Residual Tritium Research Articles

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100–600°C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300–600°C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150°C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.

Evaluation of residual tritium in stainless steel irradiated at SINQ target 3

A MW spallation target will need an accurate amount of tritium generation in materials for post irradiation experiments. A high-energy proton of 580 MeV generates a variety of constituents in the SINQ target mainly through spallation reaction, and these constituents included helium and hydrogen isotopes. For the SINQ target hydrogen and helium are generated to hundreds appm H and tens appm He, respectively. However, tritium gas production has not yet been reported. Evaluation of tritium gas generation in the specimens of target 3 was done by calculation. Results were compared with gamma spectrum measurement of austenitic and martensitic–ferritic steels three and a half years later after irradiation. Good agreement was obtained for activity of the main isotope. Residual tritium measurement was carried for the austenitic stainless steel specimens irradiated below 250 °C by the thermal desorption method four and a half years later after irradiation. An ion chamber monitored the tritium concentration. Water bubblers collected all tritium after oxidation of tritium by the oxidation catalysts. It is found that a release of tritium began over 250 °C and the ratio of residual tritium to generated one is estimated to be less than 20%.

Tritium release from neutron-irradiated Li 2O: Transport in porous sintered pellets

The tritium release behavior from Li 2O sintered pellets (81–88% T.D.) is examined by isothermal heating tests. (1) For the 88% T.D. specimens, the fraction of residual tritium is found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (2) For the 81% T.D. specimens, which are annealed after irradiation at 630 K for 4 h, the fraction of residual tritium is also found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (3) For the 81% T.D. specimens as irradiated, the tritium release rate is found to follow the square-root law of the annealing time. The rate-determining process is controlled by Kohlrauch stretched exponential form. Tritium trapped in irradiation defects released with recovering the defects by isothermal heating.

Influence of heat treatments on residual tritium amount in tritiated stainless steel waste

In the framework of waste management strategy for future fusion reactor, it has been demonstrated that one of the major issue is to reduce the high level waste. The tritium activity strongly contributes to the waste classification. A better knowledge of the mechanisms involved in tritium trapping and desorption will allow to choose the most appropriate procedure to reduce the tritium constraint. This paper deals with the management of tritiated steels and the way to reduce tritium transfer. This experimental work concerns steel obtained by remelting under vacuum of tritiated austenitic stainless steel wastes. The goal of this study is to develop heat treatments in solid phase to lower the residual tritium concentration and its desorption kinetics at room temperature. The surface activity, the desorption of residual tritium at room temperature and the residual concentration have been measured for each condition (temperature, time, atmosphere, etc.). For reheating temperatures below 600 °C, the surface activity increases by one or two orders of magnitude; this is the consequence of the formation of a tritium rich oxide layer. After a 20 h annealing above 300 °C and removing of the tritium rich oxide layer, the surface activity is strongly reduced (80%) when compared with the value measured before annealing. For such annealing conditions, the amount of desorbed tritium is reduced by a factor 25 and the desorption kinetics yield a reduction factor of 40–50 after annealing at 400 °C for 20 h.

Simplified Estimation of Tritium Inventory in Stainless Steel

An important part of tritium facility waste management is estimating the residual tritium inventory in stainless steel. This was needed as part of the decontamination and decommissioning associated with the Tritium Systems Test Assembly at Los Alamos National Laboratory. In particular, the disposal path for three, large tanks would vary substantially depending on the tritium inventory in the stainless steel walls. For this purpose the time-dependant diffusion equation was solved using previously measured parameters. These results were compared to previous work that measured the tritium inventory in the stainless steel wall of a 50-L tritium container. Good agreement was observed. These results are reduced to a simple algebraic equation that can readily be used to estimate tritium inventories in room temperature stainless steel based on tritium partial pressure and exposure time. Results are available for both constant partial pressure exposures and for varying partial pressures. Movies of the time dependant results were prepared which are particularly helpful for interpreting results and drawing conclusions.

New Systems for Waste Processing of Tritium-Containing Gases at the Savannah River Site

A project to relocate and consolidate tritium processing activities from old, second generation buildings to newer buildings was initiated in the late 1990’s at the Savannah River Site. The new waste gas processing systems located in the newer facility utilize recent technology, including metal getters, an innovative permeator design, and TCAP (Thermal Cycling Absorption Process) technology for removal of residual tritium prior to releasing the effluent to the environment. Startup testing results (using protium and deuterium) and corresponding lessons learned for these systems are presented. These systems have since successfully completed tritium startup testing and are operational.

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20–40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84–30 kBq/cm 2. The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm 2. It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.

No evidence for a forced-unfolding mechanism during ATP/GroES binding to substrate-bound GroEL: no observable protection of metastable Rubisco intermediate or GroEL-bound Rubisco from tritium exchange

In tritium–hydrogen exchange experiments, the large GroEL substrate Rubisco was unfolded and exchanged in urea/acid/tritiated water, then diluted into either protic buffer or protic buffer containing GroEL. The respective Rubisco metastable folding intermediate or Rubisco-GroEL binary complex was then separated from residual tritium after varying times of exchange by centrifugation through P-10 or G-25 resin. No significant tritium was recovered in either case, in contrast to an earlier report. Thus, although the earlier-proposed forced unfolding mechanism for the action of GroEL on a bound polypeptide, occurring during ATP/GroES binding, remains an attractive hypothesis, the data here do not provide any indication that it is involved in the folding of Rubisco.

An inventory of long-lived radionuclides residual from underground nuclear testing at the Nevada test site, 1951–1992

An inventory of long-lived radionuclides produced by 828 underground nuclear tests conducted at the Nevada test site (NTS) from 1951 to 1992 includes residual tritium, fission products, actinides, and activation products. Recently, the US Department of Energy approved the declassification of the NTS radionuclide inventory by principal geographic test centers. This permits unclassified publication of radionuclide totals for the Yucca Flat, Pahute Mesa—Area 19, Pahute Mesa—Area 20, Frenchman Flat, and Rainier Mesa/Shoshone Mountain testing locations. Activities are reported as of September 23, 1992, the date of the last underground nuclear test conducted at the NTS, and September 23, 2492, after 500 years of radioactive decay. The availability of these data affords an opportunity for the analysis of the radiologic source term within the boundaries of local hydrogeologic units and provides insight to where radionuclides are sited relative to potential exposure pathways.

Research into tritium impact on the models of cryosorption panels under conditions simulating operation of a vacuum pump of fusion rectors

The paper presents the results of research in the tritium impact on sorption and thermomechanical properties of model cryopanels (MCP) of the ITER vacuum pumps. For the first time amount of residual tritium in a sorption layer of cryopanels was determined experimentally under conditions simulating the operation of the cryopump (sorption of a tritium-containing mixture at 5 K, desorption—at 80 K).

Deformation-induced defects controlling fracture toughness of steel revealed by tritium desorption behaviors

Defects induced by plastic deformation have been revealed by means of room temperature desorption and thermal desorption spectroscopy of tritium with regards to the ductile crack growth resistance and brittle fracture initiation in steels in the ductile-to-brittle transition region. Tritium, as a probe for detecting defects, was introduced into non-deformed or deformed samples. Three steels, the microstructures of which are characterized by the constraint factor for slip extension, were employed. The residual tritium in a specimen after three days at room temperature increased to the extent corresponding to the constraint factors when plastic deformation was applied. The thermally desorbed tritium, with a peak desorption rate around 150°C, also increased according to the constraint factors. Referring to the previous analysis of the R-curves, the constraint for the extension of slip across grain boundaries is shown to control both the ductile crack growth resistance and the brittle fracture initiation through the deformation microstructures that evolve vacancy clusters or microvoids.

Health physics aspects of tritium operation at JET

JET has completed its first stage of active operations, processing more than 100 g of tritium during the deuterium–tritium experiment (DTE1), the neutral beam intervention and remote tile exchange shutdown. The radiological safety of tritium handling operations has been of particular interest. A wide range of tritium operations has been carried out in the period 1995–1998. This paper describes some of the radiological protection measures for work with tritium, and discusses Health Physics operational experience of handling tritium in this period. Descriptions are given of active operations in the gas handling plant; in the torus hall during DTE1; in related interventions, and of the remote exchange of in-vessel divertor modules. Workplace contamination levels over 100 DAC (HTO) have been encountered, tritiated water with activity of 2 TBq/l and tritiated carbon with activities of ∼4 TBq/g has been handled. Control measures involving the use of purge and extract ventilation, and of personal protection using air-fed pressurised suits are described. The project imposes tight limits on radiation exposures. Tritium doses to staff in this period have been very low (individual doses <170 μSv/year, collective doses ≤2.1 mSv/year). Aerial discharges have been <3% of annual authorised limits, and average environmental (HTO) levels have been a few Bq/m 3. Lessons have been learnt concerning exposure control, large-scale permeation effects, and the appearance of residual tritium on exposed surfaces. Tritium operations at JET have been conducted without incident and with very low personnel exposures. The methodology of using containment and ventilation systems and tight radiological control has been successful in limiting doses. JET experience shows that large-scale tritium handling and exposure control can be achieved within stringent dose limits.

Management strategy to reduce the radioactive waste amount in fusion

A waste management strategy for fusion activated materials aimed at reducing the amount of permanent nuclear waste is described. Three options are investigated: (1) clearance (declassification to nonactive waste). (2) recycling, (3) disposal as radioactive waste. Conditions allowing recycling are based on surface dose rates, those allowing clearance depend on the residual radioactivity, both after 50 years of interim storage. Analyses are carried out on two power reactor designs, the SEAFP Model 1 (SM-1) and Model 2 (SM-2). The activated waste arising from SM-2, is 69000 t: 48% could be recycled, 39% cleared, 13% needs permanent disposal. The SM-1 activated waste is 58000 t. No waste should be disposed of, 70% could be recycled, 30% could be cleared. The presence of residual tritium within the waste has been considered. An alternative to V-5Ti recycling has been analyzed, dealing with reprocessing of the alloy with elimination of noxious nuclides and subsequent clearance.

Tritium imaging of JET wall tile exposed to D-T shots

Tritium imaging was used to determine the lateral distribution of residual tritium in an X-point carbon-carbon composite tile taken from the JET tokamak after the Preliminary Tritium Experiment of November 9, 1991. Substantial variation in image brightness was observed in different parts of the tile. Features of the fine scale morphology that were visible by optical microscopy were also observed in the tritium images. Alternate sputtering and imaging of tile speciments indicates the presence of tritium at depths beyond the implantation range that would be expected under tokamak operating conditions, suggesting that tritium is being codeposited with carbon. The total tritium content of two specimens were assessed by destructive means. Analysis suggests that most of the tritium present in the tile specimens was codeposited during the post-PTE cleanup.

Efficient HTO reduction using a ZrFeMn alloy

Abstract Tritiated water production is a consequence of tritium gas handling. If tritium emission to the environment is to be minimized, then tritiated water must be processed so that the chemically bound tritium is recovered. For DT burning fusion machines, tritiated-water-processing systems are under development. These systems, however, are intended for large-scale applications and cannot be scaled down practically or economically. HTO vapour reduction to HT over hot metal getters offers a simple alternative. In such cases the metal getter selected should reduce HTO efficiently and retain negligible residual tritium inventory. HTO reduction over a ZrFeMn alloy is reported in this paper. Conversion efficiencies exceeding 90% are achievable for gas residence times above 0.7 s and alloy temperatures above 400 °C over the useful life of the alloy. HTO-to-HT conversion rates greater than 6 μmol min −1 per gram of alloy can be attained for oxygen-to-alloy mole ratios from 0 up to 2.6 without a significant reduction in the reaction rate.

Absolute calibration of TFTR helium proportional countersa)

The TFTR helium proportional counters are located in the central five (5) channels of the TFTR multichannel neutron collimator. These detectors were absolutely calibrated using a 14 MeV neutron generator positioned at the horizontal midplane of the TFTR vacuum vessel. The neutron generator position was scanned in centimeter steps to determine the collimator aperture width to 14 MeV neutrons and the absolute sensitivity of each channel. Neutron profiles were measured for TFTR plasmas with time resolution between 5 and 50 ms depending upon count rates. The He detectors were used to measure the burnup of 1 MeV tritons in deuterium plasmas, the transport of tritium in trace tritium experiments, and the residual tritium levels in plasmas following 50:50 DT experiments.

Absolute calibration of TFTR helium proportional counters (abstract)a)

The TFTR helium proportional counters are located in the central five (5) channels of the TFTR multichannel neutron collimator. These detectors were absolutely calibrated using a 14 MeV neutron generator positioned at the horizontal midplane of the TFTR vacuum vessel. The neutron generator position was scanned in centimeter steps to determine the collimator aperture width to 14 MeV neutrons and the absolute sensitivity of each channel. Neutron profiles were measured for TFTR plasmas with time resolution between 5 and 50 ms depending upon count rates. The He detectors were used to measure the burnup of 1 MeV tritons in deuterium plasmas, the transport of tritium in trace tritium experiments, and the residual tritium levels in plasmas following 50:50 DT experiments.

Tritium removal from various lithium aluminates irradiated by fast and thermal neutrons (COMPLIMENT experiment)

Abstract Within the frame of the COMPLIMENT experiment, γ-LiAlO 2 specimens with different microstructures (grain size distributions) were tested in the same environmental conditions to compare the effects caused by 6 Li(n, α)T reaction and by fast neutron scattering, the damaging dose being held at about the same level (1.6–1.8 dpa). The tritium retention times were obtained by the tritium removal of isothermal annealing under He + 0.1% H 2 sweeping gas. In spite of the different Li burnups (2.5% and 0.25%) and the residual tritium concentrations which were found in the irradiated specimens (4.3 Ci/g and 0.09 Ci/g, respectively, for specimens held at 450°C during the irradiations), the kinetics of tritium removal was not found to be discriminated by the two different irradiations. Moreover, the results were found to agree with those previously obtained by the “in-situ” TEQUILA experiment, performed on the same type of Li ceramics. Hence, the apparent first order desorption mechanism has been confirmed to control the kinetics of tritium removal from the porous fine grain γ-LiAlO 2 ceramics.

Neutron emission profile measurements during the first tritium experiments at JET

During a series of experiments with tritium (T) in deuterium (D) plasmas in the Joint European Torus (JET), the temporal evolution and the two dimensional (2-D) spatial profiles of the 2.5 and 14 MeV neutron emissivities from D-D and D-T fusion reactions were inferred from measurements with the JET neutron emission profile monitor. These experiments, involving triton production from D-D fusion, beam deposition and diffusion, D-T fusion, and tritium removed from wall tiles, were investigated in four plasma scenarios: (i) In high performance deuterium plasmas with deuterium neutral beam injection, the 14 MeV neutron emissivity due to triton burnup was observed. (ii) In discharges with 1% tritium beam injection, neutron emissivity ratios showed that approximately the same deposition profiles resulted from tritium as from deuterium beams. A thermalized tritium diffusion experiment was performed in which the T-D density ratio was found to be spatially constant across the plasma; in conjunction with similar particle source profiles, this indicates that deuterium and tritium have similar particle transport properties. (iii) In two high performance discharges for which two of the sixteen neutral beam sources operated with 100% tritium, the production rate of 14 MeV neutrons reached 6 × 1017 n.s-1 The alpha particle 2-D birth profile was directly inferred from the measured 14 MeV neutron emissivity profile. Both the axial 14 MeV neutron emissivity and the axial ion temperature saturated before the maximum global emission was reached. (iv) During the tritium cleanup phase, residual tritium entering the plasma produced a spatially constant ratio of tritium to deuterium, confirming the similarity of their particle transport properties

Residual tritium effects during spill clean-up

The specific activity of the moisture in the exhaust gas from a dryer was measured at various conditions. It approximately equalled that of the residual moisture left on the adsorbent during the previous regeneration. Modeling was used to demonstrate that this residual tritium can affect emissions and occupational dose during chronic or acute detritiation, and should be accounted for during ADS design and analysis.