Decommissioning of uranium handling facility for development of nuclear fuel manufacturing equipment

Kageyama, Tomio; Denuma, Akio; Koizumi, Jin*; Odakura, Manabu*; Haginoya, Masahiro*; Isaka, Shinichi*; Kadowaki, Hiroyuki*; Kobayashi, Shingo*; Morimoto, Taisei*; Kato, Yoshiaki*; et al.

JAEA-Technology 2022-033, 130 Pages, 2023/03

JAEA-Technology-2022-033.pdf:9.87MB

Uranium handling facility for development of nuclear fuel manufacturing equipment (Mockup room) was constructed in 1972. The Mockup room has a weak seismic resistance and is deteriorating with age. Also, the original purpose with this facility have been achieved and there are no new development plans using this facility. Therefore, interior equipment installed in this facility had been dismantled and removed since March 2019. After that, the Mockup room was inspected for contamination, and then controlled area in the Mockup room was cancelled on March 29th 2022. A total of 6,549 workers (not including security witnesses) were required for this work. The amount of non-radioactive waste generated by this work was 31,300 kg. The amount of radioactive waste generated by this work was 3,734 kg of combustible waste (103 drums), 4,393 kg of flame resistance waste (61 drums), 37,790 kg of non-combustible waste (124 drums, 19 containers). This report describes about the dismantling and removing the interior equipment in the Mockup room, the amount of waste generated by this work, and procedure for cancellation the controlled area in the facility.

Investigation on natural circulation behavior for decay heat removal in reactor vessel of sodium-cooled fast reactor under severe accident condition, 2; Transient behavior under operations of multiple decay heat removal systems

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10

In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents. Thus, it is required to evaluate the cooling capability of DHRSs including the natural circulation behavior inside the reactor vessel during heat-removal phase that the fuel debris relocated in the reactor vessel is cooled by DHRSs. In this study, the experiments which simultaneously operations of the dipped-type DHX and the penetrated-type DHX were conducted to investigate the effect of operating multiple decay heat removal system on the natural circulation behavior in the reactor vessel. After achieving the stable conditions by operating the dipped-type DHX or the penetrated-type DHX, the other DHX was operated and the transient behavior was clarified by the temperature measurements. The clear temperature rise in the reactor vessel was confirmed by operating the penetrated-type DHX as second DHX operation under the condition of the dipped-type DHX operation at the beginning and the high heater power of fuel debris on the core catcher. Therefore, it was confirmed that the inhibition of the cooling for the decay heat occurred by operating multiple DHXs.

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.89 - 96, 2021/10

Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.

Study on cooling process in a reactor vessel of sodium-cooled fast reactor under severe accident; Velocity measurement experiments simulating operation of decay heat removal systems

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08

https://doi.org/10.1115/ICONE2020-16796

The water experiments using a 1/10 scale experimental apparatus simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Study on multi-dimensional core cooling behavior of sodium-cooled fast reactors under DRACS operating conditions

Ezure, Toshiki; Onojima, Takamitsu; Tanaka, Masaaki; Kobayashi, Jun; Kurihara, Akikazu; Kameyama, Yuri*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.3355 - 3363, 2019/08

Steady-state sodium experiments under the operating conditions of a decay heat removal system (DHRS) were carried out as part of the safety enhancement of sodium-cooled fast reactors using the PLANDTL 2 facility, which has 30 heated channels with electric heaters and 25 no-heated channels as the simulated core. In the experiments, a direct reactor auxiliary cooling system (DRACS) with a dipped type direct heat exchanger (DHX) in the upper plenum was used as the DHRS. This paper reports on the preliminary experimental results of the PLANDTL 2 experiments under the DRACS operating conditions without flow in the primary circuit, including the thermal hydraulic interactions between the upper plenum and the core under the DHX operating conditions and the resulting core cooling behavior from the outside of the multiple rows of the fuel assemblies

Introduction to radiation physics; Third revised edition

Tada, Junichiro*; Nakashima, Hiroshi; Hayano, Ryugo*; Kobayashi, Hitoshi*; Asano, Yoshihiro*

Wakariyasui Hoshasen Butsurigaku; Kaitei 3-Han, 305 Pages, 2018/03

This book is an introduction to radiation physics. Under the concept of "linking physics of high school and radiation physics" for readers with high school graduation degree, we are doing simple commentary on the basis of qualitative explanation as much as possible. This book begins with "What is Radiation Physics", and consists of 12 chapters, including introductory special relativity, introductory quantum theory, structures of atoms and nuclei, radiations, radioactivity, interactions between radiation and matter, accelerators, radiation dose and so on.

Establishment of numerical estimation method for high cycle thermal fatigue in sodium-cooled fast reactor, 2; Benchmark analysis using planar triple parallel jet sodium test for fundamental validation

Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*

Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06

In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed for estimation of the thermal fatigue issue. In fundamental validation, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions at Vr=1, 1.56, and 5.56 were employed for the benchmark analyses according to the knowledge in the literatures. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.

Stabilization of MOX dissolving solution at STACY

Kobayashi, Fuyumi; Sumiya, Masato; Kida, Takashi; Kokusen, Junya; Uchida, Shoji; Kaminaga, Jota; Oki, Keiichi; Fukaya, Hiroyuki; Sono, Hiroki

JAEA-Technology 2016-025, 42 Pages, 2016/11

JAEA-Technology-2016-025.pdf:17.88MB

A preliminary test on MOX fuel dissolution for the STACY critical experiments had been conducted in 2000 through 2003 at Nuclear Science Research Institute of JAEA. Accordingly, the uranyl / plutonium nitrate solution should be reconverted into oxide powder to store the fuel for a long period. For this storage, the moisture content in the oxide powder should be controlled from the viewpoint of criticality safety. The stabilization of uranium / plutonium solution was carried out under a precipitation process using ammonia or oxalic acid solution, and a calcination process using a sintering furnace. As a result of the stabilization operation, recovery rate was 95.6% for uranium and 95.0% for plutonium. Further, the recovered oxide powder was calcined again in nitrogen atmosphere and sealed immediately with a plastic bag to keep its moisture content low and to prevent from reabsorbing atmospheric moisture.

An Experimental study on natural circulation decay heat removal system for a loop type fast reactor

Ono, Ayako; Kamide, Hideki; Kobayashi, Jun; Doda, Norihiro; Watanabe, Osamu*

Journal of Nuclear Science and Technology, 53(9), p.1385 - 1396, 2016/09

https://doi.org/10.1080/00223131.2015.1121844

Decay heat removal by natural circulation is a significant passive safety measure of a fast reactor against station blackout. The decay heat removal system (DHRS) of the loop type sodium fast reactor being designed in Japan comprises a direct reactor auxiliary cooling system and primary reactor auxiliary cooling system (PRACS). The thermal hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant. The experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, the experiments varying the pressure loss coefficients of the loop as the experimental parameters showed robustness of the PRACS.