Loss Of Tritium Research Articles

Preliminary tritium transport analysis for the water cooled ceramic breeder blanket system based on a system level code

This paper presents a dynamic analysis tool for evaluating the tritium inventory, tritium extraction and tritium losses for the Water Cooled Ceramic Breeder (WCCB) blanket system of Chinese Fusion Engineering Test Reactor (CFETR). We have developed a program for the tritium transport analysis at the system level, which aims to estimate the tritium related quantities for the WCCB blanket system. The tritium inventory in the material defects is likewise taken into account. The sensitivity analysis is conducted on key parameters is carried out. The key output quantities of the WCCB blanket system include the tritium inventory in structural material, purge gas and coolant loops, tritium extracted by the TES and CPS, and tritium losses.

Development of a tritium release system for ceramic tritium breeder pebbles irradiated by high intensity DT fusion neutron Generator at INEST

For ceramic breeder pebbles used in fusion energy, the tritium release data under fusion neutron irradiation is very important and valuable. In this study, a tritium release system for ceramic tritium breeder is introduced, which is matched with the High Intensity DT Fusion Neutron Generator (HINEG-CAS). The HINEG-CAS has been built and operated by the Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences (INEST, HFIPS, CAS), achieving a fusion neutron yield of 1012n/s since 2016. Innovative design and fabrication of sample container enables minimization of the unexpected tritium loss. The dedicated tritium release system can monitor the tritium release behavior over time for both tritiated water and tritium gas, using the bubblers-based tritium trapping system with automatic replacement technology, coupled with highly efficient catalytic oxidants and liquid scintillation counter (LSC). We have performed a tritium release experiment with the Li2TiO3 pebbles as ceramic tritium breeder, and obtained the tritium releasing data at room temperature over time and clarified the tritium chemical form in recovery. It is concluded that this newly established system can serve as a good sharing platform for evaluating prime candidate breeder materials, as well as facilitating detailed design or optimization of tritium involved systems (such as the tritium extraction system).

Tritium-titanium target degradation due to deuterium irradiation for DT neutron production

In the present article, we have investigated tritium removal from tritium-titanium targets during fusion neutron production and the impact of tritium degradation on neutron production. The removal of tritium from the target is predicted for deuterium ion irradiation with the SDTrimSp code. We adopt the binary collision approximation method to simulate the recoils and projectile trajectories and concentration of constituents in the target. We have modelled four phenomena in our simulations; ion exchange, sputtering, outgassing of tritium, and thermal diffusion of hydrogen isotopes in the target caused by deuterium irradiation. Insignificant contributors such as burn-up of tritium in neutron production and loss of tritium due to radioactive decay are not included in our model. This tritium removal results in the nonuniform distribution of tritium in the target. A Python-based script is developed to investigate the effects of tritium removal on neutron production with these pristine and irradiated targets. This script uses the layered composition of the constituents’ atoms, DT reaction cross-section, and stopping power of deuterium ions in the target. The script is validated with the NeuSdesc code for the pristine target. Using the layered composition of tritium atoms in the target obtained from the SDTrimSp simulations, the script predicts the degradation in neutron production for different irradiation scenarios.

Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

• T inventory in the ITER EU Water-Cooled Lithium-Lead Test Blanket Module investigated. • T from vacuum vessel (gas, neutrals) and solute T in PbLi included in modeling. • Effect of neutron damaging on trap concentrations included in models. • T retention in the range of 0.75 g is expected over plasma operation period. • Comparison with the TMAP7 code performed. Tritium self-sufficiency is a key requirement for future fusion power plants. Therefore, it will be necessary to minimize tritium losses to the surrounding systems, such as the tritium breeding modules, plasma-facing components, cooling system, etc. These components and systems will act also as a tritium sink, as tritium will be retained in the metal walls in traps produced during manufacturing or induced by neutron irradiation. The design of the tritium breeding systems and plasma-facing components will therefore have a direct impact on the performance, operational safety and cost of any future power plant. Consequently, accurate modeling of tritium transport and retention in these systems is needed for future engineering designs of the tritium breeding systems and supporting safety requirements. In this work, tritium permeation and retention was modeled with the diffusion-transport code TESSIM-X for the current European WCLL (Water Cooled Lithium Lead) Test Blanket Module (TBM) for the ITER TBM Programme, with the inclusion of neutron-induced traps. Trap saturation is considered for dpa values of 0.25 and above, leading to an increase in trap concentrations relative to undamaged material of roughly a factor 3. The code was also benchmarked against the well-established TMAP7 code.

Fabrication and test of a conceptual first wall containing titanium nitride as a tritium permeation barrier

The control of tritium buildup of plasma-facing components (PFCs) is essential for the safety regulation of tritium and maximizing the efficiency of the reactors. Tungsten (W) is a promising material of PFCs owing to its attractive benefits of high melting point, high thermal conductivity, low tritium retention and low sputtering yield by particles. However, the hydrogen isotope mobility of W is high, which leads to the loss of tritium of the reactor in the fueling cycle. A conceptual first wall with the capacity of tritium prevention was designed and fabricated, which has a sandwich-like structure consisting of a W top layer deposited by chemical vapor deposition (CVD-W), a reduced activation ferritic/martensitic (RAFM) steel called CLF-1 as substrate, and a titanium nitride (TiN) interlayer as a tritium barrier. This work aims to investigate the effects of substrate surface roughness and interlayer thickness, which are essential to the adhesive strength of the sandwich-like structure. Samples with various interlayer thicknesses and substrate roughnesses were prepared and analyzed by using a scanning electron microscope (SEM) equipped with an energy dispersive X-ray spectrometry (EDS), X-ray diffraction (XRD), the thermal cycling test, and the pull-off test. The results show that proper roughness of the substrate and suitable thickness of the TiN interlayer could crucially contribute to the adhesive strength. Among all the samples, the samples with the substrate roughness of 0.8–1.6 μm and the TiN interlayer thickness of 1.5±0.5 μm have the best adhesive strength. In addition, the gas-driven permeation (GDP) test is employed to evaluate the deuterium permeation resistance of the sandwich-like structure.

On the modelling of tritium transport phenomena at fluid-structure interfaces

One of the main functions of the DEMO Breeding Blanket (BB) system is to ensure the tritium breeding inside the reactor. Tritium is a beta emitter radioactive isotope, subjected to several processes that determine its permeation across materials and its leakage towards the environment, posing potential safety issues in terms of radiological hazard. Thus, the evaluation of tritium inventories inside components and tritium losses towards the environment plays a key role in the fulfilment of the pertinent BB safety requirements. In this regard, a research activity has been carried out, in close cooperation between the University of Palermo and the Karlsruhe Institute of Technology, focussing on the development of a multiphysical model that might realistically simulate 3D tritium transport phenomena across complex fluid-structure interfaces. Models, source terms and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Tritium transport code development based on COMSOL and MATLAB and its application to the WCCB blanket

Water cooled ceramic breeder (WCCB) blanket was proposed as one of the candidate blanket schemes of the Chinese Fusion Engineering Test Reactor (CFETR). In this paper, a code simulating tritium transport in WCCB blanket and its external loops has been developed, where zero and two dimensional models are integrated to get more reliable results. The purge gas flow in the breeder zone, heat transfer in the blanket and tritium isotopes mass transport are considered. The isotope exchange in the purge gas and coolant loop are also taken into account in the model. The two dimensional model of the WCCB blanket and zero dimensional model of external loops are integrated to calculate tritium related quantities. The amount of tritium inventory in different components, tritium losses, tritium extracted by the tritium extraction system (TES) and purified by the coolant purification system (CPS) are obtained. The sensitivity analysis about key parameters has also been conducted.

Preliminary estimates of tritium permeation and retention in the first wall of DEMO due to ion bombardment

• Tritium losses in the DEMO first wall were investigated with T gas and ion loading. • Code comparison with TMAP and TESSIM showed good agreement in a DEMO-like scenario. • Hundreds of g of T may be retained in wall. Up to 160 mg T could permeate per day. • With bare EUROFER wall, retention below 100 g but T permeation up to several g/day. • Impact of T losses on T fuel cycle assessed with simplified particle balance model. Tritium self-sufficiency presents a critical engineering challenge for DEMO, requiring efficient breeding and extraction systems, as well as minimizing tritium losses to the surrounding systems, such as plasma-facing components, vacuum vessel, cooling system, etc. Structural and plasma-facing components will act as a tritium sink, as tritium will be accumulated in the bulk of these components due to energetic particle bombardment and may permeate out of the vacuum system. The design of the plasma-facing components will consequently directly influence the plant lifetime, operational safety and cost of any future power plant. Therefore, modeling of tritium retention and permeation in these components is required for the engineering designs of the tritium breeding and safety systems. In this work, the diffusion-transport code TESSIM-X is benchmarked against the well-established TMAP7 code and a comparison with a simplified DEMO-relevant test case is performed. The use of either code for modeling of DEMO conditions is discussed. Following this, TESSIM-X is used to provide a preliminary assessment of tritium permeation and retention in the DEMO first wall, based on the current WCLL (Water Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) breeding blanket designs.

Optimization of target lifetime for production of 14 MeV neutrons

Two methods are examined for extending the life of tritium targets for production of 14 MeV neutrons by the 3H(2H,n)4He nuclear reaction. With thick film targets the neutron production rate decreases with time due to isotope exchange of tritium in the film with implanted deuterium. In this case, the target life is maximized by operating the target at elevated temperature where the implanted deuterium mixes by thermal diffusion throughout the entire thickness of the film. The number of neutrons obtained from a target is then proportional to the initial tritium content of the film. A novel thin-film target design was also developed and tested. With these thin-film targets, the incident deuterium is implanted through the tritide into the underlying substrate material. A thin permeation barrier layer between the tritide film and substrate, reduces the rate of tritium loss from the tritide film. Good thin-film target performance was achieved using W and Fe for the barrier and substrate materials respectively. Thin-film targets were fabricated and tested and shown to produce similar number of neutrons as thick-film targets while using only a small fraction of the amount of tritium.

Tritium permeation modelling in DEMO WCLL cooling system

• Tritium permeation in DEMO represents a risk for safety and plant performance. • H 2 addition and CPS sizing in DEMO WCLL have strong impact on tritium release. • PRF barriers have to be taken into account to maintain tritium losses at acceptable levels. Tritium management is one of the most critical aspects in DEMO design. Knowledge of tritium permeation is a priority both in terms of safety and plant performance, as it may represent a radiological risk for the employed personnel and a significance loss of plasma fuel. Generally, high pressures and temperatures promotes the permeation phenomenon, which is mainly driven by partial pressure of tritium in gas phase and depends on system geometry, operative conditions and permeability of involved materials. This contribution presents a preliminary model for the calculation of partial pressures and tritium permeation flow rates inside the cooling system and the steam generator of a WCLL (Water Cooled Lead-Lithium) system. A parametric analysis was made to test the variation of key design parameters and preliminary results are shown.

Tritium transport analysis for tokamak exhaust processing system of tritium plant

Abstract As an important part of tritium plant, tokamak exhaust processing (TEP) system is used for extracting hydrogen isotopes from exhaust gas of tokamak which contains high concentration of tritium and deuterium. A detailed tritium transportation analysis for TEP system can help to obtain more precise and accurate results of tritium behavior in the tritium plant. In this paper, a detailed model of TEP system is established for the particular tritium transportation simulation during its operation. Mean residence time method is mainly used to simulate tritium transport in TEP, and a new calculating method is developed as optimization for Pb-Ag membrane processing. The tritium inventories and tritium losses of components in TEP system are obtained. The influences of reflux ratio of one membrane permeator on the tritium transportation behavior are also analyzed.

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

India, under its breeding blanket R&D program for DEMO, is focusing on the development of two tritium breeding blanket concepts; namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder (HCCB). The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket. The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER. The Indian HCCB blanket having lithium titanate (Li2TiO3) as the tritium breeder and beryllium (Be) as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket. The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket. It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm, respectively, can give a tritium breeding ratio (TBR) >1.3, with 60% 6Li enrichment, which is assumed to be sufficient to cover potential tritium losses and associated uncertainties. The results also demonstrated that the Be packing fraction (PF) has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li2TiO3.

The RMC Method to Handle Tritium Efficiently and Safely

When it comes to the task of handling gaseous tritium, the challenge is to reduce losses of this precious, gaseous, hydrogen isotope. The driving force to achieve this is based on three requests: 1. Improve the safety and efficiency by spotting losses of gaseous tritium. 2. Embed the real-time tritium monitoring in the process and safety automation. 3. Be transparent in the whole workflow for its own safety and for auditable compliance. Many good and accepted single devices and working procedures have been proposed and used already. By introducing the Smolsys Ltd.® Radio Medical Container (RMC) method, a team at Smolsys Ltd. has brought the efficient and safe handling of tritium to a new performance level. The idea of the RMC method is to combine many of those approved single devices for gaseous tritium handling and link the workflow logically and digitally in a well-controllable confinement. In the case of the RMC, the working space or room is a container in which the working places and machines are run; hence, the room itself becomes part of the production process and tritium machine. It is monitored and controlled by the process logic and as such becomes a smart and digitized RMC for more safety and efficiency in tritium handling. This paper presents the RMC based on a realized tritium-processing factory in Switzerland. This RMC is a fully engineered tritium facility with a designed and engineered safety factor and is very flexible to be customized. The RMC is also transportable since standard container sizes are used.

Tritiated water (HTO) uptake and loss in maize plant after short-term exposure of atmospheric HTO vapor at daytime and nighttime

Maize plants were exposed to atmospheric tritiated water (HTO) in the field for 1 h during the day and night under different growth periods. The uptake of HTO vapor by leaves and tissue free water tritium (TFWT) loss in maize plants were investigated. Rate constants of uptake of HTO from the atmosphere by maize leaves were 0.10 ± 0.02–0.21 ± 0.01 h−1 during the daytime exposure and 0.035 ± 0.001–0.13 ± 0.01 h−1 in the nighttime exposure. The uptake rate tended to decrease as the exposure increased and harvest neared. At the end of exposure, the TFWT concentration was highest in the leaf and lowest in the ear, the TFWT concentration in different parts of maize varied by a factor of 3–28 for daytime release and 2–30 for nighttime release. After the end of every exposure, TFWT was lost rapidly from different maize tissues for one day and then at a much slower rate until to harvest. At harvest, the TFWT concentrations varied with different maize tissues by factors of 1–4. The decrease was by factors of 4–351 for daytime exposure and 6–605 for nighttime exposure. The reduction of TFWT varied with the exposure time by the factors of 1–10 depending on plant parts. The reduction factor was lowest in kernel. This result suggests that the HTO vapor uptake rates in maize leaves and TFWT loss rates in the ear may be lower than those in C3 plants.

RIPER: An irradiation facility to test Radiation Induced Permeation and release of deuterium for fusion blanket materials

An efficient reduction of gaseous hydrogen isotope permeation through a metal wall is essential in several applications for the tritium inventory control of fusion energy devices. Tritium is also a safety and biological hazard and tritium losses by permeation through the materials is a critical point to be avoided to ensure safety in fusion energy. Ceramic coatings are candidates to reduce tritium permeation, into the breeder blankets of future devices, due to its low hydrogen permeation rate, excellent corrosion resistance and good mechanical properties. Furthermore, these functional materials of future fusion reactors are expected to be subjected to elevated temperature and high level of radiation fluxes and their impact on the tritium inventory has not been properly assessed mainly due to the lack of data available in the literature on this effect. Radiation induced changes in the composition and microstructure of the materials may alter tritium control in future devices. In this context, a new permeation facility “Radiation Induced Permeation” (RIPER) has been developed at CIEMAT in order to obtain relevant data on permeation of hydrogen isotopes in structural/functional materials under irradiation. This system allows one to differentiate and characterize gas concentration driven permeation, and permeation due to different parameters applied to the sample conditions like temperature, irradiation dose or electric field. The reliability of the system has been checked by means of permeation measurements for different materials in comparison with values found in literature and different measurements during irradiation have been performed. The experimental challenges of these types of measurements are reflected in this paper.

Tritium transportation modeling and analysis for tritium extraction system of HCSB blanket

Tritium is essential to achieve self-sufficiency as the indispensable fuel of DT reactions for future fusion power plant. Tritium plant is intended for recycling the tritium bred by breeder blankets and the unburned fuel from core plasma, and maintaining tritium cycle transportation. As a critical system in tritium plant, tritium extraction system (TES) is used for extracting tritium from blanket breeders and purifying hydrogen isotopes in the purge gas from blankets. In order to obtain the detailed behavior for tritium self-sufficiency analysis and safety performance assessment of TES during its operation, tritium transportation analysis in TES needs to be carried out. In this paper, a detailed TES model is developed referring to the tritium extraction processing for helium cooled solid breeder (HCSB) blanket. Several behaviors of tritium including tritium permeation, tritium leakage and chemical equilibrium movement including HT and HTO, are taken into consideration. The results and analyses of tritium inventories and tritium losses in TES are obtained. PRF (Permeation Reduction Factor) sensitivity analysis of components in TES is also carried out. The requirements of flow rate in glove box (GB) for atmosphere detritiation and makeup flow for He-H2 makeup unit are given.

Oxide layer formation in reduced activation ferritic steel F82H under DEMO reactor blanket condition

Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic steels have been investigated; however, oxidation of the steel surface is expected under an actual DEMO reactor condition, and then the tritium permeation behavior will be changed. In this study, deuterium permeation through the steels heat-treated under simulated environment conditions has been investigated for more precise predictions of tritium loss at DEMO reactor blankets. Reduced activation ferritic/martensitic steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen at 300, 400 and 500 °C for 100 and 200 h to simulate a solid breeder DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. An iron oxide layer was formed on the sample surface, and the thickness of the layer was 50 nm‒12 μm. The oxide layer on the sample surface heat-treated at 500 °C for 100 h decreased deuterium permeation by a factor of 5. After the permeation tests, dissipation of the oxide layers was confirmed.

Permeation Data Analysis Considering a Nonzero Hydrogen Concentration on the Low Pressure Detector Side for a Purged Permeation Experiment

Currently available diffusion constant and Sieverts constant experimental results are based on time dependent permeation experiments. The common principle is an analysis which is expecting that the permeating hydrogen is “transported” from the retentate chamber to the permeate chamber through the connecting membrane, with a vanishing hydrogen partial pressure on the permeate side. But reality shows a different behaviour caused by the fact that a nonzero hydrogen partial pressure in the permeate chamber is necessary for detection purposes. This nonzero pressure is mostly not considered by analysis. This issue is solved (approximatively) numerically by the procedure as described in this paper. This work is rooted in the field of fusion research, where so called purge gas with low partial pressure of tritium is contacting the structural materials (300-550°C) of the fusion reactor (blanket) and of process equipment, where the tritium losses are of interest. The developed algorithms are intended for the evaluation of an experiment termed “Q-PETE” (Q for any hydrogen isotope, PEermeation Transport Experiment), which abstracts the hydrogen transport conditions of the fusion blanket, and where the effect of nonzero hydrogen concentration on the permeate side is relevant. The algorithms are useful for all experiments, where the ratio of hydrogen pressures between retentate and peremeate side are far from infinite.

Radiation effects on deuterium permeation for PLD alumina coated Eurofer steel measured during 1.8 MeV electron irradiation

Tritium permeation through structural materials in future fusion reactors is an extremely important issue due to both radiological concerns and tritium self-sufficiency. In the past many efforts to minimize tritium losses have been made and permeation barriers are considered the best method to reduce them. Alumina is one of the main candidates for such barriers due to its chemical stability and hydrogen isotope permeation reduction factor. In this work pulsed laser deposited (PLD) alumina on Eurofer steel has been studied in terms of deuterium permeation. A reduction factor of about 1000 compared to bare material in the temperature range RT-450 °C was measured. In order to study deuterium permeation during irradiation the Radiation Induced Permeation and Release (RIPER) facility at Ciemat was used. Contrary to expectations the permeation through PLD alumina coated Eurofer was observed to decrease during irradiation. This radiation effect was explained making use of thermostimulated desorption (TSD) experiments. TSD measurements were carried out for both electron irradiated and unirradiated alumina samples and the results showed that radiation induced deuterium back desorption from alumina occurred in this way reducing the overall permeation value during irradiation. Finally, the microstructural study of the irradiated coatings showed that some surface damage appears but not cracking or coating debonding was observed in agreement with the good behaviour observed in terms of deuterium permeation for the irradiated coatings.

Sensitivity study on tritium transport in water cooled solid blanket of China Fusion Engineering Test Reactor

The Water Cooled Solid Blanket (WCSB) is selected as a basic choice of the CFETR blanket design, where the tritium would be bred by lithium and extracted subsequently in the tritium cycling loop. However, there always some inevitable losses exist during the tritium cycle due to the complexity of the system, which is considered as a big concern of the tritium self-consistency. In order to realize the tritium self-consistency of CFETR, a mathematical-physical model is developed to analyze the tritium transport in the WCSB. In the present work, the sensitivity study of some key parameters on tritium losses and inventories were conducted. It was found that the tritium transport in the WCSB is influenced by the (a) thickness of the tungsten which covers the first wall, (b) the concentration of the hydrogen in the coolant water and (c) the content of the water in the purge gas, etc.