extraction side is produced by the ionization of hydrogen, because the concentration of tritium in the cathodic electrolyte is very small. The amount of permeated hydrogen was obtained by integrating the potentiostatic current, and that of permeated tritium was determined by measuring the radioactivity of the electrolyte sampled from the anodic side. The separation factor for permeation obtained under steady state conditions (the ratio of permeation rates of hydrogen to tritium divided by the ratio of the concentration of hydrogen to tritium in the cathodic electrolyte) is 12 at 288 K. This value is independent of cathodic current density. Diffusion coefficients of tritium (D T ) and hydrogen (D H ) in iron were determined from the tritium and hydrogen permeation by using time lag technique. For annealed iron at 286 K, D T = 9 x 10 -10 m 2 /s and D H = 4 x 10 -9 m 2 /s, and for 9 % cold-worked iron at 284 K, D T = 3 x 10 -10 m 2 /s and D H = 4 x 10 -10 m 2 /s. (author)

Torus evacuation and tritium handling on NET

International Nuclear Information System (INIS)

Dinner, P.; Chazalon, M.; Iseli, M.

1986-08-01

The use of tritium as a fuel affects the design of many systems, as well as requiring several new systems not needed on non DT-burning Tokamaks. This paper summarizes: major tritium process interconnections, tritium flows and inventories; primary requirements, preferred design alternatives, and related development issues; design philosophy for tritium and primary vacuum systems. 14 refs

Method and plant to remote tritium from the cooling water of a nuclear reactor

International Nuclear Information System (INIS)

O'Brien, C.J.

1976-01-01

A method is proposed for the extraction of tritium from the cooling water of a nuclear reactor, based on the principle of concentrating the tritium by a multi-stage transfer process. The cooling water is brought into contact in each stage with basic, labile, hydrogen-containing material with high pH value, whereby the tritium is transfered into an intermediate solid product and can be separated off. The technical details of the plant are described. Cellulose materials, such as cotton and wood as well as protein-containing material, such as muscle tissue are mentioned as examples of materials with a high affinity to tritium, greater than the affinity of water to tritium. They extract tritium from the cooling water. (HK) [de

System support software for TSTA [Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Claborn, G.W.; Mann, L.W.; Nielson, C.W.

1987-10-01

The fact that Tritium Systems Test Assembly (TSTA) is an experimental facility makes it impossible and undesirable to try to forecast the exact software requirements. Thus the software had to be written in a manner that would allow modifications without compromising the safety requirements imposed by the handling of tritium. This suggested a multi-level approach to the software. In this approach (much like the ISO network model) each level is isolated from the level below and above by cleanly defined interfaces. For example, the subsystem support level interfaces with the subsystem hardware through the software support level. Routines in the software support level provide operations like ''OPEN VALVE'' and CLOSE VALVE'' to the subsystem level. This isolates the subsystem level from the actual hardware. This is advantageous because changes can occur in any level without the need for propagating the change to any other level. The TSTA control system consists of the hardware level, the data conversion level, the operator interface level, and the subsystem process level. These levels are described

Initial experience of tritium exposure control at JET

International Nuclear Information System (INIS)

Patel, B.; Campling, D.C.; Schofield, P.A.; Macheta, P.; Sandland, K.

1998-01-01

Some of the safety procedures and controls in place for work with tritium are described, and initial operational experience of handling tritium is discussed. A description is given of work to rectify a water leak in a JET neutral beam heating component, which involved man-access to a confined volume to perform repairs, at tritium levels about 100 DAC (80 MBq/m 3 . HTO). Control measures involving use of purge and extract ventilation, and of personal protection using air-fed pressurized suits are described. Results are given of the internal doses to project staff and of atmospheric discharges of tritium during the repair outage. (P.A.)

Estimation of the detection limit of an experimental model of tritium storage bed designed for 'in-situ' accountability

International Nuclear Information System (INIS)

Bulubasa, Gheorghe; Bidica, Nicolae; Stefanescu, Ioan; Bucur, Ciprian; Deaconu, Mariea

2009-01-01

During the water detritiation process most of the tritium inventory is transferred from water into the gaseous phase, then it is further enriched and finally extracted and safely stored. The control of tritium inventory is an acute issue from several points of view: - Financially - tritium is an expensive material; - Safeguard - tritium is considered as nuclear material of strategic importance; - Safety - tritium is a radioactive material: requirements for documented safety analysis report (to ensure strict limits on the total tritium allowed) and for evaluation of accident consequences associated with that inventory. Large amounts of tritium can be stored, in a very safely manner, as metal tritides. A bench-scale experiment of a tritium storage bed with integrated system for in-situ tritium inventory accountancy was designed and developed at ICSI Rm. Valcea. The calibration curve and the detection limit for this experimental model of tritium storage bed were determined. The experimental results are presented in this paper. (authors)

Tritium system for a tokamak reactor with a self-pumped limiter

International Nuclear Information System (INIS)

Hassanein, A.M.; Sze, D.K.

1986-01-01

The self-pumping concept was proposed as a means of simplifying the impurity control system in a fusion reactor. The idea is to remove helium in-situ by trapping in freshly deposited metal surface layers of a limiter or divertor. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. Some of the key issues for this concept are the tritium inventory in the trapping material and the permeation of protium and recycling of tritium. These quantities are shown to be acceptable for the reference design. The tritium issues for a helium-cooled solid breeder reactor design with vanadium alloy as a structural material are also examined. Models are presented for tritium permeation and inventory calculation for structure materials with the effect of a thin layer of coating material

Advanced design of the Mechanical Tritium Pumping System for JET DTE2

International Nuclear Information System (INIS)

Giegerich, T.; Bekris, N.; Camp, P.; Day, Chr.; Gethins, M.; Lesnoy, S.; Luo, X.; Müller, R.; Ochoa, S.; Pfeil, P.; Smith, R.; Strobel, H.; Stump, H.

2016-01-01

For tritium processing in JET during the next Deuterium-Tritium-Experiment (DTE2), a fully tritium compatible and continuously working vacuum pumping system has been developed. This pump train will be used as roughing pump to cover a pressure regime between 10 −1 Pa and ambient pressure. Therefore, a two-stage liquid ring pump in combination with a booster vapor diffusion pump will be applied. In this paper, a close-to-final design of the pumps is being described. Finite element (FEM) simulation results of components where high mechanical stresses due to thermal gradients are expected are presented. Furthermore, the final design of the control and data acquisition system is shown and explained.

Advanced design of the Mechanical Tritium Pumping System for JET DTE2

Energy Technology Data Exchange (ETDEWEB)

Giegerich, T., E-mail: thomas.giegerich@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bekris, N. [EUROfusion Program Management Unit (PMU), ITER Physics Department, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Camp, P. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Day, Chr. [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gethins, M.; Lesnoy, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Luo, X.; Müller, R.; Ochoa, S.; Pfeil, P. [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Smith, R. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Strobel, H.; Stump, H. [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2016-11-01

For tritium processing in JET during the next Deuterium-Tritium-Experiment (DTE2), a fully tritium compatible and continuously working vacuum pumping system has been developed. This pump train will be used as roughing pump to cover a pressure regime between 10{sup −1} Pa and ambient pressure. Therefore, a two-stage liquid ring pump in combination with a booster vapor diffusion pump will be applied. In this paper, a close-to-final design of the pumps is being described. Finite element (FEM) simulation results of components where high mechanical stresses due to thermal gradients are expected are presented. Furthermore, the final design of the control and data acquisition system is shown and explained.

Tritium analysis at TFTR

International Nuclear Information System (INIS)

Voorhees, D.R.; Rossmassler, R.L.; Zimmer, G.

1995-01-01

The tritium analytical system at TFRR is used to determine the purity of tritium bearing gas streams in order to provide inventory and accountability measurements. The system includes a quadrupole mass spectrometer and beta scintillator originally configured at Monsanto Mound Research Laboratory in the late 1970's and early 1980's. The system was commissioned and tested between 1991 and 1992 and is used daily for analysis of calibration standards, incoming tritium shipments, gases evolved from uranium storage beds and measurement of gases returned to gas holding tanks. The low resolution mass spectrometer is enhanced by the use of a metal getter pump to aid in resolving the mass 3 and 4 species. The beta scintillator complements the analysis as it detects tritium bearing species that often are not easily detected by mass spectrometry such as condensable species or hydrocarbons containing tritium. The instruments are controlled by a personal computer with customized software written with a graphical programming system designed for data acquisition and control. A discussion of the instrumentation, control systems, system parameters, procedural methods, algorithms, and operational issues will be presented. Measurements of gas holding tanks and tritiated water waste streams using ion chamber instrumentation are discussed elsewhere

Preparation of pyronaridine labelled with tritium

Energy Technology Data Exchange (ETDEWEB)

Shangen, Jiang; Liufang, Zhang; Dongzhu, Zheng; Zheng, Feng; Zufan, Wu

1987-12-01

Pyronaridine is a high efficient and low toxic new antimalarial drug. /sup 3/H-pyronaridine was prepared by catalytic isotopic exchange in solution with tritium gas using PdO/BaSO/sub 4/ as catalyst. That crude product was purified by extraction. /sup 3/H-NMR spectra of pyronaridine showed that tritium was labelled at the 6-position. Specific activity of /sup 3/H-pyronaridine was 5.5 Ci/mmol and radiochemical purity over 95%.

Retrospective evaluation of tritium fallout by tree-ring analysis

International Nuclear Information System (INIS)

Kozak, K.; Biro, T.; Golder, F.; Rank, D.; Rajner, V.; Staudner, F.

1993-01-01

Tritium analyses of tree-ring cellulose were made to test its suitability for retrospective evaluation of a local tritium fallout. Several spruce trees were taken from an Austrian alpine area where tritium contamination of May 1974 precipitation had been detected. Wood from the annual growth rings of 1973, 1974 and 1975 was separated and the cellulose extracted. After isotopic equilibration with dead water, cellulose was combusted to yield water, whose tritium concentration was measured by liquid scintillation counting. Rigorous statistical treatment proved the significance of the increased tritium concentration caused by the tritium anomaly, which occurred during the growing season. The long-term trends of local atmospheric tritium, including the 1974 peak, were also well reflected by analysis of a 24-year ring sequence from a single tree in the contaminated area. The tritium data gained by the given method can be used at present qualitatively and a better understanding of the possible sources of contamination is required in order that the quantitative criteria be satisfied. (Author)

Technology developments for improved tritium management

International Nuclear Information System (INIS)

Miller, J.M.; Spagnolo, D.A.

1994-06-01

Tritium technology developments have been an integral part of the advancement of CANDU reactor technology. An understanding of tritium behaviour within the heavy-water systems has led to improvements in tritium recovery processes, tritium measurement techniques and overall tritium control. Detritiation technology has been put in place as part of heavy water and tritium management practices. The advances made in these technologies are summarized. (author). 20 refs., 5 figs

Tritium extraction methods proposed for a solid breeder blanket. Subtask WP-B 6.1 of the European Blanket Program 1996

International Nuclear Information System (INIS)

Albrecht, H.

1997-04-01

Ten different methods for the extraction of tritium from the purge gas of a ceramic blanket are described and evaluated with respect to their applicability for ITER and DEMO. The methods are based on the conditions that the purge gas is composed of helium with an addition of up to 0.1% of H 2 or O 2 and H 2 O to facilitate the release of tritium, and that tritium occurs in the purge gas in two main chemical forms, i.e. HT and HTO. Individual process steps of many methods are identical; in particular, the application of cold traps, molecular sieve beds, and diffusors are proposed in several cases. Differences between the methods arise mainly from the ways in which various process steps are combined and from the operating conditions which are chosen with respect to temperature and pressure. Up to now, none of the methods has been demonstrated to be reliably applicable for the purge gas conditions foreseen for the operation of an ITER blanket test module (or larger ceramic blanket designs such as for DEMO). These conditions are characterized by very high gas flow rates and extremely low concentrations of HT and HTO. Therefore, a proposal has been made (FZK concept) which is expected to have the best potential for applicability to ITER and DEMO and to incorporate the smallest development risk. In this concept, the extraction of tritium and excess hydrogen is accomplished by using a cold trap for freezing out HTO/H 2 O and a 5A molecular sieve bed for the adsorption of HT/H 2 . (orig.) [de

Experimental study of permeation and selectivity of zeolite membranes for tritium processes

Energy Technology Data Exchange (ETDEWEB)

Borisevich, Olga; Antunes, Rodrigo; Demange, David, E-mail: david.demange@kit.edu

2015-10-15

Highlights: • We report about new experimental results on advanced membranes for tritium processing especially for the DEMO breeding blanket. • High permeances are measured on different zeolite MFI membranes made by film deposition or pore plugging. • Selectivity for H{sub 2}/He is limited requiring a multi-stage membrane process. • Selectivity of H{sub 2}O/He seems high enough to operate one single module. - Abstract: Zeolites are known as tritium compatible inorganic materials widely used in packed beds as driers in detritiation systems and are also suggested for tritium removal from helium at cryogenic temperature. The Tritium Laboratory Karlsruhe (TLK) proposed a new fully continuous approach for tritium extraction from the solid breeding blanket of fusion machines that improves the overall tritium management and minimizes both the tritium inventory and processing time. It is based on membrane permeation as a pre-concentration stage upstream of a final tritium recovery stage using a catalytic Pd-based membrane reactor. Zeolite membranes were identified as the most promising candidates for the pre-concentration stage. In the present work the tubular zeolite MFI membrane provided by the Institute for Ceramic Technologies and Systems (IKTS, Hermsdorf, Germany) is studied to consolidate the proposed approach. The permeation measurements for single gases hydrogen (replacing radioactive tritium) and helium, for binary mixtures H{sub 2}/He and H{sub 2}O/He at different concentrations and temperatures are presented. The tested membrane demonstrates a high performance, almost independent from the inlet composition in the case of a gaseous mixture, while the transport in the presence of water vapour is strongly related to the temperature of the mixture and component concentrations.

Radioactive waste tank ventilation system incorporating tritium control

Energy Technology Data Exchange (ETDEWEB)

Rice, P.D. [ICF Kaiser Hanford Company, Richland, WA (United States)

1997-08-01

This paper describes the development of a ventilation system for radioactive waste tanks at the U.S. Department of Energy`s (DOE) Hanford Site in Richland, Washington. The unique design of the system is aimed at cost-effective control of tritiated water vapor. The system includes recirculation ventilation and cooling for each tank in the facility and a central exhaust air clean-up train that includes a low-temperature vapor condenser and high-efficiency mist eliminator (HEME). A one-seventh scale pilot plant was built and tested to verify predicted performance of the low-temperature tritium removal system. Tests were conducted to determine the effectiveness of the removal of condensable vapor and soluble and insoluble aerosols and to estimate the operating life of the mist eliminator. Definitive design of the ventilation system relied heavily on the test data. The unique design features of the ventilation system will result in far less release of tritium to the atmosphere than from conventional high-volume dilution systems and will greatly reduce operating costs. NESHAPs and TAPs NOC applications have been approved, and field construction is nearly complete. Start-up is scheduled for late 1996. 3 refs., 4 figs., 2 tabs.

Development status of the cryogenic distillation system in Cernavoda Tritium Removal Facility

International Nuclear Information System (INIS)

Draghia, Mirela; Ana, George; Pasca, Gheorghe; Porcariu, Florina

2009-01-01

Full text: The reference design technology for the heavy water detritiation plant of Cernavoda CANDU station is based on combination of Liquid Phase Catalytic Exchange (LPCE) and Cryogenic Distillation (CD) processes. Based on this technology, tritium is transferred from the heavy water to a deuterium stream in the catalyzed isotopic exchange process, LPCE, followed by a final enrichment within the cryogenic distillation cascade. The final step is the tritium storage on metallic hydride. The basic function of the Cryogenic Distillation System (CDS) is the separation of tritium from the tritiated deuterium coming from the LPCE column in the following conditions: - the final product has to be tritium with a concentration of at least 99%; - it must be provided a detritiation factor of at least 100 (the ration between the tritium concentration in the deuterium stream fed to the CD system and the tritium concentration in the returned stream to the LPCE); - the deuterium must be enriched up to 99.995%, by removing the protium; - provisions for safe discharge of the entire inventory of the CD cascade into buffer vessels shall be implemented. To summarize, the present status of the project consists of technical documentation for all the components of CDS, including the P and ID (Pipping and Instrumentation Diagram), preliminary data sheets, technical specifications, drawings for the major components as the buffer vessels, coldbox, etc, and 3D models as well for almost all the components. (authors)

Comparison of Tritium Component Failure Rate Data

International Nuclear Information System (INIS)

Lee C. Cadwallader

2004-01-01

Published failure rate values from the US Tritium Systems Test Assembly, the Japanese Tritium Process Laboratory, the German Tritium Laboratory Karlsruhe, and the Joint European Torus Active Gas Handling System have been compared. This comparison is on a limited set of components, but there is a good variety of data sets in the comparison. The data compared reasonably well. The most reasonable failure rate values are recommended for use on next generation tritium handling system components, such as those in the tritium plant systems for the International Thermonuclear Experimental Reactor and the tritium fuel systems of inertial fusion facilities, such as the US National Ignition Facility. These data and the comparison results are also shared with the International Energy Agency cooperative task on fusion component failure rate data

Protection against tritium radiations

International Nuclear Information System (INIS)

Bal, Georges

1964-05-01

This report presents the main characteristics of tritium, describes how it is produced as a natural or as an artificial radio-element. It outlines the hazards related to this material, presents how materials and tools are contaminated and decontaminated. It addresses the issue of permissible maximum limits: factors of assessment of the risk induced by tritium, maximum permissible activity in body water, maximum permissible concentrations in the atmosphere. It describes the measurement of tritium activity: generalities, measurement of gas activity and of tritiated water steam, tritium-induced ionisation in an ionisation chamber, measurement systems using ionisation chambers, discontinuous detection of tritium-containing water in the air, detection of surface contamination [fr

Tritium practices past and present

International Nuclear Information System (INIS)

Gede, V.P.; Gildea, P.D.

1980-01-01

History of the production and use of tritium, as well as handling techniques, are reviewed. Handling techniques first used at Lawrence Livermore National Laboratory made use of glass vacuum systems and relatively crude ion chambers for monitoring airborne activity. The first use of inert atmosphere glove boxes demonstrated that uptake through the skin could be a serious personnel exposure problem. Growing environmental concerns in the early 1970's resulted in the implementation by the Atomic Energy Commission of a new criteria to limit atmospheric tritium releases to levels as low as practicable. An important result of the new criteria was the development of containment and recovery systems to capture tritium rather than vent it to the atmosphere. The Sandia National Laboratories, Livermore, Tritium Research Laboratory containment and decontamination systems are presented as a typical example of this technology. The application of computers to control systems is expected to provide the greatest potential for change in future tritium handling practices

The tokamak fusion test reactor tritium systems test contractor operational readiness review

International Nuclear Information System (INIS)

Gentile, C.A.; Levine, J.; Norris, M.; Rehill, F.; Such, C.

1993-01-01

In preparation for D-T operations at TFTR, the TFTR project has successfully completed the C-ORR process which has led to the introduction of 200 curies of tritium to the site. Preparations for the C-ORR began approximately 2 years ago. During July 1992 a one-week Site Assistance Review was conducted by the C-ORR , and C-ORR Team consisting of 12 persons, all of whom were outside experts, many of whom were from other facilities within the DOE complex. During the July 1992 Site Assistance Review 201 findings were documented which fell into one of three categories. All of the 109 category one findings which were generated were required to be resolved prior to the introduction of tritium to the TFTR site. On April 5, 1993, the TFTR Tritium System Test C-ORR commenced. The results of the C-ORR as documented in the final report by the C-ORR was that category 1 findings were resolved, and it was the recommendation of the C-ORR Team to the PPPL ES ampersand H Board that TFTR initiate the Tritium Systems Test. DOE (Chicago Operations, Princeton Area Office) concurred with the C-ORR final report and on April 29, 1993, at 12:15 pm tritium was introduced to the TFTR site

Overview of tritium fast-fission yields

International Nuclear Information System (INIS)

Tanner, J.E.

1981-03-01

Tritium production rates are very important to the development of fast reactors because tritium may be produced at a greater rate in fast reactors than in light water reactors. This report focuses on tritium production and does not evaluate the transport and eventual release of the tritium in a fast reactor system. However, if an order-of-magnitude increase in fast fission yields for tritium is confirmed, fission will become the dominant production source of tritium in fast reactors

Radiological safety system based on real-time tritium-in-air monitoring in room and effluents

Energy Technology Data Exchange (ETDEWEB)

Bidica, N.; Sofalca, N.; Balteanu, O.; Stefan, I. [National Institute of Cryogenics and Isotopes Technologies, Ramnicu Valcea (Romania)

2006-07-01

The conceptual design of the radiological safety system based on real time-in-air monitoring in room and effluents is intended to provide the maximum achievable safety level, basing no the ALARA concept. the capabilities of this system are not only to inform any time personnel about tritium in air concentration level, but it will be able to: initiate the shut down procedure and drain off the plant, as well to start the Air cleaning System when the tritium-in-air concentration exceed pre-established threshold; estimate tritium effective dose rate before starting an activity into the monitored area, or during this activity, or soon as the activity was finished; estimate tritium effective dose and instantly record and update individual effective doses, using a special computer application called 'dose record'; lock access into the radiological area for individuals when tritium dose rate in the monitoring area will exceed the pre-established thresholds, or when any individual dose data provided by 'dose records' application ask for, or for other protection consideration; calculate the total tritium activity released to the environment (per day, week, or month). (N.C.)

Radiological safety system based on real-time tritium-in-air monitoring in room and effluents

International Nuclear Information System (INIS)

Bidica, N.; Sofalca, N.; Balteanu, O.; Stefan, I.

2006-01-01

The conceptual design of the radiological safety system based on real time-in-air monitoring in room and effluents is intended to provide the maximum achievable safety level, basing no the ALARA concept. the capabilities of this system are not only to inform any time personnel about tritium in air concentration level, but it will be able to: initiate the shut down procedure and drain off the plant, as well to start the Air cleaning System when the tritium-in-air concentration exceed pre-established threshold; estimate tritium effective dose rate before starting an activity into the monitored area, or during this activity, or soon as the activity was finished; estimate tritium effective dose and instantly record and update individual effective doses, using a special computer application called 'dose record'; lock access into the radiological area for individuals when tritium dose rate in the monitoring area will exceed the pre-established thresholds, or when any individual dose data provided by 'dose records' application ask for, or for other protection consideration; calculate the total tritium activity released to the environment (per day, week, or month). (N.C.)

Tritium

International Nuclear Information System (INIS)

Anon.

1975-01-01

The role played the large amount supply of tritium and its effects are broadly reviewed. This report is divided into four parts. The introductory part includes the history of tritium research. The second part deals with the physicochemical properties of tritium and the compounds containing tritium such as tritium water and labeled compounds, and with the isotope effects and self radiation effects of tritium. The third part deals with the tritium production by artificial reaction. Attention is directed to the future productivity of tritium from B, Be, N, C, O, etc. by using the beams of high energy protons or neutrons. The problems of the accepting market and the accuracy of estimating manufacturing cost are discussed. The expansion of production may bring upon the reduction of cost but also a large possibility of social impact. The irradiation problem and handling problem in view of environmental preservation are discussed. The fourth part deals with the use of tritium as a target, as a source of radiation or light, and its utilization for geochemistry. The future development of the solid tritium target capable of elongating the life of neutron sources is expected. The rust thickness of the surface of iron can be measured with the X-ray of Ti-T or Zr-T. The tritium can substitute self-light emission paint or lamp. The tritium is suitable for tracing the movement of sea water and land surface water because of its long half life. (Iwakiri, K.)

Tritium in metals

International Nuclear Information System (INIS)

Schober, T.

1990-01-01

In this Chapter a review is given of some of the important features of metal tritides as opposed to hydrides and deuterides. After an introduction to the topics of tritium and tritium in metals information will be presented on a variety of metal-tritium systems. Of main interest here are the differences from the classic hydrogen behavior; the so called isotope effect. A second important topic is that of aging effects produced by the accumulation of 3 He in the samples. (orig.)

Concept of a tritium extraction facility for a reprocessing plant

International Nuclear Information System (INIS)

Tunaboylu, K.; Paulovic, M.; Ulrich, D.

1991-01-01

There are several alternatives for reducing the release of tritium to the environment originating from the wastewater of a reprocessing plant. Such alternatives, which are applicable for sites not located by the sea or by large rivers, are limited to either injection of tritiated wastewater into suitable deep geological formations, or final disposal into a deep underground repository after adequate treatment similar to other low and intermediate active waste. Removal of tritium from the wastewater by enrichment represents a further feasible option of the second alternative, which allows reduction of the huge volume of tritiated water to be treated before disposal. A significant volume reduction increases the safety of the subsequent steps such as transport, interim storage and final disposal of tritiated waste, furthermore, decreases the corresponding overall waste management cost. The projected Wackersdorf reprocessing plant has been considered as a reference for assessing the permitted tritium releases and other site characteristics. (orig.)

Tritium inventory prediction in a CANDU plant

International Nuclear Information System (INIS)

Song, M.J.; Son, S.H.; Jang, C.H.

1995-01-01

The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983--1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors

Formation of radiation-induced defects and their influence on tritium extraction from lithium silicates in out-of-pile experiments

International Nuclear Information System (INIS)

Abramenkovs, A.A.; Tiliks, J.E.

1991-01-01

Formation and properties of radiation-induced defects and radiolysis products in lithium silicates irradiated in nuclear reactor till absorbed doses 1000 MGy were studied. Radiation-induced defects (RD) and radiolysis products (RP) were qualitatively and quantitatively determinated by methods of chemical scavengers (MHS), electron-spin resonance (ESR) and optical spectroscopy. Colloidal silicon and lithium, lithium and silicon oxides, oxygen, silicon and lithium peroxides are the final products of the lithium silicates radiolysis at absorbed energy doses D abs = 1000 MGy. The concentration of radiation defects and products of radiolysis strongly depend on the temperature of irradiation, humidity, granural size. The thermostimulated extraction of tritiated water (95-98% of the released tritium is in chemical form of water) from lithium silicates ceramics proceeds according to two independent mechanisms: a) chemidesorption of surface localized tritiated water (the first order chemical reaction); b) formation of the tritium water molecules limited by triton diffusion to the near-surface layer of grains. It has been found that the concentration of radiation-induced defects considerably affects the tritium localization and releasing processes from lithium silicates. (orig.)

Analysis of tritium migration and deposition in fusion-reactor systems

International Nuclear Information System (INIS)

Holland, D.F.; Merrill, B.J.

1981-01-01

EG and G Idaho, Inc., is developing a safety analysis code, TMAP (Tritium Migration Analysis Program), to determine tritium loss into the environment and tritium buildup in components, coolants, and walls during normal and accident conditions. TMAP determines the thermal response of structures, solves equations for hydrogen movement through surface and in bulk materials, and also includes equations for chemical reactions. TMAP calculations of tritium movement through metal barriers at low tritium pressure agree closely with experimental measurements. The code has been used to predict inventory buildup and loss to the coolant of tritium implanted in the first wall of a fusion device, and concentrations during cleanup of tritium released into an enclosure

Computer aided accountancy for tritium handling systems

International Nuclear Information System (INIS)

Spannagel, G.; Schmid, C.

1993-03-01

A program is presented which can be used in tritium bearing systems for inventory taking and accountancy purposes. In particular, a detailed description is given of the environment in which this program has been integrated. It is explained in which way a high user friendliness has been attained and which structures contribute to achieving the flexibility required. (orig.) [de

Tritium in rad waste management

International Nuclear Information System (INIS)

Gandhi, P.M.; Ali, S.S.; Mathur, R.K.; Rastogi, R.C.

1990-01-01

Radioactive waste arising from PHWR's are invariably contaminated with tritium activity. Their disposal is crucial as it governs the manner and extent of radioactive contamination of human environment. The technique of tritium measurement and its application plays an important role in assessing the safety of the disposal system. Thus, typical applications involving tritium measurements include the evaluation of a site for solid waste burial facility and evaluation of a water body for liquid waste dispersal. Tritium measurement is also required in assessing safe air route dispersal of tritium. (author)

ITER Dynamic Tritium Inventory Modeling Code

International Nuclear Information System (INIS)

Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

2005-01-01

A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

Regeneration and tritium recovery from the large JET neutral injection cryopump system after the FTE

International Nuclear Information System (INIS)

Obert, W.; Bell, A.; Davies, J.; Mayaux, C.; Perinic, G.; Saibene, G.; Sartori, R.; Thompson, E.; Anderson, J.; Jenkins, E.; Walthers, C.

1992-01-01

Neutral Beam Injection (NBI) was used to introduce tritium into the plasma for the First Tritium Experiment In addition to the decisive advantage of depositing the tritium into the centre of the plasma, the use of NBI also minimized the total quantity of tritium introduced into the Torus and the contamination of the vacuum vessel. However, because of the relatively low gas efficiency of the positive ion injection system approximately 95% of the total quantity of tritium introduced was pumped by the large condensation cryopumps which form an integral part of the injector. Several hardware and associated software changes were implemented in order to making provision for possible fault scenarios during operation with tritium and to ensure complete regeneration of the tritium from the cryopumps. The tritium released after all subsequent regeneration's has been monitored carefully in order to determine the amount of tritium retained by the black anodized liquid nitrogen panel surfaces of the cryopump and to compare it with experiments at TSTA on JET samples before the FTE

Commercial Light Water Reactor Tritium Extraction Facility. Geotechnical Summary report (U)

International Nuclear Information System (INIS)

McHood, M.D.

2000-09-01

A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

The research status and prospect of tritium migration in soil-plant system

International Nuclear Information System (INIS)

Qiu Guohua

2011-01-01

In this paper, the research significance and progress of tritium migration in soil-plant system are briefly introduced, which includes spatial and temporal distribution, migration pattern and influence factors, chemical forms, mathematical models of tritium migration in different soil and plant. The research results are summarized, and the existing problems in research process are analyzed and discussed. (authors)

Timing tests: automatic valve closure for tritium leaks

International Nuclear Information System (INIS)

Hanel, S.

1976-01-01

How fast can an automotive valve be closed after a tritium leak occurs in a system. Tests described found that a valve can be closed within fifteen seconds of leakage. In one practical example considered, this delay would limit loss of tritium from a plumbing leak in a tritium system to 1 1 / 4 g. The tests were made in a typical LLL air-flush hood in which a tritium handling system had been installed. Incidental observations suggest that further study be made of a possible leak-actuated recovery system for an entire tritium facility

Purification of tritium-free water

International Nuclear Information System (INIS)

Hussain, S.D.

1982-10-01

Ground water which has been out of contact with the atmosphere for a long time as compared to the half life of tritium (12.43 years) does not contain any measureable amount of tritium. Such water is called tritium-free water. It may contain dissolved and suspended impurities and has to be purified before it can be used for the preparation of blanks and standards required in the routine measurement of low level tritium in water samples. The purification of tritium-free water by distillation in a closed system has been described. The quality of processed tritium-free water was precisely checked at International Atomic Energy Agency (IAEA) Vienna and found satisfactory. (authors)

A design assessment of tritium removal systems for the mirror advanced reactor study

International Nuclear Information System (INIS)

Sood, S.K.; Kveton, O.K.

1983-01-01

This study investigates the available processes for removing tritium from light water, and selects the most appropriate process for recovering tritium from the various tritiated water streams identified in the Mirror Advanced Reactor Study (MARS). A simplified flowsheet is shown for the process and the main process parameters are identified. Previous experience is utilized to predict direct capital costs and power requirement for the Tritiated Water Removal Unit (TWRU). A number of possibilities are discussed for lowering the cost of the TWRU. An estimate is made of the direct capital cost for the Air Detritiation System that has already been selected as the reference design by MARS personnel. The leakage from the MARS coolant loop is estimated, based on the experience obtained with Ontario Hydro's coolant systems. Design targets are identified for tritium levels in the reactor hall atmosphere and in water and air emissions. Tritium levels are predicted for these and are assessed against the previously identified targets

Tritium Systems Test Facility. Volume II. Appendixes

International Nuclear Information System (INIS)

Anderson, G.W.; Battleson, K.W.; Bauer, W.

1976-10-01

This document includes the following appendices: (1) vacuum pumping, (2) tritium migration into the power cycle, (3) separation of hydrogen isotopes, (4) tritium research laboratory, (5) TSTF containment and cleanup, (6) instrumentation and control, (7) gas heating in torus, and (8) TSTF fuel loop operating procedures

A system for the analysis of tritium content in natural waters, through benzene

International Nuclear Information System (INIS)

Bocchi, N.

1980-01-01

A system is described for the analysis of tritium ( 3 H) in natural waters. The system consists of an electrolytic enrichment equipment and a vacuum line for benzene synthesis. The benzene is mixed with a scintillating solution and so used in tritium activity measurements by liquid scintillation spectrometry. The characteristcs of the system, as well as its performance, are pointed out through analysis of ground and rain waters. The precision and reproducibility of the measurements are discussed. (Author) [pt

Tritium sorption in lithium-bismuth and lithium-aluminum alloys

International Nuclear Information System (INIS)

Talbot, J.B.; Smith, F.J.; Land, J.F.; Barton, P.

1976-01-01

The sorption of tritium by molten lithium-bismuth, Li-Bi (15 at.% lithium), and solid equiatomic lithium-aluminum, Li-Al, was investigated to evaluate the potential application of both materials in controlled thermonuclear reactors. The solubility of tritium in molten Li-Bi is less than 0.1 ppb at 500 - 700 0 C and tritium partial pressures of 10 -1 - 10 -3 Torr. Therefore, extraction of tritium from molten Li 2 BeF 4 salt with Li-Bi is not practical. The solubility of hydrogen in solid Li-Al (50 - 50 at.%) at 500 0 C follows Sieverts' Law; the Sieverts' constant was measured to be 1.9 (+-0.1) X 10 4 Torrsup(1/2)/atomic fraction. Tritium sorption in Li-Al ranged from 0.01 to 7 ppm at 400 - 600 0 C at respective tritium partial pressures of 0.14 - 0.52 Torr. (Auth.)

Tritium system design for the mirror reactors FPD-I, FPD-II, and FPD-III

International Nuclear Information System (INIS)

Finn, P.A.

1985-01-01

The tritium system design for the Fusion Power Demonstration Reactor (FPD-I, II, and III) is described. The device operates at 25% availability. For FPD-II, an engineering mode using tritium neutral beams is part of the design

Preliminary analysis of the safety and environmental impact of the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Carlson, R.V.; Jalbert, R.A.

1980-01-01

The Tritium Systems Test Assembly (TSTA) is a facility dedicated to the development of technologies associated with the D-T fuel cycle of future fusion reactors while demonstrating that TSTA can be operated safely with no significant losses to the environment. During the initial design stage of TSTA, a safety analysis was performed which investigated the effects of major subsystem component failure, the meteorology and seismicity of the site and their possible effect on the facility, and accident scenarios which result in tritium releases. Major releases of tritium to the environment are considered highly improbable since they require a compound failure of primary and secondary containment, along with either a breach of the building or a failure of the Emergency Tritium Cleanup system. Accidental releases caused by natural phenomena (earthquake, tornado, etc.) are considered highly improbable

Simplified fuel cycle tritium inventory model for systems studies -- An illustrative example with an optimized cryopump exhaust system

International Nuclear Information System (INIS)

Kuan, W.; Ho, S.K.

1995-01-01

It is desirable to incorporate safety constraints due to fuel cycle tritium inventories into tokamak reactor design optimization. An optimal scenario to minimize tritium inventories without much degradation of plasma performance can be defined for each tritium processing component. In this work, the computer code TRUFFLES is used exclusively to obtain numerical data for a simplified model to be used for systems studies. As an illustration, the cryopump plasma exhaust subsystem is examined in detail for optimization purposes. This optimization procedure will then be used to further reduce its window of operation and provide constraints on the data used for the simplified tritium inventory model

Experiments on a ceramic electrolysis cell and a palladium diffuser at the tritium systems test assembly

International Nuclear Information System (INIS)

Konishi, Satoshi; Yoshida, Hiroshi; Ohno, Hideo; Naruse, Yuji; Coffin, D.O.; Walthers, C.R.; Binning, K.E.

1985-01-01

A ceramic electrolysis cell and a palladium diffuser are developed in Japan and is tested with tritium in Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory, in order to confirm the feasibility as possible upgrades for the fuel cleanup system (PCU). The ceramic electrolysis cell made of stabilized zirconia was operated at 630 0 C for an extended period with a mixture of 3% T 2 O in He carrier gas in the circulation system with oxidizing catalyst bed. The palladium diffuser was tested with circulated pure tritium gas at 280 0 C to verify the compatibility of the alloy with tritium, since the 3 He produced in the metal could cause a degradation. The isotopic effects were also measured for both devices

Tritium Waste Treatment System component failure data analysis from June 18, 1984--December 31, 1989

International Nuclear Information System (INIS)

Cadwallader, L.C.; Stolpe Gavett, M.A.

1990-09-01

This document gives the failure rates for the major tritium-bearing components in the Tritium Waste Treatment System at the Tritium Systems Test Assembly, which is a fusion research and technology facility at the Los Alamos National Laboratory. The failure reports, component populations, and operating demands/hours are given in this report, and sample calculations for binomial demand failure rates and poisson hourly failure rates are given in the appendices. The failure rates for tritium-bearing components were on the order of the screening failure rate values suggested for fusion reliability and risk analyses. More effort should be directed toward collecting and analyzing fusion component failure data, since accurate failure rates are necessary to refine reliability and risk analyses. 15 refs., 4 figs., 4 tabs

Tritium separation from light and heavy water by bipolar electrolysis

International Nuclear Information System (INIS)

Petek, M.; Ramey, D.W.; Taylor, R.D.

1981-01-01

Using multiple bipolar electrolytic separation of hydrogen isotopes with Pd-25%Ag electrodes, the mathematical feasibility of this method for tritium separation was shown and experimentally verified. Separation factors were measured on single bipolar electrodes and were found to be approximately equivalent to those associated with individual ordinary electrolytic systems. Multibipolar separations were experimentally achieved in single cascaded cells in which each bipolar electrode was of equal area to others in a series arrangement. Factors measured for multibipolar H-D separation were close to the values measured in single-stage cell measurements; for H-T separation, interstage leakage reduced the measured separation factor. However, in both cases separation of sufficient magnitude was achieved to show feasibility for real application to the extraction of tritium from large-volume systems at high current density. (author)

Use of tritium-labeled PCBs for investigation of PCBs biodegradation by soil bacteria

International Nuclear Information System (INIS)

Kim, A.A.; Djuraeva, G.T.; Takhtobiri, K.S.; Yadgarov, H.T.; Zinovev, P. V.; Abdukarimov, A.A.

2002-01-01

The method for tritium labelling of polychlorinated biphenyls (PCBs) was developed. The strains of soil bacteria - destructors of chloro organic compounds was studied with the help of test-system based on the using of tritium-labeled PCBs. The strains of bacteria were grown on the agar synthetic medium and then were introduced into the synthetic medium containing tritium-labeled mixture of PCBs (commercial mark - SOVOL) as alone source of carbon. The samples were analysed after one and two months period of incubation. PCBs were extracted by hexane from fraction of bacteria and fraction of cultural medium and radioactivity was measured. The samples were analyzed by thin layer chromatography (TLC) with following radioautography. Additionally samples were analyzed by gas chromatography. It was found that all selected strains survived in the medium with PCBs as alone source of carbon and bacteria accumulated PCBs from cultural medium. Accumulation of PCBs by strains of bacteria was different. The TLC analysis detected additional compounds labeled by tritium, that prove the degradation of PCBs in presence of bacteria. The gas chromatography analysis of cultural medium and bacteria detected redistribution in the system and qualitative changes of PCBs in bacteria. The strains of bacteria also were grown in model condition on the soil with tritium labeled PCBs. We found that some strains effectively destroy PCBs with decreasing level of tritium label in the soil. The using of tritium labeled PCBs' allows to introduce precise quantitative characteristics for study of accumulation and biodegradation PCBs by soil bacteria strains. Developed test-system is very useful tool for selection of new strains of soil bacteria - destructors of PCBs

A Visual Detection System for Determining Tritium Surface Deposition Employing Phosphor Coated Materials

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.; Zweben, S.J.

1999-01-01

A method for visually observing tritium deposition on the surface of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles is being investigated at the Princeton Plasma Physics Laboratory. A green phosphor (P31, zinc sulfide: copper) similar to that used in oscilloscope screens with a wavelength peak of 530 nm was positioned on the surface of a TFTR D-T tile. The approximately 600 gram tile, which contains approximately 1.5 Ci of tritium located on the top approximately 1-50 microns of the surface, was placed in a two liter lexan chamber at Standard Temperature and Pressure (STP). The phosphor plates and phosphor powder were placed on the surface of the tile which resulted in visible light being observed, the consequence of tritium betas interacting with the phosphor. This technique provides a method of visually observing varying concentrations of tritium on the surface of D-T carbon tiles, and may be employed (in a calibrated system) to obtain quantitative data

Study on conceptual design system of tritium production fusion reactor

International Nuclear Information System (INIS)

He Kaihui

2004-11-01

Conceptual design of an advanced tritium production reactor based on spherical torus, which is intermediate application of fusion energy, was presented. Different from traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can within vacuum vessel in order to produce 1 kg excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented. Besides systematical analyses; design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (author)

Tritium waste package

Science.gov (United States)

Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

1995-01-01

A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

Development of tritium-handling technique

International Nuclear Information System (INIS)

Ohmura, Hiroshi; Hosaka, Akio; Okamoto, Takahumi

1988-01-01

The overview of developing activities for tritium-handling techniques in IHI are presented. To establish a fusion power plant, tritium handling is one of the key technologies. Recently in JAERI, conceptual design of FER (Fusion Experimental Reactor) has been carried out, and the FER system requires a processing system for a large amount of tritium. IHI concentrate on investigation of fuel gas purification, isotope separation and storage systems under contract with Toshiba Corporation. Design results of the systems and each components are reviewed. IHI has been developing fundamental handling techniques which are the ZrNi bed for hydrogen isotope storage and isotope separation by laser. The ZrNi bed with a tritium storage capacity of 1000 Ci has been constructed and recovery capability of the hydrogen isotope until 10 -4 Torr {0.013 Pa} was confirmed. In laser isotope separation, the optimum laser wave length has been determined. (author)

Tritium handling experience in vacuum systems at TSTA [Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Anderson, J.L.; Jenkins, E.M.; Walthers, C.R.; Yoshida, H.; Fukui, H.; Naruse, Y.

1989-01-01

Compound cryopumps have been added to the Tritium Systems Test Assembly (TSTA) integrated fusion fuel loop. Operations have been performed which closely simulate an actual fusion reactor pumping scenario. In addition, performance data have been taken that support the concept of using coconut charcoal as a sorbent at 4K for pumping helium. Later tests show that coconut charcoal may be used to co-pump D,T and He mixtures on a single 4K panel. Rotary spiral pumps have been used successfully in several applications at TSTA and have acquired more than 9000 hours of maintenance-free operation. Metal bellows pumps have been used to back the spiral pumps and have been relatively trouble free in loop operations. Bellows pumps also have more than 9000 hours of maintenance-free operation. 5 refs., 6 figs

Design of a Cryogenic Distillation Column for JET Water Detritiation System for Tritium Recovery

International Nuclear Information System (INIS)

Parracho, A.I.; Camp, P.; Dalgliesh, P.; Hollingsworth, A.; Lefebvre, X.; Lesnoj, S.; Sacks, R.; Shaw, R.; Smith, R.; Wakeling, B.

2015-01-01

A Water Detritiation System (WDS) is currently being designed and manufactured to be installed in the Active Gas Handling System (AGHS) of JET, currently the largest magnetic fusion experiment in the world. JET has been designed and built to study fusion operating conditions with the plasma fuelling done by means of a deuterium-tritium gas mixture. AGHS is a plant designed and built to safely process gas mixtures and impurities containing tritium recovered from the JET torus exhaust gases. Tritium is removed from these gas mixtures and recycled. Tritium depleted gases are sent to Exhaust Detritiation System (EDS) for final tritium removal prior to discharge into the environment. In EDS, tritium and tritiated species are catalytically oxidized into water, this tritiated water is then adsorbed onto molecular sieve beds (MSB). After saturation the MSBs are heated and the water is desorbed and collected for tritium recovery. The WDS facility is designed to recover tritium from water with an average activity of 1.9 GBq/l, and is able to process water with activities of 85 GBq/l and higher. Tritiated water is filtered and supplied to the electrolyser where the water is converted into gaseous oxygen and tritiated hydrogen. The hydrogen stream is first purified by selective diffusion through membranes of palladium alloy and then is fed to two cryogenic distillation columns (CD). These operate in parallel or in series depending on the water activity. In the CD columns, hydrogen isotopes containing tritium are recovered as the bottom product and hydrogen, the top product, is safely discarded to a stack. The CD columns are foreseen to have a throughput between 200 and 300 mole/h of hydrogen isotopes vapour and they operate at approximately ≈21.2K and 105 kPa. The design of the CD columns will be presented in this work. This work has been carried out within the framework of the Contract for the Operation of the JET Facilities and has received funding from the European Union�

Tritium Management Loop Design Status

Energy Technology Data Exchange (ETDEWEB)

Rader, Jordan D. [ORNL; Felde, David K. [ORNL; McFarlane, Joanna [ORNL; Greenwood, Michael Scott [ORNL; Qualls, A L. [ORNL; Calderoni, Pattrick [Idaho National Laboratory (INL)

2017-12-01

This report summarizes physical, chemical, and engineering analyses that have been done to support the development of a test loop to study tritium migration in 2LiF-BeF2 salts. The loop will operate under turbulent flow and a schematic of the apparatus has been used to develop a model in Mathcad to suggest flow parameters that should be targeted in loop operation. The introduction of tritium into the loop has been discussed as well as various means to capture or divert the tritium from egress through a test assembly. Permeation was calculated starting with a Modelica model for a transport through a nickel window into a vacuum, and modifying it for a FLiBe system with an argon sweep gas on the downstream side of the permeation interface. Results suggest that tritium removal with a simple tubular permeation device will occur readily. Although this system is idealized, it suggests that rapid measurement capability in the loop may be necessary to study and understand tritium removal from the system.

Control system for technological processes in tritium processing plants with process analysis

International Nuclear Information System (INIS)

Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu; Bucur, Ciprian

2005-01-01

Integration of a large variety of installations and equipment into a unitary system for controlling the technological process in tritium processing nuclear facilities appears to be a rather complex approach particularly when experimental or new technologies are developed. Ensuring a high degree of versatility allowing easy modifications in configurations and process parameters is a major requirement imposed on experimental installations. The large amount of data which must be processed, stored and easily accessed for subsequent analyses imposes development of a large information network based on a highly integrated system containing the acquisition, control and technological process analysis data as well as data base system. On such a basis integrated systems of computation and control able to conduct the technological process could be developed as well protection systems for cases of failures or break down. The integrated system responds to the control and security requirements in case of emergency and of the technological processes specific to the industry that processes radioactive or toxic substances with severe consequences in case of technological failure as in the case of tritium processing nuclear plant. In order to lower the risk technological failure of these processes an integrated software, data base and process analysis system are developed, which, based on identification algorithm of the important parameters for protection and security systems, will display the process evolution trend. The system was checked on a existing plant that includes a removal tritium unit, finally used in a nuclear power plant, by simulating the failure events as well as the process. The system will also include a complete data base monitoring all the parameters and a process analysis software for the main modules of the tritium processing plant, namely, isotope separation, catalytic purification and cryogenic distillation

Mercury and tritium removal from DOE waste oils

Energy Technology Data Exchange (ETDEWEB)

Klasson, E.T. [Oak Ridge National Lab., TN (United States)

1997-10-01

This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

Tritium containment in fusion facilities

International Nuclear Information System (INIS)

Anderson, J.L.

1978-01-01

The key environmental control systems that have been identified and are being developed are listed. A brief description of each of the following systems is given: primary process materials, permeation barriers, secondary containment, tritium waste treatment, emergency tritium cleanup, maintenance procedures, and tertiary containment

Tritium proof-of-principle pellet injector results

International Nuclear Information System (INIS)

Fisher, P.W.; Fehling, D.T.; Gouge, M.J.; Milora, S.L.

1989-01-01

The tritium proof-of-principle (TPOP) experiment was built by Oak Ridge National Laboratory (ORNL) to demonstrate the feasibility of forming solid tritium pellets and accelerating them to high velocities for fueling future fusion reactors. TPOP used a pneumatic pipe-gun with a 4-mm-i.d. by 1-m-long barrel. Nearly 1500 pellets were fired by the gun during the course of the experiment; about a third of these were tritium or mixtures of deuterium and tritium. The system also contained a cryogenic 3 He separator that reduced the 3 He level to <0.005%. Pure tritium pellets were accelerated to 1400 m/s. Experiments evaluated the effect of cryostat temperature and fill pressure on pellet size, the production of pellets from mixtures of tritium and deuterium, and the effect of aging on pellet integrity. The tritium phase of these experiments was performed at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. About 100 kCi of tritium was processed through the apparatus without incident. 8 refs., 7 figs

Survey of pumps for tritium gas

International Nuclear Information System (INIS)

Dowell, T.M.

1983-05-01

This report considers many different types of pumps for their possible use in pumping tritium gas in the low, intermediate and high vacuum ranges. No one type of pump is suitable for use over the wide range of pumping pressure required in a typical pumping system. The favoured components for such a system are: bellows pump (low vacuum); orbiting scroll pump (intermediate vacuum); magnetically suspended turbomolecular pump (high vacuum); cryopump (high vacuum). Other pumps which should be considered for possible future development are: mound modified vane pump; SRTI wobble pump; roots pump with canned motor. It is proposed that a study be made of a future tritium pumping system in a Canadian tritium facility, e.g. a tritium laboratory

Analysis of trace levels of impurities and hydrogen isotopes in helium purge gas using gas chromatography for tritium extraction system of an Indian lead lithium ceramic breeder test blanket module.

Science.gov (United States)

Devi, V Gayathri; Sircar, Amit; Yadav, Deepak; Parmar, Jayraj

2018-01-12

In the fusion fuel cycle, the accurate analysis and understanding of the chemical composition of any gas mixture is of great importance for the efficient design of a tritium extraction and purification system or any tritium handling system. Methods like laser Raman spectroscopy and gas chromatography with thermal conductivity detector have been considered for hydrogen isotopes analyses in fuel cycles. Gas chromatography with a cryogenic separation column has been used for the analysis of hydrogen isotopes gas mixtures in general due to its high reliability and ease of operation. Hydrogen isotopes gas mixture analysis with cryogenic columns has been reported earlier using different column materials for percentage level composition. In the present work, trace levels of hydrogen isotopes (∼100 ppm of H 2 and D 2 ) have been analyzed with a Zeolite 5A and a modified γ-Al 2 O 3 column. Impurities in He gas (∼10 ppm of H 2 , O 2 , and N 2 ) have been analyzed using a Zeolite 13-X column. Gas chromatography with discharge ionization detection has been utilized for this purpose. The results of these experiments suggest that the columns developed were able to separate ppm levels of the desired components with a small response time (<6 min) and good resolution in both cases. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

TFTR tritium operations lessons learned

International Nuclear Information System (INIS)

Gentile, C.A.; Raftopoulos, S.; LaMarche, P.

1996-01-01

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980's The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America's first operational D-T fusion reactor. This paper will discuss negative pressure employing 'elephant trunks' in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and Δ pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs

Tritium breeding blanket

International Nuclear Information System (INIS)

Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

1991-01-01

The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

Tritium in water monitor for measurement of tritium activity in the process water

International Nuclear Information System (INIS)

Rathnakaran, M.; Ravetkar, R.M.; Abani, M.C.; Mehta, S.K.

1999-01-01

This paper presents the evaluation of a tritium in water monitor for measurement of tritium activity in the secondary coolant in pressurised heavy water reactor used for power generation. For this purpose it uses a plastic scintillator flow cell detector in a continuous on-line mode. It is observed that the sensitivity of the system depends on the transparency of the detector, which gradually reduces with use because of the collection of dirt around the scintillator. A simple type of sample conditioner based on polypropylene candle filter and filter paper is developed and installed at RAPS along with tritium in water monitor. The functioning of this system is reported here. (author)

Combined gettering and molten salt process for tritium recovery from lithium

International Nuclear Information System (INIS)

Sze, D.K.; Finn, P.A.; Bartlit, J.; Tanaka, S.; Teria, T.; Yamawaki, M.

1988-02-01

A new tritium recovery concept from lithium has been developed as part of the US/Japan collaboration on Reversed-Field Pinch Reactor Design Studies. This concept combines the γ-gettering process as the front end to recover tritium from the coolant, and a molten salt recovery process to extract tritium for fuel processing. A secondary lithium is used to regenerate the tritium from the gettering bed and, in the process, increases the tritium concentration by a factor of about 20. That way, the required size of the molten salt process becomes very small. A potential problem is the possible poisoning of the gettering bed by the salt dissolved in lithium. 16 refs., 6 figs

Tissue free water tritium separation from foodstuffs by azeotropic distillation

International Nuclear Information System (INIS)

Constantin, F.; Ciubotaru, A.; Popa, D.

1998-01-01

In this paper the tritium separation from tissue free water in foodstuffs by azeotropic distillation is described. Tritium in tissue water is assayed by liquid scintillation counting using well-established quenched correction method. The mean value of the tritium concentration in tissue water from foodstuffs is about 6-12 Bq/l very similar to the tritium mean concentration measured in the surface waters of the area where the samples have been collected (about 12 Bq/l. Therefore, the tritium content in the water fraction of the food samples can be considered in equilibrium with the local environmental water sources. The azeotropic distillation it is an accessible separation method which does not need a sophisticated and expansive distillation apparatus. It is a fast method of separation tissue free water from foodstuffs being very important in the surveillance activity of the environmental within nuclear electric plant. It is suitable for processing a small quantity of samples and for a production type facility when a large number of samples must be processed because the solvent can be purified and reused. The azeotropic distillation has some limits being used to separate water from samples with high content of water (85-90%) and simple a simple chemical structures as: vegetables, fruits, cereal, soil, vegetation. According to the results obtained, the organic substituents of milk, wine, meat (casein, lactose, milk fat, alcohol, esters) may enhance the chemisorption of tritium on through exchange organic hydrogen as -OH, -SH, -NH, -COOH with tritium. Also, the tissue water separation by azeotropic distillation is not complete and can not guarantee the absence of the vaporization isotope effect of the HTO/H 2 O system., However, the azeotropic distillation is the preferred method of the water extraction from food samples, which makes it useful for the tritium transfer from soil to foodstuffs. (authors)

Tissue free water tritium separation from foodstuffs by azeotropic distillation

Energy Technology Data Exchange (ETDEWEB)

Constantin, F; Ciubotaru, A; Popa, D [Inspectorate of Public Health of Bucharest (Romania)

1999-12-31

In this paper the tritium separation from tissue free water in foodstuffs by azeotropic distillation is described. Tritium in tissue water is assayed by liquid scintillation counting using well-established quenched correction method. The mean value of the tritium concentration in tissue water from foodstuffs is about 6-12 Bq/l very similar to the tritium mean concentration measured in the surface waters of the area where the samples have been collected (about 12 Bq/l. Therefore, the tritium content in the water fraction of the food samples can be considered in equilibrium with the local environmental water sources. The azeotropic distillation it is an accessible separation method which does not need a sophisticated and expansive distillation apparatus. It is a fast method of separation tissue free water from foodstuffs being very important in the surveillance activity of the environmental within nuclear electric plant. It is suitable for processing a small quantity of samples and for a production type facility when a large number of samples must be processed because the solvent can be purified and reused. The azeotropic distillation has some limits being used to separate water from samples with high content of water (85-90%) and simple a simple chemical structures as: vegetables, fruits, cereal, soil, vegetation. According to the results obtained, the organic substituents of milk, wine, meat (casein, lactose, milk fat, alcohol, esters) may enhance the chemisorption of tritium on through exchange organic hydrogen as -OH, -SH, -NH, -COOH with tritium. Also, the tissue water separation by azeotropic distillation is not complete and can not guarantee the absence of the vaporization isotope effect of the HTO/H{sub 2}O system., However, the azeotropic distillation is the preferred method of the water extraction from food samples, which makes it useful for the tritium transfer from soil to foodstuffs. (authors) 2 tabs.

Catalytic oxidation efficiencies for tritium and tritiated methane in a mature, industrial-scale decontamination system

International Nuclear Information System (INIS)

Mintz, J.M.; Gildea, P.D.

1981-01-01

Almost all tritium decontamination systems proposed for fusion facilities employ catalytic oxidation to water, followed by drying, to remove tritium and tritiated hydrocarbons from gas streams. One such large-scale system, the gas purification system (GPS), has been operating in the Tritium Research Laboratory (TRL) at Sandia National Laboratories, Livermore, CA, since October 1977. A series of experiments have recently been conducted there to assesss the current operating characteristics of the GPS catalyst. The experiments used tritium and tritiated methane and covered a range of temperatures, flow rates, and concentration levels. When contrasted with 1977 data, the results indicate that no measurable degradation of catalyst function had occurred. However, some reduction in active metal surface area, as indicated by B.E.T. surface area measurements (approx. 100 → 90m 2 /g) and AES scans (approx. 1.4 → 0.9 at. % Pt), had occurred. Kinetic rate coefficients were also derived and a rough temperature dependence obtained

Catalytic oxidation efficiencies for tritium and tritiated methane in a mature, industrial-scale decontamination system

International Nuclear Information System (INIS)

Mintz, J.M.; Gildea, P.D.

1980-10-01

Almost all tritium decontamination systems proposed for fusion facilities employ catalytic oxidation to water, followed by drying, to remove tritium and tritiated hydrocarbons from gas streams. One such large-scale system, the gas purification system (GPS), has been operating in the Tritium Research Laboratory (TRL) at Sandia National Laboratories, Livermore, CA, since October 1977. A series of experiments have recently been conducted there to assess the current operating characteristics of the GPS catalyst. The experiments used tritium and tritiated methane and covered a range of temperatures, flow rates, and concentration levels. When contrasted with 1977 data, the results indicate that no measurable degradation of catalyst function had occurred. However, some reduction in active metal surface area, as indicated by B.E.T. surface area measurements (approx. 100 → 90 m 2 /g) and AES scans (approx. 1.4 → 0.9 at% Pt), had occurred. Kinetic rate coefficients were also derived and a rough temperature dependence obtained

Storage and Assay of Tritium in STAR

International Nuclear Information System (INIS)

Longhurst, Glen R.; Anderl, Robert A.; Pawelko, Robert J.; Stoots, Carl J.

2005-01-01

The Safety and Tritium Applied Research (STAR) facility at the Idaho National Engineering and Environmental Laboratory (INEEL) is currently being commissioned to investigate tritium-related safety questions for fusion and other technologies. The tritium inventory for the STAR facility will be maintained below 1.5 g to avoid the need for STAR to be classified as a Category 3 nuclear facility. A key capability in successful operation of the STAR facility is the ability to receive, inventory, and dispense tritium to the various experiments underway there. The system central to that function is the Tritium Storage and Assay System (SAS).The SAS has four major functions: (1) receiving and holding tritium, (2) assaying, (3) dispensing, and (4) purifying hydrogen isotopes from non-hydrogen species.This paper describes the design and operation of the STAR SAS and the procedures used for tritium accountancy in the STAR facility

Tritium research activities in Korea

Energy Technology Data Exchange (ETDEWEB)

Jung, Ki Jung, E-mail: kjjung@nfri.re.kr [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Yun, Sei-Hun, E-mail: shyun@nfri.re.kr [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Chang, Min Ho; Kang, Hyun-Goo; Chung, Dongyou; Cho, Seungyon; Lee, Hyeon Gon [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Chung, Hongsuk; Choi, Woo-Seok [Korea Atomic Energy Research Institute, Yusung-gu, Daejeon 305-353 (Korea, Republic of); Song, Kyu-Min; Moon, Chang-Bae [Korea Hydro & Nuclear Power Central Research Institute, Yusung-gu, Daejeon 305-343 (Korea, Republic of); Lee, Euy Soo [Dongguk University, Jung-gu, Seoul, 100-715 (Korea, Republic of); Cho, Jungho; Kim, Dong-Sun [Kongju National University, Cheonan, Chungnam, 330-717 (Korea, Republic of); Moon, Hung-Man [Daesung Industrial Gases Co., Ltd., Danwon-gu, Ansan-si, Gyeonggi-do, 425-090 (Korea, Republic of); Noh, Seung Jeong [Dankook University, Suji-gu, Yongin-si, Gyeonggi-do, 448-701 (Korea, Republic of); Ju, Hyunchul [Inha University, Nam-gu, Incheon, 402-751 (Korea, Republic of); Hong, Tae-Whan [Korea National University of Transportation, Chungju, Chungbuk, 380-702 (Korea, Republic of)

2016-12-15

Highlights: • NFRI, KAERI and KHNP CRI are major leading group for the ITER tritium SDS design; studying engineering, simulation of hydride bed, risk analysis (on safety, HAZOP), basic study, control logic & sequential operation, and others. KHNP has WTRF which gives favorable experiences for collaboration researchers. • Supplementary research partners: Five Universities (Dongguk University and POSTECH, Inha University, Dankook University, Korea National Transport University, and Kongju National University) and one industrial company (Daesung Industrial Gases Co., Ltd.); studying on basic and engineering, programming & simulation on the various topics for ITER tritium SDS, TEP, ISS, ADS, and etc. - Abstract: Major progress in tritium research in the Republic of Korea began when Korea became responsible for ITER tritium Storage and Delivery System (SDS) procurement package which is part of the ITER Fuel Cycle. To deliver the tritium SDS package, a variety of research institutes, universities and industry have respectively taken roles and responsibilities in developing technologies that have led to significant progress. This paper presents the current work and status of tritium related technological research and development (R&D) in Korea and introduces future R&D plans in the area of fuel cycle systems for fusion power generation.

Tritium transfer studies in cellulose-HTO system

International Nuclear Information System (INIS)

Jayaraman, A.P.; Misra, B.M.

1986-01-01

This paper describes some aspects of studies on transfer of tritium to cellulose from tritiated water at six different specific activities and discusses the generalized tritiation pattern. Cellulose was irradiated in steps to 10 M Rads and the tritium transfer was determined at each stage. Experimental results signify substantial increase of tritiation in cellulose at higher dose of irradiation. (author). 8 refs

The Safety and Tritium Applied Research (STAR) Facility: Status-2004

International Nuclear Information System (INIS)

Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Sharpe, J.P.; Schuetz, S.T.; Petti, D.A.

2005-01-01

The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems

Subsystem software for TSTA [Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Mann, L.W.; Claborn, G.W.; Nielson, C.W.

1987-01-01

The Subsystem Control Software at the Tritium System Test Assembly (TSTA) must control sophisticated chemical processes through the physical operation of valves, motor controllers, gas sampling devices, thermocouples, pressure transducers, and similar devices. Such control software has to be capable of passing stringent quality assurance (QA) criteria to provide for the safe handling of significant amounts of tritium on a routine basis. Since many of the chemical processes and physical components are experimental, the control software has to be flexible enough to allow for trial/error learning curve, but still protect the environment and personnel from exposure to unsafe levels of radiation. The software at TSTA is implemented in several levels as described in a preceding paper in these proceedings. This paper depends on information given in the preceding paper for understanding. The top level is the Subsystem Control level

Experiences with decontaminating tritium-handling apparatus

International Nuclear Information System (INIS)

Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

1992-01-01

Tritium-handling apparatus has been decontaminated as part of the downsizing of the LLNL Tritium Facility. Two stainless-steel glove boxes that had been used to process lithium deuteride-tritide (LiDT) slat were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. In this paper the details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium, in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given

Tritium monitoring in environment at ICIT Tritium Separation Facility

International Nuclear Information System (INIS)

Varlam, Carmen; Stefanescu, I.; Vagner, Irina; Faurescu, I.; Toma, A.; Dulama, C.; Dobrin, R.

2008-01-01

Full text: The Cryogenic Pilot is an experimental project developed within the national nuclear energy research program, which is designed to develop the required technologies for tritium and deuterium separation by cryogenic distillation of heavy water. The process used in this installation is based on a combination between liquid-phase catalytic exchange (LPCE) and cryogenic distillation. Basically, there are two ways that the Cryogenic Pilot could interact with the environment: by direct atmospheric release and through the sewage system. This experimental installation is located 15 km near the region biggest city and in the vicinity - about 1 km, of Olt River. It must be specified that in the investigated area there is an increased chemical activity; almost the entire Experimental Cryogenic Pilot's neighborhood is full of active chemical installations. This aspect is really essential for our study because the sewerage system is connected with the other three chemical plants from the neighborhood. For that reason we progressively established elements of an environmental monitoring program well in advance of tritium operation in order to determine baseline levels. The first step was the tritium level monitoring in environmental water and wastewater of industrial activity from neighborhood. In order to establish the base level of tritium concentration in the environment around the nuclear facilities, we investigated the sample preparation treatment for different types of samples: onion, green beams, grass, apple, garden lettuce, tomato, cabbage, strawberry and grapes. We used azeotropic distillation of all types of samples, the carrier solvent being toluene from different Romanian providers. All measurements for the determination of environmental tritium concentration were performed using liquid scintillation counting (LSC), with the Quantulus 1220 spectrometer. (authors)

Tritium proof-of-principle injector experiment

International Nuclear Information System (INIS)

Fisher, P.W.; Milora, S.L.; Combs, S.K.; Carlson, R.V.; Coffin, D.O.

1988-01-01

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 fision 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 Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). 18 refs., 3 figs

The operation of the Tokamak Fusion Test Reactor Tritium Facility

International Nuclear Information System (INIS)

Gentile, C.A.; LaMarche, P.H.

1995-01-01

The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

Results of tritium experiments on ceramic electrolysis cells and palladium diffusers for application to fusion reactor fuel cleanup systems

International Nuclear Information System (INIS)

Carlson, R.V.; Binning, K.E.; Konishi, S.; Yoshida, H.; Naruse, Y.

1987-01-01

Tritium tests at the Tritium Systems Test Assembly have demonstrated that ceramic electrolysis cells and palladium alloy diffuser developed in Japan are possible components for a fusion reactor fuel cleanup system. Both components have been successfully operated with tritium for over a year. A failure of the first electrolysis cell was most likely the result of an over voltage on the ceramic. A simple circuit was developed to eliminate this mode of failure. The palladium diffusers tubes exhibited some degradation of mechanical properties as a result of the build up of helium from the tritium decay, after 450 days of operation with tritium, however the effects were not significant enough to affect the performance. New models of the diffuser and electrolysis cell, providing higher flow rates and more tritium compatible designs are currently being tested with tritium. 8 refs., 5 figs

Tritium emissions from a detritiation facility

International Nuclear Information System (INIS)

Rodrigo, L.; El-Behairy, O.; Boniface, H.; Hotrum, C.; McCrimmon, K.

2010-01-01

Tritium is produced in heavy-water reactors through neutron capture by the deuterium atom. Annual production of tritium in a CANDU reactor is typically 52-74 TBq/MW(e). Some CANDU reactor operators have implemented detritiation technology to reduce both tritium emissions and dose to workers and the public from reactor operations. However, tritium removal facilities also have the potential to emit both elemental tritium and tritiated water vapor during operation. Authorized releases to the environment, in Canada, are governed by Derived Release Limits (DRLs). DRLs represent an estimate of a release that could result in a dose of 1 mSv to an exposed member of the public. For the Darlington Nuclear Generating Station, the DRLs for airborne elemental tritium and tritiated water emissions are ~15.6 PBq/week and ~825 TBq/week respectively. The actual tritium emissions from Darlington Tritium Removal Facility (DTRF) are below 0.1% of the DRL for elemental tritium and below 0.2% of the DRL for tritiated water vapor. As part of an ongoing effort to further reduce tritium emissions from the DTRF, we have undertaken a review and assessment of the systems design, operating performance, and tritium control methods in effect at the DTRF on tritium emissions. This paper discusses the results of this study. (author)

A system for tritium analysis in natural water

International Nuclear Information System (INIS)

Mozeto, A.A.

1977-01-01

A method for the analysis, by scintillation counting, of tritium in natural water enriched electrolytically, is presented. The characteristics of the proposed system are indicated by experimental parameters, and by the performance obtained in the analysis of rain and under ground waters. An evaluation of the precison and reproducibility of the measurements is also made [pt

The tritium content of precipitation and groundwater at Yola, Nigeria ...

African Journals Online (AJOL)

Tritium is a radioactive isotope of hydrogen which occurs in precipitation. In groundwater studies tritium measurements give information on the time of recharge to the system; the tritium content of precipitation being used to estimate the input of tritium to the groundwater system. At Yola, the tritium ontents in precipitation and ...

Tritium pellet injector for TFTR

International Nuclear Information System (INIS)

Gouge, M.J.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foust, C.R.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Wilgen, J.B.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

1992-01-01

The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

Subsystem cost data for the tritium systems test assembly

International Nuclear Information System (INIS)

Bartlit, J.R.; Anderson, J.L.; Rexroth, V.G.

1983-01-01

Details of subsystem costs are among the questions most frequently asked about the $14.4 million Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. This paper presents a breakdown of cost components for each of the 20 major subsystems of TSTA. Also included are details to aid in adjusting the costs to other years, contracting conditions, or system sizes

Effects of tritium in elastomers

International Nuclear Information System (INIS)

Zapp, P.E.

1982-01-01

Elastomers are used as flange gaskets in the piping system of the Savannah River Plant tritium facilities. A number of elastomers is being examined to identify those compounds more radiation-resistant than the currently specified Buna-N rubber and to study the mechanism of tritium radiation damage. Radiation resistance is evaluated by compression set tests on specimens exposed to about 1 atm tritium for several months. Initial results show that ethylene-propylene rubber and three fluoroelastomers are superior to Buna-N. Off-gassing measurements and autoradiography show that retained surface absorption of tritium varies by more than an order of magnitude among the different elastomer compounds. Therefore, tritium solubility and/or exchange may have a role in addition to that of chemical structure in the damage process. Ongoing studies of the mechanism of radiation damage include: (1) tritium absorption kinetics, (2) mass spectroscopy of radiolytic products, and (3) infrared spectroscopy

Accelerator Production of Tritium Programmatic Environmental Impact Statement Input Submittal

International Nuclear Information System (INIS)

Miller, L.A.; Greene, G.A.; Boyack, B.E.

1996-02-01

The Programmatic Environmental Impact Statement for Tritium Supply and Recycling considers several methods for the production of tritium. One of these methods is the Accelerator Production of Tritium. This report summarizes the design characteristics of APT including the accelerator, target/blanket, tritium extraction facility, and the balance of plant. Two spallation targets are considered: (1) a tungsten neutron-source target and (2) a lead neutron-source target. In the tungsten target concept, the neutrons are captured by the circulating He-3, thus producing tritium; in the lead target concept, the tritium is produced by neutron capture by Li-6 in a surrounding lithium-aluminum blanket. This report also provides information to support the PEIS including construction and operational resource needs, waste generation, and potential routine and accidental releases of radioactive material. The focus of the report is on the impacts of a facility that will produce 3/8th of the baseline goal of tritium. However, some information is provided on the impacts of APT facilities that would produce smaller quantities

Tritium safety issues for TFCX

International Nuclear Information System (INIS)

Reilly, H.J.; Piet, S.J.; Merrill, B.J.

1985-01-01

Estimated tritium releases from the Tokamak Fusion Core Experiment are compared to the expected limits. A reaction kinetics model is described that predicts the conversion of tritium to the oxide form in free space. An analysis of the required capacity of the Emergency Tritium Cleanup System is also presented. The conclusions of this work are expected to be applicable to other experimental fusion devices that are now being considered

Tritium distribution ratios between the 30 % tributyl phosphate(TBP)-normal dodecane(nDD) organic phase and uranyl nitrate-nitric acid aqueous phase

International Nuclear Information System (INIS)

Fujine, Sachio; Uchiyama, Gunzou; Sugikawa, Susumu; Maeda, Mitsuru; Tsujino, Takeshi.

1989-10-01

Tritium distribution ratios between the organic and aqueous phases were measured for the system of 30 % tributyl phosphate(TBP)-normal dodecane(nDD)/uranyl nitrate-nitric acid water. It was confirmed that tritium is extracted by TBP into the organic phase in both chemical forms of tritiated water (HTO) and tritiated nitric acid (TNO 3 ). The value of tritium distribution ratio ranged from 0.002 to 0.005 for the conditions of 0-6 mol/L nitric acid, 0.5-800 mCi/L tritium in aqueous phase, and 0-125 g-U/L uranium in organic phase. Isotopic distribution coefficient of tritium between the organic and aqueous phases was observed to be about 0.95. (author)

Considerations for tritium protection at a fusion reactor

International Nuclear Information System (INIS)

Easterly, C.E.

1981-01-01

The purpose of this paper is to indicate the general direction of current fusion reactor concepts regarding tritium, and to indicate some options in tritium control strategies. Certain strategies, in addition to providing reduced potential health hazard, afford the potential for engineering alternatives for in-plant tritium control systems. The overall coupling of containment cleanup systems and health protection must continue to develop with increased knowledge of the health effects of different tritium species and the consequent systems options available subsequent to this understanding

LIBRETTO-3: Performance of tritium permeation barriers under irradiation at the HFR Petten

International Nuclear Information System (INIS)

Conrad, R.; Fuetterer, M.A.; Giancarli, L.; May, R.; Perujo, A.; Sample, T.

1994-01-01

The LIBRETTO-3 irradiation was performed at the HFR Petten during 77 full power days in three cycles to compare the efficiency of three different tritium permeation barriers in presence of Pb-17Li to uncoated AISI 316L steel. For this purpose four steel capsules (arnothing o =10 mm, arnothing i =8 mm, l=300 mm) were filled with 28 g Pb-17Li. The coatings included CVD TiC (outside), CVD TiC+Al 2 O 3 (inside), and pack cementation aluminisation (inside). The generated tritium was partly extracted by bubbling, partly it permeated through the capsules. Permeated and extracted tritium were measured as a function of temperature (280-450 C), H 2 doping (0-1 vol%) and purge gas flow rate. The driving partial pressure in the coated capsules were from an extraction model calibrated by the uncoated capsule for which tritium partial pressure could be calculated. In LIBRETTO-3 conditons, the best barrier was pack cementation aluminisation. The first interpretation of the experiment could, however, not confirm permeation reduction factors of 100 or more expected from this barrier. ((orig.))

Tritium transport calculations for the IFMIF Tritium Release Test Module

Energy Technology Data Exchange (ETDEWEB)

Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

2014-10-15

Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

Tritium transport calculations for the IFMIF Tritium Release Test Module

International Nuclear Information System (INIS)

Freund, Jana; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

2014-01-01

Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

Consideration of disposal alternatives for tritium-contaminated wastewater streams at Hanford

International Nuclear Information System (INIS)

Waters, E.D.

1988-03-01

Small quantities of tritium are produced as an undesirable by-product of the operation of light-water reactors. At the US Department of Energy Hanford Site in Washington State, some tritium has been discharged to the environment in low-level liquid and gaseous wastes from the N Reactor plant, but more than 97% of the tritium stays typically within the irradiated fuel as it is delivered for reprocessing. During fuel reprocessing, the tritium is distributed in the process streams, and most of the tritium is presently released to the soil column with excess process condensates from the Plutonium-Uranium Extraction (PUREX) Plant. On an annual basis, approximately 1 g of tritium is discharged in more than 1 x 10 6 L of process condensate water. Principal tritium release points and quantities are presented in section 4.0. The present study is intended to identify and evaluate alternate methods of tritium control and disposal that might merit additional study or development for potential application to Hanford Site effluents. 30 refs., 15 figs., 5 tabs

Use of system code to estimate equilibrium tritium inventory in fusion DT machines, such as ARIES-AT and components testing facilities

International Nuclear Information System (INIS)

Wong, C.P.C.; Merrill, B.

2014-01-01

Highlights: • With the use of a system code, tritium burn-up fraction (f burn ) can be determined. • Initial tritium inventory for steady state DT machines can be estimated. • f burn of ARIES-AT, CFETR and FNSF-AT are in the range of 1–2.8%. • Respective total tritium inventories of are 7.6 kg, 6.1 kg, and 5.2 kg. - Abstract: ITER is under construction and will begin operation in 2020. This is the first 500 MW fusion class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a system code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively

TSTA loop operation with 100 grams-level of tritium

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Hirata, Shingo; Naito, Taisei

1988-10-01

The first loop operation tests of Tritium Systems Test Assembly (TSTA) with 100 grams-level of tritium were carried out at Los Alamos National Laboratory(LANL) on June and July, 1987. The tests were one of the milestones for TSTA goal scheduled in June, 1987 through June, 1988. The objectives were (i) to operate TSTA process loop composed of tritium supply system, fuel gas purification system, hydrogen isotope separation system, etc, (ii) to demonstrate TSTA safety subsystems such as secondary containment system, tritium waste treatment system and tritium monitoring system, and (iii) to accumulate handling experience of a large amount of tritium. This report describes the plan and procedures of the milestone run done in June and the summary results especially on the safety aspects. Analysis of the emergency shutdown of the process loop, which happened in the June run, is also reported. A brief description of the process and safety subsystems as well as the summary of the TSTA safety analysis report is included. (author)

Tritium production and processing in a Tokamak reactor

International Nuclear Information System (INIS)

Leger, D.

1986-09-01

Important aspects of the tritium system in Tokamak reactors that have to be controlled are overviewed in this paper. The doubling time is one of them, that is to say the time required to produce, in addition to the tritium burned enough tritium to be able to supply the initial tritium inventory. Another one is the tritium permeation through walls. In addition to the permeation phenomena, large tritium inventories are trapped in the reactor structural material. Finally, the different atmospheres of halls, etc.., that can be contaminated with tritium, have to be reprocessed

TSTA loop operation with 100 grams-level of tritium

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Fukui, Hiroshi; Hirata, Shingo

1988-12-01

A fully integrated loop operation test of Tritium systems Test Assembly (TSTA) with 107 grams of tritium was completed at Los Alamos National Laboratory (LANL) in June, 1988. In this test, a compound cryopump with a charcoal panel was incorporated into the main process loop for the first time. The objectives were (i) to demonstrate the compound cryopump system with different flow rates and impurities, (ii) to demonstrate the regeneration of the compound cryopump system, (iii) to accumulate operating experience with other process systems such as the fuel cleanup system, the isotope separation system, the tritium supply and recovery system, etc. and (iv) to improve the data-base on TSTA safety systems such as the secondary containment system, tritium waste treatment system and tritium monitoring system. This report briefly describes characteristics of the main subsystems observed during the milestone run. (author)

Removal of contaminating tritium and tritium pressure measurement by a secondary electron multiplier

International Nuclear Information System (INIS)

Ichimura, K.; Watanabe, K.; Nishizawa, K.; Fujita, J.

1984-01-01

A ceramic secondary electron multiplier (SEM), Ceratron, was used to study impairment of the SEM performance due to adsorbed tritium, its decontamination, and the applicability of the SEM to measure tritium pressure. The background level of the SEM increased significantly, up to its counting limit, due to tritium adsorption. Heating it to 300 0 C in vacuo and/or in the presence of reactive gases such as D 2 and CO at 1 x 10 -4 Pa was not effective to decontaminate the SEM, whereas photon irradiation was extremely powerful for the decontamination. The tritium (HT) pressure in a range of 1 x 10 -6 - 1 x 10 -3 Pa could be measured with no significant impairment of the SEM performance with the aid of photon irradiation. It is revealed that a particle flux as low as 1 particle/s will be able to measure in the presence of tritium if suitable photon sources are installed in the systems. (orig.)

Treatment of tritiated exhaust gases at the Tritium Laboratory Karlsruhe

Energy Technology Data Exchange (ETDEWEB)

Hutter, E.; Besserer, U. [Kernforschungszentrum Karlsruhe GmbH (Germany); Jacqmin, G. [NUKEM GmbH, Industreistr, Alzenau (Germany)

1995-02-01

The Tritium Laboratory Karlsruhe (TLK) accomplished commissioning; tritium involving activities will start this year. The laboratory is destined mainly to investigating processing of fusion reactor fuel and to developing analytic devices for determination of tritium and tritiated species in view of control and accountancy requirements. The area for experimental work in the laboratory is about 800 m{sup 2}. The tritium infrastructure including systems for tritium storage, transfer within the laboratory and processing by cleanup and isotope separation methods has been installed on an additional 400 m{sup 2} area. All tritium processing systems (=primary systems), either of the tritium infrastructure or of the experiments, are enclosed in secondary containments which consist of gloveboxes, each of them connected to the central depressurization system, a part integrated in the central detritiation system. The atmosphere of each glovebox is cleaned in a closed cycle by local detritiation units controlled by two tritium monitors. Additionally, the TLK is equipped with a central detritiation system in which all gases discharged from the primary systems and the secondary systems are processed. All detritiation units consist of a catalyst for oxidizing gaseous tritium or tritiated hydrocarbons to water, a heat exchanger for cooling the catalyst reactor exhaust gas to room temperature, and a molecular sieve bed for adsorbing the water. Experiments with tracer amounts of tritium have shown that decontamination factors >3000 can be achieved with the TLK detritiation units. The central detritiation system was carefully tested and adjusted under normal and abnormal operation conditions. Test results and the behavior of the tritium barrier preventing tritiated exhaust gases from escaping into the atmosphere will be reported.

Software design for the Tritium System Test Assembly

International Nuclear Information System (INIS)

Claborn, G.W.; Heaphy, R.T.; Lewis, P.S.; Mann, L.W.; Nielson, C.W.

1983-01-01

The control system for the Tritium Systems Test Assembly (TSTA) must execute complicated algorithms for the control of several sophisticated subsystems. It must implement this control with requirements for easy modifiability, for high availability, and provide stringent protection for personnel and the environment. Software techniques used to deal with these requirements are described, including modularization based on the structure of the physical systems, a two-level hierarchy of concurrency, a dynamically modifiable man-machine interface, and a specification and documentation language based on a computerized form of structured flowcharts

Software design for the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Claborn, G.W.; Keaphy, R.T.

1983-01-01

The control system for the Tritium Systems Test Assembly (TSTA) must execute complicated algorithms for the control of several sophisticated subsystems. It must implement this control with requirements for easy modifiability, for high availability, and provide stringent protection for personnel and the environment. Software techniques used to deal with these requirements are described, including modularization based on the structure of the physical systems, a two-level hierarchy of concurrency, a dynamically modifiable manmachine interface, and a specification and documentation language based on a computerized form of structured flowcharts

Optimization of tritium management within the ITER project

International Nuclear Information System (INIS)

Cortes, P.; Elbez-Uzan, J.; Glugla, M.; Rosanvallon, S.; Ciattaglia, S.; Iseli, M.; Rodriguez-Rodrigo, L.

2009-01-01

The authors describe the tritium cycle existing within the ITER project and which has been considered since its beginning. They indicate how confinement systems ensure tritium confinement, how tritium is recovered and processed. They indicate the different tritium management optimization ways which have been identified and integrated into the ITER design

Tritium storage

International Nuclear Information System (INIS)

Hircq, B.

1990-01-01

This document represents a synthesis relative to tritium storage. After indicating the main storage particularities as regards tritium, storages under gaseous and solid form are after examined before establishing choices as a function of the main criteria. Finally, tritium storage is discussed regarding tritium devices associated to Fusion Reactors and regarding smaller devices [fr

Advancement Of Tritium Powered Betavoltaic Battery Systems FY16 EOY Report

Energy Technology Data Exchange (ETDEWEB)

Staack, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Gaillard, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hitchcock, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-10-12

The goal of this work is to increase the power output of tritium-powered betavoltaic batteries and investigate the change in power output and film resistance in real-time during tritium loading of adsorbent films. To this end, several tritium-compatible test vessels with the capability of measuring both the resistivity of a tritium trapping film and the power output of a betavoltaic device in-situ have been designed and fabricated using four electrically insulated feedthroughs in tritium-compatible load cells. Energy conversion devices were received from Widetronix, a betavoltaic manufacturing firm based in Ithaca, NY. Thin films were deposited on the devices and capped with palladium to facilitate hydrogen loading. Gold contacts were then deposited on top of the films to allow resistivity measurements of the film during hydrogen loading. Finally, the chips were wire bonded and installed in the test cells. The cells were then baked-out under vacuum and leak checked at temperature to reduce the chances of tritium leaks during loading. Following the bake-out, IV curves were measured to verify no internal wires were compromised, and the cells were delivered to Tritium for loading. Tritium loading is anticipated in October, 2017.

Use of system code to estimate equilibrium tritium inventory in fusion DT machines, such as ARIES-AT and components testing facilities

Energy Technology Data Exchange (ETDEWEB)

Wong, C.P.C., E-mail: wongc@fusion.gat.com [General Atomics, San Diego, CA (United States); Merrill, B. [Idaho National Laboratory, Idaho Falls, ID (United States)

2014-10-15

Highlights: • With the use of a system code, tritium burn-up fraction (f{sub burn}) can be determined. • Initial tritium inventory for steady state DT machines can be estimated. • f{sub burn} of ARIES-AT, CFETR and FNSF-AT are in the range of 1–2.8%. • Respective total tritium inventories of are 7.6 kg, 6.1 kg, and 5.2 kg. - Abstract: ITER is under construction and will begin operation in 2020. This is the first 500 MW{sub fusion} class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a system code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively.

Tritium proof-of-principle pellet injector

International Nuclear Information System (INIS)

Fisher, P.W.

1991-07-01

The tritium proof-of-principle (TPOP) experiment was designed and built by Oak Ridge National Laboratory (ORNL) to demonstrate the formation and acceleration of the world's first tritium pellets for fueling of future fusion reactors. The experiment was first used to produce hydrogen and deuterium pellets at ORNL. It was then moved to the Tritium Systems Test Assembly at Los Alamos National Laboratory for the production of tritium pellets. The injector used in situ condensation to produce cylindrical pellets in a 1-m-long, 4-mm-ID barrel. A cryogenic 3 He separator, which was an integral part of the gun assembly, was capable of lowering 3 He levels in the feed gas to <0.005%. The experiment was housed to a glovebox for tritium containment. Nearly 1500 pellets were produced during the course of the experiment, and about a third of these were pure tritium or mixtures of deuterium and tritium. Over 100 kCi of tritium was processed through the experiment without incident. Tritium pellet velocities of 1400 m/s were achieved with high-pressure hydrogen propellant. The design, operation, and results of this experiment are summarized. 34 refs., 44 figs., 3 tabs

Conceptual design of tritium treatment facility

International Nuclear Information System (INIS)

Tachikawa, Katsuhiro

1982-01-01

In connection with the development of fusion reactors, the development of techniques concerning tritium fuel cycle, such as the refining and circulation of fuel, the recovery of tritium from blanket, waste treatment and safe handling, is necessary. In Japan Atomic Energy Research Institute, the design of the tritium process research laboratory has been performed since fiscal 1977, in which the following research is carried out: 1) development of hydrogen isotope separation techniques by deep cooling distillation method and thermal diffusion method, 2) development of the refining, collection and storage techniques for tritium using metallic getters and palladium-silver alloy films, and 3) development of the safe handling techniques for tritium. The design features of this facility are explained, and the design standard for radiation protection is shown. At present, in the detailed design stage, the containment of tritium and safety analysis are studied. The building is of reinforced concrete, and the size is 48 m x 26 m. Glove boxes and various tritium-removing facilities are installed in two operation rooms. Multiple wall containment system and tritium-removing facilities are explained. (Kako, I.)

Tips for the fabrication of temporary tritium experiments

International Nuclear Information System (INIS)

Binning, K.E.; Jenkins, E.M.

1988-01-01

The Tritium System Test Assembly (TSTA) is a facility built for the demonstration of tritium handling systems necessary for tritium-burning fusion reactors. The facility has been in operation handling tritium for four years. The current inventory of tritium is approximately one hundred grams, with DOE approval for a maximum inventory of two hundred grams. Not all experiments performed at TSTA require the operation of the main process loop. During the last four years, many small scale experiments have been performed to test the compatibility and operation of tritium processing components in small self-contained experimental packages. These packages are fabricated inside secondary containment gloveboxes and can be operated for hours or months with little monitoring. Construction of these packages need to be tritium compatible, inexpensive, easy to build, and versatile. This paper discusses some of the problems and remedies encountered during the building of temporary experiments

Tips for the fabrication of temporary tritium experiments

International Nuclear Information System (INIS)

Binning, K.E.; Jenkins, E.M.

1988-01-01

The Tritium System Test Assembly (TSTA) is a facility built for the demonstration of tritium handling systems necessary for tritium-burning fusion reactors. The facility has been in operation handling tritium for four years. The current inventory of tritium is approximately one hundred grams with DOE approval exists for a maximum inventory of two hundred grams. Not all experiments performed at TSTA require the operation of the main process loop. During the last four years, many small scale experiments have been performed to test the compatibility and operation of tritium processing components in small self-contained experimental packages. These packages are fabricated inside secondary containment gloveboxes and can be operated for hours or months with little monitoring. Construction of these packages need to be tritium compatible, inexpensive, easy to build, and versatile. This paper discusses some of the problems and remedies encountered during the building of temporary experiments

Study on tritium recovery from breeder materials

International Nuclear Information System (INIS)

Moriyama, H.; Moritani, K.

1997-01-01

For the development of fusion reactor blanket systems, some of the key issues on the tritium recovery performance of solid and liquid breeder materials were studied. In the case of solid breeder materials, a special attention was focussed on the effects of irradiation on the tritium recovery performance, and tritium release experiments, luminescence measurements of irradiation defects and modeling studies were systematically performed. For liquid breeder materials, tritium recovery experiments from molten salt and liquid lithium were performed, and the technical feasibility of tritium recovery methods was discussed. (author)

Tritium waste control: April--September 1977

International Nuclear Information System (INIS)

1978-01-01

A pilot scale system was used in an initial experiment to investigate the combined-electrolysis-catalytic-exchange process (CECE) for the detritiation of water. Data taken during the experiment indicate the process does indeed strip tritium from gaseous hydrogen at the top and concentrate it in water at the bottom of the catalyst-filled column. A high activity tritiated water electrolysis system was designed and built using a solid polymer electrolyte (SPE) cell. The system was successfully operated at currents up to 50 A using deionized tap water. Triplicate samples of cement, cement-plaster (1:1 ratio by weight), and cement-plaster (1:1 ratio by volume) were injected with 386 Ci of tritium. Preliminary results indicate Type III Portland cement retains the tritium slightly better than the cement-plaster mixtures. The tritium release study of actual waste burial packages is continuing. The fractional release is 1 x 10 -5 on a 4-y old package, only 4 x 10 -7 on 3-y old packages, and 1 x 10 -9 on a 1-y old package. Pressure increase and gas composition determinations were repeated for octane (activity = 1000 Ci/liter) with and without tritium fixation using argon as the initial overgas. Pressure buildup measurements for octane without fixation were repeated a third time using hydrogen gas. The rate of pressure increase did not change significantly from previously determined values. Four elemental tritium samples were released into a laboratory to determine the efficiency of the Emergency Containment System. The ventilation system was modified during the fourth experiment to minimize leakage

Tritium processing in JT-60U

International Nuclear Information System (INIS)

Miya, Naoyuki; Masaki, Kei

1997-01-01

Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 10 19 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm 3 for radiation work permit requirements. (author)

Simulation study of intentional tritium release experiments in the caisson assembly for tritium safety at the TPL/JAERI

International Nuclear Information System (INIS)

Iwai, Y.; Hayashi, T.; Kobayashi, K.; Nishi, M.

2001-01-01

At the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI), Caisson assembly for tritium safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate the tritium behavior in the case, where the tritium leak accident 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 accident should happen in a ventilated room. As for the understanding of initial tritium behavior until the tritium concentration become steady, the precise estimation of local flow rate in a room and time-dependent release behavior from the leak point are essential to predict the tritium behavior by simulation code. The three-dimensional eddy flow model considering, tritium-related phenomena was adopted to estimate the local flow rate in the 50 m 3 /h ventilated Caisson. The time-dependent tritium release behavior from the sample container was calculated by residence time distribution function. The calculated tritium concentrations were in good agreement with the experimental observations. The primary removal tritium behavior was also investigated by another code. Tritium gas concentrations decreased logarithmically to the time by ventilation. These observations were understandable by the reason that the flow in the ventilated Caisson was regarded as the perfectly mixing flow. The concentrations of tritiated water measured, and indications of tritium concentration by tritium monitors became gradually flat. This phenomena called 'tritium soaking effect' was found to be reasonably explained by considering the contribution of the exhaustion velocity by ventilation system, and the adsorption and desorption reaction rate of tritiated water on the wall material which is SUS 304. The calculated tritium concentrations were in good agreement with the experimental observations

Radiolysis aspects of the aqueous self-cooled blanket concept and the problem of tritium extraction

International Nuclear Information System (INIS)

Bruggeman, A.; Snykers, M.; DeRegge, P.; Embrechts, M.J.

1988-01-01

In the Aqueous Self-Cooled Blanket (ASCB) concept, an aqueous 6 Li solution in a metallic structure is used as a fusion reactor shielding-breeding blanket. Radiolysis effects could be very important for the design and the use of an ASCB. Although many aspects of the radiation chemistry of water and dilute aqueous solutions are now reasonably well understood, it is not possible to predict the radiochemical behaviour of the concentrated candidate ASCB solutions quantitatively. However, by means of a worst case calculation for a possible ASCB for the Next European Torus (NET) it is shown that even with an important rate of water decomposition the ASCB concept is still workable. Gas bubbles and explosive mixtures can be avoided by increasing the pressure in the neutron irradiated zone and by extracting and/or recombining the radiolytically produced hydrogen and oxygen. This could require an additional inert gas loop, which could also be used as part of the tritium extraction installation

Turkey Point tritium. Progress report

International Nuclear Information System (INIS)

Ostlund, H.G.; Dorsey, H.G.

1976-01-01

In 1972-73 the Florida Power and Light Company (FPL) began operation of two nuclear reactors at Turkey Point on lower Biscayne Bay. One radioactive by-product resulting from the operation of the nuclear reactors, tritium, provides a unique opportunity to study transport and exchange processes on a local scale. Since the isotope in the form of water is not removed from the liquid effluent, it is discharged to the cooling canal system. By studying its residence time in the canal and the pathways by which it leaves the canals, knowledge of evaporative process, groundwater movement, and bay exchange with the ocean can be obtained. Preliminary results obtained from measurement of tritium levels, both in the canal system and in the surrounding environment are discussed. Waters in lower Biscayne Bay and Card and Barnes Sounds receive only a small portion of the total tritium produced by the nuclear plant. The dominating tritium loss most likely is through evaporation from the canals. The capability of measuring extremely low HTO levels allows the determination of the evaporation rate experimentally by measuring the tritium levels of air after having passed over the canals

Tritium contamination and monitoring at Frascati Neutron Generator

Energy Technology Data Exchange (ETDEWEB)

Lucci, F.; Sandri, S.; Ianni, A. [ENEA, Frascati (Italy). Dipartimento Ambiente; Vasselli, R. [ANPA, Roma (Italy); Pillon, M.; Bettinali, L. [ENEA, Frascati (Italy). Dipartimento Energia

1994-11-01

The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10{sup 1}1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed.

Tritium contamination and monitoring at Frascati Neutron Generator

International Nuclear Information System (INIS)

Lucci, F.; Sandri, S.; Ianni, A.; Pillon, M.; Bettinali, L.

1994-11-01

The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10 1 1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed

Tritium conference days

International Nuclear Information System (INIS)

Garnier-Laplace, J.; Lebaron-Jacobs, L.; Sene, M.; Devin, P.; Chretien, V.; Le Guen, B.; Guetat, Ph.; Baglan, N.; Ansoborlo, E.; Boyer, C.; Masson, M.; Bailly-Du-Bois, P.; Jenkinson, St.; Wakeford, R.; Saintigny, Y.; Romeo, P.H.; Thompson, P.; Leterq, D.; Chastagner, F.; Cortes, P.; Philippe, M.; Paquet, F.; Fournier, M.

2009-01-01

This document gathers the slides of the available presentations given during this conference day. Twenty presentations out of 21 are assembled in the document and deal with: 1 - tritium in the environment (J. Garnier-Laplace); 2 - status of knowledge about tritium impact on health (L. Lebaron-Jacobs); 3 - tritium, discrete but present everywhere (M. Sene); 4 - management of tritium effluents from Areva NC La Hague site - related impact and monitoring (P. Devin); 5 - tritium effluents and impact in the vicinity of EDF's power plants (V. Chretien and B. Le Guen); 6 - contribution of CEA-Valduc centre monitoring to the knowledge of atmospheric tritiated water transfers to the different compartments of the environment (P. Guetat); 7 - tritium analysis in environment samples: constraints and means (N. Baglan); 8 - organically-linked tritium: the analyst view (E. Ansoborlo); 9 - study of tritium transfers to plants via OBT/HTO air and OBT/HTO free (C. Boyer); 10 - tritium in the British Channel (M. Masson and P. Bailly-Du-Bois); 11 - tritium in British coastal waters (S. Jenkinson); 12 - recent results from epidemiology (R. Wakeford); 13 - effects of tritiated thymidine on hematopoietic stem cells (P.H. Romeo); 14 - tritium management issue in Canada: the point of view from authorities (P. Thompson); 15 - experience feedback of the detritiation process of Valduc centre (D. Leterq); 16 - difficulties linked with tritiated wastes confinement (F. Chastagner); 17 - optimisation of tritium management in the ITER project (P. Cortes); 18 - elements of thought about the management of tritium generated by nuclear facilities (M. Philippe); 19 - CIPR's position about the calculation of doses and risks linked with tritium exposure (F. Paquet); 20 - tritium think tanks (M. Fournier). (J.S.)

Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

International Nuclear Information System (INIS)

Talbot, J.B.

1976-03-01

The sorption of tritium by molten lithium--bismuth (Li--Bi, approx. 15 at. % lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility ( 0 C, the extraction process is not attractive

A compact, low cost, tritium removal plant for CANDU-6 reactors

International Nuclear Information System (INIS)

Sood, S.K.; Fong, C.; Kalyanam; Woodall, K.B.

1997-01-01

Tritium concentrations in CANDU-6 reactors are currently around 40 Ci/kg in moderator systems and around 1.5 Ci/kg in primary heat transport (PHT) systems. It is expected that tritium concentrations in moderator systems will continue to rise and will reach about 80 Ci/kg at maturity. A more detailed description of the increase in tritium concentrations in the moderator and PHT systems of CANDU-6 reactors is given in the next section of this paper. While moderator systems currently contribute more than 50% to tritium emissions, the impact of acute releases of moderator water is more severe at higher tritium concentrations. This impact can be substantially reduced by the addition of an isotope separation system for lowering the tritium level in the moderator system. In addition, lower tritium levels in CANDU systems will inevitably result in reduced occupational exposures, or will provide economic benefits due to ease of maintenance because less protective measures are required and maintenance activities can be more efficient

Simulation of tritium behavior after intended tritium release in ventilated room

International Nuclear Information System (INIS)

Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko; Kobayashi, Kazuhiro; Nishi, Masataka

2001-01-01

At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m 3 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 50 m 3 /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 m 3 /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 released in the 3,000 m 3 of tritium handling room was investigated experimentally under a US-Japan collaboration. The tritium concentration history calculated with the same method was consistent with the experimental observations, which proves that the present developed method can be applied to the actual scale of tritium handling room. (author)

Estimation of Biological Effects of Tritium.

Science.gov (United States)

Umata, Toshiyuki

2017-01-01

Nuclear fusion technology is expected to create new energy in the future. However, nuclear fusion requires a large amount of tritium as a fuel, leading to concern about the exposure of radiation workers to tritium beta radiation. Furthermore, countermeasures for tritium-polluted water produced in decommissioning of the reactor at Fukushima Daiichi Nuclear Power Station may potentially cause health problems in radiation workers. Although, internal exposure to tritium at a low dose/low dose rate can be assumed, biological effect of tritium exposure is not negligible, because tritiated water (HTO) intake to the body via the mouth/inhalation/skin would lead to homogeneous distribution throughout the whole body. Furthermore, organically-bound tritium (OBT) stays in the body as parts of the molecules that comprise living organisms resulting in long-term exposure, and the chemical form of tritium should be considered. To evaluate the biological effect of tritium, the effect should be compared with that of other radiation types. Many studies have examined the relative biological effectiveness (RBE) of tritium. Hence, we report the RBE, which was obtained with radiation carcinogenesis classified as a stochastic effect, and serves as a reference for cancer risk. We also introduce the outline of the tritium experiment and the principle of a recently developed animal experimental system using transgenic mouse to detect the biological influence of radiation exposure at a low dose/low dose rate.

Tritium solubility and permeation in high retention fusion reactor breeder elements

International Nuclear Information System (INIS)

Jakeman, D.

1979-11-01

As an alternative to the current philosophy of reducing the tritium inventory to a minimum by continuously extracting tritium from the breeder of a fusion reactor, an alternative design philosophy is examined in which tritium is contained within high retention breeder elements which can remain in the reactor for a substantial time. To prevent tritium diffusion through the clad of the element it is necessary to maintain a low tritium pressure within the element. Pressures of between 10 4 Pa and 1 Pa appear possible with an element containing a high solubility material provided it is kept below about 400 0 C. This should lead to a leakage into the coolant of between 10 Ci day -1 and 10 4 Ci day -1 which is considerably less than the 10 7 Ci day -1 in present designs. (author)

Internal dose from tritium at Wolsung nuclear power plant

International Nuclear Information System (INIS)

Hee Geun Kim; Jeong Yull Dho; Myung Jae Song

1995-01-01

Tritium is produced in large quantities at heavy water nuclear power reactors via the neutron activation reaction 2 H(n,γ) 3 H. At Wolsung nuclear power plant which has a CANDU reactor, the tritium concentrations in coolant and in moderator systems are 1.5 Ci/Kg-D 2 O and 35 Ci/kg-D 2 O, respectively, after 12 years of operation. The airborne tritium concentration in main access area is normally less than 5 MPCa except short-term peaks. The average tritium concentrations in main access controlled areas are normally less than 100 MPCa. Tritium is mainly present in the air of workplace of CANDU reactors as a tritiated water vapour. Airborne tritiated water vapour enters the workers body via inhalation and absorption through skin and can result in a significant dose. The occupational doses from tritium at Wolsung NPP have been maintained below 1 man-Sv per year so far. The tritium contribution to the total plant man-Sv changes between 30 percent and 50 percent. For the mitigation of tritium inhalation, various protective equipment are being used at Wolsung NPP. The respirator system was devised at Wolsung NPP in order to remove tritiated water vapours from the inhaled air. A respirator is connected to a small plastic bottle filled with ice cubes. The system devised shows a good tritium removal efficiency. The air pressure drop through the ice cubes is minimal. The operation cost of the system is also very cheap. Further mitigation of tritium inhalation is heavily dependant on the source term reduction. One of the ultimate solutions is to introduce a tritium removal facility. (author). 7 figs., 3 tabs

Simulation and optimisation of the data acquisition system for tritium removal pilot plant

International Nuclear Information System (INIS)

Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu

2004-01-01

Optimization and simulation of systems especially in science and engineering can help to reduce risk and cost of design and testing processes. A huge number of codes has been developed to support modeling and simulation efforts. All of these software tools support the use of one or more mathematical model classes. Despite all of these efforts, it is hard to find simulation software, which is capable of combining several model classes in a real industry standard environment. The paper presents a simulation software product for controlling and data acquisition system of cryogenic installation process in the tritium removal pilot plant, using an industry standard programming environment widely applied to data acquisition, process control and data visualization, namely LabView. One of the problems in a tritium separation installation is controlling the temperature. To solve this problem it is necessary to develop a simulation system which includes the mathematical model for cryogenic distillation. Also with this simulation system we can approach the safety system which ensures the monitoring of radiations and toxic gases from installation. All elements used in controlling, modeling and simulation of the process, as well as, in the datalogging and supervisory control module from tritium removal installation are new. (authors)

FDMH - The tritium model in RODOS

International Nuclear Information System (INIS)

Galeriu, D.; Mateescu, G.; Melintescu, A.; Turcanu, C.; Raskob, W.

2000-01-01

Under the auspices of its RTD (Research and Technological Development) Framework Programmes, the European Commission has supported the development of the RODOS (Real-time On-line DecisiOn Support) system for off-site emergency management. The project started in 1989 focusing on PWR/LWR type accidents and using experience from the Chernobyl accident. In 1996 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exist in Europe and to allow a potential expansion of the RODOS system for application on future fusion reactor accidents. The National Institute for Physics and Nuclear Engineering (IFIN-HH) in Romania - in close co-operation with the Research Centre Karlsruhe (FZK) - was charged to develop the tritium module, based on previous experience in environmental tritium modelling and the operation of CANDU reactor-based NPP in Romania (with potential tritium accidents). Tritium, being an isotope of hydrogen, is incorporated immediately in the life cycle and its transport into the biosphere differs considerably from other radionuclides treated by the RODOS system. Concentrations in the individual compartments may change very rapidly (hours) under varying environmental conditions and conversion to organic forms by biochemical and metabolic processes takes place in plants and animals. Consequently, the tritium code in RODOS was developed as a separate module and harmonisation in data sets and interfaces with other food chain modules integrated in RODOS was ensured. Presently, the tritium module - FDMH- is integrated and documented in the RODOS system, delivering time dependent tritium concentration (as tritiated water or organically bound tritium) in plant and animal products, inhalation dose and ingestion dose for various groups of population, after an accident emitting tritiated water and for up to 2520 locations around the source. FDMH incorporates many improved techniques in radiological assessment and makes

Gas-handling system for studies of tritium-containing materials

International Nuclear Information System (INIS)

Carstens, D.H.W.

1975-01-01

A gas handling system for preparation and study of tritium containing compounds and materials is described. The system at any one time can handle amounts of DT gas up to about 3 moles and has provisions for purification, storage, and measurement of the gas. Experimental conditions covering the ranges 20 to 800 0 C and 0.1 Pa to 137 MPa (10 -2 torr to 20,000 psi) can be maintained. (auth)

System study application to the safety analysis of the exhaust and the tritium systems of a fusion reactor

International Nuclear Information System (INIS)

Djerassi, H.; Rouillard, J.; Leger, D.; Zappellini, G.; Gambi, G.

1988-01-01

An applicative example of the general methodology system study to the safety analysis of the exhaust and tritium systems, in a tokamak device, is shown. The framework of the study is split into the following tasks: initial information collection, functional analysis, failure scenarios identification and description, reliability data assessment, accident sequence quantification, consequence seriousness evaluation, risk assessment. Results concerning risk contribution from direct failures show that, in the exhaust system and in the tritium system, the risk contribution to public is rather smaller than the safety design targets. Nevertheless, if the reactor building is not taken into account, the risk contribution from the exhaust system can be significant. Risk contribution to the workers seems to be not to heavy

Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.

1994-04-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet trademark system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of ∼ 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed

Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Jaeger, E.F.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Laughlin, M.J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D.E.; Marchlik, M.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M.P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Milora, S.L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Paul, S.F.; Pavlov, Y.I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Pitcher, S.; Popovichev, S.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.

1994-01-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert TM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of ∼10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed

Tritium contamination experience in an operational D-T fusion reactor

International Nuclear Information System (INIS)

Gentile, C.A.; Ascione, G.

1994-01-01

During December 1993, the Tokamak Fusion Test Reactor (TFTR) injected a mixture of deuterium and tritium in the TFTR vacuum vessel for the purpose of creating D-T plasmas. The tritium used in these D-T plasmas was stored, delivered and processed in the TFTR tritium facility that includes the tritium vault, waste handling area, clean-up area, and gas holding tank room. During this time period, several components in the tritium process system were found to have tritium leaks which led to tritium deposition on process skids, components and floor area. Radiological surveys of surfaces contaminated with tritium oxide indicate a decrease in surface contamination in time (on the order of 12 to 36 hours) as the result of room ventilation. In instances where the facility HVAC system was maintained in the purge mode, a dramatic decrease in surface contamination was observed. Areas contaminated with tritium oxide (> 16.6 Bq/100 cm 2 ) were found to be clean ( 2 ) after several hours of continuous purging by the facility HVAC system. In instances where relative humidity was not decreased, the tritium surface contamination was found to be attenuated. During the months of December 1993, January and February 1994 tritium leaking components were either replaced, redesigned or repaired. During this time period, data were collected in the form of contamination surveys, real time tritium monitor output, and HVAC configuration indicating the correlation of purge ventilation leading to a decrease in tritium oxide surface contamination

Tritium breeding and release-rate kinetics from neutron-irradiated lithium oxide

International Nuclear Information System (INIS)

Quanci, J.F.

1989-01-01

The research encompasses the measurement of the tritium breeding and release-rate kinetics from lithium oxide, a ceramic tritium-breeding material. A thermal extraction apparatus which allows the accurate measurement of the total tritium inventory and release rate from lithium oxide samples under different temperatures, pressures and carrier-gas compositions with an uncertainty not exceeding 3% was developed. The goal of the Lithium Blanket Module program was to determine if advanced computer codes could accurately predict the tritium production in the lithium oxide blanket of a fusion power plant. A fusion blanket module prototype was built and irradiated with a deuterium-tritium fusion-neutron source. The tritium production throughout the module was modeled with the MCNP three dimensional Monte Carlo code and was compared to the assay of the tritium bred in the module. The MCNP code accurately predicted tritium-breeding trends but underestimated the overall tritium breeding by 30%. The release rate of tritium from small grain polycrystalline sintered lithium oxides with a helium carrier gas from 300 to 450 C was found to be controlled by the first order surface desorption of monotritiated water. When small amounts of hydrogen were added to the helium carrier gas, the first order rate constant increased from the isotopic exchange of hydrogen for tritium at the lithium oxide surface occurring in parallel with the first order desorption process. The isotopic-exchange first order rate constant temperature dependence and hydrogen partial pressure dependence were evaluated

Diffusion and autoradiographic investigations of the tritium--304 stainless steel system

International Nuclear Information System (INIS)

Downs, G.L.; Braun, J.D.; Chaney, K.F.; Powell, G.W.

1975-01-01

The diffusion coefficient of tritium in 304-stainless steel at low temperatures (100 to 300 0 C) was determined. Autoradiography was used to establish the concentration as well as the distribution of tritium in the alloy. The autoradiographic study shows that tritium is distributed heterogeneously at room temperature in the cold-worked alloy and also in the fusion zone of weldments. Tritium partitions preferentially to the delta ferrite in weldments and to martensite produced by the cold working of 304-stainless steel. (auth)

Tritium Management In HCLL-PPCS Model AB Blanket

International Nuclear Information System (INIS)

Ricapito, I.; Aiello, A.; Benamati, G.; Utili, M.; Ciampichetti, A.; Zucchetti, M.

2006-01-01

One the main issues in the HCLL blanket development for a prototype fusion reactor is the technical feasibility of the bred tritium processing system. The basis of such concern lies in the very low tritium-Pb17Li Sieverts' constant, as measured by different scientists in the past years. In the PPCS reactor 650 g/d of tritium must be generated in the breeding blanket while less than 1 g/y of tritium has to be released to the environment through the secondary cooling circuit. As a consequence, CPS (Coolant Purification System) plays a fundamental role because it has to keep at an acceptable level the tritium partial pressure in the primary HCS (Helium Cooling Circuit) limiting, therefore, the tritium environmental release through leakage and permeation into the secondary cooling circuit. On the other hand, the He mass flow-rate to be processed by CPS is linear with the tritium permeation rate from the breeder into HCS. Therefore, with the above mentioned low Sieverts' constant values and the consequent high tritium partial pressure in the liquid metal, the possibility to keep acceptable the CPS capacity depends on a highly efficient and stable performance of tritium permeation barriers, to be applied not only on the blanket cooling plates but also on the steam generator walls. However, the experimental results on the tritium permeation barriers under relevant operative conditions were so far quite disappointing. The new data on the Sieverts' constant achieved at ENEA CR Brasimone, one order of magnitude higher than those founding the past, have a big impact in relaxing the above mentioned requirements for the tritium management in PPCS model AB reactor. Besides presenting and discussing these recent experimental results, an updated assessment of the tritium permeation rate from the liquid breeder into HCS through the cooling plates and from HCS into the environment through the steam generators is given in this paper. The consequent new constraints in terms of tritium

Experiences with decontaminating tritium-handling apparatus

International Nuclear Information System (INIS)

Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

1991-07-01

Tritium-handling apparatus has been decontaminated as part of the shutdown of the LLNL Tritium Facility. Two stainless-steel gloveboxes that had been used to process lithium deuteride-tritide (LiDT) salt were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. Further surface decontamination was performed by scrubbing the interior with paper towels and ethyl alcohol or Swish trademark. The surface contamination, as shown by swipe surveys, was reduced from 4x10 4 --10 6 disintegrations per minute (dpm)/cm 2 to 2x10 2 --4x10 4 dpm/cm 2 . Details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given

Tritium control: October 1982-March 1983

International Nuclear Information System (INIS)

Lamberger, P.H.; Rogers, M.L.

1983-01-01

Surveys in gloveboxes indicated surface activity on stainless steel and its apparent dependence on time and atmospheric tritium levels. Surveys in fumehoods were completed to investigate the extent of surface contamination on surfaces of various materials. Gas generation rates caused by radiolysis of tritiated waste materials were determined for polymer and nonpolymer-impregnated tritiated concrete and fixated and nonfixated tritiated waste vacuum pump oil. In addition, the pressure change of hydrogen cover gas over tritiated water on cement-plaster was determined. The test program to measure and compare the release of tritium from tritiated concrete with and without styrene impregnation continued. Tritium permeation data from small test blocks are given. The drum study monitoring the release of tritium from actual burial packages continued. The maximum fractional release rate for the three types of high activity, tritiated liquid waste generated is 5.1 x 10 -5 , and the maximum total permeation is 179 mCi after 8.5 yr. These two values represent a 13% increase for the past 6 months. Tritium release from the polymer-impregnated, tritiated concrete (PITC) and from the control (non-PITC) remains very low. The Emergency Containment System (ECS), an automatically actuated system developed at Mound to remove tritium from room air, has been modified and upgraded to support new applications. The leakage rate in the ECS area has been lowered, a fast-start system installed for greater conversion efficiency at startup, and the alumina beds regenerated

A Low-Level Real-Time In Situ Monitoring System for Tritium in Groundwater and Vadose Zone

Science.gov (United States)

Santo, J. T.; Levitt, D. G.