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Journal Articles

Integration of pool scrubbing research to enhance source-term calculations (IPRESCA) project

Gupta, S.*; Herranz, L. E.*; Lebel, L. S.*; Sonnenkalb, M.*; Pellegrini, M.*; Marchetto, C.*; Maruyama, Yu; Dehbi, A.*; Suckow, D.*; K$"a$rkel$"a$, T.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

JAEA Reports

Proceedings of the 2019 Symposium on Nuclear Data; November 28-30, 2019, Kyushu University, Chikushi Campus, Fukuoka, Japan

Watanabe, Yukinobu*; Shigyo, Nobuhiro*; Kin, Tadahiro*; Iwamoto, Osamu

JAEA-Conf 2020-001, 236 Pages, 2020/12

JAEA-Conf-2020-001.pdf:13.75MB

The 2019 Symposium on Nuclear Data was held at Chikushi Campus Cooperation Building (C-Cube), Kyushu University, on November 28 to 30, 2019. The symposium was organized by the Nuclear Data Division of the Atomic Energy Society of Japan (AESJ) in cooperation with Sigma Investigative Advisory Committee of AESJ, Nuclear Science and Engineering Center of Japan Atomic Energy Agency, Kyushu Branch of AESJ, and Center for Accelerator and Beam Applied Science of Kyushu University. In the symposium, there were one tutorial, "From the resonance theory to statistical model", and five sessions, "Study on Nuclear Data and related topics", "Reactor physics", "International Cooperation", "Nuclear Physics", and "High Energy Nuclear Data and their Application". In addition, recent research progress on experiments, nuclear theory, evaluation, benchmark and applications was presented in the poster session. Among 85 participants, all presentations and following discussions were very active and fruitful. This report consists of total 42 papers including 13 oral and 29 poster presentations.

Journal Articles

Systematic measurements and analyses for lead void reactivity worth in a plutonium core and two uranium cores with different enrichments

Fukushima, Masahiro; Goda, J.*; Oizumi, Akito; Bounds, J.*; Cutler, T.*; Grove, T.*; Hayes, D.*; Hutchinson, J.*; McKenzie, G.*; McSpaden, A.*; et al.

Nuclear Science and Engineering, 194(2), p.138 - 153, 2020/02

 Times Cited Count:2 Percentile:17.09(Nuclear Science & Technology)

To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worth was conducted systematically in three fast spectra with different fuel compositions on the Comet critical assembly of the National Criticality Experiments Research Center. Previous experiments in a high-enriched uranium (HEU)/Pb and a low-enriched uranium (LEU)/Pb systems had been performed in 2016 and 2017, respectively. A follow-on experiment in a plutonium (Pu)/Pb system has been completed. The Pu/Pb system was constructed using lead plates and weapons grade plutonium plates that had been used in the Zero Power Physics Reactor (ZPPR) of Argonne National Laboratory until the 1990s. Furthermore, the HEU/Pb system was re-examined on the Comet critical assembly installed newly with a device that can guarantee the gap reproducibility with a higher accuracy and precision, and then the experimental data was re evaluated. Using the lead void reactivity worth measured in these three cores with different fuel compositions, the latest nuclear data libraries, JENDL 4.0 and ENDF/B VIII.0, were tested with the Monte Carlo calculation code MCNP version 6.1. As a result, the calculations by ENDF/B-VIII.0 were confirmed to agree with lead void reactivity worth measured in all the cores. It was furthermore found that the calculations by JENDL 4.0 overestimate by more than 20% for the Pu/Pb core while being in good agreements for the HEU/Pb and LEU/Pb cores.

JAEA Reports

Proceedings of the 2018 Symposium on Nuclear Data; November 29-30, 2018, Tokyo Institute of Technology, Ookayama Campus, Tokyo, Japan

Chiba, Satoshi*; Ishizuka, Chikako*; Tsubakihara, Kosuke*; Iwamoto, Osamu

JAEA-Conf 2019-001, 203 Pages, 2019/11

JAEA-Conf-2019-001.pdf:18.86MB

The 2018 Symposium on Nuclear Data was held at Multi-Purpose Digital Hall and Collaboration Room of Tokyo Institute of Technology, on November 29 and 30, 2018. The symposium was organized by the Nuclear Data Division of the Atomic Energy Society of Japan (AESJ) in cooperation with Sigma Special Committee of AESJ, Nuclear Science and Engineering Center of Japan Atomic Energy Agency, and Laboratory for Advanced Nuclear Energy of Institute of Innovative Research, Tokyo Institute of Technology. In the symposium, there were one tutorial, "Development of nuclear data processing code FRENDY", one special lecture "What the future holds for Nuclear Energy" and seven oral sessions, "Nuclear Data and Future Perspectives", "Current Status and Future Perspectives of Reactor Physics", "Topics", "Nuclear Data Applications", "International Session", "Nuclear Data Measurements and New Technology for Nuclear Reactor Diagnosis", and "Data Needs from New Fields". In addition, recent research progress on experiments, evaluation, benchmark and application was presented in the poster session. Among 82 participants, all presentations and following discussions were very active and fruitful. This report consists of total 35 papers including 13 oral and 22 poster presentations.

Journal Articles

Lead void reactivity worth in two critical assembly cores with differing uranium enrichments

Fukushima, Masahiro; Goda, J.*; Bounds, J.*; Cutler, T.*; Grove, T.*; Hutchinson, J.*; James, M.*; McKenzie, G.*; Sanchez, R.*; Oizumi, Akito; et al.

Nuclear Science and Engineering, 189, p.93 - 99, 2018/01

 Times Cited Count:6 Percentile:57.41(Nuclear Science & Technology)

To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worths was conducted in a high-enriched uranium (HEU)/Pb system and a low enriched uranium (LEU)/Pb system using the Comet Critical Assembly at NCERC. The critical experiments were designed to provide complementary data sets having different sensitivities to scattering cross sections of lead. The larger amount of the $$^{238}$$U present in the LEU/Pb core increases the neutron importance above 1 MeV compared with the HEU/Pb core. Since removal of lead from the core shifts the neutron spectrum to the higher energy region, positive lead void reactivity worths were observed in the LEU/Pb core while negative values were observed in the HEU/Pb core. Experimental analyses for the lead void reactivity worths were performed with the Monte Carlo calculation code MCNP6.1 together with nuclear data libraries, JENDL 4.0 and ENDF/B VII.1. The calculation values were found to overestimate the experimental ones for the HEU/Pb core while being consistent for the LEU/Pb core.

Journal Articles

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

Fission rate ratios of FCA-IX assemblies as integral experiment for assessment of TRU's fission cross sections

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

EPJ Web of Conferences, 111, p.07002_1 - 07002_5, 2016/03

 Times Cited Count:2 Percentile:84.04

In the IX-th series at FCA, central fission rate ratios for TRU such as $$^{237}$$Np, $$^{238}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm and $$^{239}$$Pu were measured in the seven uranium-fueled assemblies with systematically changed neutron spectra. The FCA-IX assemblies were constructed with simplicity both in geometry and composition. Taking their advantage, benchmark models with respect to central fission rate ratios had been recently established for the assessment of the TRU's fission cross sections. By virtue of these FCA-IX assemblies where the simple combinations of uranium fuel and diluent in their core regions were systematically varied, the neutron spectra of these benchmark models cover those of various reactor types, from fast to epithermal reactors. As an application of these benchmark models, JENDL-4.0 was utilized by a Monte Carlo calculation code. Several results show obvious discrepancies between calculation and experimental values. The benchmark models would be well suited for the assessment and improvement of the nuclear data for TRU's fission cross sections.

Journal Articles

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 52(2), p.282 - 293, 2015/02

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

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the "finite neutron multiplication factor", $$k^ast$$, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and $$k^ast$$ on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Journal Articles

Journal Articles

Effects of secondary depressurization on core cooling in PWR vessel bottom small break LOCA experiments with HPI failure and gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01

 Times Cited Count:10 Percentile:58.85(Nuclear Science & Technology)

Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.

JAEA Reports

Proceedings of the 11th International Workshop on Ceramic Breeder Blanket Interactions; December 15 - 17, 2003, Tokyo, Japan

Enoeda, Mikio

JAERI-Conf 2004-012, 237 Pages, 2004/07

JAERI-Conf-2004-012.pdf:44.1MB

This report is the Proceedings of "the Eleventh International Workshop on Ceramic Breeder Blanket Interactions" which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li$$_{2}$$TiO$$_{3}$$ and other breeders, fabrication technology developments and characterization of the Li$$_{2}$$TiO$$_{3}$$ and Li$$_{4}$$SiO$$_{4}$$ pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li$$_{2}$$TiO$$_{3}$$ and Li$$_{4}$$SiO$$_{4}$$ pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system.

JAEA Reports

Present status of chemical analysis of uranyl nitrate solution used for the criticality experiments in NUCEF

Haga, Takahisa*; Gunji, Kazuhiko; Fukaya, Hiroyuki; Sonoda, Takashi; Sakazume, Yoshinori; Sakai, Yutaka; Niitsuma, Yasushi; Togashi, Yoshihiro; Miyauchi, Masakatsu; Sato, Takeshi; et al.

JAERI-Tech 2004-005, 54 Pages, 2004/02

JAERI-Tech-2004-005.pdf:2.06MB

Criticality experiments using uranyl nitrate solution fuel are being conducted at STACY (the Static Experiment Critical Facility) and TRACY (the Transient Experiment Critical Facility) in NUCEF (the Nuclear Fuel Cycle Safety Engineering Research Facility). Chemical analyses of the solution have been carried out to take necessary data for criticality experiments, for treatment and control of the fuel, and for safeguards purpose at the analytical laboratory placed in NUCEF. About 300 samples are analyzed annually that provide various kinds of data, such as uranium concentration, isolation acid concentration, uranium isotopic composition, concentration of fission product (FP) nuclides, tri-butyl phosphoric acid (TBP) concentration, impurities in the solution fuel and so on. This report summarizes the analytical methods and quality management of the analysis for uranyl nitrate solution relating to the criticality experiments.

JAEA Reports

Summary of the 6th Workshop on the Reduced-Moderation Water Reactor; March 6, 2003, JAERI, Tokai

Nabeshima, Kunihiko; Nakatsuka, Toru; Ishikawa, Nobuyuki; Uchikawa, Sadao

JAERI-Conf 2003-020, 240 Pages, 2003/11

JAERI-Conf-2003-020.pdf:27.66MB

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since 1998 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The workshop began with five lectures on status of research and development on RMWRs in JAERI entitled "Status and Future Program of Research and Development on Reduced-Moderation Water Reactors", "Design of Small Reduced-Moderation Water Reactors", "Critical Experiments for Reduced-Moderation Water Reactors", "Critical Heat Flux Experiments in Tight Lattice Core" and "Development of High Performance Cladding". Then two lectures followed: "Status of Phase II of Feasibility Studies on Commercialized Fast Breeder Reactor System" by JNC and "Present Status of Study on Super-critical water Cooled Power Reactor" by Toshiba Corporation.

JAEA Reports

Annual report of Department of Research Reactor 2001; April 1, 2001 - March 31, 2002

Department of Research Reactor

JAERI-Review 2002-033, 215 Pages, 2002/12

JAERI-Review-2002-033.pdf:16.19MB

no abstracts in English

Journal Articles

Nuclear characteristics evaluation for a supercritical experiment facility using low-enriched uranium solution fuel, TRACY

Nakajima, Ken

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 8 Pages, 2002/10

The nuclear characteristics of TRACY, such as the criticality, the $$beta$$$$_{eff}$$/$$Lambda$$ ratio, the peak power, the energy of pulse, and the total energy, have been evaluated using the experimental data. TRACY is a supercritical reactor fueled with low-enriched uranyl nitrate aqueous solution to simulate criticality accidents in a fuel processing facility, such as a spent-fuel reprocessing plant. In this evaluation, the availability of criticality calculation and the models to evaluate the power and energy have been studied.

Journal Articles

Depressurization effects of vacuum vessel pressure supression systems in fusion reactors at multiple first wall pipe break events

Takase, Kazuyuki; Akimoto, Hajime

Applied Electromagnetics in Materials, p.177 - 178, 2001/00

no abstracts in English

Journal Articles

Analysis of WITCH/LINER experiments on heat transfer between gas-agitated steel melt and vertical wall

Maruyama, Yu; Sugimoto, Jun

Journal of Nuclear Science and Technology, 36(10), p.914 - 922, 1999/10

 Times Cited Count:3 Percentile:29.36(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Experimental and analytical studies on high-speed plane jet along concave wall simulating IFMIF Li target flow

Nakamura, Hideo; Ito, Kazuhiro*; Kukita, Yutaka*; *; ; Maekawa, Hiroshi; Katsuta, Hiroji

Journal of Nuclear Materials, 258-263, p.440 - 445, 1998/00

 Times Cited Count:7 Percentile:54.57(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Current studies on low temperature irradiation experiments

Iwase, Akihiro

Materia, 37(6), p.470 - 478, 1998/00

no abstracts in English

Journal Articles

Small break LOCA tests at ROSA-V/LSTF on next generation PWR designs

Yonomoto, Taisuke; ; ; Anoda, Yoshinari; Kukita, Yutaka*

Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 1, p.535 - 542, 1997/00

no abstracts in English

43 (Records 1-20 displayed on this page)