Mathematical models for tritium permeation analysis in liquid metal flows with helium bubbles

In LIBRETTO-2 test, evidence was obtained that helium bubbles nucleated and grew in the neutron irradiated PbLi probes. If such phenomenon occurs inside liquid metal (LM) breeding blanket channels, the study of its effect on tritium permeation and heat transfer in the near wall region will acquire utmost importance. The T4F research group has developed in the past a nucleation, growth and transport model for helium bubbles in LM flows, as well as a tritium transport model in such a multi-fluid system. In the present study, we are focused on the near-wall region analysis in order to obtain a wall function that allow reproducing the tritium permeation with coarse meshes and, hence, reduce the computational time. First, we perform some detailed CFD simulations of the near-wall region where bubbles might be attached. In these simulations, tritium diffusion processes as well as tritium recombination and dissociation are modelled. The analysis of such simulations allows us to further understand the complex phenomena and justify the use of simplified models. As a result, a new model for tritium transport across a LM–solid interface partially covered by helium bubbles is developed, implemented and validated. This simplified model can be seen as a wall function for the CFD simulation which substantially reduces computational time.     

Option for spallation neutron sources

Spallation reactions are a very important option for efficient neutron sources appropriate for fusion materials testing. An option of this option is the EURAC concept, which makes use of short-term accelerator technology in the cheapest way and is proved to provide the needed neutron flux to verify fast experiments on fusion materials performance. Its flexible conception allows an optimum combination of very high fluxes of about 10¹⁶ n/cm²/s, with decreasing fluxes along the testing zones in enough volume to perform the correct irradiations. With this assumption, the rate effect can be perfectly analyzed together with the end-of-life conditions assumed in the structural material of the future fusion environments. The possible negative effects of the high-energy neutrons in the Spallation spectrum have been taken into account, concluding their non-significance in the desired damage parameters. The EURAC concept can also be considered in light of other purposes like incineration processes, production, and, with the appropriate booster, high-flux cold neutron source.     

Feasibility of uranium spallation target in accelerator-driven system

A feasibility study on natural uranium spallation target in accelerator-driven system (ADS) for minor actinide (MA) transmutation was performed. As a result of comparative study of uranium and lead-bismuth (PbBi) targets in the bare case without blanket surrounding, it was found that uranium target had better neutron generation performance, but limited by the geometrical size due to high neutron absorption in ²³⁸U. In ADS for MA transmutation, uranium used as target instead of PbBi also absorbs neutrons passing the target area.More realistic concept of pin type uranium spallation target cooled by liquid PbBi was considered aiming at enhancing spallation target performance in terms of neutron generation efficiency and operation temperature. The uranium pin target design had nothing better effects on neutron balance of such system than a conventional PbBi target in ADS and it was concluded that uranium target was not suitable for the full-scale ADS.     

Analysis of tritium production in a 2 MW liquid-fueled molten salt experimental reactor and its environmental impact

Tritium release is one of the most concerning topics in nuclear power plants. Here, the tritium production in a 2 MW liquid-fueled molten salt experimental reactor(TMSR-LF1) was calculated by ORIGEN-S with an updated cross-section library generated by TRITON in SCALE 6.1.3 code system. The results show that the tritium production rate and normalized tritium production rate of TMSR-LF1 are 8.90x10~(11)Bq/day and4.45x10~(11)Bq/(MW day), respectively. The environmental impact of tritium was analyzed via PC-CREAM 08 with an assumed 36 % release rate of tritium referring to the molten salt reactor experiment. During normal operations, the maximum tritium concentration is 1.4 Bq/m~3 under normal condition, and the corresponding individual dose to the public is about 1 μSv/a; under extreme conditions, the maximum concentration and corresponding individual doses are 11.8 Bq/m~3 and 9 μSv/a, respectively.Ingestion is the main exposure pathway and accounts for62 % of the total dose. Of this, 35 % is from organically bound tritium.     

产氚回路中氚的在线监测研究

描述一套产氚回路中氚的在线监测装置——流气式正比计数系统。测试该系统的坪曲线、本底、探测效率等性能后,开展了氚在线监测实验研究。结果表明,该系统性能稳定,控制好测量条件,可作产氚回路中氚的在线监测设备。     

Investigation of neutronic parameters of natU spallation target irradiated by low-energy protons

Accelerator-based neutron sources could outstandingly compete with the reactor-based ones, which are widely used for research aims and radioisotope production.Spallation neutron sources are used by many research centers. In this work, the potential of natural uranium spallation target irradiated by low-energy protons for production of an external neutron source was investigated.MCNPX code was used to model the spallation target. The results showed using 30-Me V protons of 100 μA current a neutron flux in order of 10~7n/s cm~2 leaks from an optimized-dimension target. Different physical models available in the computational code do not result in significant relative discrepancies for neutron yield and deposited heat calculations. Water with a velocity of 0.6 m/s can be used as coolant for the spallation target to keep the surface temperature under 100 °C at atmospheric pressure.     

Measurement of tritium production rate distribution in natural LiAlO2/HDPE assembly irradiated by D-T neutrons

A neutronics experiment was performed to measure the tritium production rate (TPR) profile in the breeder assembly with LiAlO₂ as breeder and high density polyethylene (HDPE) as neutron reflector. The breeder assembly was irradiated with 14 MeV neutrons from DT neutron generator at IPR Neutronics Laboratory. The objective of the experiment was to validate the tritium production prediction capability of the Monte-Carlo code MCNP and FENDL 2.1 data library. The tritium production rate profile in the breeding assembly was measured by irradiating Li₂CO₃ pellets kept at various locations and then tritium counting liquid scintillation technique. Experiment was analyzed with 3D Monte-Carlo code MCNP with FENDL 2.1 cross-section data library. The calculation results were found to agree with the measured tritium production rates except one point near to the source. This experiment is a starting experiment in the series of benchmark experiments for the Indian Demo breeding blanket.     

用DCE/CEM和LAHET软件计算ADS标准散裂靶的比较

利用蒙卡程序DCM／CEM和LAHET对加速器驱动的次临界系统ADS标准散裂中子靶进行了计算。长为0．6m、直径为0．2m的圆柱形^208Pb靶在0．15—1．6GeV的高能质子轰击的情况下,利用两软件对质子在靶内的能量沉积、Pb靶发生散裂反应产生的中子及靶内和表面的中子注量、中子能谱分布等进行了比较计算。结果表明：两者模拟结果在中能区(0．8—1．0GeV)符合很好,但在质子能量较高或较低时,两者略有差别。     

Neutronic performance of the MEGAPIE spallation target under high power proton beam

The MEGAPIE project, aiming at the construction and operation of a megawatt liquid lead-bismuth spallation target, constitutes the first step in demonstrating the feasibility of liquid heavy metal target technologies as spallation neutron sources. In particular, MEGAPIE is meant to assess the coupling of a high power proton beam with a window-concept heavy liquid metal target. The experiment has been set at the Paul Scherrer Institute (PSI) in Switzerland and, after a 4-month long irradiation, has provided unique data for a better understanding of the behavior of such a target under realistic irradiation conditions. A complex neutron detector has been developed to provide an on-line measurement of the neutron fluency inside the target and close to the proton beam. The detector is based on micrometric fission chambers and activation foils. These two complementary detection techniques have provided a characterization of the neutron flux inside the target for different positions along its axis. Measurements and simulation results presented in this paper aim to provide important recommendations for future accelerator driven systems (ADS) and neutron source developments.     

Preliminary design of inner radius of spallation target of an accelerator driven subcritical reactor for waste transmutation (ADS-NWT)

This paper presented preliminary design studies of inner radius of spallation target for lead-bismuth cooled accelerator driven sub-critical reactor, aimed at reducing proton beam current while simultaneously maintaining high Minor Actinides transmutation performance. Lead-bismuth eutectic (LBE) and uranium-free metallic dispersion fuel were used as the coolant and the fuel for this 1000 MWt reactor. By taking Proton Transmutation Support Ratio (PTSR) as an evaluation index, the inner radius of the spallation target was designed for this reference reactor and the optimized design proposal could obtain the higher transmutation performance. The preliminary study found that when the inner radius of the spallation target ran up to a certain value, the spallation neutron yield would have greater influence on the transmutation performance.     

大米样品总氚与其组织自由水氚的相关性

大米样品的总氚(TT)与其组织自由水氚(TFWT)具有非常显著的相关性,其线性回归方程为TT=1.0671+1.3605TFWT。该方程的应用为大米样品总氚测定提供了一个简便的方法。     

The integral experiment on beryllium with D-T neutrons for verification of tritium breeding

A clean benchmark experiment on beryllium was performed with D-T neutrons at the FNS facility of the Japan Atomic Energy Agency. The main objective was to verify the integral data related to the tritium production on lithium isotopes. Tritium production rates, as well as activation reaction rates were measured inside the beryllium assembly that was shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628 mm and a thickness of 355 mm. Experimental results were analyzed with a three-dimensional Monte Carlo transport code MCNP-4C and FENDL/MC-2.0, JENDL-3.2/3.3 neutron transport libraries. Evaluation of reaction rates was based on the cross section data taken from the JENDL Dosimetry File and ENDF B-VI data libraries. Analysis shows that all calculation combinations (transport and activation cross section libraries) used for evaluation of reaction rates give data that is agreeable with measured values within 10%.     

基于铝箔封装碳酸锂探测片产氚率测量的液闪样品制备方法

氚增殖包层中产氚率的测量是聚变核能系统中需要研究的重要问题之一,本文开展了用于产氚率测量的Al箔封装碳酸锂探测片液闪样品制备化学处理方法的研究。结果表明,首先采用氢氧化钠溶液来溶解Al箔,然后再用盐酸溶解碳酸锂探测片的溶解方式,能制成透明度高且无分层的液闪样品。为了提高测氚计数效率和保证样品兼容性,对20 m L的标准液闪样品,闪烁液体积应至少取12 m L,同时还应将液闪样品保持在10-20°C范围进行储存和测量。     

Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different ⁶Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.     

Determination of the concentrations of tritium and helium in titanium tritide targets

Atom ratios of T to Ti in T-Ti targets have been determined by the ⁴He⁺ Rutherford backscattering (RBS) and the proton elastic scattering (PES) techniques, and that of ³He to T in aged T-Ti targets by the ³He(d, p)⁴He and T(d, n)⁴He reactions. The results from RBS show that the depth distribution of T is uniform in a region of 8000 Å under the target surface. For the results from the nuclear reaction analysis (NRA), a discrepancy between the experimentally measured and the calculated concentration of ³He in a heavily bombarded T-Ti target obviously exists.     

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。     

氦在钛靶及Al2O3中的行为研究

报道了利用质子增强背散射和扫描电镜等方法，研究不同工艺制作的钛膜靶和不同纯度的Ａｌ２Ｏ３中注入的＾４Ｈｅ在室温下的行为。研究结果表明，离子束增强淀积新工艺制备的钛膜与基体的结合强度优于原来热蒸发镀制的钛靶；氦在Ａｌ２Ｏ３与在钛靶中具有不同的释放机制；钛靶的起泡注入剂量在４×１０＾１７－７×１０＾１７Ｈｅ／ｃｍ＾２之间，而且起泡剂量与泡破裂剂量相差不大；较之新工艺制备的钛靶，氦在Ａｌ２Ｏ３中更不易形     

Neutron production in bombardments of thin and thick W, Hg, Pb targets by 0.4, 0.8, 1.2, 1.8 and 2.5 GeV protons

Neutron experimental data relevant to the design of the target of neutron spallation sources are presented and discussed. The data include the reaction cross-sections for W, Hg and Pb investigated with 0.4, 0.8, 1.2, 1.8 and 2.5 GeV proton beams as well as the neutron production, neutron multiplicity distribution, as determined event per event using a high-efficiency detector. The production as a function of target material is investigated for both thin (with a single reaction) and thick targets (multiple reactions). Comparisons are made with the predictions of a high-energy transport code.     

散裂中子源靶站和中子散射谱仪的概念设计

本文介绍了可应用于多学科应用的散裂中子源(CSNS)靶站和中子散射谱仪概念设计的进展。CSNS靶站将由重水冷却多片钨靶,铍/铁反射体和铁/重混凝土生物屏蔽体组成。采用三个WING型慢化器:水(室温),液体甲烷(100K)和液体氢(20K),设有18个水平中子孔道。MonteCarlo模拟显示优化的靶截面高宽比为1:2.5左右。额定的100kW核功率的质子束轰击后,慢化器处钨靶溢出的脉冲中子通量约为2.4×1016cm?2·s?1。有限元方法计算表明,钨靶体内的总发热量是47kJ/s。即使使用截面较小的钨靶,在通常的水冷速率下,靶体温度也仅略高于90°C。靶体的热应力形变最大不超过0.2mm。根据经济实用原则选择建造粉末衍射仪、小角散射仪、反射仪及直接几何非弹性散射仪等四类有代表性的中子散射谱仪,就能覆盖>80%的中子散射研究领域。     

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.