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Journal Articles

Systematic effects on cross section data derived from reaction rates in reactor spectra and a re-analysis of $$^{241}$$Am reactor activation measurements

$v{Z}$erovnik, G.*; Schillebeeckx, P.*; Becker, B.*; Fiorito, L.*; Harada, Hideo; Kopecky, S.*; Radulovic, V.*; Sano, Tadafumi*

Nuclear Instruments and Methods in Physics Research A, 877, p.300 - 313, 2018/01

 Times Cited Count:2 Percentile:57.14(Instruments & Instrumentation)

Methodologies to derive cross section data from spectrum integrated reaction rates were studied. The Westcott convention and some of its approximations were considered. The accuracy of the results strongly depends on the assumptions that are made about the neutron energy distribution, which is mostly parameterised as a sum of a thermal and an epi-thermal component. Resonance integrals derived from such data can be strongly biased. When the energy dependence of the cross section is known and information about the neutron energy distribution is available, a method to correct for a bias on the cross section at thermal energy is proposed. Reactor activation measurements to determine the thermal $$^{241}$$Am(n, $$gamma$$) cross section reported in the literature were reviewed, where the results were corrected to account for possible biases. These data combined with results of time-of-flight measurements give a capture cross section 720 (14) b for $$^{241}$$Am(n, $$gamma$$) at thermal energy.

Journal Articles

Improving nuclear data accuracy of $$^{241}$$ Am and $$^{237}$$ Np capture cross sections

$v{Z}$erovnik, G.*; Schillebeeckx, P.*; Cano-Ott, D.*; Jandel, M.*; Hori, Junichi*; Kimura, Atsushi; Rossbach, M.*; Letourneau, A.*; Noguere, G.*; Leconte, P.*; et al.

EPJ Web of Conferences, 146, p.11035_1 - 11035_4, 2017/09

 Times Cited Count:4 Percentile:2.98

Journal Articles

RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project

Abe, Satoshi; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo

Journal of Nuclear Science and Technology, 51(10), p.1164 - 1176, 2014/10

 Times Cited Count:5 Percentile:52.58(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Meeting nuclear data needs for advanced reactor systems

Harada, Hideo; Shibata, Keiichi; Nishio, Katsuhisa; Igashira, Masayuki*; Plompen, A.*; Hambsch, F.-J.*; Schillebeeckx, P.*; Gunsing, F.*; Ledoux, X.*; Palmiotti, G.*; et al.

NEA/NSC/WPEC/DOC(2014)446, 111 Pages, 2014/02

Journal Articles

RELAP5/MOD3.2 sensitivity analysis using OECD/NEA ROSA-2 project 17% cold leg intermediate-break LOCA test data

Takeda, Takeshi; Watanabe, Tadashi; Maruyama, Yu; Nakamura, Hideo

NEA/CSNI/R(2013)8/PART2 (Internet), p.173 - 183, 2013/12

An experiment simulating a PWR cold leg IBLOCA with 17% break at cold leg was conducted. The post-test analysis by RELAP5/MOD3.2.1.2 code revealed that cladding surface temperature was underpredicted due to later major core uncovery. The post-test analysis conditions were considered as Base Case assuming the discrepancies were caused by uncertainties in the code predictability and input data. Key phenomena and related important parameters, which may affect the cladding surface temperature, were selected based on the LSTF test data analysis and post-test analysis results. Sensitivity analyses were performed by changing the parameters relevant to the key phenomena within the ranges to investigate influences of the parameters onto the cladding surface temperature. It was confirmed that both constant C of Wallis CCFL correlation at the core exit and inter-phase drag in the core are more sensitive to the cladding surface temperature.

Journal Articles

Measurement of non-condensable gas in a PWR small-break LOCA simulation test with LSTF for OECD/NEA ROSA project and RELAP5 post-test analysis

Takeda, Takeshi; Owada, Akihiko; Nakamura, Hideo

Experimental Thermal and Fluid Science, 51, p.112 - 121, 2013/11

 Times Cited Count:10 Percentile:46.45(Thermodynamics)

Journal Articles

A Preliminary 3D steam flow analysis for CET behavior during LSTF SBLOCA experiment using FLUENT code

Irwanto, D.; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo

Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07

Journal Articles

OECD/NEA ROSA project experiment on steam condensation in PWR horizontal legs during large-break LOCA

Takeda, Takeshi; Otsu, Iwao; Nakamura, Hideo

Journal of Energy and Power Engineering, 7(6), p.1009 - 1022, 2013/06

Journal Articles

LSTF test on cet performance during PWR hot leg small-break LOCA and RELAP5 analysis

Takeda, Takeshi; Otsu, Iwao; Nakamura, Hideo

Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 12 Pages, 2013/05

Journal Articles

Major outcomes from OECD/NEA ROSA and ROSA-2 projects

Nakamura, Hideo; Takeda, Takeshi; Satou, Akira; Ishigaki, Masahiro; Abe, Satoshi; Irwanto, D.

Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 21 Pages, 2013/05

Journal Articles

Guidelines for thermodynamic sorption modelling in the context of radioactive waste disposal

Payne, T. E.*; Brendler, V.*; Ochs, M.*; Baeyens, B.*; Brown, P. L.*; Davis, J. A.*; Ekberg, C.*; Kulik, D.*; Lutzenkirchen, J.*; Missana, T.*; et al.

Environmental Modelling & Software, 42, p.143 - 156, 2013/04

 Times Cited Count:27 Percentile:21.86(Computer Science, Interdisciplinary Applications)

Thermodynamic sorption models (TSMs) can be utilised to provide a scientific basis for Kd setting in the safety case, and for assessing the response of Kd to changes in chemical conditions. The TSM development involves a series of decisions on model features such as surface sites, sorption reactions and electrostatic corrections. There is a lack of consensus on the best ways to develop TSMs, and the NEA has therefore co ordinated an international project to assess the strategies and processes for building a TSM. This paper presents recommendations from the project on a number of aspects of TSM development in the context of radioactive waste disposal. Key recommendations include: definition of modelling objectives, identification of major decision points, a clear decision making rationale with reference to experimental or theoretical evidence, a suitable consultative and iterative model development process, testing to the maximum practicable extent, and documentation of key decisions.

Journal Articles

OECD/NEA ROSA Project experiment on steam condensation in PWR horizontal legs during large-break LOCA

Takeda, Takeshi; Otsu, Iwao; Nakamura, Hideo

Proceedings of 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference (ICONE-20 & POWER 2012) (DVD-ROM), 11 Pages, 2012/07

Journal Articles

RELAP5 analyses of OECD/NEA ROSA-2 project experiments on intermediate-break LOCAs at hot leg or cold leg

Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

Journal of Power and Energy Systems (Internet), 6(2), p.87 - 98, 2012/06

Journal Articles

RELAP5 analysis of OECD/NEA ROSA project experiment simulating a PWR loss-of-feedwater transient with high-power natural circulation

Takeda, Takeshi; Asaka, Hideaki*; Nakamura, Hideo

Science and Technology of Nuclear Installations, 2012, p.957285_1 - 957285_15, 2012/00

 Times Cited Count:10 Percentile:31.69(Nuclear Science & Technology)

A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater transient with assumptions of failure of scram and total failure of HPI system. AFW was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and SG secondary-side pressures were maintained respectively at around 16 and 8 MPa by cycle opening of PZR PORV and SG relief valves. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis CCFL correlation. Problems remain in the predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

Journal Articles

RELAP5 analyses of OECD/NEA ROSA-2 project experiments on intermediate-break LOCAs at hot leg or cold leg

Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10

Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 project using LSTF. In the hot leg IBLOCA test, core uncovery appeared simultaneously with loop seal clearing (LSC). Water remained on upper core plate in upper plenum due to CCFL. In the cold leg IBLOCA test, core dryout took place before LSC. Liquid was accumulated in upper plenum, SG U-tube upflow-side and SG inlet plenum before the LSC due to CCFL. The RELAP5/MOD3.2.1.2 post-test analyses were performed. In the hot leg IBLOCA case, cladding surface temperature was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

Journal Articles

Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA Project

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

Journal of Power and Energy Systems (Internet), 3(1), p.146 - 157, 2009/00

Presented are experiment results of the LSTF with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break LOCA simulation experiment. The break size is equivalent to 1.9% cold leg break. The accident management (AM) action to rapidly open the SG relief valves was initiated when CET temperature rose up to 623 K. The core overheat, however, was detected with a time delay of about 230 s and a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarified the reasons of time delay and temperature discrepancy between the CETs and heated core including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to PWR conditions and a possibility of alternative indicators for earlier AM action is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

Journal Articles

RELAP5 post-test analyses of OECD/NEA ROSA project experiments on steam generator depressurization with or without non-condensable gas inflow

Takeda, Takeshi; Asaka, Hideaki*; Watanabe, Tadashi; Nakamura, Hideo

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11

Two LSTF experiments were conducted for OECD/NEA ROSA Project simulating PWR 0.5% cold leg small break LOCA. Steam generator (SG) secondary-side depressurization was performed by fully opening the relief valves at 10 minutes after safety injection signal with or without non-condensable gas (air) inflow from accumulator tanks with total failure of high pressure injection system. Further assumptions were made to conduct enhanced SG depressurization by fully opening the safety valves when the primary pressure decreased to 2 MPa and no actuation of low pressure injection system, both to well observe natural circulation (NC) phenomena at low pressures. The primary depressurization rate decreased when non-condensable gas started to enter primary loops because of degradation in the condensation heat transfer in SG U-tubes, while two-phase flow NC has continued even after non-condensable gas inflow. Asymmetric NC behaviors appeared between two loops due probably to different number of forward flow SG U-tubes which would have been under influences of non-condensable gas. Post-test analyses by using JAEA-modified RELAP5/MOD3.2.1.2 code indicated that the code has remaining problems in proper prediction of primary loop flow rate and SG U-tube liquid level behaviors especially after non-condensable gas inflow. The improvement of the condensation heat transfer model under non-condensable gas mixture condition and the SG U-tube model may be necessary for correct analysis of the LSTF SG depressurization transients.

Journal Articles

RELAP5 analysis of OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

Takeda, Takeshi; Asaka, Hideaki*; Nakamura, Hideo

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

OECD/NEA ROSA Project experiment with the Large Scale Test Facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR analysis with coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment, while it overpredicted the break flow rate.

Journal Articles

RELAP5 analysis of ROSA/LSTF vessel upper head break LOCA experiment

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

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 3 Pages, 2007/09

RELAP5 code analysis was performed to validate the code predictability by using ROSA/LSTF experiment data that simulated a PWR vessel upper head small break loss-of-coolant accident (SBLOCA) with a break equivalent to 1% cold leg break. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient (Cd) of 0.61 for two-phase break flow. In the experiment, liquid level in the upper head was found to control break flow rate as coolant in the upper plenum entered the upper head through control rod guide tubes (CRGTs) until the penetration holes at the CRGT bottom were exposed to steam in the upper plenum. The upper head noding and flow paths between the upper plenum and the CRGT were thus modeled to simulate well the liquid level and coolant flow around the upper portion of pressure vessel. The code, however, overpredicted the break flow rate due to the underprediction of break-upstream void fraction especially during two-phase flow discharge period. Cd for two-phase break flow was thus adjusted to be 0.58. Effects of break area on the core cooling were investigated further. The parameter analyses showed that peak cladding temperature (PCT) is the maximum at 1% break case, while the PCT would be lower than 1200 K in the larger break size cases because vapor condensation on injected accumulator coolant induces loop seal clearing and effectively enhances core cooling thereafter.

Oral presentation

OECD/NEA ROSA Project, 5; Experiment on PWR PV bottom break LOCA and post-test analysis

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

no journal, , 

no abstracts in English

36 (Records 1-20 displayed on this page)