Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2022

Kokubun, Yuji; Nakada, Akira; Seya, Natsumi; Nagaoka, Mika; Koike, Yuko; Kubota, Tomohiro; Hirao, Moe; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; et al.

JAEA-Review 2023-052, 118 Pages, 2024/03

JAEA-Review-2023-052.pdf:3.67MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2022. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Thermophysical properties of dense molten AlO determined by aerodynamic levitation

Sun, Y.*; Takatani, Tomoya*; Muta, Hiroaki*; Fujieda, Shun*; Kondo, Toshiki; Kikuchi, Shin; Kargl, F.*; Oishi, Yuji*

International Journal of Thermophysics, 45(1), p.11_1 - 11_19, 2024/01

https://doi.org/10.1007/s10765-023-03302-2

no abstracts in English

Study on criticality safety control of fuel debris for validation of methodology applied to the safety regulation

Suyama, Kenya; Ueki, Taro; Gunji, Satoshi; Watanabe, Tomoaki; Araki, Shohei; Fukuda, Kodai; Yamane, Yuichi; Izawa, Kazuhiko; Nagaya, Yasunobu; Kikuchi, Takeo; et al.

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 6 Pages, 2023/10

To remove and store safely the fuel debris generated by the severe accident of the Fukushima Daiichi Nuclear Power Station in 2011 is one of the most important and challenging topics for decommissioning of the damaged reactors in Fukushima. To validate the adopted method for the evaluation of criticality safety control of the fuel debris through comparison with the experimental data obtained by the criticality experiments, the Nuclear Regulation Authority (NRA) of Japan funds a research and development project which was entrusted to the Nuclear Safety Research Center (NSRC) of Japan Atomic Energy Agency (JAEA) from 2014. In this project, JAEA has been conducting such activities as i) comprehensive computation of the criticality characteristics of the fuel debris and making database (criticality map of the fuel debris), ii) development of new continuous energy Monte Carlo code, iii) evaluation of criticality accident and iv) modification of the critical assembly STACY for the experiments for validation of criticality safety control methodology. After the last ICNC2019, the project has the substantial progress in the modification of STACY which will start officially operation from May 2024 and the development of the Monte Carlo Code "Solomon" suitable for the criticality calculation for materials having spatially random distribution complies with the power spectrum. We present the whole picture of this research and development project and status of each technical topics in the session.

Magnetic excitation in the =1/2 Ising-like antiferromagnetic chain CsCoCl in longitudinal magnetic fields studied by high-field ESR measurements

Kimura, Shojiro*; Onishi, Hiroaki; Okunishi, Koichi*; Akaki, Mitsuru*; Narumi, Yasuo*; Hagiwara, Masayuki*; Kindo, Koichi*; Kikuchi, Hikomitsu*

Journal of the Physical Society of Japan, 92(9), p.094701_1 - 094701_9, 2023/09

https://doi.org/10.7566/JPSJ.92.094701

Validation practices of multi-physics core performance analysis in an advanced reactor design study

Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08

https://dx.doi.org/10.13182/NURETH20-40118

An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.

Anomalous behavior of liquid molecules near solid nanoparticles; Novel interpretation on thermal conductivity enhancement in nanofluids

Hashimoto, Shunsuke*; Yamaguchi, Satoshi*; Harada, Masashi*; Nakajima, Kenji; Kikuchi, Tatsuya*; Oishi, Kazuki*

Journal of Colloid and Interface Science, 638, p.475 - 486, 2023/05

https://doi.org/10.1016/j.jcis.2023.01.101

Recently, it has been reported that anomalous improvement in the thermal conductivity of nanofluid composed of base liquids and dispersed solid nanoparticles, compared to the theoretically predicted value calculated from the particle fraction. Generally, the thermal conductivity values of gases and liquids are dominated by the mean free path of the molecules during translational motion. Herein, we present solid evidence showing the possible contribution of the vibrational behavior of liquid molecules around nanoparticles to increasing these thermal conductivities.

Validation study of thermal-hydraulics analysis code SPIRAL to a large-scale wire-wrapped fuel assembly sodium test at a low Reynolds number flow regime

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

Removal of core decay heat by utilizing natural circulation is expected as a significant measure to enhance the safety of sodium-cooled fast reactors (SFRs). Accurate evaluation of the temperature distribution in the fuel assembly (FA) at the low Re regime in natural circulation operation is demanded. A detailed thermal-hydraulics analysis code named SPIRAL has been developed to clarify thermal-hydraulic phenomena in the FA at various operation conditions. In this study, SPIRAL with the hybrid turbulence model was applied to analyze a large-scale fuel assembly experiment of a 91-pin bundle for two cases at the mixed and the natural convection conditions respectively in low Re regime with heat transfer from outside of the FA. The applicability of the SPIRAL to the thermal-hydraulics evaluation of FA at mixed and natural convection conditions was confirmed by the comparisons of temperatures predicted by SPIRAL with those measured in the experiment.

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2021

Nakada, Akira; Kanai, Katsuta; Kokubun, Yuji; Nagaoka, Mika; Koike, Yuko; Yamada, Ryohei*; Kubota, Tomohiro; Hirao, Moe; Yoshii, Hideki*; Otani, Kazunori*; et al.

JAEA-Review 2022-079, 116 Pages, 2023/03

JAEA-Review-2022-079.pdf:2.77MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2021. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Overhaul of the Primary cooling system heat exchanger in JRR-3

Uno, Yuki; Ouchi, Yasuhiro; Ouchi, Satoshi; Baba, Ryota; Kikuchi, Masanobu; Kawamata, Satoshi

JAEA-Technology 2021-046, 39 Pages, 2023/02

JAEA-Technology-2021-046.pdf:3.76MB

JRR-3 (Japan Research Reactor No.3) is a light water research reactor cooling pool type light water deceleration of low-enriched uranium up to 20MW thermal power. November 1990, begin to operation in modified that we are provided to users as a high neutron flux form reactor facility in various types of irradiation facilities and neutron beam experiment equipment. Currently, JRR-3 has completed the period of facility inspections, which had been extended due to the effects of the Great East Japan Earthquake of March 11, 2011, and has been able to conformity to the New Regulatory Requirements. It has also resumed operation for the first time in about 10 years. FY 2017, overhauled the primary cooling heat exchanger No.1 and No.2 based on a maintenance plan. This is report for take advantage what inspection and maintenance of future about overhaul of the primary cooling system heat exchanger for collect of inspection records and performance.

Seismic evaluation of the CRDM and the CRDM guide tube for JRR-3

Kikuchi, Masanobu; Kawamura, Sho; Hosoya, Toshiaki

JAEA-Technology 2021-040, 86 Pages, 2023/02

JAEA-Technology-2021-040.pdf:3.26MB

In JRR-3, in response to new regulatory standard for research and test reactor which is enforced December 2013, we established new design basis ground motion for confirming new regulatory standard and carried out seismic evaluations of the appointments, instruments and structures which are installed in JRR-3 by using that earthquake motion. This report shows that the result of evaluations by fatigue strength evaluation, which is more detailed evaluation approach, about Control Rod Drive Mechanism (CRDM) and the CRDM Guide Tube that have gotten the serious result of seismic safety margin by using time history response analysis method. As a result, it was confirmed that CRDM and the CRDM Guide Tube have sufficient seismic safety margin.