Search results for: tritium

Authors: Igor P. Maximkin, Arkady A. Yukhimchuk, Victor V. Baluev, Igor L. Malkov, Rafael K. Musyaev, Damir T. Sitdikov, Alexey V. Buchirin, Vasily V. Tikhonov

Abstract:

Minimization of tritium diffusion leakage when developing devices handling tritium-containing media is key problems whose solution will at least allow essential enhancement of radiation safety and minimization of diffusion losses of expensive tritium. One of the ways to solve this problem is to use Al₂O₃ high-strength non-porous ceramics as a structural material of the bed body. This alumina ceramics offers high strength characteristics, but its main advantages are low hydrogen permeability (as against the used structural material) and high dielectric properties. The latter enables direct induction heating of an hydride-forming metal without essential heating of the pressure and containment vessel. The use of alumina ceramics and induction heating allows: - essential reduction of tritium extraction time; - several orders reduction of tritium diffusion leakage; - more complete extraction of tritium from metal hydrides due to its higher heating up to melting in the event of final disposal of the device. The paper presents computational and experimental results for the tritium bed designed to absorb 6 liters of tritium. Titanium was used as hydrogen isotope sorbent. Results of hydrogen realize kinetic from hydride-forming metal, strength and cyclic service life tests are reported. Recommendations are also provided for the practical use of the given bed type.

Keywords: aluminum oxide ceramic, hydrogen pressure, hydrogen isotope storage, titanium hydride

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Authors: Rodrigo Antunes, Olga Borisevich, David Demange

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In future nuclear fusion reactors, tritium self-sufficiency will be ensured by tritium (3H) production via reactions between the fusion neutrons and lithium. To favor tritium breeding, a neutron multiplier must also be used. Both tritium breeder and neutron multiplier will be placed in the so-called Breeding Blanket (BB). For the European Helium-Cooled Pebble Bed (HCPB) BB concept, the tritium production and neutron multiplication will be ensured by neutron bombardment of Li4SiO4 and Be pebbles, respectively. The produced tritium is extracted from the pebbles by purging them with large flows of He (~ 104 Nm3h-1), doped with small amounts of H2 (~ 0.1 vol%) to promote tritium extraction via isotopic exchange (producing HT). Due to the presence of oxygen in the pebbles, production of tritiated water is unavoidable. Therefore, the purging gas downstream of the BB will be composed by Q2/Q2O/He (Q = 1H, 2H, 3H), with Q2/Q2O down to ppm levels, which must be further processed for tritium recovery. A two-stage continuous approach, where zeolite membranes (ZMs) are followed by a catalytic membrane reactor (CMR), has been recently proposed to fulfil this task. The tritium recovery from Q2/Q2O/He is ensured by the CMR, that requires a reduction of the gas flow coming from the BB and a pre-concentration of Q2 and Q2O to be efficient. For this reason, and to keep this stage with reasonable dimensions, ZMs are required upfront to reduce as much as possible the He flows and concentrate the Q2/Q2O species. Therefore, experimental activities have been carried out at the Tritium Laboratory Karlsruhe (TLK) to test the separation performances of different zeolite membranes for H2/H2O/He. First experiments have been performed with binary mixtures of H2/He and H2O/He with commercial MFI-ZSM5 and NaA zeolite-type membranes. Only the MFI-ZSM5 demonstrated selectivity towards H2, with a separation factor around 1.5, and H2 permeances around 0.72 µmolm-2s-1Pa-1, rather independent for feed concentrations in the range 0.1 vol%-10 vol% H2/He. The experiments with H2O/He have demonstrated that the separation factor towards H2O is highly dependent on the feed concentration and temperature. For instance, at 0.2 vol% H2O/He the separation factor with NaA is below 2 and around 1000 at 5 vol% H2O/He, at 30°C. Overall, both membranes demonstrated complementary results at equivalent temperatures. In fact, at low feed concentrations ( ≤ 1 vol% H2O/He) MFI-ZSM5 separates better than NaA, whereas the latter has higher separation factors for higher inlet water content ( ≥ 5 vol% H2O/He). In this contribution, the results obtained with both MFI-ZSM5 and NaA membranes for H2/He and H2O/H2 mixtures at different concentrations and temperatures are compared and discussed.

Keywords: nuclear fusion, gas separation, tritium processes, zeolite membranes

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Authors: Gheorghe Ionita, Sebastian Brad, Ioan Stefanescu

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In any fusion device, tritium plays a key role both as a fuel component and, due to its radioactivity and easy incorporation, as tritiated water (HTO). As for the ITER project, to reduce the constant potential of tritium emission, there will be implemented a Water Detritiation System (WDS) and an Isotopic Separation System (ISS). In the same time, during operation of fission CANDU reactors, the tritium content increases in the heavy water used as moderator and cooling agent (due to neutron activation) and it has to be reduced, too. In Romania, at the National Institute for Cryogenics and Isotopic Technologies (ICIT Rm-Valcea), there is an Experimental Pilot Plant for Tritium Removal (Exp. TRF), with the aim of providing technical data on the design and operation of an industrial plant for heavy water depreciation of CANDU reactors from Cernavoda NPP. The selected technology is based on the catalyzed isotopic exchange process between deuterium and liquid water (LPCE) combined with the cryogenic distillation process (CD). This paper presents an updated review of activities in the field carried out in Romania after the year 2000 and in particular those related to the development and operation of Tritium Removal Experimental Pilot Plant. It is also presented a comparison between the experimental pilot plant and industrial plant to be implemented at Cernavoda NPP. The similarities between the experimental pilot plant from ICIT Rm-Valcea and water depreciation and isotopic separation systems from ITER are also presented and discussed. Many aspects or 'opened issues' relating to WDS and ISS could be checked and clarified by a special research program, developed within ExpTRF. By these achievements and results, ICIT Rm - Valcea has proved its expertise and capability concerning tritium management therefore its competence may be used within ITER project.

Keywords: ITER project, heavy water detritiation, tritium removal, isotopic exchange

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Authors: Rodrigo Antunes, Olga Borisevich, David Demange

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The Tritium Laboratory Karlsruhe (TLK) has identified zeolite membranes as most promising for tritium processes in the future fusion reactors. Tritium diluted in purge gases or gaseous effluents, and present in both molecular and oxidized forms, can be pre-concentrated by a stage of zeolite membranes followed by a main downstream recovery stage (e.g., catalytic membrane reactor). Since 2011 several membrane zeolite samples have been tested to measure the membrane performances in the separation of hydrogen and water vapor from helium streams. These experiments were carried out in the ZIMT (Zeolite Inorganic Membranes for Tritium) facility where mass spectrometry and cold traps were used to measure the membranes’ performances. The membranes were tested at temperatures ranging from 25 °C up to 130 °C, at feed pressures between 1 and 3 bar, and typical feed flows of 2 l/min. During this experimental campaign, several zeolite-type membranes were studied: a hollow-fiber MFI nanocomposite membrane purchased from IRCELYON (France), and tubular MFI-ZSM5, NaA and H-SOD membranes purchased from Institute for Ceramic Technologies and Systems (IKTS, Germany). Among these membranes, only the MFI-based showed relevant performances for the H2/He separation, with rather high permeances (~0.5 – 0.7 μmol/sm2Pa for H2 at 25 °C for MFI-ZSM5), however with a limited ideal selectivity of around 2 for H2/He regardless of the feed concentration. Both MFI and NaA showed higher separation performances when water vapor was used instead; for example, at 30 °C, the separation factor for MFI-ZSM5 is approximately 10 and 38 for 0.2% and 10% H2O/He, respectively. The H-SOD evidenced to be considerably defective and therefore not considered for further experiments. In this contribution, a comprehensive analysis of the experimental methods and results obtained for the separation performance of different zeolite membranes during the past four years in inactive environment is given. These results are encouraging for the experimental campaign with molecular and oxidized tritium that will follow in 2017.

Keywords: gas separation, nuclear fusion, tritium processes, zeolite membranes

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Authors: Longkui Zhu, Zhengcao Li

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High temperature gas-cooled reactors (HTGRs) are considered as one of the next-generation advanced nuclear reactors, in which porous nuclear graphite is used as neutron moderators, reflectors, structure materials, and cooled by inert helium. Radioactive tritium and ¹⁴C are generated in terms of reactions of thermal neutrons and ⁶Li, ¹⁴N, ¹⁰B impurely within nuclear graphite and the coolant during HTGRs operation. Currently, hydrogen and nitrogen diffusion behavior together with nuclear graphite microstructure evolution were investigated to minimize the radioactive waste release, using thermogravimetric analysis, X-ray computed tomography, the BET and mercury standard porosimetry methods. It is found that the peak value of graphite weight loss emerged at 573-673 K owing to nitrogen diffusion from graphite pores to outside when the system was subjected to vacuum. Macropore volume became larger while porosity for mesopores was smaller with temperature ranging from ambient temperature to 1073 K, which was primarily induced by coalescence of the subscale pores. It is suggested that the porous nuclear graphite should be first subjected to vacuum at 573-673 K to minimize the nitrogen and the radioactive 14°C before operation in HTGRs. Then, results on hydrogen diffusion show that the diffusible hydrogen and tritium could permeate into the coolant with diffusion coefficients of > 0.5 × 10⁻⁴ cm²·s⁻¹ at 50 bar. As a consequence, the freshly-generated diffusible tritium could release quickly to outside once formed, and an effective approach for minimizing the amount of radioactive tritium is to make the impurity contents extremely low in nuclear graphite and the coolant. Besides, both two- and three-dimensional observations indicate that macro and mesopore volume along with total porosity decreased with temperature at 50 bar on account of synergistic effects of applied compression strain, sharpened pore morphology, and non-uniform temperature distribution.

Keywords: advanced high temperature gas-cooled reactor, hydrogen and nitrogen diffusion, microstructure evolution, nuclear graphite, radioactive waste management

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Authors: Olga A. Momot, Boris I. Synzynys, Alla A. Oudalova

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Obninsk is situated in the Kaluga region 100 km southwest of Moscow on the left bank of the Protva River. Several enterprises utilizing nuclear energy are operating in the town. A special attention in the region where radiation-hazardous facilities are located has traditionally been paid to radioactive gas and aerosol releases into the atmosphere; liquid waste discharges into the Protva river and groundwater pollution. Municipal intakes involve 34 wells arranged 15 km apart in a sequence north-south along the foot of the left slope of the Protva river valley. Northern and southern water intakes are upstream and downstream of the town, respectively. They belong to river valley intakes with mixed feeding, i.e. precipitation infiltration is responsible for a smaller part of groundwater, and a greater amount is being formed by overflowing from Protva. Water intakes are maintained by the Protva river runoff, the volume of which depends on the precipitation fallen out and watershed area. Groundwater contamination with tritium was first detected in a sanitary-protective zone of the Institute of Physics and Power Engineering (SRC-IPPE) by Roshydromet researchers when realizing the “Program of radiological monitoring in the territory of nuclear industry enterprises”. A comprehensive survey of the SRC-IPPE’s industrial site and adjacent territories has revealed that research nuclear reactors and accelerators where tritium targets are applied as well as radioactive waste storages could be considered as potential sources of technogenic tritium. All the above sources are located within the sanitary controlled area of intakes. Tritium activity in water of springs and wells near the SRC-IPPE is about 17.4 – 3200 Bq/l. The observed values of tritium activity are below the intervention levels (7600 Bq/l for inorganic compounds and 3300 Bq/l for organically bound tritium). The risk has being assessed to estimate possible effect of considered tritium concentrations on human health. Data on tritium concentrations in pipe-line drinking water were used for calculations. The activity of 3H amounted to 10.6 Bq/l and corresponded to the risk of such water consumption of ~ 3·10-7 year-1. The risk value given in magnitude is close to the individual annual death risk for population living near a NPP – 1.6·10-8 year-1 and at the same time corresponds to the level of tolerable risk (10-6) and falls within “risk optimization”, i.e. in the sphere for planning the economically sound measures on exposure risk reduction. To estimate the chemical risk, physical and chemical analysis was made of waters from all springs and wells near the SRC-IPPE. Chemical risk from groundwater contamination was estimated according to the EPA US guidance. The risk of carcinogenic diseases at a drinking water consumption amounts to 5·10-5. According to the classification accepted the health risk in case of spring water consumption is inadmissible. The compared assessments of risk associated with tritium exposure, on the one hand, and the dangerous chemical (e.g. heavy metals) contamination of Obninsk drinking water, on the other hand, have confirmed that just these chemical pollutants are responsible for health risk.

Keywords: radiation-hazardous facilities, water intakes, tritium, heavy metal, health risk

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Authors: Long Wang, Yongjin Feng, Xiaofang Luo

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Tritium permeation through structural materials is a significant issue for fusion demonstration (DEMO) reactor blankets in terms of fuel cycle efficiency and radiological safety. Reduced activation ferritic (RAFM) steel CLF-1 is a prime candidate for the China’s CFETR blanket structural material, facing high permeability of hydrogen isotopes at reactor operational temperature. To confine tritium as much as possible in the reactor, surface modification of the steels including fabrication of tritium permeation barrier (TPB) attracts much attention. As a new alloy system, high entropy alloy (HEA) contains at least five principal elements, each of which ranges from 5 at% to 35 at%. This high mixing effect entitles HEA extraordinary comprehensive performance. So it is attractive to lead HEA into surface alloying for protective use. At present, studies on the hydrogen isotope permeability of HEA coatings is still insufficient and corresponding mechanism isn’t clear. In our study, we prepared three kinds of HEA coatings, including AlCrTaTiZr, (AlCrTaTiZr)N and (AlCrTaTiZr)O. After comprehensive characterization of SEM, XPS, AFM, XRD and TEM, the structure and composition of the HEA coatings were obtained. Deuterium permeation tests were conducted to evaluate the hydrogen isotope permeability of AlCrTaTiZr, (AlCrTaTiZr)N and (AlCrTaTiZr)O HEA coatings. Results proved that the (AlCrTaTiZr)N and (AlCrTaTiZr)O HEA coatings had better hydrogen isotope permeation resistance. Through analyzing and characterizing the hydrogen isotope permeation results of the corroded samples, an internal link between hydrogen isotope permeation behavior and structure of HEA coatings was established. The results provide valuable reference in engineering design of structural and TPB materials for future fusion device.

Keywords: high entropy alloy, hydrogen isotope permeability, tritium permeation barrier, fusion demonstration reactor

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Authors: Egor Atan Obeten, Abong Augustine Agwul, Bissong A. Samson

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The position of the aquifer, its protective capability, and the quality of the groundwater beneath the dumpsite were all investigated. The techniques employed were laboratory, tritium tagging, electrical resistivity tomography (ERT), and vertical electrical sounding (VES). With a maximum electrode spacing of 500 meters, fifteen VES stations were used, and IPI2win software was used to analyze the data collected. The resistivity map of the dumpsite was determined by deploying six ERT stations for the 2 D survey. To ascertain the degree of soil infiltration beneath the dumpsite, the tritium tagging method was used. Using a conventional laboratory procedure, groundwater samples were taken from neighboring boreholes and examined. The findings showed that there were three to five geoelectric layers, with the aquifer position being inferred to be between 24.2 and 75.1 meters deep in the third, fourth, and fifth levels. Siemens with values in the range of 0.0235 to 0.1908 for the load protection capacity were deemed to be, at most, weakly and badly protected. The obtained porosity values ranged from 44.45 to 89.75. Strong calculated values for transmissivity and porosity indicate a permeable aquifer system with considerable storativity. The area has an infiltration value between 8 and 22 percent, according to the results of the tritium tagging technique, which was used to evaluate the level of infiltration from the dumpsite. Groundwater samples that have been analyzed reveal levels of NO2, DO, Pb2+, magnesium, and cadmium that are higher than what the NSDWQ has approved. Overall analysis of the results from the above-described methodologies shows that the study area's aquifer system is porous and that contaminants will circulate through it quickly if they are contaminated.

Keywords: aquifer, transmissivity, dumpsite, groundwater

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Authors: Alexey Yu. Postnikov, Nikolay T. Kazakovskiy, Valery V. Mokrushin, Irina A. Tsareva, Andrey A. Potekhin, Valentina N. Golubeva, Yuliya V. Potekhina, Maxim V. Tsarev

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The development of tritium and deuterium targets for neutron tubes and generators is a part of the activities in All-Russia Research Institute of Experimental Physics (RFNC-VNIIEF). These items contain a metal substrate (for example, copper) with a titanium film with a few microns thickness deposited on it. Then these metal films are saturated with tritium, deuterium or their mixtures. The significant problem in neutron tubes and neutron generators is the characterization of substrate surface before a deposition of titanium film on it, and analysis of the deposited titanium film’s surface before hydrogenation and after a saturation of the film with hydrogen isotopes. The performance effectiveness of neutron tube and generator also depends on upon the quality parameters of the surface of the initial substrate, deposited metal film and hydrogenated target. The objective of our work is to study the target prototype samples, that have differ by various approaches to the preliminary chemical processing of a copper substrate, and to analyze the integrity of titanium film after its saturation with deuterium. The research results of copper substrate and the surface of deposited titanium film with the use of electron microscopy, X-ray spectral microanalysis and laser-spark methods of analyses are presented. The causes of surface defects appearance have been identified. The distribution of deuterium and some impurities (oxygen and nitrogen) along the surface and across the height of the hydrogenated film in the target has been established. This allows us to evaluate the composition homogeneity of the samples and consequently to estimate the quality of hydrogenated samples. As the result of this work the propositions on the advancement of production technology and characterization of target’s surface have been presented.

Keywords: tritium and deuterium targets, titanium film, laser-spark methods, electron microscopy

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Authors: Yury S. Shpanskiy, Boris V. Kuteev

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Development of a fusion-fission hybrid facility based on superconducting conventional tokamak DEMO-FNS runs in Russia since 2013. The main design goal is to reach the technical feasibility and outline prospects of industrial hybrid technologies providing the production of neutrons, fuel nuclides, tritium, high-temperature heat, electricity and subcritical transmutation in Fusion-Fission Hybrid Systems. The facility should operate in a steady-state mode at the fusion power of 40 MW and fission reactions of 400 MW. Major tokamak parameters are the following: major radius R=3.2 m, minor radius a=1.0 m, elongation 2.1, triangularity 0.5. The design provides the neutron wall loading of ~0.2 MW/m², the lifetime neutron fluence of ~2 MWa/m², with the surface area of the active cores and tritium breeding blanket ~100 m². Core plasma modelling showed that the neutron yield ~10¹⁹ n/s is maximal if the tritium/deuterium density ratio is 1.5-2.3. The design of the electromagnetic system (EMS) defined its basic parameters, accounting for the coils strength and stability, and identified the most problematic nodes in the toroidal field coils and the central solenoid. The EMS generates toroidal, poloidal and correcting magnetic fields necessary for the plasma shaping and confinement inside the vacuum vessel. EMC consists of eighteen superconducting toroidal field coils, eight poloidal field coils, five sections of a central solenoid, correction coils, in-vessel coils for vertical plasma control. Supporting structures, the thermal shield, and the cryostat maintain its operation. EMS operates with the pulse duration of up to 5000 hours at the plasma current up to 5 MA. The vacuum vessel (VV) is an all-welded two-layer toroidal shell placed inside the EMS. The free space between the vessel shells is filled with water and boron steel plates, which form the neutron protection of the EMS. The VV-volume is 265 m³, its mass with manifolds is 1800 tons. The nuclear blanket of DEMO-FNS facility was designed to provide functions of minor actinides transmutation, tritium production and enrichment of spent nuclear fuel. The vertical overloading of the subcritical active cores with MA was chosen as prospective. Analysis of the device neutronics and the hybrid blanket thermal-hydraulic characteristics has been performed for the system with functions covering transmutation of minor actinides, production of tritium and enrichment of spent nuclear fuel. A study of FNS facilities role in the Russian closed nuclear fuel cycle was performed. It showed that during ~100 years of operation three FNS facilities with fission power of 3 GW controlled by fusion neutron source with power of 40 MW can burn 98 tons of minor actinides and 198 tons of Pu-239 can be produced for startup loading of 20 fast reactors. Instead of Pu-239, up to 25 kg of tritium per year may be produced for startup of fusion reactors using blocks with lithium orthosilicate instead of fissile breeder blankets.

Keywords: fusion-fission hybrid system, conventional tokamak, superconducting electromagnetic system, two-layer vacuum vessel, subcritical active cores, nuclear fuel cycle

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Authors: Jie Yang, Qifu Cheng, Yafang Liu, Zhijie Gu

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Controlled emissions of effluent from nuclear power plants are an important means of ensuring environmental safety. In order to fully grasp the actual discharge level of nuclear power plant in China's nuclear power plant in the pressurized water reactor and heavy water reactor, it will use the global average nuclear power plant effluent discharge as a reference to the standard analysis of China's nuclear power plant environmental discharge status. The results show that the average normalized emission of liquid tritium in PWR nuclear power plants in China is slightly higher than the global average value, and the other nuclides emissions are lower than the global average values.

Keywords: radioactive effluent, HWR, PWR, nuclear power plant

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Authors: Chunduru Amareswara Prasad

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The main aim of this proposal it to reveal the secrets of the universe by accelerating neutrons. The proposal idea in its abridged version speaks about the possibility of making neutrons accelerate with help of thermal energy and magnetic energy under controlled conditions. Which is helpful in revealing the hidden secrets of the universe namely dark energy and in finding properties of Higgs boson. The paper mainly speaks about accelerating neutrons to near velocity of light in a LINAC, using magnetic energy by magnetic pressurizers. The center of mass energy of two colliding neutron beams is 94 GeV (~0.5c) can be achieved using this method. The conventional ways to accelerate neutrons has some constraints in accelerating them electromagnetically as they need to be separated from the Tritium or Deuterium nuclei. This magnetic gradient method provides efficient and simple way to accelerate neutrons.

Keywords: neutron, acceleration, thermal energy, magnetic energy, Higgs boson

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Authors: Adem Acır

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In this study, time dependent neutronic analysis of incineration of minor actinides of a Laser Fusion Inertial Confinement Fusion Fission Energy (LIFE) engine was performed. The calculations were carried out by using MCNP codes with ENDF/B.VI neutron data library. In the neutronic calculations, TRISO particles fueled with minor actinides with natural lithium coolant were performed. The natural lithium cooled LIFE engine used 10 % TRISO fuel minor actinides composition. Tritium breeding ratios (TBR) and energy multiplication factor (M) burnup values were computed as 1.46 and 3.75, respectively. The reactor operation time was calculated as ~ 21 years. The burnup values were obtained as ~1060 GWD/MT, respectively. As a result, the very higher burnup were achieved of LIFE engine.

Keywords: Monte Carlo, minor actinides, nuclear waste, LIFE engine

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Authors: Marta Malo, Carlota Soto, Carmen García-Rosales, Teresa Hernández

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SiC based composites are candidates for possible use as structural and functional materials in the future fusion reactors, the main role is intended for the blanket modules. In the blanket, the neutrons produced in the fusion reaction slow down and their energy is transformed into heat in order to finally generate electrical power. In the blanket design named Dual Coolant Lead Lithium (DCLL), a PbLi alloy for power conversion and tritium breeding circulates inside hollow channels called Flow Channel Inserts (FCIs). These FCI must protect the steel structures against the highly corrosive PbLi liquid and the high temperatures, but also provide electrical insulation in order to minimize magnetohydrodynamic interactions of the flowing liquid metal with the high magnetic field present in a magnetically confined fusion environment. Due to their nominally high temperature and radiation stability as well as corrosion resistance, SiC is the main choice for the flow channel inserts. The significantly lower manufacturing cost presents porous SiC (dense coating is required in order to assure protection against corrosion and as a tritium barrier) as a firm alternative to SiC/SiC composites for this purpose. This application requires the materials to be exposed to high radiation levels and extreme temperatures, conditions for which previous studies have shown noticeable changes in both the microstructure and the electrical properties of different types of silicon carbide. Both initial properties and radiation/temperature induced damage strongly depend on the crystal structure, polytype, impurities/additives that are determined by the fabrication process, so the development of a suitable material requires full control of these variables. For this work, several SiC samples with different percentage of porosity and sintering additives have been manufactured by the so-called sacrificial template method at the Ceit-IK4 Technology Center (San Sebastián, Spain), and characterized at Ciemat (Madrid, Spain). Electrical conductivity was measured as a function of temperature before and after irradiation with 1.8 MeV electrons in the Ciemat HVEC Van de Graaff accelerator up to 140 MGy (~ 2·10 -5 dpa). Radiation-induced conductivity (RIC) was also examined during irradiation at 550 ºC for different dose rates (from 0.5 to 5 kGy/s). Although no significant RIC was found in general for any of the samples, electrical conductivity increase with irradiation dose was observed to occur for some compositions with a linear tendency. However, first results indicate enhanced radiation resistance for coated samples. Preliminary thermogravimetric tests of selected samples, together with posterior XRD analysis allowed interpret radiation-induced modification of the electrical conductivity in terms of changes in the SiC crystalline structure. Further analysis is needed in order to confirm this.

Keywords: DCLL blanket, electrical conductivity, flow channel insert, porous SiC, radiation damage, thermal stability

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Authors: S. Visetpotjanakit, C. Khrautongkieo

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Radiation monitoring in the environment and foodstuffs is one of the main responsibilities of Office of Atoms for Peace (OAP) as the nuclear regulatory body of Thailand. The main goal of the OAP is to assure the safety of the Thai people and environment from any radiological incidents. Various radioanalytical methods have been developed to monitor radiation and radionuclides in the environmental and foodstuff samples. To validate our analytical performance, several proficiency test exercises from the International Atomic Energy Agency (IAEA) have been performed. Here, the results of a proficiency test exercise referred to as the Proficiency Test for Tritium, Cobalt, Strontium and Caesium Isotopes in Seawater 2017 (IAEA-RML-2017-01) are presented. All radionuclides excepting ³H were analysed using various radioanalytical methods, i.e. direct gamma-ray counting for determining ⁶⁰Co, ¹³⁴Cs and ¹³⁷Cs and developed radiochemical techniques for analysing ¹³⁴Cs, ¹³⁷Cs using AMP pre-concentration technique and 90Sr using di-(2-ethylhexyl) phosphoric acid (HDEHP) liquid extraction technique. The analysis results were submitted to IAEA. All results passed IAEA criteria, i.e. accuracy, precision and trueness and obtained ‘Accepted’ statuses. These confirm the data quality from the OAP environmental radiation laboratory to monitor radiation in the environment.

Keywords: international atomic energy agency, proficiency test, radiation monitoring, seawater

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Authors: S. Sandri, G. M. Contessa, C. Poggi

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An experimental nuclear fusion device could be seen as a step toward the development of the future nuclear fusion power plant. If compared with other possible solutions to the energy problem, nuclear fusion has advantages that ensure sustainability and security. In particular considering the radioactivity and the radioactive waste produced, in a nuclear fusion plant the component materials could be selected in order to limit the decay period, making it possible the recycling in a new reactor after about 100 years from the beginning of the decommissioning. To achieve this and other pertinent goals many experimental machines have been developed and operated worldwide in the last decades, underlining that radiation protection and workers exposure are critical aspects of these facilities due to the high flux, high energy neutrons produced in the fusion reactions. Direct radiation, material activation, tritium diffusion and other related issues pose a real challenge to the demonstration that these devices are safer than the nuclear fission facilities. In Italy, a limited number of fusion facilities have been constructed and operated since 30 years ago, mainly at the ENEA Frascati Center, and the radiation protection approach, addressed by the national licensing requirements, shows that it is not always easy to respect the constraints for the workers' exposure to ionizing radiation. In the current analysis, the main radiation protection issues encountered in the Italian Fusion facilities are considered and discussed, and the technical and legal requirements are described. The licensing process for these kinds of devices is outlined and compared with that of other European countries. The following aspects are considered throughout the current study: i) description of the installation, plant and systems, ii) suitability of the area, buildings, and structures, iii) radioprotection structures and organization, iv) exposure of personnel, v) accident analysis and relevant radiological consequences, vi) radioactive wastes assessment and management. In conclusion, the analysis points out the needing of a special attention to the radiological exposure of the workers in order to demonstrate at least the same level of safety as that reached at the nuclear fission facilities.

Keywords: fusion facilities, high energy neutrons, licensing process, radiation protection

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Authors: Gayathri Devi V, Aravamudan Kannan, Amit Sircar

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In the inner fuel cycle system of a nuclear fusion reactor, the Hydrogen Isotopes Removal System (HIRS) plays a pivoted role. It enables the effective extraction of the hydrogen isotopes from the breeder purge gas which helps to maintain the tritium breeding ratio and sustain the fusion reaction. One of the components of HIRS, Cryogenic Molecular Sieve Bed (CMSB) columns with zeolites adsorbents are considered for the physisorption of hydrogen isotopes at 1 bar and 77 K. Even though zeolites have good thermal stability and reduced activation properties making them ideal for use in nuclear reactor applications, their modest capacity for hydrogen isotopes adsorption is a cause of concern. In order to enhance the adsorbent capacity in an informed manner, it is helpful to understand the adsorption phenomena at the quantum electronic structure level. Physicochemical modifications of the adsorbent material enhances the adsorption capacity through the incorporation of active sites. This may be accomplished through the incorporation of suitable metal ions in the zeolite framework. In this work, molecular hydrogen isotopes adsorption on the active sites of functionalized zeolites are investigated in detail using Density Functional Theory (DFT) study. This involves the utilization of hybrid Generalized Gradient Approximation (GGA) with dispersion correction to account for the exchange and correlation functional of DFT. The electronic energies, adsorption enthalpy, adsorption free energy, Highest Occupied Molecular Orbital (HOMO), Lowest Unoccupied Molecular Orbital (LUMO) energies are computed on the stable 8T zeolite clusters as well as the periodic structure functionalized with different active sites. The characteristics of the dihydrogen bond with the active metal sites and the isotopic effects are also studied in detail. Validation studies with DFT will also be presented for adsorption of hydrogen on metal ion functionalized zeolites. The ab-inito screening analysis gave insights regarding the mechanism of hydrogen interaction with the zeolites under study and also the effect of the metal ion on adsorption. This detailed study provides guidelines for selection of the appropriate metal ions that may be incorporated in the zeolites framework for effective adsorption of hydrogen isotopes in the HIRS.

Keywords: adsorption enthalpy, functionalized zeolites, hydrogen isotopes, nuclear fusion, physisorption

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Authors: Nicole Virgili, Marco Utili

Abstract:

The design of the Breeding Blanket in Tokamak fusion energy systems has to guarantee sufficient availability in addition to its functions, that are, tritium breeding self-sufficiency, power extraction and shielding (the magnets and the VV). All these function in the presence of extremely harsh operating conditions in terms of heat flux and neutron dose as well as chemical environment of the coolant and breeder that challenge structural materials (structural resistance and corrosion resistance). The movement and activation of fluids from the BB to the Ex-vessel components in a fusion power plant have an important radiological consideration because flowing material can carry radioactivity to safety-critical areas. This includes gamma-ray emission from activated fluid and activated corrosion products, and secondary activation resulting from neutron emission, with implication for the safety of maintenance personnel and damage to electrical and electronic equipment. In addition to the PbLi breeder activation, it is important to evaluate the contribution due to the activated corrosion products (ACPs) dissolved in the lead-lithium eutectic alloy, at different concentration levels. Therefore, the purpose of the study project is to evaluate the PbLi activity utilizing the FISPACT II inventory code. Emphasis is given on how the design of the EU-DEMO WCLL, and potential recycling of the breeder material will be impacted by the activation of PbLi and the associated active corrosion products (ACPs). For this scope the following Computational Tools, Data and Geometry have been considered: • Neutron source: EU-DEMO neutron flux < 1014/cm2/s • Neutron flux distribution in equatorial breeding blanket module (BBM) #13 in the WCLL BB outboard central zone, which is the most activated zone, with the aim to introduce a conservative component utilizing MNCP6. • The recommended geometry model: 2017 EU DEMO CAD model. • Blanket Module Material Specifications (Composition) • Activation calculations for different ACP concentration levels in the PbLi breeder, with a given chemistry in stationary equilibrium conditions, using FISPACT II code. Results suggest that there should be a waiting time of about 10 years from the shut-down (SD) to be able to safely manipulate the PbLi for recycling operations with simple shielding requirements. The dose rate is mainly given by the PbLi and the ACP concentration (x1 or x 100) does not shift the result. In conclusion, the results show that there is no impact on PbLi activation due to ACPs levels.

Keywords: activation, corrosion products, recycling, WCLL BB., PbLi

Procedia PDF Downloads 122

Authors: Monica Jong, Antonios Banos, Tom Scott, Chris Webster, David Fletcher

Abstract:

Hydrogen, as a clean alternative to methane, is relatively easy to make, either from water using electrolysis or from methane using steam reformation. However, hydrogen is much trickier to store than methane, and without effective storage, it simply won’t pass muster as a suitable methane substitute. Physical storage of hydrogen is quite inefficient. Storing hydrogen as a compressed gas at pressures up to 900 times atmospheric is volumetrically inefficient and carries safety implications, whilst storing it as a liquid requires costly and constant cryogenic cooling to minus 253°C. This is where DU steps in as a possible solution. Across the periodic table, there are many different metallic elements that will react with hydrogen to form a chemical compound known as a hydride (or metal hydride). From a chemical perspective, the ‘king’ of the hydride forming metals is palladium because it offers the highest hydrogen storage volumetric capacity. However, this material is simply too expensive and scarce to be used in a scaled-up bulk hydrogen storage solution. Depleted Uranium is the second most volumetrically efficient hydride-forming metal after palladium. The UK has accrued a significant amount of DU because of manufacturing nuclear fuel for many decades, and that is currently without real commercial use. Uranium trihydride (UH3) contains three hydrogen atoms for every uranium atom and can chemically store hydrogen at ambient pressure and temperature at more than twice the density of pure liquid hydrogen for the same volume. To release the hydrogen from the hydride, all you do is heat it up. At temperatures above 250°C, the hydride starts to thermally decompose, releasing hydrogen as a gas and leaving the Uranium as a metal again. The reversible nature of this reaction allows the hydride to be formed and unformed again and again, enabling its use as a high-density hydrogen storage material which is already available in large quantities because of its stockpiling as a ‘waste’ by-product. Whilst the tritium storage credentials of Uranium have been rigorously proven at the laboratory scale and at the fusion demonstrator JET for over 30 years, there is a need to prove the concept for depleted uranium hydrogen storage (HyDUS) at scales towards that which is needed to flexibly supply our national power grid with energy. This is exactly the purpose of the HyDUS project, a collaborative venture involving EDF as the interested energy vendor, Urenco as the owner of the waste DU, and the University of Bristol with the UKAEA as the architects of the technology. The team will embark on building and proving the world’s first pilot scale demonstrator of bulk chemical hydrogen storage using depleted Uranium. Within 24 months, the team will attempt to prove both the technical and commercial viability of this technology as a longer duration energy storage solution for the UK. The HyDUS project seeks to enable a true by-product to wonder material story for depleted Uranium, demonstrating that we can think sustainably about unlocking the potential value trapped inside nuclear waste materials.

Keywords: hydrogen, long duration storage, storage, depleted uranium, HyDUS

Procedia PDF Downloads 147