Assessment Of Tritium Inventory Research Articles

A system dynamics model for tritium cycle of pulsed fusion reactor

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: 1h). 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.

TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γimax of >1.0×1023m−2s−1 and qheat of >1MWm−2 with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

Tritium inventory assessment for ITER using TRIMO

Currently, the strategy for determination of ITER in-vessel tritium inventory envisages that at predetermined intervals, tritiated gases in all systems of fuel cycle will be transferred to the storage and delivery system (SDS) and tritium quantities measured by in-bed calorimetry. The isotope separation system (ISS) is the system used to separate hydrogen isotopes at the quality required to be stored in SDS, and is one of the systems with highest tritium inventory within the fuel cycle. Therefore, during tritium inventory procedure, ISS has to be ‘milked down’ of tritium, mainly as DT molecular species. Based on the dynamic modelling code TRIMO of the tritium content in the main sub-systems of ITER Fuel Cycle, the procedure for tritium extraction from ISS is presented and numerical examples given to assess the necessary time of transferring the tritium from ISS to SDS, and the residual amount of tritium in ISS after different milking scenarios. Consequently a fuel handling strategy during tritium inventory assessment in the ISS and SDS is described, with the constraint of mobilizable tritium inventory minimisation.

Jet tritium inventory-control and measurement during DT, shutdown and DD operational phases

DT plasma operations took place at Joint European Torus (JET) from May to November 1997. The site tritium inventory was 20 g, and this was recycled several times using the active gas handling system (AGHS) to recover tritium from mixed exhaust gases. The amount of tritium available for use at any time was determined by the hold-up in the in-vessel components, in-process inventory, and losses via permeation and transfer to waste. It was therefore necessary, to continually assess the tritium inventory through on-line and batch measurements. During the shutdown periods, tritium continued to be evolved from the machine and its trapped volumes. This meant that a large number of specific measurements were necessary to ensure that the safety requirements were met and that the waste could be properly characterised. During DD operations, the exhaust gas from the machine contained quantities of tritium, too great to be discharged to atmosphere and must continue to be processed by the AGHS. The techniques for tritium inventory assessment are reviewed, taking into account continuing uncertainties and plans for resolutions of these. The JET experiment is used to highlight issues of tritium inventory control, which will apply to future fusion reactors.

Deuterium retention in codeposited layers and carbon materials exposed to high flux D-plasma

A ceramic BCN target with samples of dense RG–Ti–91 without boron, RG–Ti–B with boron (0.1 at.%) and porous POCO AXF-5Q graphites was exposed in a stationary D-plasma of the `Lenta' device with an ion energy of 200 eV and an ion flux of (3 − 6) × 10 17 D/cm 2s at 1040 and 1400 K to a fluence of ∼1 × 10 22 D/cm 2. Codeposited layers were obtained for comparison on the target surface. Thermal desorption spectroscopy (TDS) showed that the amount of deuterium in RG–Ti after exposure at 1040 K was more than an order of magnitude higher than in POCO (9 × 10 17 and 7 × 10 16 D/cm 2, respectively). The retention took place preferentially in a surface layer about 100 μm thick. The bulk deuterium concentration in both RG–Ti and POCO was lower than 1 appm. The irradiated RG–Ti surface was subjected to strong erosion and consisted of `columnar' grains covered with TiC at their tips. The deuterium in RG–Ti irradiated at 1400 K was located in the surface layer (1.5 × 10 16 D/cm 2). The value of the bulk concentration did not exceed 0.1 appm while in POCO it was equal to about 20 appm. TDS for deuterium in RG–Ti demonstrated a spectrum similar to that for codeposited layers on a target surface. The differences in deuterium retention in the graphites are explained on the basis of structural differences. Considering tritium inventory assessment for ITER, dense graphites like RG–Ti are preferred for working divertor plates at high temperatures.

Assessment of Tritium Inventory in Waste by Outgassing Measurements

The production of large volumes of low specific activity waste is a byproduct of working with tritium. Reliable, labour minimized techniques for assessing the tritium bound in the waste is essential for an effective waste management program. This work examines the use of tritium outgassing rates from waste as a means of determining the bound inventory. A simple and inexpensive technique has been developed which, when combined with a sampling methodology, leads to accurate measurements of tritium outgassing rates. These outgassing rates when used in conjuction with the appropriate database will yield the inventory bound in the waste. The results of this study support the viability of the proposed approach.

Tritium retention in plasma facing components for NET/ITER fusion devices

This paper presents the derivation and the results of a tritium inventory assessment carried out for prototypical NET/ITER first-wall and divertor carbon-based armors, and discusses the impact of the uncertainties associated with these estimates. Transient effects associated with fusion device pulsed operation were modelled using two tritium retention models: (i) a new model primarily suitable for tritium retention analysis in porous media, and (ii) a diffusion-type model developed for metallic materials. The influence of several key factors such as the material microstructure, the flux density, the temperature range, the burn operation scenario, and the burn duration which are expected to strongly affect tritium retention is analyzed. A parametric assessment is performed for a 50/50% D-T plasma, particle fluxes in the range of 10 16–10 19 particles cm −2 s −1, temperatures in the range of 300–1900°C, and burn-times in the range of 400–1000 s.