Tritium Retention Research Articles

A Comparison of Tritium Dispersion Methodology for Accident Analysis in U.S. Department of Energy Complex Facilities

A comparison of three United States (U.S.) Department of Energy (DOE) Standard DOE-STD-3009–2014 dispersion modeling protocol options has been performed assuming a ground-level release of tritium oxide source term. The options are characterized by differing sets of assumptions and inputs that allow incorporating greater user flexibility and realism into the modeling and subsequent analysis. The three options used to evaluate atmospheric dispersion include: (1) Use of U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145; (2) Application of a DOE-approved toolbox code and application of conservative input parameters; and (3) Use of site-specific methods and parameters as defined in a site/facility specific DOE-approved modeling protocol.Option 1 dose results are the lowest of the three sets of results at close-in distances, but are the highest for distances beyond approximately 3,000 m, reflecting the distance-dependent NRC plume meander model. Option 1 doses also reflect a lower minimum wind speed and consideration of G stability. Option 3 dose results are consistently lower than the Option 2 results by a factor of 2.2 reflecting the higher vertical dispersion values calculated from the crediting site-specific surface roughness. Option 2 and 3 results are obtained with DOE Central Registry computer software reflect default parameters in Option 2, and more site-specific input with Option 3. An averaging time of two hours leads to dose results that are lower than those obtained with an averaging time of three minutes by a factor of 2.5 due to the higher crosswind dispersion parameter values. This effect is due to the larger crosswind dimension of the plume with increasing averaging time using the Gifford meander model. A sensitivity case study indicates appreciable differences are observed between results obtained with the NRC Regulatory Guide 1.145 temperature difference (ΔT) method and those with U.S. Environmental Protection Agency (EPA) EPA-454/R-99–005 methodology for stability class categorization. A second sensitivity case suggests that crediting deposition, hold-up or other retention of tritium may be difficult to defend from a regulatory perspective, recognizing region of transport characteristics and accounting for reemission phenomenon. In terms of recommending one of the three options for modeling tritium releases in Documented Safety Analysis (DSA) applications, the Option 2 approach (Application of a DOE-approved toolbox code and conservative input parameters – without crediting tritium deposition) is the simplest model for source to receptor distances of 500 m or greater. Option 3 requires additional resource commitment and DOE authority approval, but may provide regulatory relief for certain accident scenarios. These recommendations apply to deterministic DSA dispersion analysis but are not extended to best estimate, realistic analyses such as those supporting probabilistic safety analyses.

Prospects for a dominantly microwave-diagnosed magnetically confined fusion reactor

Compared to present experiments, tokamak and stellarator reactors will be subject to higher heat loads, sputtering, erosion and subsequent coating, tritium retention, higher neutron fluxes, and a number of radiation effects. Additionally, neutral beam penetration in tokamak reactors will only be limited to the plasma edge. As a result, several optical, beam-based and magnetic diagnostics of today's plasmas might not be applicable to tomorrow's reactors, but the present discussion suggests that reactors could largely rely on microwave diagnostics, including techniques based on mode conversions and Collective Thomson Scattering.

Numerical study of thermal erosion and topographical change of divertor target plates induced by type-I edge-localized modes

The high-Z material tungsten (W) is a promising candidate of the plasma facing components (PFCs) for the future tokamak reactors due to its high melting point (3683 K), low tritium retention and low sputtering yield. However, there are still many problems about W PFCs. One of them is the material melting under off-normal transient heat fluxesit is one of the most outstanding open questions associated with the use of W divertor targets in international thermonuclear experimental reactor (ITER). This requires us urgently to understand the W melting behavior under high power flux deposition condition. In this paper, a two-dimensional (2D) fluid dynamic model is employed by solving the liquid hydrodynamic Navier-Stokes equation together with the 2D heat conduction equation for studying the erosion of the divertor tungsten targets and its resulting topographical modification during a type I-like edge-localized mode (ELM) in ITER with a Gaussian power density profile heat load. In the present model, major interaction forces, including surface tension, pressure gradient and magnetic force responsible for melt layer motion, are taken into account. The simulation results are first benchmarked with the calculated results by other code to validate the present model and code. Simulations are carried out in a wide range of fusion plasma performance parameters, and the results indicate that the lifetime of W plate is determined mainly by the evolution of the melt layer. As a consequence of the melt layer motion, melted tungsten is flushed to the periphery, a rather deep erosion dent appears, and at the dent edges two humps of tungsten form during the ELM. The humps at both edges are almost at the same height. Calculated results show the topographical modification becomes noticeable when the W plate is exposed to a heat flux of 2000 MWm-2 for 0.8 ms (in the simulation, the parameter k=ə/əT is taken to be -9.010-5 Nm-1K-1, where is the surface tension coefficient and T is the temperature). The values of the humps are both about 2.1 m, and the surface roughness is about 1.1 m. The longer the duration of the ELM, the more rapidly the humps rise. The melt flow may account for the higher surface temperature at the pool periphery, and for the larger melt thickness. It is found that when the energy flux is under 3000 MWm-2 the surface tension is a major driving force for the motion of melt layer. Under the same heat flux, the bigger the k used in the simulation, the more severe the surface topography of the target becomes; while at the same k, the higher the heat flux, the more severe the surface topography of the target becomes. In addition, a modified numerical method algorithm for solving the governing equations is proposed.

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

Tritium transport analysis of HCPB blanket for CFETR

Based on the HCPB (Helium-Cooled Pebble Bed) blanket of CFETR (Chinese Fusion Engineering Testing Reactor), a more detailed tritium cycle model of HCPB blanket was developed by modeling the breeding parts inside the blanket separately. Utilizing the detailed model, tritium transport analysis of HCPB blanket was carried out. The outputs of the model were the tritium concentrations and tritium retentions at different positions, tritium extraction rate by TES (Tritium Extraction System) and by CPS (Coolant Purification System) and the tritium losses of blanket. In addition, sensitivity analysis of key parameters’ influence on tritium losses and tritium extraction rate by TES was performed. The tritium transport analysis provides valuable reference for the design of HCPB blanket.

Tracking of tritium charged into stainless steel by BIXS

Retention and release behavior of tritium charged into stainless steel type 316 by two different methods were tracked by β-ray-induced X-ray spectrometry (BIXS). Exposure to tritium gas and irradiation of tritium ions were employed. Effects of pre-irradiation of helium ions on tritium retention were also investigated. After charging tritium by these methods and tracking by BIXS, thermal release behavior of tritium was examined by BIXS. The observed X-ray spectrum showed clearly that thermally charged tritium at a high temperature diffused into the bulk of SS316, but tritium ions irradiated at room temperature were trapped in surface layers of the sample irrespective of tritium ion energy. Thermally charged tritium was not able to remove completely even heating at high temperature of 973K. On the contrary, it was seen that the ion-irradiated tritium decreased almost linearly with temperature irrespective of ion energy. Furthermore, rapid decrease of tritium appeared in the helium pre-irradiated samples.

Elucidating the Stereochemistry of Enzymatic Benzylsuccinate Synthesis with Chirally Labeled Toluene.

Benzylsuccinate synthase is a glycyl radical enzyme that initiates anaerobic toluene metabolism by adding fumarate to the methyl group of toluene to yield (R)-benzylsuccinate. To investigate whether the reaction occurs with retention or inversion of configuration at the methyl group of toluene, we synthesized both enantiomers of chiral toluene with all three H isotopes in their methyl groups. The chiral toluenes were converted into benzylsuccinates preferentially containing (2) H and (3) H at their benzylic C atoms, owing to a kinetic isotope effect favoring hydrogen abstraction from the methyl groups. The configuration of the products was analyzed by enzymatic CoA-thioester synthesis and stereospecific oxidation using enzymes involved in benzylsuccinate degradation. Assessment of the configurations of the benzylsuccinate isomers based on loss or retention of tritium showed that inversion of configuration at the methyl group occurs when the chiral toluenes react with fumarate.

The influence of neutron activated products in BIXS method for tungsten

Tritium β-decay Induced X-ray Spectroscopy (BIXS) is a promising method to obtain tritium retention details both on surface and in bulk of plasma facing materials (PFMs) in magnetic fusion devices. Influence of neutron activated products (NAPs) is one of the primary problems in BIXS, since the energy of Compton scattering photons and characteristic X-rays is close to the range of tritium β-rays. In this paper, Monte Carlo simulation was introduced to evaluate effects of NAPs in BIXS measurements for tungsten. Three steps are included in simulation, neutron flux simulation to reckon the neutron flux in PFMs based on the simplified model of ITER, activated products calculation to compute the distribution of NAPs in PFMs and radiation spectra simulation to calculate the spectra of NAPs in PFMs sample. Six typical nuclides (including W-185, W-181 and Re-188, etc.) produced by neutron irradiation were considered in this study. Finally, influence of NAPs was estimated quantitatively with the assumption that 1% of DT fuel retained in PFMs uniformly. It is found that the intensity of spectrum devoted by NAPs is as much as 10.6 times larger than it by tritium in the same region of interest (ROI) directly after reactor shutdown, while it will reduce to 0.83% after 19 months.

Tritium release from advanced beryllium materials after loading by tritium/hydrogen gas mixture

Comparison of different beryllium samples on tritium release and retention properties after high-temperature loading by tritium/hydrogen gas mixture and following temperature-programmed desorption (TPD) tests has been performed. The I-220-H grade produced by hot isostatic pressing (HIP) having the smallest grain size, the pebbles of irregular shape with the smallest grain size (10–30μm) produced by the crushing method (CM), and the pebbles with 1mm diameter produced by the fluoride reduction method (FRM) having a highly developed inherent porosity show the highest release rate. Grain size and porosity are considered as key structural parameters for comparison and ranking of different beryllium materials on tritium release and retention properties.

A new vision of plasma facing components

This paper advances a vision for plasma facing components (PFCs) that includes the following points. The solution for plasma facing materials likely consists of engineered structures in which the layer of plasma facing material (PFM) is integrated with an engineered structure that cools the PFM and may also transition with graded composition. The key to achieving this PFC architecture will likely lie in advanced manufacturing methods, e.g., additive manufacturing, that can produce layers with controlled porosity and features such as micro-fibers and/or nano-particles that can collect He and transmutation products, limit tritium retention, and do all this in a way that maintains adequate robustness for a satisfactory lifetime. This vision has significant implications for how we structure a development program.

Technological exploitation of Deuterium–Tritium operations at JET in support of ITER design, operation and safety

Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned Deuterium–Tritium experiment on JET (DTE2) with the objective of maximising the scientific and technological return of DT operations at JET in support of ITER. This paper presents the progress since the start of the project in 2014 in the preparatory experiments, analyses and studies in the areas of neutronics, neutron induced activation and damage in ITER materials, nuclear safety, tritium retention, permeation and outgassing, and waste production in preparation of DTE2.

Tritium retention measurements by accelerator mass spectrometry and full combustion of W-coated and uncoated CFC tiles from the JET divertor

Accelerator mass spectrometry (AMS) and the full combustion method (FCM) followed by liquid scintillation counting were applied to quantitatively determine the tritium retention in the tungsten-coated carbon fibre composites (CFC), in comparison to uncoated CFC tiles from the JET divertor. The tiles were adjacent and exposed to plasma operations between 2007 and 2009. The tritium depth profiles are showing that the tritium retention on the W-coated tile was reduced by a factor of 13.5 in comparison to the uncoated tile whereas the bulk tritium concentration is approximately the same for both tiles.

In situ monitoring hydrogen isotope retention in ITER first wall

Tritium retention inside the vacuum vessel is a potentially serious constraint in the operation of large-scale fusion machines like ITER. An in situ diagnostics for first wall H/D/T retention by laser induced desorption spectroscopy (LIDS) is proposed for use between plasma discharges. The technique is based on local baking of the first wall by laser irradiation and subsequent analysis of the in-vessel gas by optical emission spectroscopy of plasma radiation. The local heating implementation, kinetics of H/D/T thermal extraction and the accuracy of optical emission spectroscopy measurements are analysed. To resolve the H/D/T lines spectroscopically, their thermal broadening should be minimized to prevent overlapping of the line shapes. A comparative performance analysis of several types of plasma sources with relatively cold ions is made including the following types of discharges: Penning, RF multipactor, laser torch and ECR. All these radiation sources require rather low power and could be used for remote in situ measurements of relative densities of the thermally extracted hydrogen isotopes.

Simulating the nitrogen migration in Be/W tokamaks with WallDYN

The migration of wall material or seeding impurities plays an important role in the formation of mixed materials, the impurity contamination of the plasma and tritium retention. First, this work presents an improved model for the sputtering from mixed material surfaces in WallDYN. Second, we present dynamic SDTrimSP and WallDYN simulations of the nitrogen implantation in Be and the migration of nitrogen in tokamaks with a Be main wall. The simulations with the binary collision code SDTrimSP predict that N accumulates directly at the surface and that the Be erosion decreases with increasing N surface content. A first application of WallDYN to the nitrogen migration with an ITER-like wall indicates that the Be main wall may cause wall pumping of N by co-deposition with Be.

Influence of helium on deuterium retention in reduced activation ferritic martensitic steel (F82H) under simultaneous deuterium and helium irradiation

Deuterium and helium retention in Japanese reduced activation ferritic martensitic (RAFM) steel (F82H) under simultaneous D–He irradiation at 500, 625, 750, and 818 K was studied. This study aims to clarify tritium retention behavior in RAFM steels to assess their use as plasma facing materials. The irradiation fluence was kept constant at 1 × 1024 D m−2. Four He desorption peaks were observed with He retention greatest at 625 K. At T > 625 K a monotonic decrease in He retention was observed. At all temperatures a systematic reduction in D retention was observed for the simultaneous D–He case in comparison to D-only case. This suggests that He implanted at the near surface in RAFM steels may reduce the inward penetration of tritium in RAFM steels that would result in lower tritium inventory for a given fluence.

Impact of temperature during He+ implantation on deuterium retention in tungsten, tungsten with carbon deposit and tungsten carbide

Temperature dependence on deuterium (D) retention for He+ implanted tungsten (W) was studied by thermal desorption spectroscopy (TDS) to evaluate the tritium retention behavior in W. The activation energies were evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code and found to be 0.55 eV, 0.65 eV, 0.80 eV and 1.00 eV. The heating scenarios clearly control the D retention behavior and, dense and large He bubbles could work as a D diffusion barrier toward the bulk, leading to D retention enhancement at lower temperature of less than 430 K, even if the damage was introduced by He+ implantation. By comparing the D retention for W, W with carbon deposit and tungsten carbide (WC), the dense carbon layer on the surface enhances the dynamic re-emission of D as hydrocarbons, and induces the reduction of D retention. However, by He+ implantation, the D retention was increased for all the samples.

Tritium distributions on tungsten and carbon tiles used in the JET divertor

Tritium distributions on the W-coated divertor tiles used with Be wall in JET 2011–2012 ITER-like wall (JET-ILW) campaign were measured using an imaging plate (IP) technique. The high intensity of photo-stimulated luminescence (PSL) from IP was observed at the regions covered by deposited Be layers. However, the PSL intensity was not simply proportional to the thickness of the deposited Be layers; the shadowed region of Tile 4 showed the highest PSL intensity though the thickness of deposited Be layer on this region was smaller than that on Tile 0 and the apron of Tile 1 by an order of magnitude. These observations indicated the influence of impurities such as oxygen on tritium retention in the deposited Be layers. The C tiles used in the 2007–2009 JET carbon wall (JET-C) campaign were also examined. The high PSL intensity was observed for the regions covered with deposited C layers in this case. The area of tile surfaces covered by the deposited tritium-rich layers on the W-coated-tiles used in the JET-ILW campaign was significantly smaller than that on the C tiles used in the JET-C campaign.

Tritium retention in individual metallic dust particles examined by a tritium imaging plate technique

Tritium imaging plate technique (TIPT) has been applied to examine tritium (T) retention in individual particles made of titanium (Ti) with 30 and 100 μm in diameter and tungsten (W) with 50 μm in diameter. Distribution of T radioactivity observed by TIPT corresponded well to spatial distribution of the particles. In a limited case of uniform and high T concentration in the bulk of the individual particle, the amount of T is directly quantified from T radioactivity by a master curve method. Density and size of the particle and T concentration profiles in the bulk of the particle are important factors to change emission behavior of T β-ray and thus accurate quantification of the amount of T in the individual particle.

Tritium desorption and tritium removal from tungsten pre-irradiated with helium

In this study, 1keV DT+ ion irradiation was performed on tungsten pre-irradiated with helium. The thermal desorption behavior and the reduction of tritium retention during vacuum preservation at room temperature, as well as isochronal annealing were investigated using an IP technique, taking advantage of the fact that tritium detection is nondestructive and is highly sensitive. At a pre-irradiated helium fluence of 1×1017He/cm2, retained tritium tended to be desorbed at higher temperatures when compared to no helium irradiation case. Tritium retention during preservation in vacuum and during isochronal annealing became smaller with increasing helium fluence up to 1×1017He/cm2. At a helium fluence of 1×1018He/cm2, the reduction of tritium retention was found to be greater compared to 1×1017He/cm2. The results indicate that helium irradiation has a significant influence not only on the thermal desorption temperature of tritium but on longtime tritium reduction at room and elevated temperatures.

Study of tritium desorption by energetic ion bombardment from tungsten material

To solve the problem of tritium retention in a nuclear fusion reactor, the removal of hydrogen isotopes retained in a tungsten material by exposure to the other hydrogen isotope plasma was investigated. We proposed the balance equation between the absorption and desorption of the hydrogen isotope atoms in the tungsten material. Absorption ratio and desorption yield were calculated by binary collision approximation simulation. Simulation results showed that desorption yield was proportional to the concentration of the retained hydrogen isotope. Therefore, the decrease in the retained hydrogen isotope atoms by hydrogen plasma irradiation can be regarded as approximately exponential decay. The decay rate in the case that retained deuterium atoms are removed by light hydrogen irradiation was two thirds of the decay rate in the case that retained tritium atoms are removed by deuterium irradiation.