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The channel pipes of the control and safety systems of RBMK-1000 reactors are made of a Zr–2.5% Nb (E125) alloy and must operate for 45 years. The evolution of the alloy structure under the action of long-term low-temperature neutron irradiation is studied. The grain structure of the alloy is shown to be stable; annealing of initial dislocations and the formation of radiation-induced defect take place. Radiation-induced diffusion processes, which lead to a decrease in the niobium content in β-Nd precipitates and the appearance of new fine-grained precipitates in the alloy matrix, are revealed.

An electrodeposited thorium target and a ThO2 sample were irradiated together with fast neutrons using the Chinese Fast Burst Reactor II. The activities of 233Pa in both samples were measured and c...

A modern approach to nuclear energy involves reprocessing like transmutations of spent nuclear fuel products to reduce their radiotoxicity and time needed for their storage. For this purpose, they are immobilized in inert matrices made of zirconia and can be "burned" in fast neutron reactor or Accelerator Driven System. These matrices in spherical form can be obtained by sol-gel process. The paper presents a method of microspheres fabrication based on the combined Complex Sol-Gel Process and double extraction process consisting in the preparation of zirconium-ascorbate sol and simultaneous extraction of water and nitrates. The procedure allows obtaining gel microspheres with a diameter of 50 μm, which after heat treatment are processed into the final product. The synthesis of zirconia microspheres with Yttrium by internal gelation process is well known for over a decade now. However, the explanation and characterization of synthesis of such material by extraction of water process is rarely found. Parameters such as: pH, viscosity, shape, sphericity and crystal structure have been determined for synthesized products and semi-products. In addition, preliminary research consisting in irradiation of the obtained materials in fast and thermal neutron flux was carried out. The obtained results are presented and described in this work.

TEM, SEM, and APT techniques have been used to analyze radiation-induced components of metal structure of fragments cut from the pressure vessel internals of Novovoronezh NPP Unit No 3 after 45 years of operation. The fragments differed in the neutron damaging doses (from 14 to 43 dpa) and the irradiation temperature (from 285 to 315°C). The density and dimensions of titanium carbides and carbonitrides, dislocation loops, radiation-induced voids, segregations, and nanoscale precipitates were determined. The contributions of structural components to the radiation hardening of the investigated fragments of 18Cr-10Ni-Ti stainless steel were estimated.

The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain ...

Organic-rich shale often receives high doses of radiation from radionuclides, mostly uranium (U). Natural decay of U releases particles and energy simultaneously. Thermal stress is considered to be...

For deuterium–tritium fusion reactor, the fuel of tritium is not available naturally. Tritium should be produced by the reaction of Li (n, ɑ) T. Li4SiO4 is one of promising candidates due to its high lithium density. In present work, effects of tritium release behavior from Li4SiO4 have been investigated and compared. The main release peak at lower temperature was attributed to tritium located on/near the grain surface and chemical adsorption on pore surface. The small peak at higher temperature was designated as tritium diffusion from grain including lattice and boundary. The tritium release behavior was simulated by diffusion model and surface reaction model. Tritium behavior is controlled by both diffusion and surface reaction for different stages. Based on the desorption theory, the kinetic parameters were obtained. The effects on tritium release behavior were studied. The porosity has significant effects on tritium release. Higher porosity has large specific surface providing more activation sites and fast channels for tritium release which results in lower temperature of tritium release. The main tritium release form was considered as tritium water. The release peak of deuterium moved towards lower temperature after adsorbing water vapor for 1800 h. It was demonstrated that adsorbing water was beneficial for tritium release due to isotope exchange. Compared the effects of porosity and water vapor, higher porosity has larger effects on tritium release in the present work. Based on the work, the tritium release is controlled by both diffusion and surface reaction at different stage and affected by comprehensive effects including porosity and adsorption water.

An increase of neutron irradiation fluence caused a decrease of Si radiation detector efficiency that was exceptionally well seen at 1017 neutron/cm2 fluence when the observed I–V characteristic of p-n junction under forward bias and under reverse bias became similar. Therefore the investigation of free carrier mobility could be a key experiment to understand the change of heavily irradiated silicon. The electron mobility was investigated by magnetoresistance means in microstrip silicon samples at temperature range T = 200–276 K. The analysis included the free carrier scattering by phonons, ionized impurities, dipoles and clusters and a contribution of each process was found by fitting the mobility dependence on temperature. The analysis of experimental data clearly demonstrated that the applied model did not explain the mobility in the samples irradiated to the highest fluence. Therefore a new concept of carrier transport is needed, and, as a conclusion, it could be stated that Si irradiated above 1016 cm–2 fluence (and up to 1020 cm–2) is a disordered material with the clusters.

We investigated deuterium (D) retention in three W samples irradiated with MeV neutrons at high damage level of 0.39 ~ 0.74 displacements per atom (dpa) at high temperatures, 894 K, 1074 K and 1379 K. The W specimens were exposed to high-flux (~1 × 1022 m−2 s−1) and high-fluence (~5 × 1025 m−2) D plasma at 873 K in the Tritium Plasma Experiment. Broad desorption peaks extended from 900 K to 1200 K were observed for the neutron-irradiated W by thermal desorption spectroscopy (TDS). The retention in neutron-irradiated specimens was much larger than for an un-irradiated specimen. The highest D retention was obtained for a specimen irradiated at 894 K. With increasing neutron irradiation temperature, the retention was reduced about by half at 1074 K and further increase of the temperature (1379 K) resulted in comparable retention. In addition, one-dimensional diffusion calculations (D desorption in TDS and D depth distribution in plasma exposure) were performed to derive retention parameters (the detrapping energy, the depth occupied by D atoms and D/W ratio) from experimental D retention properties of neutron-irradiated W. By TDS simulation calculation, simple dependences of the peak temperature, height and width of TDS peaks on the retention parameters were obtained with total retention in the orders of 1019 ~ 1022 m−2. The calculation of the depth distribution of trapped D atoms made a relationship between the D/W ratio and the depth occupied by D atoms after plasma exposure at relevant conditions. By comparing the relationship (the D/W and the depth) with that obtained from the experimental results, we estimate each retention parameters for the specimens irradiated by high-dose neutrons at the high temperatures. And, we discussed the neutron-irradiation temperature dependence of D retentions.

The paper presents microstructural studies of specimens cut from fuel elements made of E110 spongy zirconium-based alloy after operation in a VVER-1000 before reaching the burnup of ~35 MWd/kg U. As a result of exposure to high temperatures and neutron irradiation, significant changes in the phase composition of fuel cladding materials appear: change in the size, density, and composition of β-Nb particles; change in the composition of the Laves phase; formation of dislocation loops of α type, as well as δ and γ hydrides. The main structural elements determining the degradation of the mechanical properties of the E110 alloy under irradiation are dislocation loops and fine-phase precipitates owing to their relatively large density. The data obtained can be used to construct dose dependences of microstructural changes with the aim of predicting the residual life of claddings and fuel assemblies as a whole.

This paper presents the results of the experiments on hydrogen isotopes interaction with lithium capillary-porous systems (CPS) under reactor irradiation, when sample’s temperatures were from 180 to 550 °C. A special ampoule device with a tubular sample of lithium CPS was developed for experiments. The IVG.1 M reactor (Kurchatov) was used as a source of radiation. The experiments were carried out by the method of dynamic sorption: a constant flow of deuterium was swept through the tube sample with constant degassing and fixation of gas composition in the experimental ampoule device. The paper presents initial data on temperature dependencies of tritium release from lithium CPS under reactor irradiation. The initial analysis showed some features of tritium release: insignificant tritium release in solid lithium (samples temperature below 180 °C), decrease of tritium release under temperature of cell’s center ˜ 250℃, stop of tritium release under temperature of cell’s center ˜ 300℃; then tritium release increased with increase of temperature. The conclusion was made that tritium capture by lithium has a significant influence on the processes of tritium release from lithium CPS under reactor irradiation. The work was performed within the framework of the ISTC #K-2204 project.

The production of I-131 for use in medicine can be accomplished by the neutron irradiation of tellurium, typically in the form of TeO2. Unfortunately, the literature contains conflicting data concerning the I-131 yield as a function of neutron fluence, target mass, irradiation time and post-irradiation decay. In this work, the activity of the I-131 was determined using calculations and experimental verifications based on the interplay of these variables.

Boron neutron capture therapy (BNCT) is a targeted therapy, which consists of preferential accumulation of boron carriers in tumor followed by neutron irradiation. Each oral cancer patient has different risks of developing one or more carcinomas and/or oral mucositis induced after treatment. Our group proposed the hamster oral cancer model to study the efficacy of BNCT and associated mucositis. Translational studies are essential to the advancement of novel boron delivery agents and targeted strategies. Herein, we review our work in the hamster model in which we studied BNCT induced mucositis using three different cancerization protocols, mimicking three different clinical scenarios. The BNCT-induced mucositis increases with the aggressiveness of the carcinogenesis protocol employed, suggesting that the study of different oral cancer patient scenarios would help to develop personalized therapies.

The United States and Japan have collaborated on fusion materials research in a series of agreements reaching back to 1981. The PHENIX collaboration is the latest U.S.-Japan project which s...

The paper presents microstructural studies of specimens cut from fuel elements made of E110 spongy zirconium-based alloy after operation in WWER-1000 before reaching the burnout of ~35 MW per day/kg U. As a result of exposure to high temperatures and neutron irradiation significant changes in the phase composition of the material of fuel rods claddings appear in particles β-Nb’ size, density, and composition; composition of the Laves phase, formation of dislocation loops of α-type, as well as δ and γ hydrides. The main structural elements determining the degradation of the mechanical properties of the E110 alloy under irradiation are dislocation loops and fine-phase precipitates due to their relatively large density. The data obtained can be used to construct dose dependences of microstructural changes with the aim of predicting the residual life of claddings and fuel assemblies as a whole.

PURPOSE To examine the effect of a change in reactor power on the response of solid tumors, referring to impact on quiescent (Q) tumor cell population. MATERIALS AND METHODS Tumor-bearing mice received 5-bromo-2'-deoxyuridine (BrdU) to label all proliferating (P) tumor cells, and were treated with boronophenylalanine-10B (BPA) or sodium mercaptododecaborate-10B (BSH). After reactor neutron beam irradiation at a power of 1 or 5 MW with an identical beam spectrum, cells from tumors were isolated and incubated with a cytokinesis blocker. The responses of BrdU-unlabeled Q and total (P + Q) tumor cells were assessed based on the frequencies of micronucleation using immunofluorescence staining for BrdU. RESULTS After neutron irradiation with or without 10B-carrier, radio-sensitivity was reduced by decreasing reactor power in both cells, especially in Q cells and after irradiation with BPA. The values of relative and compound biological effectiveness were larger at a power of 5 MW and in Q cells than at a power of 1 MW and in total cells, respectively. The sensitivity difference between total and Q cells was widened when combined with 10B-carrier, especially with BPA, and through decreasing reactor power. CONCLUSION 5 MW is more advantageous than 1 MW for boron neutron capture therapy.

Positron annihilation spectroscopy was used to evaluate the defects in neutron irradiated tungsten exposed at five different irradiation conditions. The variables in neutron irradiation included temperature, displacements per atom (dpa), and neutron spectrum. A set of W, Re, WRe, and WReOs control samples were used in assessing the data. Positron annihilation lifetime spectroscopy and coincidence Doppler broadening measurements revealed that samples irradiated at 500 °C had more vacancy clusters than samples irradiated at higher temperatures. This trend was observed despite some higher temperature samples having a significantly higher dpa. Positron lifetimes indicate these are divided into large (>40) and small (

Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.

The development of advanced tungsten grades able to tolerate irradiation damage combined with thermo-mechanical loads is important for design of plasma-facing components for DEMO. The material microstructure (i.e. grain size, dislocation density, sub grains, texture) is defined by manufacturing and post heat treatment processes. In turn, the initial microstructure might have an important influence on the accumulation of neutron damage because irradiation defects interact with microstructural defects evolving into a new microstructural state. In this work, the microstructure and hardness of four tungsten grades is assessed before and after neutron irradiation performed at 600, 1000 and 1200 °C, up to a dose of ∼1.2 dpa. Experimental characterization involves hardness testing, energy dispersive spectroscopy, electron backscatter diffraction, and transmission electron microscopy. The investigated grades include Plansee and AT&M ITER specification tungsten, as well as fine grain tungsten produced by spark plasma sintering, and ultra-fine grain tungsten reinforced with 0.5 wt% ZrC particles.

International audience; In nuclear research reactors, aluminium alloys are corroded and an aluminium hydroxide film covers their surface. Defects created by neutron irradiation can have a detrimental effect on this corrosion. In this study, two ion irradiations are performed on a 6061-T6 aluminium alloy. The first irradiation experiment aims at studying the effect of metal matrix irradiation on aluminium corrosion. The second irradiation experiment aims at studying the effect of hydroxide irradiation on aluminium corrosion. The displacement per atom is at most 2.5 dpa in the hydroxide film irradiated with Al ions of 1.2 and 5 MeV and at most 8 dpa in the aluminium matrix irradiated with Al ions of 1.8 MeV (calculated with SRIM). The effects of ion irradiation are investigated by TEM and SEM observations. The crystalline structure of aluminium hydroxide is studied by electron diffraction, μ-Raman and X-rays diffraction. In the aluminium hydroxide film (composed of a mix of bayerite, α-Al(OH) 3 , and boehmite, γ-AlOOH, before irradiation), ion irradiation causes formation of voids and dehydrates the aluminium hydroxide (nanocrystallites of η-Al 2 O 3 are observed in the irradiated film). In the aluminium matrix, irradiation increases density of dislocations and amorphizes dispersoids. After ion irradiation, samples are corroded at 70 °C in 2.8L of demineralised water. The two irradiations increase aluminium corrosion. The second part of this study is about samples corroded in the Osiris nuclear research reactor, in water at 42 °C, with a pH of 6 and for 18 months. The aluminium hydroxide film observed on these samples has a layered microstructure composed of a compact inner layer, a thick intermediate layer and an outer layer of cuboid microcrystals. Silicon enrichment is observed in the inner layer. The effect of ion and neutron irradiation on aluminium corrosion is compared in this study and similarities are observed between ion and neutron irradiations.; Dans les réacteurs nucléaires de recherche, les alliages d'aluminium sont corrodés et un film d'hydroxyde d'aluminium recouvre leur surface. Les défauts créés par l'irradiation neutronique peuvent avoir un effet néfaste sur cette corrosion. Dans cette étude, deux irradiations ioniques sont réalisées sur un alliage d'aluminium 6061-T6. La première expérience d'irradiation vise à étudier l'effet de l'irradiation de la matrice métallique sur la corrosion de l'aluminium. La seconde expérience d'irradiation vise à étudier l'effet de l'irradiation de l'hydroxyde sur la corrosion de l'aluminium. Le déplacement par atome est d'au plus 2,5 dpa dans le film d'hydroxyde irradié avec des ions Al de 1,2 et 5 MeV et d'au plus 8 dpa dans la matrice d'aluminium irradiée avec des ions Al de 1,8 MeV (calculé avec le SRIM). Les effets de l'irradiation ionique sont étudiés par des observations MET et SEM. La structure cristalline de l'hydroxyde d'aluminium est étudiée par diffraction électronique, Raman et diffraction des rayons X. Dans le film d'hydroxyde d'aluminium (composé d'un mélange de bayérite, -Al(OH)3 , et de boehmite, -AlOOH, avant irradiation), l'irradiation ionique provoque la formation de vides et déshydrate l'hydroxyde d'aluminium (nanocristallites de η-Al2O3 sont observés dans le film irradié). Dans la matrice d'aluminium, l'irradiation augmente la densité des dislocations et amorphise les dispersoïdes. Après irradiation ionique, les échantillons sont corrodés à 70°C dans 2,8L d'eau déminéralisée. Les deux irradiations augmentent la corrosion de l'aluminium. La deuxième partie de cette étude porte sur des échantillons corrodés dans le réacteur nucléaire de recherche Osiris, dans de l'eau à 42 °C, avec un pH de 6 et pendant 18 mois. Le film d'hydroxyde d'aluminium observé sur ces échantillons présente une microstructure stratifiée composée d'une couche interne compacte, d'une couche intermédiaire épaisse et d'une couche externe de microcristaux cuboïdes. Un enrichissement en silicium est observé dans la couche interne. L'effet de l'irradiation ionique et neutronique sur la corrosion de l'aluminium est comparé dans cette étude et des similitudes sont observées entre les irradiations ionique et neutronique

Co-deposition of carbon atoms with hydrogen isotopes and hydrogenated carbon radicals and molecules is recognized as the main mechanism for tritium retention in the graphite walls of the previous tokamak fusion devices. Significant tritium retention would be a serious concern for safe and economic long-term operation of future fusion test reactors and fusion energy systems. Similar deposits are observed on the surface of the engineered components in a tritium-producing assembly, known as a Tritium-Producing Burnable Absorber Rod (TPBAR). Characterization of the deposits can help understand the tritium transport, accumulation history and distribution in TPBARs. This study reports our recent results from the carbonaceous deposits formed on an aluminide-coated cladding in the lower plenum of a TPBAR following thermal neutron irradiation. The observed deposits are amorphous in nature, consisting of flakes of interconnected nanoscale features. They contain primarily double-bonded carbon (e.g., alkene) and carbonyl carbon, as well as a minor fraction of aliphatic carbon, all of which are likely tritiated. A similar co-deposition process that occurred in previous fusion devices is responsible for the formation and growth of the carbonaceous deposits.

INTRODUCTION AND OBJECTIVE:Boron neutron capture therapy (BNCT) is based on selective accumulation of B10 carriers in tumors followed by neutron irradiation. As the major limitation of BNCT was dru...

The high temperature gas-cooled reactor (HTGR) has advantages for irradiation applications such as large space available for irradiation at reflector region and high thermal neutron spectrum with the graphite moderator. High temperature engineering test reactor (HTTR), a prismatic type of the HTGR, has been constructed to establish and upgrade the basic technologies for the HTGRs. Many irradiation regions are reserved in the HTTR to be served as a potential tool for an irradiation test reactor in order to promote innovative basic researches such as materials, fusion reactor technology, and radiation chemistry. This study shows the overview of some possible irradiation applications at the HTTRs including neutron transmutation doping silicon (NTD-Si) and Iodine-125 (125I) productions. The HTTR has possibility to produce about 40 tons of doped Si-particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8 × 105 GBq/yr of 125I isotope, comparing to 3.0 × 103 GBq of total 125I supplied in Japan in 2016.

The effects of the neutron irradiation treatment on indium-gallium-zinc oxide (IGZO) are investigated as a function of the neutron irradiation time. With an increase in neutron irradiation time, the oxygen vacancies associated the oxygen deficient states increase, and both shallow and deep band edge states below the conduction band also increase. Moreover, the conduction band offset continuously decreases because of the increase in the oxygen vacancies with increasing the neutron irradiation time. In IGZO TFTs with the neutron irradiation time for 10 s, superior device performance demonstrates such as the lower threshold voltage, higher field effect mobility, smaller sub-threshold gate swing, larger on-off current ratio, and improved bias stability, comparing those of other IGZO TFTs.

More effective boron-containing compounds are needed for use in boron neutron capture therapy (BNCT). Here, borate esters were synthesized by heating and dehydrating nucleotides adenosine triphosphate (ATP), adenosine diphosphate (ADP), adenosine monophosphate (AMP), the nucleoside (inosine) or glycerol in the presence of boric acid (H

This chapter provides a comprehensive and detailed review of the topic of He effects in irradiated structural alloys. Following an introductory overview section, the chapter reviews: (a) experimental approaches to studying He effects; (b) He effects on void swelling and high-temperature embrittlement, focusing on austenitic stainless steels, as well as a rate theory (RT)-critical bubble model (CBM) based framework for understanding and mitigating deleterious He effects; (c) the literature on spallation proton irradiation data on bulk helium effects in ferritic–martensitic and austenitic steels; (d) recent electronic–atomistic models of He behavior in steels; (e) nanostructured ferritic alloys (NFA) and in situ He injection data showing that these remarkable high-temperature materials hold great promise of managing He in a way that could turn it from a liability to an asset as well as a very promising comparison of this data to a master multiscale model of He transport and fate in real ferritic–martensitic and NFA microstructures. The chapter concludes with a summary and list of some outstanding issues.

Production and supply of tungsten for the first wall fusion application is becoming an important aspect given the progress of ITER construction. Exploration of advanced routes alternative to the conventional powder metallurgy is currently undertaken. In this work we have assessed a potential of the spark plasma sintering (SPS) production route to deliver well controlled microstructure, chemistry and mechanical properties of bulk tungsten as a first step. SPS-produced tungsten was sintered at 2000 °C and was characterized in terms of mechanical properties, namely: tensile, three point bending and fracture toughness data in the temperaure range of 250–600 °C. Then, neutron irradiation was performed at 600 °C and the change of the fracture toughness was measured after irradiation together with the characterization of the fracture surface. The results are compared with those obtained for the commerically produced swaged tungsten irradiated and tested in equivalent conditions. The obtained results show that SPS technology offers the production of bulk tungsten with a good potential for further optimization (by e.g. swaging/rolling). Neutron irradiation causes the reduction of the fracture toughness comparable to the one induced in the commercially produced tungsten.

Zinc-65 has been of great interest in medical, biomedical, agricultural, and industrial applications due to its suitable half-life and decay properties. The 65Zn was produced via neutron irradiation on natural zinc oxide and natural zinc oxide nanoparticles targets in Tehran Research Reactor (TRR) at a thermal neutron flux of 4.5 × 1013 n cm−2 s−1 for 30 min. The excitation function of 64Zn(n,γ)65Zn reaction was calculated via the TALYS-1.8 code. The MCNPX code was used to calculate the thermal neutron distribution. The 65Zn theoretical production yield was estimated using calculated cross sections and the calculated thermal neutron distribution. The obtained experimental data and simulated value of production yield for 65Zn were in reasonable agreement.

The paper reviews the current status of development and qualification of the HCPB TBM’s functional materials, namely, Li-ceramic breeder and Be multiplier materials. The main functional and performance requirements for both functional materials are overviewed. The main results and outcomes of the post-irradiation examinations (PIE) of Li-ceramic breeder and Be/beryllides materials irradiated in HIDOBE and HICU irradiation campaigns are presented aiming at determination of the materials’ performance, properties and characteristics under neutron irradiation, like e.g. morphology and microstructure, tritium retention/release characteristics, mechanical properties. Fabrication of the advanced ceramic breeder material, containing lithium orthosilicate with addition of lithium metatitanate phase in order to improve its mechanical properties, is described. Results of characterization of this advanced ceramic breeder in terms of mechanical properties, porosity, morphology and effective thermal conductivity of a pebble bed are overviewed. In addition, results of characterization of the reference 1 mm Be pebbles produced by Rotation Electrode Process are presented in terms of their microstructure, tritium release characteristics and interaction with air and steam. A necessity to realize in the future a new irradiation experiment for the functional materials, LIBERTI experiment, is discussed in the conclusions.

International audience; This paper presents a new methodology used to study transmutation damage in boron carbide (B$_4$C). B$_4$C samples were set in a specially designed sample holder and irradiated in a neutron thermal flux in the ILL reactor. After describing the sample holder and the irradiation conditions, this paper discusses the results of the first post-irradiation examinations by secondary-ion mass spectrometry (SIMS), X-ray diffraction (XRD) and scanning electron microscopy (SEM). The results obtained for a B$_4$C disc with natural boron composition prove the efficiency of the methodology.

The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors (HTGRs) was investigated. A high-temperature engineering test reactor (HTTR), which is located in Japan at Oarai-machi Research and Development Center, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the 125I production could be maximized and the 126I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 1.8 × 105GBq/y of 125I production.

New microsphere sorbents are reported, which could find application in demanding radiation environments and especially as targets for the production of nuclear medicines by neutron irradiation. An easily-synthesized Zr anionic complex was introduced into quaternary amine-functionalised polystyrene-divinylbenzene-based anion-exchange resins by batch adsorption. Upon carbothermal reduction, the precursors were converted to porous carbon matrices containing particles of ZrC and ZrO2 polymorphs. The most phase-pure material, ZrAX-1, possessed high surface area, multi-scale porosity and high mechanical strength. Adsorption of Re and W was investigated and its possible deployment as a reusable host for the production of 188W/188Re is discussed.

This manuscript describes the production of three targets obtained by implantation of different activities of 7 Be into thin aluminium disks. Two of the produced targets were used to measure the 7 Be(n, p) 7 Li cross section in the energy range of interest for the Big-Bang Nucleosynthesis. A third target was used to measure the cross sections of 7 Be(n, p) 7 Li and 7 Be(n, α ) 7 Li nuclear reactions with cold and thermal neutrons, respectively. This paper describes also the characterization of the first two targets, performed after the neutron irradiation, in terms of implanted 7 Be activities and spatial distributions.

Boron-carbide-based materials (B12X2) with two-atom instead of three-atom chains have better ductility, which indicates that they may be better alternatives to nuclear absorber materials than B4C. In this study, we investigated the migration energy of neutron-induced helium interstitials using density functional theory calculations. As a result, we discovered that the migration energy of helium in B12Si2 and B12O2 is lower than that in B4C, which suggests that these materials might be better in inhibiting the introduction of helium gas and subsequent volume expansion during the neutron irradiation. Moreover, we found that B12P2 and B12As2 have isotropic helium migration barriers, while B4C, B12Si2, and B12O2 exhibited a strong anisotropy in the helium migrations.

Owing to the importance of two HfB2 and ZrB2 materials in nuclear industry, the behavior of the two materials under neutron irradiation was examined. As Boron element plays an important role in these materials, it seems that these two substances are more likely to be similar to each other in more terms. However, the results indicated that there were differences. HfB2 is a better absorbent for neutron in all energies rather than ZrB2. However, ZrB2 may become damaged or burnt-out relatively soon. The prompt gamma rays produced in ZrB2 for low energy neutron are far less than those in HfB2, but for high energy neutron, high intensity gamma ray peaks arise in ZrB2. Therefore, if one wants to use these two materials as radiation shield, he should pay attention to that point. Alpha particle spectrum and gamma rays were obtained for different energies of neutron. Half Value Layer (HVL) values for both materials and different neutron energies were also obtained.

Archeological artifacts can be analyzed after neutron irradiation at the pulsed ISIS neutron and muon source, UK, using a newly installed high purity germanium gamma ray spectrometer to perform neutron activation analysis. In this work, the details of the measurement methods and data analysis are presented. In particular, it is explained how Monte Carlo calculations are necessary to evaluate the detection efficiency, taking into account self-shielding effects. The results for two certified bronze standards are presented. The good agreement between expected and measured compositions is promising for the use of this technique for archeological artifacts where the elemental concentration is often unknown. As an example, the analysis of a Chinese sword from the first or second century BC is presented.

In order to enhance the neutron-shielding performance of concrete in neutron-generation facilities from the viewpoint of reduction of effective dose rates in operation and radioactive wastes in dec...

Hall sensors with bismuth sensitive layer were selected to be implemented on ITER as ex-vessel steady state magnetic sensors. The paper review the design of these sensors and their manufacturing technology, including some alternative options. The sensor prototypes were extensively tested to evaluate their compatibility with the ITER requirements. Dependence of sensitivity of sensors on temperature was found to be very similar for various thickness of the sensitive layer. Results of the neutron irradiation test prove the sufficient radiation hardness of the sensors and show only very modest change of their sensitivity over their full lifetime on ITER.

In order to investigate the effect of post-weld heat treatment (PWHT) and post-irradiation annealing (PIA), electron-beam-weld specimens of the reference low activation vanadium alloy, NIFS-HEAT-2,...

The segregation of certain impurities to grain boundaries (GBs) which induced by neutron irradiation in Reactor Pressure Vessel (RPV) steels may have adverse effects upon their mechanical properties. In this work, ten symmetrical and tilt GBs in BCC Fe were created based on Coincident Site Lattice method, the GBs segregation embrittling potency was simulated by adding phosphorus, copper, nickel, and manganese to the boundaries and free surface (FS) using molecular static method. We have addressed both high- and low-angle tilt grain boundaries. Systematic Molecular Statics simulations were performed employing recent Embedded Atom Method (EAM) potentials. We show that P will strongly bind to GBs and FS; P and Cu have a detrimental effect on the GB cohesion, and segregation of these two elements will lead to GB embrittlement.

We propose an object kinetic Monte Carlo (OKMC) model for describing the microstructural evolution in pure tungsten under neutron irradiation. We here focus on low doses ( under 1 dpa), and we neglect transmutation in first approximation. The emphasis is mainly centred on an adequate description of neutron irradiation, the subsequent introduction of primary defects, and their thermal diffusion properties. Besides grain boundaries and the dislocation network, our model includes the contribution of carbon impurities, which are shown to have a strong influence on the onset of void swelling. Our parametric study analyses the quality of our model in detail, and confronts its predictions with experimental microstructural observations with satisfactory agreement. We highlight the importance for an accurate determination of the dissolved carbon content in the tungsten matrix, and we advocate for an accurate description of atomic collision cascades, in light of the sensitivity of our results with respect to correlated recombination. (C) 2017 Published by Elsevier B.V.

Various interfering reactions producing Ar isotopes during neutron irradiation from Cl, Ar, K, and Ca have been previously detailed with the significant ones being routinely corrected for in 40 Ar/ 39 Ar geochronology. Interference of double charged 80 Kr ( 80 Kr ++ ) with 40 Ar has not yet been considered. Significant amounts of 80 Kr are produced during neutron irradiation through the reaction 79 Br(n,β − ) 80 Kr. While previous workers reported a computed fission spectrum averaged cross section of ∼48 mb—compared to ∼113 mb for 39 K(n,p) 39 Ar—we determined a ∼33-fold higher production rate of 80 Kr from Br compared to 39 Ar from K in the CLICIT facility of the OSU reactor. Low-K, high-Br phases, e.g., some amphiboles, and fluid or melt inclusion rich samples are prone to 80 Kr ++ interference biasing 40 Ar/ 39 Ar dates in the ‰ to % range. 80 Kr ++ resembles excess 40 Ar in step-heating experiments. The interference can be corrected for by using the parallel reaction 81 Br(n,β − ) 82 Kr.

SiC fiber-reinforced SiC matrix composites (SiCf/SiC) are considered as one of the candidates for blanket materials in future fusion reactors and as an advanced fuel cladding material for next-gene...