Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 51

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Uncertainty analysis for source term evaluation of high temperature gas-cooled reactor under accident conditions; Identification of influencing factors in loss-of-forced circulation accidents

Honda, Yuki; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

Journal of Nuclear Engineering and Radiation Science, 4(3), p.031013_1 - 031013_11, 2018/07

There is growing interest in uncertainty analysis for probabilistic risk assessment (PRA). The focus of this research is to propose and trial investigate the new approach which identify influencing factors for uncertainty in a systematic manner for High Temperature Gas -cooled Reactor (HTGR). As a trial investigation, this approach is tested to evaluation of maximum fuel temperature in a depressurized loss-of-forced circulation (DLOFC) accident and failure of mitigation systems such as control rod systems from the view point of reactor dynamics and thermal hydraulic characteristics. As a result, 16 influencing factors are successfully selected in accordance with the suggested procedure. In the future, the selected influencing factors will be used as input parameter for uncertainty propagation analysis.

Journal Articles

Uncertainty analysis for source term evaluation of high temperature gas-cooled reactor under accident conditions; Identification of influencing factors in loss-of-forced circulation accidents

Honda, Yuki; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 9 Pages, 2017/07

There is growing interest in uncertainty analysis for probabilistic risk assessment (PRA). Our target is the uncertainty analysis method development for depressurized loss-of-forced circulation (DLOFC) accident with failure of control rod systems (CRS). As one of key elements, this paper focuses on the quantification of uncertainty for the fuel temperature which is dominant for a source term analysis. As an initial step, this paper aims to suggest a procedure to identify influencing factors which is input parameter for uncertainty analysis, and shows the results of derivation of variable parameters by expansion of dynamic equation and extraction of uncertainties in variable factors.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Nippon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

no abstracts in English

Journal Articles

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

Journal Articles

EBR-II passive safety demonstration tests benchmark analyses; Phase 2

Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08

The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.

Journal Articles

Improvement of transient analysis method of a sodium-cooled fast reactor with FAIDUS fuel sub-assemblies

Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code $$alpha$$-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.

Journal Articles

Development of risk assessment methodology of decay heat removal function against external hazards for sodium-cooled fast reactors, 3; Numerical simulations of forest fire spread and smoke transport as an external hazard assessment methodology development

Okano, Yasushi; Yamano, Hidemasa

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

Numerical simulations of forest fire propagation and smoke transport were performed with sensibility studies to weather conditions, and the effect by the smoke on the air filter was quantitatively evaluated. Forest fire propagation simulations were performed using FARSITE code. A temporal increase of a forest fire spread area, a position of the frontal fireline, "reaction intensity" and "frontal fireline intensity" are obtained and used for the smoke transport simulations by ALOFT-FT where spatial distribution of PM2.5 and PM10 are evaluated. The total amount of particle matter at the air filter at the nuclear power plant is around several hundred grams per m$$^{2}$$ which is well below the operational limit of the air filter of 15 kg/m$$^{2}$$.

Journal Articles

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 Times Cited Count:14 Percentile:11.2(Nuclear Science & Technology)

Journal Articles

Prediction of fission product release during the LOFC experiments at the HTTR

Shi, D.*; Xhonneux, A.*; Ueta, Shohei; Verfondern, K.*; Allelein, H.-J.*

Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 11 Pages, 2014/10

Demonstration tests were conducted using the High Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, to confirm the safety of HTGR technologies and assure the expected physical phenomena to occur under given conditions. As part of the OECD directed LOFC (loss of forced cooling) project, a series of three tests at the HTTR has been planned with tripping of all gas circulators while deactivating all reactor reactivity control to disallow reactor scram due to abnormal reduction of primary coolant flow rate. The tests fall into anticipated transient without scram (ATWS) with occurrence of reactor recriticality. The paper will describe the Source Term Analysis Code System (STACY) newly developed at the Research Center J$"u$lich and present the results of fission product behavior in the HTTR core under the LOFC test conditions. STACY encompasses the original verified and validated computer models for simulating fission product transport and release. For verification of the modernized and extended version, it was assured that results obtained with the original tools could be reproduced. One of the new features of STACY is its ability to also treat fuel compacts of (full) cylindrical or annular shape and a complete prismatic block reactor core, respectively, supposed sufficient input data be available. In the paper, calculations are based on time-dependent neutronics and fluid dynamics results obtained with the Serpent and MGT models.

Journal Articles

Study of metallofullerenes encapsulating actinides

Akiyama, Kazuhiko; Sueki, Keisuke*; Tsukada, Kazuaki; Yaita, Tsuyoshi; Miyake, Yoko*; Haba, Hiromitsu*; Asai, Masato; Kodama, Takeshi*; Kikuchi, Koichi*; Otsuki, Tsutomu*; et al.

Journal of Nuclear and Radiochemical Sciences, 3(1), p.151 - 154, 2002/06

The oxidation state of actinide elements encapsulated in fullerenes is studied. HPLC elution behavior of actinide-fullerenes is classified into two groups; the elution behavior of the first group, encapsulating U, Np, and Am, is similar to that of the light lanthanide-fullerenes, such as La, Ce, Pr, and Nd, while the behavior of the second group, encapsulating Th and Pa, is quite different from that of any lanthanide-fullerenes. The chemical species in the main HPLC elution peak of each group were identified as M@C82 and M@C84 (M = metal atom) from the mass of the U and Th fullerenes, respectively. The oxidation states of the U and Th atoms in the fullerenes were deduced to be 3+ and 4+, respectively, from the UV/vis/NIR absorption and XANES spectroscopy.

JAEA Reports

Transient thermal and stress analyses of the ITER shielding blanket/first wall under off-normal conditions

Furuya, Kazuyuki; Hashimoto, T.*; Sato, Satoshi; Kuroda, Toshimasa*; *; Kurasawa, Toshimasa; *; Takatsu, Hideyuki

JAERI-Tech 95-045, 53 Pages, 1995/09

JAERI-Tech-95-045.pdf:3.13MB

no abstracts in English

Journal Articles

Ceramic breeding blanket development for experimental fusion reactor in JAERI

Kurasawa, Toshimasa; Takatsu, Hideyuki; Sato, Satoshi; Mori, Seiji*; Hashimoto, T.*; Nakahira, Masataka; Furuya, Kazuyuki; Tsunematsu, Toshihide; Seki, Masahiro; Kawamura, Hiroshi; et al.

Fusion Engineering and Design, 27, p.449 - 456, 1995/00

 Times Cited Count:7 Percentile:38.9

no abstracts in English

JAEA Reports

JAEA Reports

Journal Articles

Results of OECD LOFT program

Nippon Genshiryoku Gakkai-Shi, 33(12), p.1112 - 1120, 1991/12

no abstracts in English

Journal Articles

Small break experiments LP-SB-1 and LP-SB-2 results and calculations

*; Anoda, Yoshinari; Kukita, Yutaka

The OECD/LOFT Project; Achievements and Significant Results, p.145 - 163, 1991/00

no abstracts in English

JAEA Reports

Prediction of thermal behavior of FBR fuel in the NSRR transient tests

Nakamura, Takehiko; Sobajima, Makoto; *

JAERI-M 90-140, 189 Pages, 1990/09

JAERI-M-90-140.pdf:3.97MB

no abstracts in English

JAEA Reports

Estimation of energy deposition in various test fuels in the NSRR FBR fuel tests

Nakamura, Takehiko; Fuketa, Toyoshi; Sobajima, Makoto; *

JAERI-M 90-067, 85 Pages, 1990/04

JAERI-M-90-067.pdf:2.28MB

no abstracts in English

51 (Records 1-20 displayed on this page)