Experience and technology consolidation related to dismantling sodium equipment; Technology to reduce sodium remaining in 100m grade large tanks

Hayakawa, Masato; Shimoyama, Kazuhito; Miyakoshi, Hiroyuki; Suzuki, Shigeaki*

JAEA-Technology 2021-027, 33 Pages, 2022/01

JAEA-Technology-2021-027.pdf:3.64MB

At the Oarai Research and Development Institute of the Japan Atomic Energy Agency, experimental studies in various sodium environments are being conducted in connection with the research and development of sodium-cooled fast reactors such as the experimental fast reactor Joyo and the prototype fast reactor Monju. The dismantling of sodium test facilities and equipment that have achieved their purpose has been carried out sequentially, and a wealth of experience and technology has been accumulated. On the other hand, a large amount of metallic sodium used for research and testing is being reused for new testing facilities, and the large sodium tanks that contained the metallic sodium are being dismantled. In order to dismantle these tanks safely and efficiently, it is important to reduce the residual sodium inside the tanks (especially at the bottom) as much as possible before dismantling. Therefore, we have been working on the reduction of residual sodium at the bottom of several large sodium tanks of 100 m class. This report describes the technologies and experiences related to the reduction of residual sodium that have been carried out so far.

Construction of the sodium test loop in advanced technology experiment sodium facility (AtheNa)

Imamura, Hiroaki; Suzuki, Masashi*; Shimoyama, Kazuhito; Miyakoshi, Hiroyuki

JAEA-Technology 2019-005, 163 Pages, 2019/06

JAEA-Technology-2019-005.pdf:25.24MB

For the R&D of safety enhance in future fast reactor development, the constructed the large sodium test loop (mother loop) in advanced technology experiment sodium facility (AtheNa) was completed. The sodium test loop possesses the largest capacity of about 240 tons of the world's largest sodium and can supply impurity-controlled high temperature sodium to large structural and technology demonstration test sections. It is greatly expected as R&D such as future international cooperation. For the purpose of future R&D tests, this report compiled the design specifications, fabrication and performance confirmation results of sodium test loop.

Sodium experiments on decay heat removal system of Japan sodium cooled fast reactor; Start-up transient of decay heat removal system

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Watanabe, Osamu*

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The JSFR has two loops of the main heat transport system in order to reduce number of components and the construction cost. The DHRS of JSFR consists of one DRACS and two PRACS, which have a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and also start-up transient of the DHRS loop. Heat transfer coefficient on the tube outer surface was in good agreement with a conventional correlation under operation condition in the reactor. The transient experiments for the start-up of DHRS loop showed that smooth increase of natural draft in the air duct followed by the sodium flow rate in the DHRS loop. Some delay of the flow rate increase was recognized in the DHRS loop as compared with that of the natural draft in the air cooler.

Experimental study on thermal stratification in a reactor vessel of innovative sodium-cooled fast reactor; Mitigation approach of temperature gradient across stratification interface

Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Journal of Nuclear Science and Technology, 47(9), p.829 - 838, 2010/09

https://doi.org/10.3327/jnst.47.829

Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V) of sodium cooled fast reactor. A water experiment using an 1/10 scale model was carried out. Steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at a dipped plate for a fuel handling machine, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was smaller than that in the case of the higher plug position.

Experimental study of gas entrainment phenomena in sodium cooled fast reactor by 1/1.8 scale partial model

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Nihon Kikai Gakkai Rombunshu, B, 76(763), p.457 - 459, 2010/03

Gas entrainment (GE) at free surface is one of the significant issues for the design of Japan Sodium Cooled Fast Reactor (JSFR). It is required to evaluate the occurrence of GE in the operating condition of JSFR. In the present study, a water experiment was performed in a large-scaled partial upper plenum model to simulate the startup operation of JSFR, where the sodium level was lower than in the rated condition. The onset condition of GE was clarified by the visualization of free surface. Vertical velocity distribution was also measured in order to grasp the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of JSFR.

Experimental study on high cycle thermal fatigue in T-junction; Effect of local flow velocity on transfer of temperature fluctuation from fluid to structure

Kimura, Nobuyuki; Ono, Ayako; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 16 Pages, 2009/09

A quantitative evaluation on high cycle thermal fatigue due to temperature fluctuation in fluid is of importance for structural integrity in the reactor. It is necessary for the quantitative evaluation to investigate occurrence and propagate processes of temperature fluctuation, e. g. decay of temperature fluctuation near structures and transfer of temperature fluctuation from fluid to structures. In this study, a water experiment using T-junction was performed to evaluate the transfer characteristics of temperature fluctuation from fluid to structure. In the experiment, temperatures and local velocities were measured simultaneously to evaluate the correlation between the temperature and velocity under the non-stationary fields. The large heat transfer coefficients were registered at the region where the local velocity was high.

Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor; Detailed velocity distribution in a subchannel

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/09

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. It is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle for the high burn-up core. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The velocity distribution in an inner subchannel with the wrapping wire was measured by Particle Image Velocimetry. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetrical flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed.

Experimental studies of natural circulation decay heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

Kamide, Hideki; Miyakoshi, Hiroyuki; Watanabe, Osamu*; Eguchi, Yuzuru*; Koga, Tomonari*

Dai-14-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.427 - 428, 2009/06

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS are performed. Water experiments were carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increase of natural circulation flow rates in all systems of air and sodium were confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan.

Experimental study on onset of gas entrainment by 1/1.8 scale model of sodium cooled fast reactor

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11

Gas entrainment (GE) at free surface is one of the significant issues for the design of Japanese Sodium Cooled Fast Reactor. In the previous study, authors confirmed that GE did not occur at the rated operating mode in the reactor. In the present study, a water experiment to simulate the startup operation of reactor was performed in a large-scaled partial model of the reactor upper plenum in the reactor. The onset condition of GE was observed by the visualization of free surface. Vertical velocity distribution was also measured in order to qualify the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of reactor.

Study on velocity field in a deformed fuel pin bundle; Influence of pin deformation and wrapping wire on velocity distribution

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 9 Pages, 2008/05

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease flow rate via change of flow area in the sub-assembly and influence the heat removal capability. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the influence of pin bowing and wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had the refractive index nearly the same with water and high light transmission rate. This enabled to visualize around the central pin through the outer pins. Velocity distribution was measured by using PIV in sub-channels around the central pin in reference and deformation condition. Velocity distribution around the wrapping wire was measured and the wire influenced to the velocity and RMS in the wide region near the pin surface.