Calibrating a gas chromatograph to measure tritium using calorimetry

The availability of calibrated gas mixtures in a tritium plant is an important precondition for correct analysis of unknown gas samples because a quantitative analysis with a gas chromatograph (GC) requires frequent calibration checks. To calibrate a GC for tritium, certified gas mixtures with different tritium concentrations are necessary. These mixtures can be produced directly in a laboratory. In this case an independent method is needed to determine the tritium concentration in the mixture with high accuracy. Calorimetry is the method of choice because its accuracy is better than 1% and it is non-destructive method.Using the closed tritium loop of the Tritium Laboratory Karlsruhe (TLK), different mixtures containing tritium in the range 1–100% have been produced and measured with one of the in-house calorimeters. After that the GC of the CAPER facility at TLK has been calibrated several times with an uncertainty of a few percents.     

Review of tritium confinement and atmosphere detritiation system in hot cells complex

The tritium confinement strategy adopted during the past years in the ITER hot cell building is compared to the safety requirements given by the standard ISO-17873 “Nuclear facilities - criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors”. In fact, this is the reference safety guideline recommended by French licensing authorities.Several features of the considered design of the hot cell building are not in agreement with these guidelines. Main discrepancies concern the zoning of the hot cell complex, the flow rates of ventilation, and the possibility to recycle the room atmosphere and to detritiate the effluent air. These aspects are discussed together with some proposed modifications of the design.     

Estimates of routine tritium discharges from ETHEL

The European Tritium Handling Experimental Laboratory (ETHEL) at JRC-Ispra is constructed within the ESSOR reactor complex and is therefore required to be consistent with the existing site license agreement. This paper summarizes the status of ETHEL under the Italian licensing legislation and focuses, in particular, on the problem of tritium discharges under routine conditions. First, the available margins for additional tritium releases from the ESSOR complex are established by taking into account the existing operational requirements. The discharge requirements of ETHEL are then estimated by analyzing the various in-plant routes by which tritium may be released to the environment under normal operation conditions. An analysis is also undertaken for assessing the radiological impact on the population due to exposure and ingestion pathways. It is shown that the expected gaseous releases and tritiated aqueous discharges comply comfortably with the existing site limits and the fraction available for ETHEL.     

Numerical simulation of tritium behavior under a postulated accident condition for CFETRTEP system

China Fusion Engineering Test Reactor(CFETR)is China's self-designed and ongoing next-generation fusion reactor project.Tritium confinement systems in CFETR guarantee that the radiation level remains below the safety limit during tritium handling and operation in the fuel cycle system.Our tritium technology team is responsible for studying tritium transport behavior in the CFETR tritium safety confinement systems of the National Key R&D Program of China launched in 2017,and we are conducting CFETR tritium plant safety analysis by using CFD software.In this paper,the tritium migration and removal behavior were studied under a postulated accident condition for the Tokamak Exhaust Processing system of CFETR.The quantitative results of the transport behavior of tritium in the process room and glove box during the whole accident sequence(e.g.,tritium release,alarm,isolation,and tritium removal)have been presented.The results support the detailed design and engineering demonstration-related research of CFETR tritium plant.     

The control of tritium in ETHEL

The operation of the European Tritium Handling Experimental Laboratory (ETHEL) will require the implementation of means and procedures for allowing tritium control within the facility. For that purpose, account must be taken of the particular characteristics of tritium, such as its high mobility, capacity to dissolve in materials, often limited precision when performing inventory measurements. This paper estimates the influence of these effects on the overall tritium balance in ETHEL. By employing available models for predicting tritium hold-up, it is estimated that three to four grams of tritium may potentially remain irreversibly fixed in various plant items of the standard laboratory infrastructure (exclusive of experimental circuits). On the other hand, the highest overall precision that may be attained with the present plant regarding inventory measurements is estimated to be of the order of few percent. On the basis of the above estimates, the allowable limits for the Material Unaccounted For (MUF) are discussed.     

高比活度[甲基一~3H]胸腺嘧啶核苷的制备

以５—甲酰基尿嘧啶为起始原料，采用“二步法”：第一步合成了氚标记的５—羟甲基尿嘧啶，第二步合成了氚标记的胸腺嘧啶。放射性比活度分别达到１．６ＰＢｑ／ｍｏｌ和２．２ＰＢｑ／ｍｏｌ，再经酶促反应，生成氚标记胸腺嘧啶核苷，其比活度达２．２ＰＢｑ／ｍｏｌ，放化纯度大于９５％     

高压氘氚气氛下不锈钢吸附氚的初步研究

本工作对于氚在不锈钢表面的吸附和解吸行为进行了初步研究.样品在n(D)∶n(T)=1∶1,230℃时,15 MPa下恒温8 h后,接着在27 MPa下恒温6 h的情况下进行了氚的吸附,测量了室温下和加热到1173 K时的解吸氚量和总吸附量.其结果如下:不锈钢的总吸附氚量是857.4 MBq·cm-2,不锈钢的解吸氚量是722.2 MBq·cm-2;在本实验的条件下,在室温和加热条件下,不锈钢所释放的氚中,化学成分主要是HTO和HT两部分,大部分以HT形式存在;不锈钢的自由氚占吸附总氚量的7.34%;不锈钢的热解吸谱存在三个解吸峰,其解吸温度分别为439、530和843K.     

防氚高分子复合材料的制备及其性能

利用氯化聚乙烯、聚偏二氯乙烯(Polyvinylidene chloride,PVDC)、高密度聚乙烯等高分子材料复合防氚材料。采用渗透实验测量氚气渗透率;通过测量断裂强力和剥离力测量力学性能;通过测量浸泡化学试剂前后的溶失率、断裂强力下降率、剥离力下降率等表征材料的耐化学性能。结果表明,氚气在防氚高分子复合材料中的渗透率为3.44×10-10 m3/(m2?s);与PVC相比,氚气渗透率降低约70倍。PVDC薄膜复合后断裂强力提高约20倍,经纬向断裂强力为721 N。防氚高分子复合材料浸泡5种化学试剂后,最大溶失率为0.29%、断裂强力最大下降率为5.1%,其力学性能、耐化学性能均满足《防护服装酸碱类化学品防护服》(GB24540-2009)等相关标准的要求。     

Application of High Energy Protons in Inertial Confinement Fusion by Using the Fast Ignitor Concept for DT Fusion

A short-laser-pulse driven ion flux is examined as a fast ignitor candidate for inertial confinement fusion. The main mechanism for ion acceleration is charge separation in a plasma due to high-energy electrons driven by the laser inside the target. Another very new branch of fast ignition research is the investigation of the use of laser generated proton beams. In the present paper aims to provide insights into the feasibility of the fast ignition concept with high energy beams of protons generated in laser–plasma interactions. The optimum parameters of an ion beam and laser pulse that are suitable for an ignition spark in a hot precompressed DT fuel are estimated as a rough guide. Also, in this paper we estimate the radius of Deuterium–Tritium (DT) fuel pellet that is equal to the protons range in DT plasma.     

氘和氚标记ST—1435的制备

在氘或氚化溶剂中,16-次甲基-17α-乙酰氧基-19-羧基-孕甾-4-烯-3、20-二酮(Ⅰ)通过脱羧反应制得[10-~2H]ST-1435或[10-~3H]ST-1435(Ⅱ)。氘标记ST-1435经质谱分析表明有1—3个氘原子掺入到ST-1435分子中去,其中m/e 371(M~+)的氘掺入率为22.6％。氚标记ST-1435的比活度为52.5TBq/mol,放化纯度>95％。     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

L-脯-L-亮-甘氨酰胺(PLG)的离子束氚标记及其在鼠脑内的分布

本文用离子束氚标记技术制得氚标记PLG。经磋胶纸层析法纯化,氚化三肽的比活度为6.81GBq/mmol,放化纯度>95％,具有完全的生物活性。由腹腔注射~3H-PLG,直接测定大鼠脑内部分组织的放射性分布,提示用药4.5h后,垂体、纹状体有较高的放射性,而下丘脑和中脑的放射性较低。     

First results with the upgraded TLK tritium calorimeter IGC-V0.5

At Tritium Laboratory Karlsruhe (TLK) calorimetry has been used for almost 20 years as the main accountancy method for tritium inventory. An extensive work has been carried out in order to improve the existing calorimeters. This paper covers the efforts made for the upgrade of the IGC-V0.5 calorimeter. We replaced the hardware interface – including the obsolete PC – and developed a new control and data acquisition software. The new software applies a smart automatic process control during measurements, significantly reducing measurement time and possible user errors. The three PID control loops have been re-tuned using the standard closed loop Ziegler–Nichols procedure to find the optimal PID parameters. Five calibration runs have been performed between 0.5 μW and 1 W, and their results are being presented and discussed.     

ITER中国液态锂铅实验包层模块氚渗透分析

对ITER中国液态锂铅实验包层模块的氚渗透途径进行了初步分析,并建立了氚渗透模型;在确保环境安全的前提下,通过计算LiPb中的氚分压分析了氚渗透量及氚总量的分配情况;在此基础上通过改变进入氚提取系统中LiPb比例(F)和涂层氚渗透减少因子(TPRF)对氚提取及渗透的影响做了灵敏性分析.     

Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM

Tritium behavior in the reactor such as production, diffusion and release are accompanied by their adsorption and desorption in graphite materials, which are essential to the safety of high temperature gas cooled reactor (HTGR). In order to study this important issue, hydrogen instead of tritium is experimentally used in this work and justified viable by theory. By performing multiple sets of comparative experiments, the features of hydrogen adsorption and desorption behavior changing by adsorption temperature and time in typical graphites used in HTR-PM (High Temperature Gas Cooled Reactor – Pebble Bed Module), i.e. reflective layer, fuel element and boron carbon bricks, have been observed and analyzed. Furthermore, the adsorption rates of hydrogen in the three materials as above at different conditions are also given. Based on the experimental results, tritium behavior in the HTR-PM was inferred and estimated, which is significant for the further study on the mechanism of tritium transport.     

Tritium diffusivity in crystal grain of Li2TiO3 and tritium release behavior under several purge gas conditions

It has been reported by the present authors that behavior of tritium release from solid breeder grain is consisted of diffusion in grain, tritium transfer at surface layer and surface reactions on grain surface such as adsorption or isotope exchange reactions. Tritium release curves estimated using the tritium release model gave good agreement with observed tritium release curves from Li₄SiO₄, Li₂ZrO₃ or LiAlO₂.

Tritium release behavior from Li₂TiO₃ under humid purge gas, dry purge gas and dry purge gas with hydrogen conditions is discussed in this study, tritium release curves using the release model that we proposed previously give a good agreements with experimental tritium release curves. Tritium effective diffusivity in the crystal grain of Li₂TiO₃ is also estimated in this study using a curve-fitting method applied to the release curves obtained under the humid purge gas condition. It is discussed that change of color of Li₂TiO₃ surface under hydrogen purge gas condition is observed and this phenomenon might affect tritium release behavior from Li₂TiO₃.     

Transfer of tritium in concrete coated with hydrophobic paints

An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (i) how fast HTO penetrates through concrete walls, (ii) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (iii) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The experiment is comparatively carried out using disks of cement paste which W(water)/C(cement) weight ratio is 0.6:1 with or without hydrophobic paints, and mortar disks which W/C/sand ratio is 0.6:1:2 with or without the paints. The hydrophobic paints tested in the present study are an epoxy polymer resin paint and an acrylic-silicon polymer resin one. After T exposure during specified time under a constant HTO vapor pressure in an acrylic box, the amount of water-soluble HTO on/in the disks is determined using a technique of H₂O dissolution during specified time. The results obtained here are summarized as follows: (1) HTO penetration in porous concrete can be correlated in terms of the effective diffusivity. (2) Its value in porous cement without coating is 1.2 × 10⁻¹¹ m²/s at 25 °C. (3) HTO penetrates only through pores in cement, and there is no path for HTO transfer in non-porous sand. (4) Rates of sorption and dissolution of HTO in disks of cement and mortar coated with the epoxy resin paint are correlated in terms of the effective diffusivity through the paint film which value is D_T = 1.0 × 10⁻¹⁶ m²/s. The rate-determining step is diffusion through the paint. (5) The epoxy resin paint works more effectively as an anti-HTO diffusion coating. (6) Another acrylic-silicon resin paint does not work well as anti-HTO diffusion coating. This may be because the hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H₂O vapor or liquid. (7) The HTO uptake inside the epoxy paint is greater than that of the silicon one. (8) The permeation reduction factor (PRF) of HTO for the epoxy paint at steady-state is expected large, if HTO vapor only contributes to diffusion. However, when concrete surfaces coated with the epoxy paint are under wet conditions, the PRF value becomes smaller. All those results can be used to estimate the effect of HTO soaking in concrete walls in case of accidental T release in a fusion reactor room and to decontaminate wastes of tritiated concrete.     

Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R&D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

Improved modelling of helium and tritium production for spallation targets

Reliable predictions of light charged particle production in spallation reactions are important to correctly assess gas production in spallation targets. In particular, the helium production yield is important for assessing damage in the window separating the accelerator vacuum from a spallation target, and tritium is a major contributor to the target radioactivity. Up to now, the models available in the MCNPX transport code, including the widely used default option Bertini-Dresner and the INCL4.2-ABLA combination of models, were not able to correctly predict light charged particle yields. The work done recently on both the intranuclear cascade model INCL4, in which cluster emission through a coalescence process has been introduced, and on the de-excitation model ABLA allows correcting these deficiencies. This paper shows that the coalescence emission plays an important role in the tritium and ³He production and that the combination of the newly developed versions of the codes, INCL4.5-ABLA07, now lead to good predictions of both helium and tritium cross-sections over a wide incident energy range. Comparisons with other available models are also presented.