Aerosol characterization during heating and mechanical cutting of simulated uranium containing debris; The URASOL project in the framework of Fukushima Daiichi fuel debris removal

Porcheron, E.*; Leblois, Y.*; Journeau, C.*; Delacroix, J.*; Molina, D.*; Suteau, C.*; Berlemont, R.*; Bouland, A.*; Lallot, Y.*; Roulet, D.*; et al.

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 5 Pages, 2022/10

One of the important challenges for the decommissioning of the damaged reactors of the Fukushima Daiichi Nuclear Power Station (1F) is the fuel debris retrieval. The URASOL project, which is undertaken by a French consortium consisting of ONET Technologies, CEA, and IRSN for JAEA/CLADS, is dedicated to acquiring basic scientific data on the generation and characteristics of radioactive aerosols from the thermal or mechanical processing of fuel debris simulant. Heating process undertaken in the VITAE facility simulates some representative conditions of thermal cutting by LASER. For mechanical cutting, the core boring technique is implemented in the FUJISAN facility. Fuel debris simulants have been developed for inactive and active trials. The aerosols are characterized in terms of mass concentration, real time number concentration, mass size distribution, morphology, and chemical properties. The chemical characterization aims at identifying potential radioactive particles released and the associated size distribution, both of which are important information for assessing possible safety and radioprotection measures during the fuel debris retrieval operations at 1F.

Dynamics of radiocaesium within forests in Fukushima; Results and analysis of a model inter-comparison

Hashimoto, Shoji*; Tanaka, Taku*; Komatsu, Masabumi*; Gonze, M.-A.*; Sakashita, Wataru*; Kurikami, Hiroshi; Nishina, Kazuya*; Ota, Masakazu; Ohashi, Shinta*; Calmon, P.*; et al.

Journal of Environmental Radioactivity, 238-239, p.106721_1 - 106721_10, 2021/11

https://doi.org/10.1016/j.jenvrad.2021.106721

This study was aimed at analysing performance of models for radiocesium migration mainly in evergreen coniferous forest in Fukushima, by inter-comparison between models of several research teams. The exercise included two scenarios of countermeasures against the contamination, namely removal of soil surface litter and forest renewal, and a specific konara oak forest scenario in addition to the evergreen forest scenario. All the models reproduced trend of time evolution of radiocesium inventories and concentrations in each of the components in forest such as leaf and organic soil layer. However, the variations between models enlarged in long-term predictions over 50 years after the fallout, meaning continuous field monitoring and model verification/validation is necessary.

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03

https://doi.org/10.1016/j.fusengdes.2020.112181

Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by Li(n,)H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.

Acceleration of fusion plasma turbulence simulation on Fugaku and Summit

Idomura, Yasuhiro; Ina, Takuya*; Ali, Y.*; Imamura, Toshiyuki*

Dai-34-Kai Suchi Ryutai Rikigaku Shimpojiumu Koen Rombunshu (Internet), 6 Pages, 2020/12

A new communication avoiding (CA) Krylov solver with a FP16 (half precision) preconditioner is developed for a semi-implicit finite difference solver in the Gyrokinetic Toroidal 5D full-f Eulerian code GT5D. In the solver, the bottleneck of global collective communication is resolved using a CA-Krylov subspace method, and halo data communication is reduced by the FP16 preconditioner, which improves the convergence property. The FP16 preconditioner is designed based on the physics properties of the operator and is implemented using the new support for FP16 SIMD operations on A64FX. The solver is ported also on GPUs, and the performance of ITER size simulations with trillion grids is measured on Fugaku (A64FX) and Summit (V100). The new solver accelerates GT5D by from the conventional non-CA solver, and excellent strong scaling is obtained up to 5,760 CPUs/GPUs both on Fugaku and Summit.

Communication-avoiding Krylov solvers for extreme scale nuclear CFD simulations

Idomura, Yasuhiro; Ina, Takuya*; Ali, Y.*; Imamura, Toshiyuki*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.225 - 230, 2020/10

A new communication avoiding (CA) Krylov solver with a FP16 (half precision) preconditioner is developed for a semi-implicit finite difference solver in the Gyrokinetic Toroidal 5D full-f Eulerian code GT5D. In the solver, the bottleneck of global collective communication is resolved using a CA-Krylov subspace method, while the number of halo data communication is reduced by improving the convergence property using the FP16 preconditioner. The FP16 preconditioner is designed based on the physics properties of the operator and is implemented using the new support for FP16 SIMD operations on A64FX. The solver is ported on Fugaku (A64FX) and Summit (V100), which respectively show 63x and 29x speedups in socket performance compared to the conventional non-CA Krylov solver on JAEA-ICEX (Haswell).

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

Report of summer holiday practical training 2018; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design

Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.

JAEA-Technology 2019-008, 12 Pages, 2019/07

JAEA-Technology-2019-008.pdf:2.37MB

As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.

Calculations of Tritium Recoil Release from Li and U Impurities in Neutron Reflectors (Joint research)

Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2018-010, 33 Pages, 2018/11

JAEA-Technology-2018-010.pdf:2.58MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.

Development and validation of a fracture model for the granite rocks at Mizunami Underground Research Laboratory, Japan

Kalinina, E. A.*; Hadgu, T.*; Wang, Y.*; Ozaki, Yusuke; Iwatsuki, Teruki

Proceedings of 2nd International Discrete Fracture Network Engineering Conference (DFNE 2018) (Internet), 7 Pages, 2018/06

The Mizunami Underground Research Laboratory is located in the Tono area (Central Japan). Its main purpose is providing a scientific basis for the research and development of technologies needed for deep geological disposal of radioactive waste in fractured crystalline rocks. The current work is focused on the experiments in the research tunnel (500 m depth). The collected tunnel and borehole data were shared with the participants of DEvelopment of COupled models and their VALidation against EXperiments (DECOVALEX) project. This study describes how these data were used to (1) develop the fracture model of the granite rocks around the research tunnel and (2) validate the model.

Calculation by PHITS code for recoil tritium release rate from beryllium under neutron irradiation (Joint research)

Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2016-022, 35 Pages, 2016/10

JAEA-Technology-2016-022.pdf:3.73MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 210 (cm) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.