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Kawaguchi, Maho*; Shiba, Shigeki*; Iwahashi, Daiki*; Okawa, Tsuyoshi*; Gunji, Satoshi; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
The Nuclear Regulation Authority (NRA) has been working on an experimental approach for evaluating the criticality of fuel debris produced by the Fukushima Daiichi Nuclear Power Plant (FDNP) accident since 2014, collaborating with the Japan Atomic Energy Agency (JAEA). As part of the approach, JAEA has modified the STAtic experiment Critical facilitY (STACY) for critical experiments to evaluate characteriscs of pseudo-fuel debris. As the preliminary analyses, we verified critical characteristics with major nuclear data libraries for the proposed core configuration patterns. The three-dimensional continuous-energy Monte Carlo neutron and photon transport code, SERPENT-V2.2.0 was used with the latest JENDL, JENDL-5. As a result, larger multiplication factors of JENDL-5 across the modified STACY core configuration patterns were evaluated in comparison to the other libraries. And, 
H scattering and 
U fission sensitivity coefficients of JENDL-5 were different from those of the other libraries. Comparing among analyses with those libraries, the updated S(
, 
) of JENDL-5 might affect the result of critical characteristics in the critical analyses for the modified STACY core configuration.
Shiba, Shigeki*; Iwahashi, Daiki*; Okawa, Tsuyoshi*; Gunji, Satoshi; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The Nuclear Regulation Authority (NRA) has tackled the experimental approach for determining the criticality of pseudo-fuel debris plausibly simulating actual fuel debris since 2014, collaborating with the Japan Atomic Energy Agency. To elucidate the characteristics of the pseudo-fuel debris, the Japan Atomic Energy Agency modified the STACY (STAtic experiment Critical facilitY) to conduct critical experiments simulating fuel debris. Thus, we proposed three types of modified STACY core configurations. In critical experiments in the modified STACY core, it is important to judge whether the proposed modified STACY core configurations are representative of molten core-concrete interaction debris or not. In this study, we built pseudo-fuel debris models considering a volume ratio of pseudo-fuel debris to moderation (V
/V
) and calculated uncertainty-based similarity values (C
) between the modified STACY core configurations and pseudo-fuel debris models using Tools for Sensitivity and Uncertainty Analysis Methodology Implementation-Indices and Parameters (TSUNAMI-IP) in SCALE 6.2. Consequently, the modified STACY core configuration loading structure rods we proposed completely resulted in high similarity to the pseudo-fuel debris models through V
/V
values. The main contributions to C values were 
U 
, U 
, and 
Fe (n,
), except for the pseudo-fuel debris model, including extremely high concrete components.
Okawa, Tsuyoshi; Greenspan, E.*
Nuclear Technology, 160(3), p.257 - 278, 2007/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Since all the ENHS (Encapsulated Nuclear Heat Source) cores designed so far have positive coolant void reactivity, the ENHS reactor core was investigated to have a negative coolant void reactivity feedback.
; Aita, Tsuyoshi; Murakami, Takanori; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro
JNC TN9410 2004-006, 36 Pages, 2004/03
Joyo Plant Operation Management Expert Tool system named JOYPET has developed with the aim of confirming the stable and safety operation of JOYO and improving operational reliability in future FBR plants.New JOYPET system was designed and manufactured in 2002, and began or operation in 2003, because the former system, which was designed in 1988 and operated from 1991 to 2002, was superannuated, and it was difficult to obtain alternative hardwares and replace parts.The difference between the former one and the later new one was adopted the web-online system to use lan(Lacal area network) instead of the host and the terminal computer processing system.Then the new system enabled to take unitary document management for reactor operation, and each person in one's rost was able to search, refer and wake document on line directly.This document reported new JOYPET system design, manufacturing, system constitution and operation actual result.
; Iijima, Takashi;
Journal of Nuclear Science and Technology, 41(2), p.183 - 195, 2004/02
Times Cited Count：10 Percentile：55.63(Nuclear Science & Technology)None
Okawa, Tsuyoshi; Greenspan, E.*
ANS 2004 Winter Meeting, 1(1), 31 Pages, 2004/00
This paper describes a study of the feasibility of designing the ENHS (Encapsulated Nuclear Heat Source) reactor core to have a negative coolant void reactivity feedback along with a flat power distribution using a stepped geometry core.
Okawa, Tsuyoshi; Gi Hong, Ser*; Greenspan, E.*
ARWIF, 0 Pages, 2004/00
The feasibility of designing a 125 MWth molten-salt cooled ENHS (Encapsulated Nuclear Heat Source)-like reactor cores with Pu15N -U15N nitride fuel for high temperature applications is assessed. The cores considered have uniform fuel composition and no blanket elements and solid reflectors. They are to operate for at least 20 effective full power years without refueling, without fuel shuffling and with burnup reactivity swing less than 0.5%. Six molten-salts are considered: NaF(57)-BeF2(43), 7LiF(66)-BeF2(34), LiF(46.5)-NaF(11.5)-KF(42), NaF(50)-ZrF4 (50), LiF(42)-NaF(29)-ZrF4(29) and LiF(43)-RbF(57). Only the first three are considered for the neutronic analysis. Six structural materials are considered: SS304, Hastelloy-N, HT-9, Mn-316SS, PCA, and SiC. It is found that, neutronically, ENHS-like cores can be designed for all combinations of molten-salt coolants and structural materials considered. Relative to the reference ENHS core, the molten-salt cooled cores require significantly tighter lattice, have softer neutron spectra, significantly more negative Doppler reactivity effect, much more positive coolant temperature and void reactivity effect and smaller reactivity worth of the control elements. Of the molten salts considered, LiF-NaF-KF offers the largest p/d ratio and is most suitable for natural circulation cooling. Its drawbacks include a relatively large positive coolant temperature coefficient of reactivity and a relatively small negative Doppler coefficient. Of the structural materials considered, SiC gives the largest p/d ratio, lowest plutonium loading, flattest power distribution and largest reactivity worth of the control elements. Hastelloy-N is the worst structural material; it requires the smallest p/d and gives the largest peak-to-average power density and smallest reactivity worth of the control elements.
Gi Hong, Ser*; Okawa, Tsuyoshi; Greenspan, E.*
Proceedings of 1st COE-INES International Symposium (INES-1), 68 Pages, 2004/00
The feasibility of designing Encapsulated Nuclear Heat Source (ENHS)-like reactors using molten salt rather than liquid metal coolant is being studied. The objective is to design a once-for-life reactor module for developing countries that could provide high-temperature heat for efficient production of electricity and hydrogen. At the present state of knowledge of lead alloy coolant technology, corrosion of structural materials limit the attainable heat source temperature to ~550oC. Molten salt (MS) coolant may enable to increase the heat source temperature to over 750oC.
Maeda, Seiichiro; Takashita, Hirofumi; Okawa, Tsuyoshi; Higuchi, Masashi*; Abe, Tomoyuki
JNC TN8400 2003-019, 185 Pages, 2003/08
JNC-TN8400-2003-019.pdf:11.78MB
We are studying on an in-core breeding concept as a candidate for a practical FBR fuel cycle system attainable in an early stage on the premise that sodium coolant and mixed oxide fuel should be adopted, since the technical issues with these combination are most advanced and common with the fuel cycle for a LWR-MOX system. An enhancement of fuel volume fraction using thick fuel pins enables the in-core breeding. The fuel material flow can be greatly lessened by minimizing amount of the blanket with the in-core breeding core. The low material flow leads to significant reduction of the fuel cycle cost. We investigated a 3500 MWth large-scale core adjusting several conditions presented in JNC's feasibility study program for a commercialized FBR system in this study. These were shown in this study that a discharged burnup averaged over the core and the blanket could reach approximately 130 GWd/t (core averaged about 150 GWd/t) within the maximum fast neutron fluence about 5
10
/cm
, that the small reactivity loss with burnup easily enabled long operation and that stable power distribution during operation significantly improved hydraulic property in this type core. We investigated measures to reduce sodium void reactivity, because core height enlargement to enhance neutron efficiency caused the increase of sodium void reactivity.We also investigated feasibility of a high breeding type core with low burnup considering a variety of FBR introducing scenarios and a trade-off correlation between breeding performance and burnup extension. The performance in this core design at core disruption accidents is not revealed enough. Further investigation should be made in detail to confirm that the in-core breeding concept could be accepted in a safety aspect.
Iijima, Takashi; ;
Chikyu Kankyo To Shingata Genshiryoku Puranto Ni Kansuru Kokusai Kaigi (GENES4/ANP2003), 0 Pages, 2003/00
None
; ; Iijima, Takashi
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
None
;
Proceedings of 10th International Conference on Nuclear Engineering (ICONE-10), 0 Pages, 2002/00
None
Okawachi, Yasushi; Kitano, Akihiro; Suzuki, Takayuki; Okimoto, Yutaka; Usami, Shin; Deshimaru, Takehide
no journal, ,
no abstracts in English
Okawachi, Yasushi; Kitano, Akihiro; Suzuki, Takayuki; Okimoto, Yutaka; Usami, Shin; Deshimaru, Takehide
no journal, ,
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Karube, Koji; Aita, Tsuyoshi; Okawa, Toshikatsu
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Nakashima, Yosuke*; Ichimura, Kazuya*; Takeda, Hisahito*; Iwamoto, Miki*; Hosoda, Yasunari*; Shimizu, Keita*; Oki, Kensuke*; Sakamoto, Mizuki*; Ono, Noriyasu*; Kado, Shinichiro*; et al.
no journal, ,
no abstracts in English
Aizawa, Kenji; Iseki, Atsushi; Okawa, Toshikatsu; Aita, Tsuyoshi; Kamata, Hidehisa
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no abstracts in English
