Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
Cs activity in spent nuclear fuel with calculated detector response functionSato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Kashima, Takao*; Futakami, Kazuhiro*; Suyama, Kenya
Journal of Nuclear Science and Technology, 60(6), p.615 - 623, 2022/06
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)A new non-destructive method for evaluating Cs activity in spent nuclear fuels was proposed and experimentally demonstrated for physical measurements in burnup credit implementation. Cs activities were quantified using gamma ray measurements and numerical detector response simulations without reference fuels, in which Cs activities are well known. Fuel samples were obtained from a lead use assembly (LUA) irradiated in a commercial pressurized water reactor (PWR) up to 53 GWd/t. Gamma rays emitted from the samples were measured using a bismuth germinate (BGO) scintillation detector through a collimator attached to a hot cell. The detection efficiency of gamma rays with the detector was calculated using the PHITS particle transport calculation code considering the measurement geometry. The relative activities of 
Cs, Cs, and 
Eu in the sample were measured with a high-purity germanium (HPGe) detector for more accurate simulations of the detector response for the samples. The absolute efficiency of the detector was calibrated by measuring a standard gamma ray source in another geometry. Cs activity in the fuel samples was quantified using the measured count rate and detection efficiency. The quantified Cs activities agreed well with those estimated using the MVP-BURN depletion calculation code.
Ozawa, Masaaki*; Amaya, Masaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.185 - 200, 2020/12
no abstracts in English
Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi
Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08
Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Shiba, Tomooki; Tomikawa, Hirofumi; Hori, Masato
Proceedings of IAEA Symposium on International Safeguards; Building Future Safeguards Capabilities (Internet), 6 Pages, 2018/11
Kikuchi, Takeo; Tada, Kenichi; Sakino, Takao; Suyama, Kenya
JAEA-Research 2017-021, 56 Pages, 2018/03
JAEA-Research-2017-021.pdf:2.15MBJAEA-Research-2017-021(errata).pdf:0.13MB
The criticality management of the fuel debris is one of the most important research issues in Japan. The current criticality management adopts the fresh fuel assumption. The adoption of the fresh fuel assumption for the criticality control of the fuel debris is difficult because the k
of the fuel debris could exceed 1.0 in most of cases which the fuel debris contains water and does not contain neutron absorbers such as gadolinium. Therefore, the adoption of the burnup credit is considered. The prediction accuracy of the isotopic composition of used nuclear fuel must be required to adopt the burnup credit for the treatment of the fuel debris. JAEA developed a burnup calculation code SWAT4.0 to obtain reference calculation results of the isotopic composition of the used nuclear fuel. This code is used to evaluate the composition of fuel debris. In order to investigate the prediction accuracy of SWAT4.0, we analyzed the PIE of BWR obtained from 2F2DN23.
Fukaya, Yuji; Nishihara, Tetsuo
Nuclear Engineering and Design, 307, p.188 - 196, 2016/10
Times Cited Count：4 Percentile：36.53(Nuclear Science & Technology)Reduction of High Level Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and the features are significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing, and effective waste loading method for direct disposal is proposed by applying the feature in this study. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with LWR representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.53 - 62, 2016/09
In order to evaluate adequacy of present safety criteria and safety margins in terms of advanced fuels and provide a database for future regulation on them, JAEA started an extensive research program called ALPS-II program, which has been sponsored by NRA, Japan. This program is primarily composed of tests simulating a RIA and a LOCA on the high-burnup advanced fuels irradiated in commercial PWR or BWR. Recently, the failure limits of the high-burnup advanced fuels under RIA conditions were investigated at NSRR, and post-test examinations on the fuel rods after the pulse irradiation tests are being performed. In terms of the simulated LOCA test, integral thermal shock tests and high temperature oxidation tests were carried out at RFEF, and the fracture limits, high temperature oxidation rate, etc. of the high-burnup advanced fuel cladding were investigated. This paper mainly describes some recent experimental results obtained in this program with respect to RIA and LOCA.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09
Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10
LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 
81
m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.
Sekimoto, Hiroshi*
JAERI-Tech 2005-008, 111 Pages, 2005/03
JAERI-Tech-2005-008.pdf:15.9MB
The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. When this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the bunrup of discharged fuel is about 40%. It means that the nuclear energy can be utilized for many hundreds years without new mining, enrichment and reprocessing, and the amount of spent fuel can be reduced considerably. Compared to fast reactors, application of CANDLE burnup to prismatic fuel high-temperature gas cooled reactors is very easy. In this report, the applications of CANDLE burnup to both these types of reactors are studied.
Sugiyama, Tomoyuki; Fuketa, Toyoshi; Ozawa, Masaaki*; Nagase, Fumihisa
Proceedings of 2004 International Meeting on LWR Fuel Performance, p.544 - 550, 2004/09
Two pulse irradiation experiments simulating reactivity initiated accidents were performed on high burnup (60 GWd/t) PWR UO
rods with advanced cladding alloys. Test OI-10 was performed on an MDA cladded rod with large-grain (25 m) fuel pellets with a peak fuel enthalpy condition of 435 J/g, and resulted in a peak residual hoop strain of 0.7%. On the other hand, Test OI-11 on a ZIRLO cladded rod with conventional pellets resulted in a fuel failure at a fuel enthalpy of 500 J/g due to the pellet-cladding mechanical interaction (PCMI). A long axial split was generated on the cladding over the active length. The fuel pellets were fragmented and dispersed into the coolant water. The fuel enthalpy at failure is higher than the PCMI failure criterion of 209 J/g at the corresponding burnup. The experimental results suggest that the rods with improved corrosion resistance have much safety margin against the PCMI failure compared to the conventional Zircaloy-4 rod.
Nomura, Yasushi*; Okuno, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2004-030, 64 Pages, 2004/03
JAERI-Tech-2004-030.pdf:4.59MB
no abstracts in English
Katanishi, Shoji; Takei, Masanobu; Nakata, Tetsuo*; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.67 - 75, 2004/03
no abstracts in English
Shimada, Shoichiro*; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi
JAERI-Research 2003-003, 72 Pages, 2003/03
JAERI-Research-2003-003.pdf:3.82MB
As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about -26.1pcm/%void at BOC and -21.7pcm/%void at EOC. Effects of about 10% of MA or about 2 % of FP on core performances were investigated, and they were confirmred within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used.
Department of Hot Laboratories
JAERI-Review 2002-039, 106 Pages, 2003/01
JAERI-Review-2002-039.pdf:9.46MB
no abstracts in English
Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Nakata, Tetsuo; Yan, X.; Takei, Masanobu; Kosugiyama, Shinichi; Shiozawa, Shusaku
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(4), p.352 - 360, 2002/12
no abstracts in English
Nomura, Yasushi; Mochizuki, Hiroki*
JAERI-Tech 2002-068, 131 Pages, 2002/11
JAERI-Tech-2002-068.pdf:5.59MB
no abstracts in English
Nagase, Fumihisa; Tanimoto, Masataka*; Uetsuka, Hiroshi
IAEA-TECDOC-1320, p.270 - 278, 2002/11
With a view to obtaining basic data for evaluating high burnup fuel behavior under LOCA conditions, a systematic research program is being conducted at JAERI. High-temperature oxidation tests with non-irradiated cladding have been performed to investigate separate effects of pre-oxidation and pre-hydriding on the oxidation kinetics. "Integral thermal shock tests" have been conducted simulating a LOCA condition to examine the influence of pre-hydriding on failure-bearing capability of oxidized cladding upon quenching. Test results showed almost no influence of absorbed hydrogen on the threshold value for oxidation amount under no axial restraint condition. On the other hand, it was shown that the threshold value is reduced by absorbed hydrogen for the restraint condition.
