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Journal Articles

Advanced reactor experiments for sodium fast reactor fuels (ARES) project; Transient irradiation experiments for metallic and MOX fuels

Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hiroka, Shun; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

Journal Articles

Numerical investigations on the coolability and the re-criticality of a debris bed with the density-stratified configuration

Li, C.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*

Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07

The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.

Journal Articles

Estimation of hydrogen gas production at transient criticality in uranyl nitrate solution

Yoshida, Ryoichiro; Yamane, Yuichi; Abe, Hitoshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.408 - 414, 2019/09

In a criticality accident, it is known that some kinds of radiolysis gases are generated mainly due to kinetic energy of fission fragments. Hydrogen gas (H$$_{2}$$) is one of them, which is able to initiate explosion. The rate of H$$_{2}$$ generation and its total amount can be estimated from the number of fission per second if its G value is known. In this study, it was tried to estimate G value of hydrogen gas (G(H$$_{2}$$)) by using the H$$_{2}$$ concentration measured as time-series data in Transient Experiment Critical Facility (TRACY) which was carried out by Japan Atomic Energy Agency. There was time lag in the measured H$$_{2}$$ concentration from its generation. To overcome those problems, measured profile of H$$_{2}$$ concentration was reproduced based on a hypothetical model and its total amount was evaluated. Based on the model, the obtained G(H$$_{2}$$) was 1.2.

Journal Articles

Application of transient pulse method to permeability measurement for clay

Kato, Masaji*; Nara, Yoshitaka*; Okazaki, Yuki*; Kono, Masanori*; Sato, Toshinori; Sato, Tsutomu*; Takahashi, Manabu*

Zairyo, 67(3), p.318 - 323, 2018/03

To ensure the safe geological disposal of radioactive wastes, it is important to determine the permeability of clays. The transient pulse test is suitable to apply to the low permeability materials, because it takes relatively short term to determine the permeability. Usually we increase the upstream pore pressure in the measurement with the transient pulse test. However, it is impossible to determine the permeability of clay in this procedure because of the increase of pore pressure. Therefore, the transient pulse test has never been applied to the determination of permeability of clays. In this study, we tried to apply the transient pulse test to a clay obtained in Mizunami Underground Research Laboratory to determine the permeability with decreasing the downstream pore pressure. It was clarified that the transient pulse test with decreasing downstream pore pressure is appropriate from the measurements of granite and sandstone. It was shown that the permeability of a clay was determined by the transient pulse test with decreasing the downstream pore pressure, which agreed with the permeability determined from the falling head test. The measurement time of the transient pulse test is much shorter than that of the falling head test. It is concluded that the transient pulse test is appropriate for the determination of the permeability of clays.

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

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

no abstracts in English

Journal Articles

Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka

Journal of Nuclear Materials, 480, p.386 - 392, 2016/11

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290$$^{circ}$$C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.

Journal Articles

The Effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11

 Times Cited Count:7 Percentile:66.89(Nuclear Science & Technology)

Journal Articles

The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

 Times Cited Count:4 Percentile:45.9(Nuclear Science & Technology)

Journal Articles

Prediction of heater surface temperature change at subcooled flow boiling DNB

Liu, W.; Podowski, M. Z.*

Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2015 Koen Rombunshu (CD-ROM), 2 Pages, 2015/10

This paper gives prediction to the transient heat transfer at Departure of Nucleate Boiling (DNB) point for subcooled flow boiling. The prediction is carried out by solving the heat conduction equations in cylindrical coordinates with convective boundary condition, which changes with the change of the heat transfer mode on the heated surface. DNB is assumed to happen at the complete dryout of liquid sublayer trapped between the heated wall and an elongated vapor clot, during the passing time of the vapor clot. Important parameters including initial thickness of the liquid sublayer, vapor clot length, vapor clot velocity and void fraction etc., are calculated from the Liu - Nariai model. The initial heater surface temperature is derived from the Jens-Lottes correlation. The transient changes of liquid sublayer thickness, surface temperature at DNB are reported. No obvious temperature jumping is observed at DNB. To predict temperate excursion at Critical Heat Flux (CHF), more simulations to the transient boiling and film boiling processes are needed.

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 diagnostic method for deep levels in semiconductors using charge induced by heavy ion microbeams

Kada, Wataru*; Kambayashi, Yuya*; Iwamoto, Naoya*; Onoda, Shinobu; Makino, Takahiro; Koka, Masashi; Kamiya, Tomihiro; Hoshino, Norihiro*; Tsuchida, Hidekazu*; Kojima, Kazutoshi*; et al.

Nuclear Instruments and Methods in Physics Research B, 348, p.240 - 245, 2015/04

 Times Cited Count:4 Percentile:40.75(Instruments & Instrumentation)

Journal Articles

Observation of deep levels and their hole capture behavior in p-type 4H-SiC epilayers with and without electron irradiation

Kato, Masashi*; Yoshihara, Kazuki*; Ichimura, Masaya*; Hatayama, Tomoaki*; Oshima, Takeshi

Japanese Journal of Applied Physics, 53(4S), p.04EP09_1 - 04EP09_5, 2014/04

 Times Cited Count:3 Percentile:16.04(Physics, Applied)

Journal Articles

Particle simulation of the transient behavior of one-dimensional SOL-divertor plasmas after an ELM crash

Takizuka, Tomonori; Hosokawa, Masanari*

Contributions to Plasma Physics, 46(7-9), p.698 - 703, 2006/09

 Times Cited Count:13 Percentile:44.28(Physics, Fluids & Plasmas)

Enhanced heat and particle fluxes to the divertor plates after an ELM crash in H-mode plasmas are the crucial issues for the tokamak reactor operation. Kinetic effect in the transient behaviour of SOL-divertor plasmas for this case is not yet well known. We investigate above problems with an advanced particle simulation code, PARASOL. Dependence of the particle and heat propagations on the collisionality is studied systematically. Effect of the particle recycling is also studied.

Journal Articles

Critical power prediction for tight lattice rod bundles

Liu, W.; Onuki, Akira; Tamai, Hidesada; Akimoto, Hajime

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

In this research, the newest version of critical power correlation for tight-lattice rod bundles is proposed by using 7-rod and 37-rod bundle data derived in Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux - critical quality type. For low mass velocity region, it is written in critical quality - annular flow length type. The correlation is verified by JAERI data and Bettis Atomic Power Laboratory data. It is confirmed the correlation is able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The correlation is further implemented into TRAC code to analyze flow decrease and power increase transients. It is confirmed transient BT can be predicted within the accuracy of the implemented critical power correlation.

Journal Articles

Proving test and analyze for critical power performance in the RMWR tight lattice rod bundles under transient condition

Liu, W.; Tamai, Hidesada; Onuki, Akira; Kureta, Masatoshi*; Sato, Takashi; Akimoto, Hajime

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 10 Pages, 2005/05

A major concern in the design of RMWR is that sufficient cooling capability be provided to keep fuel cladding temperature below specified values, even for a postulated abnormal transient process. In this research, centered the postulated transient cases that may be possibly met in the RMWR running, transient BT tests are performed in 7-rod and 37-rod double-humped tight lattice bundles, under the RMWR nominal operating condition (P = 7.2 MPa, Tin =556 K) for mass velocity G = 300 - 800 kg / (m$$^{2}$$s). Experiments are analyzed with TRAC code, in which JAERI critical power correlation is implemented for BT judgment. The traditional quasi-steady-state prediction of BT in transient process is confirmed being applicable for the postulated nominal transients in the RMWR cores.

Journal Articles

Temperature transient analysis of gas circulator trip test using the HTTR

Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30%(9MW). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.

JAEA Reports

Light water reactor fuel analysis code FEMAXI-6, 1; Detailed structure and user's manual

Suzuki, Motoe; Saito, Hiroaki*

JAERI-Data/Code 2003-019, 423 Pages, 2003/12

JAERI-Data-Code-2003-019.pdf:17.7MB

A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version with a number of improvements. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, permitting an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and the coupling with burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code is a versatile tool not only in the normal operation but also in transient conditions. This report describes the design, basic theory, models and numerical method, improvements, and model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output, and a sample output in an actual form are included.

Journal Articles

Time-dependent Boltzmann kinetic model of X-rays produced by ultrashort-pulse laser irradiation of argon clusters

Abdallah, J. Jr.*; Csanak, G.*; Fukuda, Yuji; Akahane, Yutaka; Aoyama, Makoto; Inoue, Norihiro*; Ueda, Hideki; Yamakawa, Koichi; Faenov, A. Y.*; Magunov, A. I.*; et al.

Physical Review A, 68(6), p.063201_1 - 063201_8, 2003/12

 Times Cited Count:52 Percentile:87.91(Optics)

The Boltzmann Equation and a detailed collisional radiative model are solved simultaneously as a function of time to model the time integrated X-ray spectra of the transient plasma produced by a high intensity ultrafast laser source. The results are in agreement with highly resolved F-like to He-like K-shell emission spectra recorded recently during ultrashort laser experiments with argon cluster targets in Japan.

JAEA Reports

Revised version of tokamak transient simulation code SAFALY, 2

Senda, Ikuo*; Fujieda, Hirobumi; Neyatani, Yuzuru; Tada, Eisuke; Shoji, Teruaki

JAERI-Data/Code 2003-012, 73 Pages, 2003/07

JAERI-Data-Code-2003-012.pdf:3.45MB

The tokamak transient simulation code, named SAFALY, was revised recently and the sensitivity analyses on the parameters in the code were carried out. This report is composed of two volumes. The formulation and the parameters in modeling the plasma and in-vessel components are described in the first volume. In this second volume, the results of the sensitivity studies are reported. The sensitivity studies were performed in two steps. In the first step, the responses of plasmas in the occurrence of plasma disturbances were analyzed for various initial conditions. For each disturbance, the initial condition of the plasma, which gave the largest increase of the fusion power, was identified. In the second step, by using initial conditions derived in the first step, the sensitivities of plasma reactions with respect to variation of the parameters in SAFALY code were analyzed. In the analyses, the increase of the fueling, the increase of the plasma confinement improvement factor and the increase of the auxiliary heating power were considered as plasma disturbances.

JAEA Reports

Mechanical properties changes of high burnup PWR fuel cladding by temperature transient

Nagase, Fumihisa; Uetsuka, Hiroshi

JAERI-Research 2002-023, 23 Pages, 2002/11

JAERI-Research-2002-023.pdf:1.94MB

To obtain basic data to evaluate fuel rod integrity during abnormal transient and accident of LWRs, high burnup PWR fuel claddings were heated for 0 to 600s at temperatures of 673 through 1173K, and the mechanical property changes were examined by using ring tensile test at room temperature. As a result of the test, it was shown that strength and ductility of the cladding are changed depending on heating temperature and time. The mechanical property changes by temperature transients are considered to be correspondent mainly to recovery of irradiation defect, recovery and recrystallization of the Zircaloy, phase transformations, and associated change of the hydride distribution and morphology. Comparison with unirradiated claddings suggested that irradiation effects are not completely annealed out by the short-term annealing at high temepratures. Radial change of hydrogen concentration was measured for the high burnup PWR fuel cladding and very high hydrogen concentration of about 2400wtppm was detected at the cladding periphery.

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