Migration of rhenium and osmium interstitials in tungsten

Suzudo, Tomoaki; Yamaguchi, Masatake; Hasegawa, Akira*

Journal of Nuclear Materials, 467(Part 1), p.418 - 423, 2015/12

AA2015-0099.pdf:0.62MB

https://doi.org/10.1016/j.jnucmat.2015.05.051

Tungsten is expected to be a promising plasma-facing material for future fusion devices, but radiation-induced precipitation (RIP), which leads the material to hardening, is a concern at their practical use. One of the keys to accurate prediction of the emergence of RIP is migration of solute atoms, rhenium and osmium, that are produced by nuclear transmutation through irradiation. We conduct a series of numerical simulations using an atomic kinetic Monte Carlo method and investigate the migration of these solute atoms in the form of tungsten-rhenium and tungsten-osmium mixed dumbbells, considered to be the most efficient "carriers" of the solute atoms. We find that the low rotation energy barrier of these mixed dumbbells leading to three-dimensional migration greatly influences their diffusivities. The result also suggests that, although these dumbbells have three-dimensional motion, one cannot simply reduce their migration behavior to that of vacancy-like spherical objects.

Proceedings of the 3rd Workshop on Molten Salts Technology and Computer Simulation

Okamoto, Yoshihiro; Minato, Kazuo

JAERI-Conf 2004-008, 228 Pages, 2004/04

JAERI-Conf-2004-008.pdf:17.05MB

Applications of molten salts technology to separation and synthesis of materials have a potential to give us a civilized life, for example aluminium refinement. Recently, much attention is given to the pyrochemical reprocessing of spent nuclear fuels in the molten salt research field. On the other hand, computer simulation technique is expected to play an important role for supporting experimental works and predicting unknown physical properties in the molten salts application studies. Research group for Actinides Science, Department of Materials Science, Japan Atomic Energy Research Institute(JAERI), together with Reprocessing and Recycle Technology Division, Atomic Energy Society of Japan, organized the 3rd Workshop on Molten Salts Technology and Computer Simulation at Tokai Research Establishment, JAERI on December 16, 2003. Many molten salts researchers in Japan participated in the workshop and many useful presentations and discussions were performed.

High-energy X-ray diffraction studies of non-crystalline materials

Ono, Hideo; Kohara, Shinji*; Umesaki, Norimasa*; Suzuya, Kentaro

Journal of Non-Crystalline Solids, 293-295, p.125 - 135, 2001/11

https://doi.org/10.1016/S0022-3093(01)00661-5

no abstracts in English

A Study on modeling and numerical simulation of extraction in the CMPO-TBP system

; ;

JNC TN8400 2001-022, 60 Pages, 2001/03

JNC-TN8400-2001-022.pdf:1.31MB

A numerical simulation code for the TRUEX (Transuranium Extraction) process was developed. Concentration profiles of americium and europium were calculated for some experiments of the counter current extraction system those were carried out in CPF (Chemical Processing Facility) by using the code. Calculation profiles were in agreement with the experimental results. Operational conditions were also examinted for the americium recovery experiment by the TRUEX process carried out in the Plutonium Fuel Center. It was shown that lowering the concentration of nitric acid in the scrub solution and decreasing the flow rate of solvent and strip solution was effective for improving the performance of the stripping step and reducing the volume of the waste solution. In order to find the optimum conditions for various experiments, this simulation code was modified to calculate the concentration profiles of other metal elements such as zirconium and iron and the effect of oxalic acid on the extraction behavior of the metal elements. The calculated concentration profiles of americium and europium were varied by this modification. In the experiment at CPF, the calculations were carried out to obtain recovery ratio of americium in the product stream with the amount of oxalic acid added to the process. This calculation result showed that it was possible to improve the performance of decontamination of fission products by increasing oxalic acid concentration added to the process. The calculation was also carried out for finding the optimum conditions of oxalic acid concentration added to the europium recovery process.

Study of the quality of vipac oxide fuel obtained by pyro-processing

Kakehi, Isao;

JNC TN9400 2000-054, 84 Pages, 2000/04

JNC-TN9400-2000-054.pdf:7.15MB

This report describes accomplishment of the study on the quality of vipac (vibro-packed) oxide fuel obtained by pyrochemical processing (molten salt electrolytic processing). This study is intended to contribute to the design study of the pyro-reprocessing-vipac fuel recycling system of oxide fuel. In this study, vibro-packing experiment has been conducted using granular U0 obtained by molten salt electrolytic processing (cold experiment). The oxide pyro process developed by Research lnstitute of Atomic Reactors (RIAR) is the method in which the sintered oxide is electrically deposited on the cathode at approximately 600C. 0xide granules for vipac fuel are obtained by crushing the oxide deposited on the cathode. This process is also developed as recycle process because it is capable of FP separation. Also in Japan, this process is studied as one of the new FBR fuel recycling systems. ln this study, we made an effort to clarify the mechanisms of vibro-packing of the electrically obtained granules, which influence on the effective parameters of vibro-packing density and fuel particles size distribution in the fuel cladding in case of non-sphere particles of the granules. As a result of the study, smear density of 75% and almost uniform distribution of U0 particles have been taken in the experiment, and much knowledge for the improvement of the vibro-packing quality has been found. And the possibility of the smear density over 80% and the uniform distribution of U0 particles has been suggested in this study.

Development of system analysis code for pyrochemical process using molten salt electrorifining

Tozawa, Katsuhiro; ; Kakehi, Isao

JNC TN9400 2000-052, 110 Pages, 2000/04

JNC-TN9400-2000-052.pdf:4.39MB

This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basis of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without ...

None

*; *; *; *; Hasegawa, Makoto; Hirano, Koichiro

JNC TY9400 2000-007, 50 Pages, 2000/03

JNC-TY9400-2000-007.pdf:1.29MB

no abstracts in English

Feasibility study on magnetic separation

Oda, Yoshihiro; Funasaka, Hideyuki; Wang, X.*; Obara, Kenji*; Wada, Hitoshi*

JNC TY8400 2000-002, 47 Pages, 2000/03

JNC-TY8400-2000-002.pdf:2.53MB

no abstracts in English

The Research on the behavior of the minor products in the PUREX Process

Koga, Jiro*; Shinzato, Takushi*

JNC TJ8400 2000-054, 48 Pages, 2000/02

JNC-TJ8400-2000-054.pdf:1.23MB

The "STELLA" which is a tool for simulation of dynamical systems applied to the numerical simulation of the behavior of minor constituent, such as hydrazoic acid, forming and extinguishing on the operation of reprocessing process. The concentration of hydrazoic acid forming by the reaction of nitrite and hydrazine were determined by use of STELLA after the determination of concentration of main constituents by MIXSET-X. The results from simulation is shown that the STELLA is applicable to the numerical simulation of the behavior of minor constituent.

The Development of MESHNOTE Code for Radionuclide Migration in the Near Field

; Makino, Hitoshi; Peter*

JNC TN8400 99-095, 69 Pages, 1999/12

JNC-TN8400-99-095.pdf:10.06MB

MESHNOTE code was developed to evaluate the engineered barrier system in collaboration with QuantiSci. This code is used to simulate glass dissolution, diffusive transport of nuclides in the buffer material and release to surrounding host rock. MESHNOTE is a one-dimensional finite difference, code, which uses cylindrical co-ordinates for the solution of a radially symmetric diffusion problem. MESHNOTE has the followig characteristics: (1) MESHNOTE can solve for diffusive transport of nuclides through an annulus shaped buffer region while accounting for multiple decay chains, linear and non-linear sorption onto the buffer materials and elemental solubility limits; (2) MESHNOTE can solve for ingrowth of plural daughter nuclides from a singular parent nuclide (branching), and the ingrowth of a singular daughter nuclide from plural parent nuclides (rejoining); (3) MESHNOTE can treat the leaching of nuclide from the vitrified waste and the release of nuclide from buffer to surrounding rock, which are boundary conditions for migration in the buffer, basing on the phenomena; (4) MESHNOTE can treat principal parameters (e.g. solubility and distribution coefficient) relevant to nuclide migration as time and space-dependence parameters; (5) The time stepping scheme in MESHNOTE is controlled by tolerance defined by the user. The time stepping will increase automatically while checking the accuracy of the numerical solution. The conceptual model, the mathematical model and the numerical implementation of the MESHNOTE code are described in this report and the characteristic functions of MESHNOTE are verified by comparing with analytical solutions or simulations produced with other calculation codes.