Japan sodium-cooled fast reactor

Suzuki, T., Kamiyama, K., Yamano, H., Kuho, S., Tobita, Y., Nakai, R., Koyama, K., January 20, 2014. A scenario of core dismptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention. Journal of Nuclear Science and Technology. 

Yamano, H., Kubo, S., Shimakawa, Y., Fujita, K., Suzuki, T., Kurisaka, K., 2012. Safety design and evaluation in a large-scale Japan sodium-cooled fast reactor. Science and Technology of Nuclear Installations 2012. Article Id 614973, 14 pages. 

With a purpose of probing a commercially feasible fast reactor system, a feasibility study on commercialized fast reactor cycle systems (FS) was initiated in 1999 (Aizawa, 2001). In the FS, survey studies were made to identify the most promising concept among various systems such as sodium-cooled fast reactors, gas-cooled fast reactors, heavy metal-cooled fast reactors (lead-cooled fast reactors and lead-bismuth cooled fast reactors), and water-cooled fast reactors with various fuels types such as oxide, nitride, and metal fuels. The FS concluded to select an advanced loop-type SFR with mixed oxide fuel named Japan sodium-cooled fast reactor (JSFR Kotake et al., 2005). 

Table 11.1 Major design specifications of demonstration Japan sodium-cooled fast reactor...

Key innovative technologies in the Japan sodium-cooled fast reactor design... 

Kamide, H., et al., 2010. Sodium experiments on decay heat removal systems of Japan sodium cooled fast reactor — start-up transient of decay heat removal system. In Presented at Seventh Korea—Japan Symp. Nuclear Thermal Hydraulics and Safety, Chuncheon, Korea, November 14—17, 2010. 

Kobayashi, J., et al., 2015. Water experiments on thermal striping in reactor vessel of Japan sodium-cooled fast reactor — countermeasures for significant temperature fluctuation generation. In Proceedings of ICONE 23, May 17—21, Chiba, Japan, 2015. 

Okubo, T., et al., 2011. Conceptual design for a large-scale Japan sodium-cooled fast reactor (3) core design in JSFR. In Presented at International Congress on Advances in Nuclear Power Plants (ICAPP ll), Nice, France, May 2-5, 2011. 

K. Aoto, et al.. Design study and R D progress on Japan sodium-cooled fast reactor, J. Nucl. Sci. Technol. 48 (2011) 463. 

T. Wakai, et al., A study for proposal of welded joint strength factors of modified 9Cr-lMo steel for Japan sodium cooled fast reactor (JSER), in Proceedings of ASME PVP2013, 2013. 

A new thermochemical and electrolytic hybrid hydrogen production system in lower temperature range has been developed by the Japan Nuclear Cycle Development Institute (JNC) to achieve the hydrogen production from water by using the heat from a sodium cooled fast reactor (SFR) [7]. 

France, Japan and the Russian Federation presented the status on sodium cooled fast reactor experience preservation made in these coimtries. The reports underlined these countries large experience with design, construction and operation of sodium cooled fast reactors. The discussions underlined the importance of the IAEA support for knowledge preservation of fast reactor experience. 

JSFR-1500 (Japan) JNC Sodium-cooled Fast Reactor... 

The technical basis for the RAPID includes general experience with sodium cooled fast reactors. Specifically, the RAPID concept includes no control rods but incorporates the passive lithium expansion modules, lithium injection modules and lithium release modules to enable an operator-free operation mode. These systems utilize Li as a liquid poison instead of B4C rods. To verify the reactivity worth of Li, the criticality test [XVII-5] using the fast critical assembly (FCA) of the Japan Atomic Research Institute (JAERI) has been conducted. Also, the manufacturing technology of the lithium modules was mastered, and the performance and neutron radiography tests of the lithium expansion and lithium injection module pilots were conducted. 

Figure 5.1 Typical homogeneous and heterogeneous sodium-cooled fast reactor core configurations (a) homogeneous core, (h) axial heterogeneous core, and (c) radial heterogeneous core. All rights reserved hy Japan Atomic Energy Agency.

Figure 5.4 Sodium-cooled fast reactor system (loop type). All rights reserved by Japan Atomic Energy Agency.

Kamide, H., Ando, M., Ito, T., 2015. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors (1). Overview. In 23rd International Conference on Nuclear Engineering, Paper 1666, May 17—21, 2015, Chiba, Japan. 

Okano, Y., Nakai, R., Kubo, S., 2014. International reviews on safety design criteria and development of safety design guidelines for generation-IV sodium-cooled fast reactors. In The Ninth Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS9), Buyeo, Korea, November 16—19, Keynote Lecture 1. 

Nakai, R., et al., December 9—12, 2012. Development of safety design criteria for the generation-IV sodium-cooled fast reactor. In Keynote Lecture at NTHAS8 The Eighth Japan—Korea Symposium on Nuclear Thermal Hydraulics and Safety, Beppu, Japan. 

Zabudko, L.M., Poplavsky, V.M., Shkaboura, I.A., et al., 2009. Euels for advanced sodium cooled fast reactors in Russia state-of-art and prospects. In International Conference on East Reactors and Related Euel Cycles (ER09). Kyoto, Japan. 

Operating with a sodium-cooled fast neutron core, ASTRID is expected to produce approximately 600 MW of electricity. Before the construction of a first-off commercial unit, a demonstration facility is needed to test innovations with respect to previous FNRs. The FNR greatly improves the amount of energy derived from depleted or reprocessed natural uranium, enables plutonium to be used and recycled several times and can recycle minor actinides if needed. Such reactors are currently being built or are on the drawing board in India, Russia, China, and Japan. 

Japan has a long-term national plan to introduce sodium-cooled fast breeder reactors (FBRs) for effective utilization of natural uranium to provide their initial fuel load, plutonium will be extracted from the spent fuel of existing light water reactors (LWRs). 

The construction of Monju, the prototype fast breeder reactor (FBR) which PNC has built as part of the Japanese FBR development programme, was completed in April 1991, and system start-up tests are presently underway. Monju is a loop-type sodium-cooled fast breeder reactor with U-Pu mixed oxide fuel. It supplies 280 MWe to the grid and is situated on the Tsuruga Peninsula facing the Sea of Japan, about 400 km west of Tokyo. 

Monju is Japan s prototype fast breeder reactor 280 MWe(714MWt), fueled with mixed oxides of plutonium and uranium, cooled tty liquid sodium. Construction was started in 1985 nd initial criticality was attained in April 1994... 

Both designs rely on certain experience in development and operation of fast sodium cooled reactors in Japan, described in Annex XIV. 

Large amounts of sodium waste arise from fast neutron reactors (Phenix and Superphenix in France, Dounreay in the UK, Monju in Japan), which are cooled by large amounts of liquid sodium, which is contaminated by 137Cs during its functioning. We shall see that it is possible to remove radioactive cesium after conversion of liquid sodium to sodium hydroxide. 

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