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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

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

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.

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

Light water reactor fuel analysis code FEMAXI-V(Ver.1)

Suzuki, Motoe

JAERI-Data/Code 2000-030, 280 Pages, 2000/09

JAERI-Data-Code-2000-030.pdf:11.06MB

no abstracts in English

JAEA Reports

Outline of operation and control system and analytical investigation of transient behavior of an out-of-pile hydrogen production system for HTTR heat utilization system

Inagaki, Yoshiyuki; Hada, Kazuhiko; Nishihara, Tetsuo; Takeda, Tetsuaki; Hino, Ryutaro; Haga, Katsuhiro

JAERI-Tech 97-050, 125 Pages, 1997/10

JAERI-Tech-97-050.pdf:2.96MB

no abstracts in English

Journal Articles

Confirmatory thermal-hydraulic analyses of safety design concept of a passive safety light water reactor JPSR

Araya, Fumimasa; *; Ochiai, Masaaki

Proc. of Post-SMiRT14 Int. Seminar 18, p.E1.26 - E1.32, 1997/00

no abstracts in English

Journal Articles

Transient analysis for design of primary coolant pump adopted to JAERI passive safety reactor JPSR

Araya, Fumimasa; Murao, Yoshio; Iwamura, Takamichi

Journal of Nuclear Science and Technology, 32(10), p.1039 - 1046, 1995/10

 Times Cited Count:3 Percentile:37.28(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Possibility of a pressurized water reactor concept with highly inherent heat removel following capability

Araya, Fumimasa; Murao, Yoshio

Journal of Nuclear Science and Technology, 32(4), p.339 - 350, 1995/04

 Times Cited Count:3 Percentile:37.28(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Conceptual design of the JAERI passive safety reactor and its thermal-hydraulic characteristics

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

Transactions of the American Nuclear Society, 71, p.527 - 529, 1995/00

no abstracts in English

Journal Articles

Conceptual design of JAERI passive safety reactor (JPSR) and its thermal-hydraulic characteristics

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

10th Proc. of Nuclear Thermal Hydraulics, 0, p.3 - 12, 1994/00

no abstracts in English

Journal Articles

CHF experiments under steady-state and transient conditions for tight lattice core with non-uniform axial power distribution

Iwamura, Takamichi; Watanabe, Hironori; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(5), p.413 - 424, 1993/05

 Times Cited Count:2 Percentile:30.22(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Accident simulation tests for high conversion light water reactor using a high pressure water loop

Iwamura, Takamichi; Watanabe, Hironori; Araya, Fumimasa; Okubo, Tsutomu; Murao, Yoshio

JAERI-M 92-050, 46 Pages, 1992/03

JAERI-M-92-050.pdf:1.24MB

no abstracts in English

JAEA Reports

Journal Articles

Thermal transient analysis for structural integrity of high temperature components in HTTR

Hada, Kazuhiko; Fujimoto, Nozomu; Sudo, Yukio; Wada, Hozumi*

Proc. of the 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering,Vol. 2, p.291 - 298, 1991/00

no abstracts in English

JAEA Reports

Critical heat flux experiments for high conversion light water reactor, 3; Heated length: 0.5$$sim$$1.0m, P/D: 1.126$$sim$$1.2, number of rods: 4$$sim$$7

Iwamura, Takamichi; Okubo, Tsutomu; *; ; Murao, Yoshio

JAERI-M 90-044, 158 Pages, 1990/03

JAERI-M-90-044.pdf:2.81MB

no abstracts in English

JAEA Reports

Implementation of reactor safety analysis code CATHARE and its use on FACOM M-380

; *; ; *

JAERI-M 86-079, 85 Pages, 1986/05

JAERI-M-86-079.pdf:1.79MB

no abstracts in English

Journal Articles

Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

; ;

Nihon Genshiryoku Gakkai-Shi, 28(9), p.838 - 849, 1986/00

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Loss-of-Feedwater Transient Calculations for the ROSA-IV LSTF and the Reference PWR with RELAP5/MOD1(cycle 1)

C.P.Fineman*; ; Tasaka, Kanji

JAERI-M 83-088, 50 Pages, 1983/06

JAERI-M-83-088.pdf:1.5MB

no abstracts in English

25 (Records 1-20 displayed on this page)