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Journal Articles

Interaction between caesium iodide particles and gaseous boric acid in a flowing system through a thermal gradient tube (1030 K-450 K) and analysis with ASTEC/SOPHAEROS

Gou$"e$llo, M.*; Hokkinen, M.*; Suzuki, Eriko; Horiguchi, Naoki; Barrachin, M.*; Cousin, F.*

Progress in Nuclear Energy, 138, p.103818_1 - 103818_10, 2021/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The present work aimed to study the transport of caesium iodide particles through a Thermal Gradient Tube (TGT) from 1023 K to 453 K. Retention inside the tube was evaluated for laminar flowrates composed of argon and steam. Higher retention of particles was highlighted for the experiments using higher steam content and lower flowrate. The second phase of the experiment aimed at identifying the possible revaporization or/and resuspension processes after the deposition. Three atmosphere compositions (Ar/H$$_{2}$$O, Ar/H$$_{2}$$ and Ar/Air) were investigated. The particles removed from what was deposited on the surface walls during the sampling phase exhibited a similar GMD in Ar/H$$_{2}$$O and Ar/H$$_{2}$$ and a bigger diameter in Ar/Air. The experimental results were then analysed with the SOPHAEROS module of the ASTEC code. Overall, the results obtained during the first phase were in agreement with the measured experimental results and during second phase led to no resuspension process.

Journal Articles

Development of a membrane reactor with a closed-end silica membrane for nuclear-heated hydrogen production

Myagmarjav, O.; Tanaka, Nobuyuki; Nomura, Mikihiro*; Noguchi, Hiroki; Imai, Yoshiyuki; Kamiji, Yu; Kubo, Shinji; Takegami, Hiroaki

Progress in Nuclear Energy, 137, p.103772_1 - 103772_7, 2021/07

 Times Cited Count:1 Percentile:81.22(Nuclear Science & Technology)

Journal Articles

Part 3, Evaluating a small modular high temperature reactor design during control rod withdrawal and a depressurised loss of coolant accidents

Atkinson, S.*; Aoki, Takeshi; Litskevich, D.*; Merk, B.*; Yan, X.

Progress in Nuclear Energy, 134, p.103689_1 - 103689_10, 2021/04

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

This article evaluates the safety features of the designed 10 MWth U-Battery concept with respect to a control rod withdrawal and a depressurised loss of coolant accident. This article provides the evaluation methodology for both transients, using a one-dimensional heat transfer model involving point reactor kinetic model to simulate reactor feedback in the control rod withdrawal. Overall, this work has shown that during the control rod withdrawal the fuel temperature rises by 110 K and at this point the excess reactivity is compensated by the negative temperature coefficient of the fuel. During the depressurised loss of coolant accident, the maximum fuel temperature reached 1455 K after 60 hours. This concludes that during both transients the temperatures maintained well below the maximum fuel operating temperature.

Journal Articles

A Review of separation processes proposed for advanced fuel cycles based on technology readiness level assessments

Baron, P.*; Cornet, S. M.*; Collins, E. D.*; DeAngelis, G.*; Del Cul, G.*; Fedorov, Y.*; Glatz, J. P.*; Ignatiev, V.*; Inoue, Tadashi*; Khaperskaya, A.*; et al.

Progress in Nuclear Energy, 117, p.103091_1 - 103091_24, 2019/11

 Times Cited Count:17 Percentile:88.22(Nuclear Science & Technology)

The results of an international review of separation processes for spent nuclear fuel (SNF) recycling in future closed fuel cycles with the evaluation of Technology Readiness Level are reported. This study was made by the Expert Group on Fuel Recycling Chemistry (EGFRC) organised by the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD). A unique feature of this study was that processes were classified according to a hierarchy of separations aimed at different elements within spent fuel (uranium; uranium-plutonium co-recovery; minor actinides; high heat generating radionuclides) and also the Head-end processes, used to prepare the SNF for chemical separation, were included. Separation processes covered both wet (hydrometallurgical) and dry (pyro-chemical) processes.

Journal Articles

STRAD project for systematic treatments of radioactive liquid wastes generated in nuclear facilities

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


 Times Cited Count:6 Percentile:75.46(Nuclear Science & Technology)

Journal Articles

Research and development activities for accelerator-driven system in JAEA

Sugawara, Takanori; Takei, Hayanori; Iwamoto, Hiroki; Oizumi, Akito; Nishihara, Kenji; Tsujimoto, Kazufumi

Progress in Nuclear Energy, 106, p.27 - 33, 2018/07

 Times Cited Count:9 Percentile:81.13(Nuclear Science & Technology)

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.

Journal Articles

A Refined analysis on the power reactivity loss measurement in Monju

Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin

Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

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.

Journal Articles

Study of experimental core configuration of the modified STACY for measurement of criticality characteristics of fuel debris

Gunji, Satoshi; Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki

Progress in Nuclear Energy, 101(Part C), p.321 - 328, 2017/11

 Times Cited Count:2 Percentile:28.88(Nuclear Science & Technology)

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 $$^{235}$$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.

Journal Articles

Diffusion and adsorption of uranyl ion in clays; Molecular dynamics study

Arima, Tatsumi*; Idemitsu, Kazuya*; Inagaki, Yaohiro*; Kawamura, Katsuyuki*; Tachi, Yukio; Yotsuji, Kenji

Progress in Nuclear Energy, 92, p.286 - 297, 2016/09

 Times Cited Count:6 Percentile:61.4(Nuclear Science & Technology)

Diffusion and adsorption behavior of uranyl (UO$$_2^{2+}$$) species is important for the performance assessment of radioactive waste disposal. The diffusion behaviors of UO$$_2^{2+}$$, K$$^{+}$$, CO$$_3^{2-}$$ and Cl$$^{-}$$ and H$$_{2}$$O in the aqueous solutions were evaluated by molecular dynamics (MD) calculations. The diffusion coefficient (De) of UO$$_2^{2+}$$ is the smallest and is 26% less than the self-diffusion coefficient of H$$_{2}$$O. For the aqueous solution with high concentration of carbonate ions, uranyl carbonate complexes: UO$$_{2}$$CO$$_{3}$$ and UO$$_{2}$$(CO$$_{3}$$)$$^{2-}$$ can be observed. For the clay (montmorillonite or illite)-aqueous solution systems, the adsorption and diffusion behaviors of UO$$_2^{2+}$$ and K$$^{+}$$ were evaluated by MD calculations. The distribution coefficients (Kd) increase with the layer charge of clay, and Kd of UO$$_2^{2+}$$ 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.

Journal Articles

Prediction of the effects of boron release kinetics on the vapor species of cesium and iodine fission products

Miwa, Shuhei; Yamashita, Shinichiro; Osaka, Masahiko

Progress in Nuclear Energy, 92, p.254 - 259, 2016/09

 Times Cited Count:14 Percentile:87.81(Nuclear Science & Technology)

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$$_{2}$$ under steam atmosphere, while the formation of CsBO$$_{2}$$ 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$$_{2}$$ formation would have resulted in a lesser formation of gaseous hydrogen iodine, HI, and a high volatile atomic I under steam-starvation atmosphere.

Journal Articles

Gas-liquid bubbly flow structure in a vertical large-diameter square duct

Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo

Progress in Nuclear Energy, 89, p.140 - 158, 2016/05

 Times Cited Count:13 Percentile:86.01(Nuclear Science & Technology)

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.

Journal Articles

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

Yoshioka, Kenichi*; Kitada, Takanori*; Nagaya, Yasunobu

Progress in Nuclear Energy, 82, p.7 - 15, 2015/07

 Times Cited Count:1 Percentile:12.03(Nuclear Science & Technology)

A reduced-moderation LWR has been developed for the reduction of spent fuel and for the efficient utilization of uranium resources. The streaming channel concept to improve the negative void reactivity coefficient is one of the features of the reactor. This concept makes the fuel assembly more heterogeneous. The geometrical heterogeneity makes azimuthal neutron flux distribution of fuel rods steep. To validate azimuthal neutron flux distribution calculation, we measured the distribution around fuel rods in reduced moderation LWR lattices. These measurements were conducted in NCA with the foil activation method. The core consisted of the central triangular tight lattice zone and the outer driver zone of a square lattice. A pile of polystyrene plates for simulating void fraction was installed into the triangular tight lattice. Azimuthal neutron flux distributions were deduced from the activity of these small foils measured with plastic scintillators. Measurements were compared to calculations by the MVP code with JENDL-3.3. It was found that calculations agreed well with measurements.

Journal Articles

An Analysis of $$beta$$-delayed neutron emission of even-even neutron-rich nuclei with proton-neutron QRPA

Minato, Futoshi; Iwamoto, Osamu

Progress in Nuclear Energy, 82, p.112 - 117, 2015/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

$$beta$$-decay and $$beta$$-delayed neutron emission of neutron-rich spherical nuclei are investigated. Our formalism adopts a self-consistent QRPA approach for $$beta$$-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 $$T=0$$ 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 $$beta$$-delayed neutron yields become smaller than that without them. This work is planned to apply to nuclear data evaluation of $$beta$$-delayed neutron yield of fission fragments in future.

Journal Articles

Oxidation and carburizing of FBR structural materials in carbon dioxide

Furukawa, Tomohiro; Rouillard, F.*

Progress in Nuclear Energy, 82, p.136 - 141, 2015/07

 Times Cited Count:23 Percentile:92.84(Nuclear Science & Technology)

The application of a supercritical carbon dioxide (SC-CO$$_{2}$$) turbine cycle to fast rectors has the potential to enhance reliability because the SC-CO$$_{2}$$ 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$$_{2}$$ at high temperatures. The authors have performed corrosion test on high-chromium martensitic and austenitic stainless steels in CO$$_{2}$$ 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$$_{2}$$ system. This paper initially reports the results of metallurgical examination of the steels after 8010 hours in SC-CO$$_{2}$$ 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.

Journal Articles

Safety design consideration for HTGR coupling with hydrogen production plant

Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

Progress in Nuclear Energy, 82, p.46 - 52, 2015/07

 Times Cited Count:8 Percentile:65.81(Nuclear Science & Technology)

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.

Journal Articles

Photonuclear reactions of calcium isotopes calculated with the nuclear shell model

Utsuno, Yutaka; Shimizu, Noritaka*; Otsuka, Takaharu*; Ebata, Shuichiro*; Homma, Michio*

Progress in Nuclear Energy, 82, p.102 - 106, 2015/07

 Times Cited Count:3 Percentile:32.14(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Removal of zirconium from spent fuel solution by alginate gel polymer

Onishi, Takashi; Koyama, Shinichi; Mimura, Hitoshi*

Progress in Nuclear Energy, 82, p.69 - 73, 2015/07

 Times Cited Count:4 Percentile:40.75(Nuclear Science & Technology)

no abstracts in English

Journal Articles

A Summary of sodium-cooled fast reactor development

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

 Times Cited Count:56 Percentile:98.8(Nuclear Science & Technology)

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 Ph$'e$nix 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.

Journal Articles

Enhancement of proliferation resistance properties of commercial FBRs by material barriers

Meiliza, Y.; Oki, Shigeo; Kawashima, Katsuyuki; Okubo, Tsutomu

Progress in Nuclear Energy, 70, p.270 - 278, 2014/01

 Times Cited Count:2 Percentile:10.7(Nuclear Science & Technology)

Journal Articles

Protected plutonium production at fast breeder reactor blanket; Chemical analysis of uranium-238 samples irradiated in experimental fast reactor Joyo

Onishi, Takashi; Koyama, Shinichi; Shiba, Tomooki*; Sagara, Hiroshi*; Saito, Masaki*

Progress in Nuclear Energy, 57, p.125 - 129, 2012/05

 Times Cited Count:1 Percentile:11.68(Nuclear Science & Technology)

In order to develop blanket fuel with high proliferation resistance in fast breeder reactor, chemical analysis of nine $$^{238}$$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 $$^{238}$$U on neutron spectra was revealed.

64 (Records 1-20 displayed on this page)