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Journal Articles

Behavior of high burnup advanced fuels for LWR during design-basis accidents

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09

Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.

JAEA Reports

Mercury flow experiments, 3; Simulation test plan under abnormal condition

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro

JAERI-Tech 2002-002, 22 Pages, 2002/02

JAERI-Tech-2002-002.pdf:5.37MB

The Neutron Scattering Facility will be utilized for advanced fields of Material and Life science using high intensity neutron generated by spallation reaction of 1MW pulsed proton beam and mercury target. Design of spallation mercury target system is in progress to obtain high neutron performance with high reliability and safety. The target system is using mercury and contains large amount of radioactive spallation products, therefore to establish the safety of the target system, transient behaviors of the system during anticipated events should be well understood. The safety protection system and an instrumentation system for detecting abnormal conditions must be designed based on the transient behaviors in order to terminate the transient events safely. Transient behaviors of the mercury system have been analyzed by using RELAP5 code. This report presents a test plan of mercury system transient phenomena during abnormal events using a mercury experimental loop and modification of the mercury experimental loop for abnormal transient simulation tests.

JAEA Reports

Water permeability test of rock specimen with natural fractures using high viscosity liquid

*;

JNC-TN8430 2001-006, 72 Pages, 2001/10

JNC-TN8430-2001-006.pdf:1.71MB

We had been conducted to study hydraulic permeability along fracture intersection by NETBLOCK system using natural rock specimen. Since the permeability of this rock specimen fracture is high, it was suggest that turbulent flow might be occurred in available range of measurement system. In case of turbulent flow, estimated permeability and fracture aperture from test data tend to be low. Therefore we should achieve laminar flow. This study was used the high viscosity liquid instead of water, and test conditions which could attain laminar flow with the rock specimen was examined. The rock specimen is granite rock, has natural Y-type fractures intersection. A solution of Methyl-cellulose is used as high viscosity liquid. Due to the high viscosity liquid, hydraulic head could be measured in the wide range, and high viscosity liquid improved the accuracy of measurement. Laminar flow could be achieved in the rock specimen by the high viscosity liquid over 0.1wt%.

JAEA Reports

Examination of hydraulic property of natural rock specimen

*; *;

JNC-TN8430 2001-003, 64 Pages, 2001/03

JNC-TN8430-2001-003.pdf:2.15MB

Handling methods and test conditions of hydraulic tests for NETBLOCK system had been examined by using acrylic and/or artificial rock specimen. A natural rock specimen (granite : excavated from Kamaishi mine) with fracture intersection was formed into practicable size for NETBLOCK system. Recently, we conducted a series of hydraulic test, in order to study the influence of fracture intersection by using the natural rock specimen. Hydraulic tests were conducted under several centimeters of head, which could be controlled by improved system because hydraulic permeability of target fractures were high. As a result, 10$$^{-4}$$$$sim$$10$$^{-5}$$(m$$^{2}$$/s) orders of hydraulic transmissivity of target fractures could be measured. A low permeability in the NW direction at the lower fracture was estimated from the heterogeneous head distribution. However, it is also expected that turbulence flow might be occurred under this study condition because fracture permeability is high and flow rate through the fracture is relatively high. In case of turbulence-flow, an estimated hydraulic transmissivity is low. High-viscosity fluid hydraulic test to achieve laminar flow will be needed for correcting an evaluated transmissivity.

JAEA Reports

Mechanical integrity of floor liner in secondaly heat transport system cells of Monju

; ; Ueno, Fumiyoshi; ; ; ;

JNC-TN2400 2000-005, 103 Pages, 2000/12

JNC-TN2400-2000-005.pdf:3.98MB

Inelastic analyses of the floor liner subjected to thermal loading due to sodium leakage and combustion were carried out, considering thinning of the liner plate due to molten salt type corrosion. Because the inelastic strain obtained by the analyses stayed below the ductility limit of the material, mechanical integrity, i.e., there exist no through-wall crack on the floor liner, was confirmed. Partial structural model tests were conducted, with a band of local thinning of the liner plate. Displacements were controlled to give specimens much larger strains than those obtained by the inelastic analyses above. No through-wall crack was observed by these tests. Mechanical integrity of the floor liner was confirmed by these results of the inelastic analyses and the partial structural model tests.

JAEA Reports

None

Kawada, Koji; ; ; ;

PNC-TN9450 97-005, 145 Pages, 1997/03

PNC-TN9450-97-005.pdf:2.48MB

None

JAEA Reports

Improvement of single-phase subchannel analysis code ASFRE-III; Modification and verification of porous blockage model

Ohshima, Hiroyuki

PNC-TN9410 96-128, 82 Pages, 1996/04

PNC-TN9410-96-128.pdf:2.42MB

The purpose of the study is to offer an analytical tool for investigations of local subassembly accidents which have been recognized as a key issue of the safety assessment of liquid metal cooled fast breeder reactors due to the dense structure of the fuel grid and the high power density. The porous blockage model, which was implemented in the single-phase subchannel analysis code ASFRE-III last year, was upgraded for the accuracy improvement. Correlation equations on pressure loss, heat conduction and heat transfer based on the packed bed theory was newly applied to the model in order to more precisely evaluate thermal hydraulic phenomena in the porous blockage regions. The model verification analysis was carried out on the basis of the data of porous blockage simulation test Scarlet-2 performed in France. In the analysis, it was confirmed that this model could reproduce coolant temperature distribution in the blockage region and could predict the peak temperature with high accuracy. The detailed model validation will be carried out by using the water/sodium out-of-pile tests which have scheduled to be conducted at PNC in the near future.

JAEA Reports

Planning study of in-pile loop tests for the evaluation of fission product transport

Nakagiri, Toshio; ; Ohno, Shuji; ; *; Koyama, Shinichi;

PNC-TN9510 94-001, 246 Pages, 1994/05

PNC-TN9510-94-001.pdf:6.38MB

None

JAEA Reports

Mockup test apparatus for the inspection system of steam generator tubes; Design and Manufacturing

; ; ; ;

PNC-TN9410 92-254, 76 Pages, 1992/07

PNC-TN9410-92-254.pdf:2.02MB

A verification test of the inspection system of Monju steam generator(SG) tubes will be performed in near future. Mockup Test Apparatus for the inspection system of SG tubes was manufactured and installed at Mechatronics Application Reserch Facility (MARF) in OEC. The test apparatus has the same specification, which is prepared for verification test, as Monju plant; for instance, which are dimension and material of tubes, and workability for the inspection equipment. About one hundred and forty SG tubes are radially arranged in tube sheets in Monju SG, however, three tubes, inner, center and outer one, are sellected in this test apparatus for testing of inspection system, It was verified that the test apparatus was manufactured with the same accuracy and dimension as Monju. System verification test is planned using this test apparatus.

JAEA Reports

Operational test (II) of falled fuel detection and location system (FFDL) of Joyo

Morimoto, Makoto; ; ; ; ; ;

PNC-TN9410 91-334, 64 Pages, 1991/10

PNC-TN9410-91-334.pdf:1.72MB

An failed fuel detection and location system (FFDL) using a sipping method is adopted as the FFDL of Joyo. FFDL has not operated since the first falled fuel simulated (FFDL-I) test in April, 1985 because Joyo has not yet experienced any operation with breached fuels. Therefore, the operational test (II) of FFDL was carried out on July 12$$sim$$19, 1991 for a preparation of the FFDL-II test which is scheduled in 1992. Main results from the test are as follows ; (1)The adequacy of the functions and operating procedure of FFDL was reaffirmed and the operating experience was gained. (2)Radioactivity measurement was conducted by FFDL for six subassemblies and their integrity was confirmed.

JAEA Reports

None

Mano, Tadashi*;

PNC-TN1410 91-035, 14 Pages, 1991/05

PNC-TN1410-91-035.pdf:0.25MB

no abstracts in English

Journal Articles

Simulation test on tube failure accident of PWC for HTTR

Hino, Ryutaro; *;

Nippon Genshiryoku Gakkai-Shi, 33(1), p.73 - 75, 1991/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

SIMMER-II Analysis of simulated core expansion experiments at purdue university

*; *

PNC-TN941 85-44, 75 Pages, 1985/03

PNC-TN941-85-44.pdf:1.83MB

In the analysis of a core expansion (or postdisassembly expansion) phase by the SIMMER-II code, it was shown that there exist various thermo-hydraulic phenomena available for mitigating effectively the mechanical energy released in a hypothetical core disruptive accident. To utilize SIMMER-II as a standard tool in future safety assessment, the experimental validation of the code is crucial especially on the energetics-mitigating effects. In this study, a series of simulated core expansion experiments performed at Purdue University was analyzed by SIMMER-II as the first effort of the code validation program in Japan. In the experiments, either the nitrogen gas at room temperature or the flashing water at high temperature was injected and expanded into the water pool simulating the outlet plenum of the reactor vessel (a 1/7-scaled model of the Clinch River Breeder Reactor vessel). In the analysis of the nitrogen expansion experiments, SIMMER-II could reproduce the experimentally measured slug impact time without adjusting input parameters. This means that the overall fluid-dynamics model of SIMMER-II is valid. In the flashing water expansion experiments, on the other hand, SIMMER did not reproduce the experimental data very well due to the presence of complex rate-limited processes including heat transfer and phase transiton. This discrepancy is ascribed to lack of modeling the entrainment phenomenon occurring at the interface of a vapor bubble. The effect of the entrainment is very important since the entrained cold liquid efficiently enhances the vapor condensation and hence reduces the slug kinetic energy. It was shown that this effect can be approximated by increasing the heat transfer coefficient between liquid components. Obviously, this result cannot be directly extrapolated to the reacor condition, but implys that the nominal SIMMER parameters are conservative from the energetics point of view because of underestimation of the vapor condensation. ...

Journal Articles

Effect of steam temperature on the deterioration of insulating materials under sequential LOCA test

; ; ; ; ; Yoshida, Kenzo

EIM-84-135, p.67 - 74, 1984/00

no abstracts in English

Journal Articles

Decomposition of EPRs in LOCA environments simulated by simultaneous and sequential procedures

Yoshida, Kenzo; ; ; ; ;

EIM-84-137, p.1 - 10, 1984/00

no abstracts in English

Journal Articles

Neutron irradiation damage in fusion reactor

Shiraishi, K.

Nippon Genshiryoku Gakkai-Shi, 20(9), p.620 - 625, 1978/09

 Times Cited Count:0

no abstracts in English

JAEA Reports

Radiation Resistance of Insulating Polymer Materials

Kuriyama, Isamu; Hayakawa, Naohiro;

JAERI-M 6751, 43 Pages, 1976/10

JAERI-M-6751.pdf:1.51MB

no abstracts in English

Oral presentation

Current status of fuel safety research at JAEA

Amaya, Masaki

no journal, , 

The objectives of the fuel safety research program at JAEA are to evaluate the adequacy of present safety criteria and safety margins, to provide a database for the regulation on improved fuels using new materials of cladding and pellet, and to provide reasonably mechanistic computer codes for regulatory application, in terms of light water reactor fuel. In this presentation, in addition to recent progress in the reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) test programs, an overview of the current status of the fuel safety research at JAEA is described.

Oral presentation

In-situ mock-up test on the hydrochemical influence of cement material to groundwater

Iwatsuki, Teruki; Murakami, Hiroaki; Watanabe, Yusuke; Fukuda, Kenji

no journal, , 

Drift closure test was carried out to understand the hydrochemical influence of cement materials to groundwater in Mizunami Underground Research Laboratory. In the end, the amount of cement material including portlandite to evolve the alkaline groundwater was quantitatively estimated 0.5mm depth of shotcrete.

Oral presentation

Development of multi-level simulation system for sodium-cooled fast reactor; Application of coupled 1D-CFD simulation to ULOF test of EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Murakami, Satoshi*; Tanaka, Masaaki

no journal, , 

The multi-level simulation system with 1D-CFD coupling method which enables to evaluate various phenomena from the whole plant dynamics to the local thermal hydraulics has been developed. The numerical simulation of the ULOF test in the experimental fast reactor EBR-II in the U.S is performed for validation study of the 1D-CFD coupling method, which combines a one-dimensional plant dynamics analysis (1D) code with a computational fluid dynamics (CFD) code. Through the numerical simulation, it was shown that the whole plant response and the multi-dimensional thermal hydraulics in the core upper plenum could be simulated. And the applicability of the 1D-CFD coupling method to plant scale analysis was confirmed in comparison with the experimental results.

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