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.     

Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm³. Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10⁻¹³–1.8 × 10⁻¹³ m²/s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.     

用于含氚废气的无机载体疏水催化剂研制

为处理含氚废气筛选无机疏水载体,通过酸萃、高温灼烧对载体改性,经附铂、焙烧、还原后制备无机载体型Pt、Pd催化剂,常温下氢(氚)气的催化氧化实验中,几种催化剂表现相似的催化性能,氚气的单程转化率均大于90%,均可用于含氚废气处理。XRD测试结果显示,载体改性提高了载体的结晶度,有利于增加载体疏水性,降低附在载体上的铂晶粒尺寸,提高铂的分散度,但附铂后灼烧再还原的方法降低了铂的分散度。ZrO2的加入降低了钯晶粒尺寸,提高了钯的分散度。     

Study on behavior of tritium in concrete wall

It is required to understand the tritium behavior in concrete for establishment of tritium safety technology of a fusion reactor or a tritium handling facility because the concrete wall is used as the final containment to prevent tritium release to the environment. This paper discusses about the effect of adsorption and diffusion of water and isotope exchange reaction between physically adsorbed water and chemically adsorbed water or structural water. It is known in this study that a large amount of tritium can be trapped to the concrete wall because cement paste has the nature of porous hydrophilic material.     

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。     

Effect of tritium reduction in determining energy gain by using R-matrix method direct laser fusion in D-T reaction

The laser fusion criterion is known as the ρR-Criterion, also called high-gain condition. This parameter is temperature dependent and can be calculated by R-matrix method. This method is applied for determining improved fusion cross-section for the reactions T(d,n)4He, 3He(d,p)4He, D(d,p)T, D(d,n)3He. In this paper the time dependent reaction rate equations for fusion reaction T(d,n)4He are solved and by using the obtained results we computed the fu- sion power density, energy gain versus temperature and ρR-parameter. The obtained results show that a suitable com- bination may be a deuterium fraction fD=0.65 and fT=0.35 which would lead 30% reduction in the tritium content of the fuel mixture, and this choice would not change the energy gain value very much. Finally, the obtained energy gain for D-T reaction by using R-matrix is in good agreement with other theories.     

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.     

Microstructure and property of SiC coating for carbon materials

SiC coating has been developed on four kinds of carbon substrates using a slurry-sintering method. The relationship between the microstructure and property of SiC coating and carbon substrates was investigated experimentally and theoretically. It was found that the pore radius of carbon substrates has marked effect on the microstructure and property of SiC coating. SiC gradient coating which is beneficial to improve the oxidation resistance for carbon materials, is expected to form on the surface of carbon substrate with the pore radius mainly over the range of 100-500 nm.     

Effect of outside tritium source on tritium balance of a D-T fusion reactor

Attainable tritium breeding ration in the blanket system must be larger than the required breeding ratio when no effective tritium resources from outside are expected. It is revealed recently that a considerable amount of tritium can be trapped to the re-deposition layer of the first wall materials and that the time constant of this phenomenon is rather long. Then, the tritium breeding ratio around 1.1 is required in the blanket system when 3 years is claimed for the tritium doubling time to prepare tritium for the initial inventory of a next reactor. Construction of an outside tritium supply is one of the possible ways to compensate the lack of tritium because it is generally considered that the attainable tritium breeding ratio in the solid breeder system is around 1.05. It is reported recently that a high-temperature gas-cooled reactor can produce 10 kg of tritium per year. The preferable amount of tritium production rate of the outer tritium supply is discussed in this study from the viewpoint of tritium balance in a D-T power reactor.     

Experimental study of efficiency of natural oxide layers for reduction of tritium permeation through Eurofer 97

Tritium permeation from the breeder through the helium coolant is a fundamental safety issue in the design of the HCLL (Helium coolant lithium lead) blanket system. The permeation of hydrogen isotopes through Eurofer in different conditions was deeply studied in the past, demonstrating that it is necessary to reduce this amount using tritium permeation barriers (TPB). A strong effort has been made to select the best technological solution for the realisation of tritium permeation barriers on complex structures not directly accessible after the completion of the manufacturing process, but after many years of activity for the qualification of different materials acting as TPBs it was demonstrated that these technologies are not yet mature for nuclear applications. An easier solution was identified in the nucleation and growth of natural oxides on the helium-exposed surface of cooling system components. The major objective of this work is the evaluation of the Permeation Reduction Factor (PRF) of natural oxides on Eurofer steel adding a known content of water and hydrogen to argon, used in substitution of helium. The PRF was measured on disk shaped specimens, in gas phase, using the PERI II apparatus, at a temperature of 550 °C. The oxide layer was produced in situ, in a well-defined range of hydrogen on water ratios. The obtained results are presented and discussed.     

Comparison of reconstruction methods of depth distribution of tritium in materials based on BIXS

Three reconstruction methods based on Tikhonov regularization are used to solve the ill-posed inverse problem when incorporating Monte Carlo calculations into the β-ray induced X-ray spectrometry (BIXS) for analyzing the tritium content and depth distribution in the material surface. These methods are implemented by the GENEREG, CONTIN and REGUTOOLS codes, respectively. The tritium sample is assumed to be a molybdenum slab of 5 μm thickness, and an argon gas of 5 mm thickness is filled in front of the molybdenum slab. The Monte Carlo code PENELOPE is used to simulate the X-ray energy spectra induced by tritium’s β-decays. The reconstruction results for tritium depth distributions by these three codes are compared and the comparison shows that the code GENEREG can give satisfactory results. Some discussions are made.     

Chemical thermodynamics of fusion reactor breeding materials and their interaction with tritium

Liquid lithium, lithium alloys (solid and liquid) and ceramic lithium compounds are candidate breeding materials for (D,T) fusion reactors.Besides their tritium breeding capability, which results from neutron capture, their thermochemical properties and their interaction with tritium are of particular interest. A good knowledge of the physical and chemical properties of liquid lithium exists; and the systems Li-LiH, Li-LiD and Li-LiT have been studied in great detail. For dilute solutions of D₂ in liquid lithium, Sieverts' law was found to be valid down to an atom fraction of $\text{x}{}_{\mathrm{D}}\text{= 10}{}^{-6}$; in the vapor, lithium polymers up to Li₄ and lithium deuterides are found.In the system liquid Li-Pb, the solubility of D₂ was measured as a function of temperature and alloy composition, and correlated with the activities of the constituent metals. The solubility of D₂ was found to obey Sieverts' law at low concentrations, and is many orders of magnitude smaller than that in liquid lithium. This holds also for solid “Li₇ Pb₂”.Vaporization studies yielded data on the thermal stability of the oxides: Li₂0, γ-LiAlO₂, β-Li_sAlO₄, LiAl₅O₈, Li₂ZrO₃, Li₄ZrO₄, Li₈ZrO₆, Li₂SiO₃ and Li₄SiO₄. Tritium diffusivity was studied in Li₂O, γ-LiAlO₂, β-Li₅AlO₄ and Li₄SiO₄. A large number of gaseous lithides were detected during these studies.     

Neutronic evaluation of coating and cladding materials for accident tolerant fuels

In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with high-temperature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10–30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.     

Repair materials and techniques for concrete structures in nuclear power plants

The paper summaries portions of work of the Structural Aging Program, sponsored by the Nuclear Regulatory Commission (NRC). The paper addresses the assessment and repair of concrete structures in nuclear power plants. It presents the results of a survey of the the nuclear power plants in the United States to identify susceptible concrete components, rates of occurrence of deterioration, and to determine the durability of repairs. The paper describes deterioration mechanisms and discusses their effect. Repair techniques are described. Evaluation techniques and nondestructive test techniques are also discussed.     

An advance process of aluminum rich coating as tritium permeation barrier on 321 steel workpiece

We have proposed an advance three-step process, Al-electroplating in ionic liquid followed by heat treating and selectively oxidation, preparing aluminum rich coating as tritium permeation barrier (TPB). In present work, the advance process was applied to 321 steel workpieces. In the Al-electroplating, pieces were coated by galvanostatic electrodeposition at 20 mA/cm² in aluminum chloride (AlCl₃)–1-ethyl-3-methylimidazolium chloride (EMIC) ionic liquid. The Al coating on those pieces all displayed attractive brightness and well adhered to surface of pieces. Within the aluminizing time from 1 to 30 h, a series of experiments were carried out to aluminize 321 steel pieces with Al 20 μm coating at 700 °C. After heat treated for 8 h, a 30 μm thick aluminized coating on piece appeared homogeneous, free of porosity, and mainly consisted of (Fe, Cr, Ni)Al₂, and then was selectively oxidized in argon gas at 700 °C for 50 h to form Al₂O₃ scale. The finally fabricated aluminum rich coating, without any visible defects, had a double-layered structure consisting of an outer γ-Al₂O₃ layer with thickness of 0.2 μm and inner (Fe, Cr, Ni)Al/(Fe, Cr, Ni)₃Al layer of 50 μm thickness. The deuterium permeation reduction factor, PRF, of piece (Φ 80 × 2, L 150 mm) with such coating increased by 2 orders of magnitude at 600–727 °C. The reproducibility of the process was also showed.     

氚气在弹性复合材料中的溶解、扩散和渗透性能的测量

本文根据氚气具有很强渗透性能及其在不同材料中渗透性能的差异,用正比计数测量技术对氚气在各种材料中的渗透性能进行测量、比较,从中选出最佳复合形式,提供生产制造防氚手套。     

Development of the models for advection-diffusion of eroded concrete into debris and concrete volume reduction in molten core-concrete interactions

Models for the three-dimensional (3D) advection, diffusion, and volume reduction of eroded concrete into molten core are being developed. As part of the assessment of the reactor interior at TEPCO's Fukushima Daiichi Nuclear Power Plant, analytical models of molten core–concrete interaction (MCCI) to predict locations and condition of molten core (debris) have been improved in the debris spreading analysis (DSA) module of the severe accident analysis code SAMPSON. In addition to the primary model for 3D natural convection with simultaneous spreading, melting, and solidification in an open space, the analysis model to treat phenomena in a closed space, such as debris eroding laterally under concrete floors at the bottom of the sump pits, had been improved. This modeling with practical applicability is referred to as the full-3D MCCI model. This paper presents modeling of the advection and diffusion of eroded concrete into debris melt and calculation processes that were installed for simulating volume reduction when concrete decomposed. They were developed and incorporated into the full-3D MCCI model. The advanced DSA module with the models noted above was validated using MCCI test data. The calculated erosion rates agreed with the test data within a margin of about 16%.     

用氢同位素估算氚在熔盐堆结构材料中的渗透

熔盐堆产生的氚在高温下透过结构材料管壁进入大气环境。为降低此危害,研究氚在堆结构材料中的渗透过程十分必要。针对熔盐堆常用的结构材料Hastelloy N、GH3535、Hastelloy C230、Hastelloy C276,本研究采用压力差驱动法,在400-700 oC和5-40 k Pa的试验条件下,获得了氕、氘的渗透系数,并初步估算氚在该结构材料中的渗透。结果表明:氕、氘在镍基合金中的渗透通量与气体压力的平方根成正比,渗透系数随温度增大,与温度倒数的关系符合阿伦尼乌斯公式;氕、氘在成分相似的GH3535和Hastelloy N中渗透系数接近,与其它两种合金材料中渗透系数都在一个数量级内,在相同温度下两者渗透系数比值接近1.4,符合经典扩散理论;在400-700 oC内,氚在4种镍金合金中渗透过程的指前因子为1.1×10-7-1.6×10-7 mol·m-1·s-1·Pa-1/2,活化能为59-62 k J·mol-1。     

Corrosion behavior of Ni-based structural materials for electrolytic reduction in lithium molten salt

In this study, the corrosion behavior of new Ni-based structural materials was studied for electrolytic reduction after exposure to LiCl-Li₂O molten salt at 650 °C for 24-216 h under an oxidizing atmosphere. The new alloys with Ni, Cr, Al, Si, and Nb as the major components were melted at 1700 °C under an inert atmosphere. The melt was poured into a preheated metallic mold to prepare an as-cast alloy. The corrosion products and fine structures of the corroded specimens were characterized by scanning electron microscope (SEM), Energy Dispersive X-ray Spectroscope (EDS), and X-ray diffraction (XRD). The corrosion products of as cast and heat treated low Si/high Ti alloys were Cr₂O₃, NiCr₂O₄, Ni, NiO, and (Al,Nb,Ti)O₂; those of as cast and heat treated high Si/low Ti alloys were Cr₂O₃, NiCr₂O₄, Ni, and NiO. The corrosion layers of as cast and heat treated low Si/high Ti alloys were continuous and dense. However, those of as cast and heat treated high Si/low Ti alloys were discontinuous and cracked. Heat treated low Si/high Ti alloy showed the highest corrosion resistance among the examined alloys. The superior corrosion resistance of the heat treated low Si/high Ti alloy was attributed to the addition of an appropriate amount of Si, and the metallurgical evaluations were performed systematically.