Watanabe, So; Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Takahatake, Yoko; Shibata, Atsuhiro; Nomura, Kazunori; Kamiya, Yuichi*; Asanuma, Noriko*; Matsuura, Haruaki*; et al.
Progress in Nuclear Energy, 117, p.103090_1 - 103090_8, 2019/11
Sugawara, Takanori; Takei, Hayanori; Iwamoto, Hiroki; Oizumi, Akito; Nishihara, Kenji; Tsujimoto, Kazufumi
Progress in Nuclear Energy, 106, p.27 - 33, 2018/07
The Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides which will be partitioned from the high level waste. There are various inherent issues for the research and development on the ADS. The recent two activities to realize a feasible and reliable ADS concept are introduced in this paper. For the feasibility, the design of a beam window which is a boundary of the accelerator and the subcritical core, is one of the most important issues. To mitigate the design condition of the beam window, namely to reduce the proton beam current, the subcritical core concept with subcriticality adjustment rods were investigated. For the reliability, the beam-trip is the inherent and serious issue for the ADS design because it induces rapid temperature change to coolant and structures in the subcritical core. To improve the beam-trip frequencies, a double-accelerator concept was proposed and its beam-trip frequency was estimated.
Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin
Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11
The present paper describes the evaluation of the power reactivity loss data obtained in the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano, et al., 2008) is updated considering the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of the updates are quantitatively identified and the most precise and probable C/E value is derived together with a thorough uncertainty evaluation. As a result, it is revealed that the analysis overestimates the measurement by 4.6% for the measurement uncertainty of 2.0%. The discrepancy is reduced to as small as 1.1% when the core bowing effect is considered, which implies the importance of the core bowing effect in the calculation of the power reactivity loss.
Gunji, Satoshi; Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki
Progress in Nuclear Energy, 101(Part C), p.321 - 328, 2017/11
Criticality safety of fuel debris, particularly MCCI (Molten-Core-Concrete-Interaction) products, is one of the major safety issues for decommissioning of Fukushima Daiichi Nuclear Power Station. Criticality or subcriticality condition of the fuel debris is still uncertain; its composition, location, neutron moderation, etc. are not yet confirmed. The effectiveness of neutron poison in cooling water is also uncertain for use as a criticality control of fuel debris. A database of computational models is being built by Japan Atomic Energy Agency (JAEA), covering a wide range of possible conditions of such composition, neutron moderation, etc., to facilitate assessing criticality characteristics once fuel debris samples are taken and their conditions are known. The computational models also include uncertainties which are to be clarified by critical experiments. These experiments are planned and will be conducted by JAEA with the modified STACY (STAtic experiment Critical facilitY) and samples to simulate fuel debris compositions. Each of the samples will be cladded by a zircalloy tube whose outer shape is compatible with the fuel rod of STACY and loaded into an array of the fuel rods. This report introduces a study of experimental core configurations to measure the reactivity worth of samples simulating MCCI products. Parameters to be varied in the computation models for the experimental series are:(1) Uranium dioxide with U enrichments of 3, 4, and 5 wt.%; (2) Concrete volume fraction in the samples of 0, 20, 40, 60, and 80%; and (3) Porosity of the samples filled from 0 to 80% where the sample void is filled with water. It is concluded that the measurement is feasible in both under- and over-moderated conditions. Additionally, the required amount of samples was estimated.
Arima, Tatsumi*; Idemitsu, Kazuya*; Inagaki, Yaohiro*; Kawamura, Katsuyuki*; Tachi, Yukio; Yotsuji, Kenji
Progress in Nuclear Energy, 92, p.286 - 297, 2016/09
Diffusion and adsorption behavior of uranyl (UO) species is important for the performance assessment of radioactive waste disposal. The diffusion behaviors of UO, K, CO and Cl and HO in the aqueous solutions were evaluated by molecular dynamics (MD) calculations. The diffusion coefficient (De) of UO is the smallest and is 26% less than the self-diffusion coefficient of HO. For the aqueous solution with high concentration of carbonate ions, uranyl carbonate complexes: UOCO and UO(CO) can be observed. For the clay (montmorillonite or illite)-aqueous solution systems, the adsorption and diffusion behaviors of UO and K were evaluated by MD calculations. The distribution coefficients (Kd) increase with the layer charge of clay, and Kd of UO might be smaller than that of K. Further, their two-dimensional diffusion coefficients were relatively small in the adsorption layer and were extremely small for illite with higher layer charge.
Miwa, Shuhei; Yamashita, Shinichiro; Osaka, Masahiko
Progress in Nuclear Energy, 92, p.254 - 259, 2016/09
Cesium (Cs) and iodine (I) vapor species formed just after release from degraded fuels were predicted by means of the chemical equilibrium calculation with focuses on the effects of boron (B) release kinetics. Modified equations for the release kinetics of Cs, I and molybdenum (Mo) were utilized for evaluation of atmospheric dependences of their releases fractions. The release kinetics of B was evaluated considering the formation of iron (Fe)-B-O-H compounds. The release of B was enhanced above approximately 2250 K with the vapor species of CsBO under steam atmosphere, while the formation of CsBO was limited under steam-starvation atmosphere due to the much lower the release of B by the formation of low volatile Fe-B compounds. This limitation of CsBO formation would have resulted in a lesser formation of gaseous hydrogen iodine, HI, and a high volatile atomic I under steam-starvation atmosphere.
Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo
Progress in Nuclear Energy, 89, p.140 - 158, 2016/05
An experimental study was performed on the local structure of upward air-water two-phase flow in a vertical large diameter square duct by using a four-sensor probe. The four-sensor probe method classifying spherical and non-spherical bubbles was applied as a key measurement way to obtain local parameters such as 3-D bubble velocity vector, bubble diameter and interfacial area concentration. Both the local void fraction and interfacial area concentration indicated radial core-peak and wall-peak distributions at low and high liquid flow rates respectively. The 2 components of the bubble velocity vector in the cross-section revealed that there exists a rotating secondary flow in the octant symmetric triangular area and the magnitude of the rotating secondary flow increases with the liquid flow rate. Some of constitutive correlations of drift-flux model and interfacial area concentration are reviewed to study their predictabilities against the present data.
Minato, Futoshi; Iwamoto, Osamu
Progress in Nuclear Energy, 82, p.112 - 117, 2015/07
-decay and -delayed neutron emission of neutron-rich spherical nuclei are investigated. Our formalism adopts a self-consistent QRPA approach for -decay and Hauser-Feshbach statistical model for particle evaporation from highly excited state of daughter (precursor) nucleus. In this work, we particularly pay attention to the effects of two contributions. One is tensor force, which is not taken into account in conventional self-consistent QRPA but is important for reproducing half-lives of closed-shell nuclei. And another is isospin finite range pairing. They play a significant role to reduce energy of excited state of precursor nuclei. We found that these effects reduce the number of decay branches above neutron threshold of precursor nuclei and consequently a predicted -delayed neutron yields become smaller than that without them. This work is planned to apply to nuclear data evaluation of -delayed neutron yield of fission fragments in future.
Furukawa, Tomohiro; Rouillard, F.*
Progress in Nuclear Energy, 82, p.136 - 141, 2015/07
The application of a supercritical carbon dioxide (SC-CO) turbine cycle to fast rectors has the potential to enhance reliability because the SC-CO turbine system is capable of replacing the steam generator turbine system of conventional sodium cooled fast reactors. A key problem in the application is the corrosion of structural material by SC-CO at high temperatures. The authors have performed corrosion test on high-chromium martensitic and austenitic stainless steels in CO under the pressure conditions from atmospheric pressure to 25 MPa at elevated temperature, and proposed corrosion allowances of the steels for preliminary design of the SC-CO system. This paper initially reports the results of metallurgical examination of the steels after 8010 hours in SC-CO which is the longest immersion data in our experiments, and then describes the behavior of the oxide growth from the view point of estimation of the corrosion allowance for the design.
Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko
Progress in Nuclear Energy, 82, p.46 - 52, 2015/07
Safety requirements and design considerations for a HTGR hydrogen production system by IS process are examined. Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results clarified that design considerations suggested for coupling IS plant to HTGR are reasonably practicable.
Utsuno, Yutaka; Shimizu, Noritaka*; Otsuka, Takaharu*; Ebata, Shuichiro*; Homma, Michio*
Progress in Nuclear Energy, 82, p.102 - 106, 2015/07
no abstracts in English
Onishi, Takashi; Koyama, Shinichi; Mimura, Hitoshi*
Progress in Nuclear Energy, 82, p.69 - 73, 2015/07
no abstracts in English
Aoto, Kazumi; Dufour, P.*; Hongyi, Y.*; Glats, J. P.*; Kim, Y.-I.*; Ashurko, Y.*; Hill, R.*; Uto, Nariaki
Progress in Nuclear Energy, 77, p.247 - 265, 2014/11
Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phnix end-of-life tests, the restart of Monju, the lifetime extension of BN-600 and the startup of CEFR. Planned startup in 2014 for BN-800 and PFBR will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicated to sustainable energy generation and actinide management, and several advanced SFR concepts are under development. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.
Meiliza, Y.; Oki, Shigeo; Kawashima, Katsuyuki; Okubo, Tsutomu
Progress in Nuclear Energy, 70, p.270 - 278, 2014/01
Onishi, Takashi; Koyama, Shinichi; Shiba, Tomooki*; Sagara, Hiroshi*; Saito, Masaki*
Progress in Nuclear Energy, 57, p.125 - 129, 2012/05
In order to develop blanket fuel with high proliferation resistance in fast breeder reactor, chemical analysis of nine U samples irradiated in experimental fast reactor Joyo and Pu contents and Pu isotopic composition of the samples were measured. As results, dependence of Pu production behavior from U on neutron spectra was revealed.
Numakura, Masahiko*; Sato, Nobuaki*; Bessada, C.*; Okamoto, Yoshihiro; Akatsuka, Hiroshi*; Nezu, Atsushi*; Shimohara, Yasuaki*; Tajima, Keisuke*; Kawano, Hirokazu*; Nakahagi, Takeshi*; et al.
Progress in Nuclear Energy, 53(7), p.994 - 998, 2011/11
X-ray absorption fine structure (XAFS) measurements on thorium fluoride in molten lithium-calcium fluoride mixtures and molecular dynamics (MD) simulation of zirconium and yttrium fluoride in molten lithium-calcium fluoride mixtures have been carried out. In the molten state, coordination number of thorium and inter ionic distances between thorium and fluorine in the first neighbor are nearly constant in all mixtures. However the fluctuation factors (Debye-Waller factor and C cumulant) increase until CaF = 0.17 and decrease by addition of excess CaF. It means that the local structure around Th is disordered until CaF=0.17 and stabilized over CaF = 0.17. The variation of fluctuation factors is related to the number density of F in ThF mixtures and the stability of local structure around Th increases with decreasing the number density of F in ThF mixtures. This tendency is common to those in the ZrF and YF mixtures.
Meiliza, Y.; Oki, Shigeo; Okubo, Tsutomu
Progress in Nuclear Energy, 53(7), p.964 - 968, 2011/09
Furukawa, Tomohiro; Inagaki, Yoshiyuki; Aritomi, Masanori*
Progress in Nuclear Energy, 53(7), p.1050 - 1055, 2011/09
Compatibility of the FBR candidate materials, 12Cr-steel and 316FR, with supercritical CO pressurized at 20 MPa for up to 8000 hours at 400-600 C has been investigated. Corrosion due to the high temperature oxidation was measured in both steels. Results showed that the behavior differed greatly. For 12Cr-steel, weight gain showed parabolic growth as exposure time increased at each temperature, and no breakaway oxidation was observed. The specimens were covered by two successive oxide layers. For 316FR, the weight gain was significantly lower than that of 12Cr-steel, and good resistance against corrosion was observed. No dependency of temperature or immersed time on weight gain was observed. Based on the metallurgical examination, corrosion formula in supercritical CO has been shown for the candidate materials for provisional design.
Usuki, Toshiyuki; Yoshida, Katsumi*; Yano, Toyohiko*; Miwa, Shuhei; Osaka, Masahiko
Progress in Nuclear Energy, 53(7), p.1078 - 1081, 2011/09
The effects of sintering additives of magnesium silicates, i.e. enstatite (MgSiO), steatite (MgSiO) and forsterite (MgSiO), on the fabrication properties and characteristics of the silicon nitride ceramics based inert matrix fuels were experimentally investigated. CeO was selected as simulating element of AmO. Sintered pellets were characterized in term of their densities, thermal conductivities and solubility to nitric acid. The densifications of sintered bodies were enhanced by using additives of magnesium silicates at relative low sintering temperature. The relative density of silicon nitride ceramics based inert matrix fuels with forsterite were achieved above 90% at 1723 K. The thermal conductivities of silicon nitride ceramics based inert matrix fuels varied according to sintering temperature, and those sintered at 1923 K were above 34 W/m K. The grain boundary phases in Silicon nitride ceramics based inert matrix fuels found to be dissolved into HNO.
Permana, S.; Suzuki, Mitsutoshi
Progress in Nuclear Energy, 53(7), p.958 - 963, 2011/09
Material attractiveness evaluation based on isotopic plutonium barrier compositions have been investigated based on intrinsic feature of proliferation resistance such as decay heat (DH), spontaneous fission neutron (SFN), as well as attractiveness concepts of figure of merit (FOM) and attractiveness concept (ATTR) as a function of diluted fraction of even mass plutonium to Pu-239 composition.