Insights from review and analysis of the Fukushima Dai-ichi accident

Hirano, Masashi; Yonomoto, Taisuke; Ishigaki, Masahiro; Watanabe, Norio; Maruyama, Yu; Sibamoto, Yasuteru; Watanabe, Tadashi; Moriyama, Kiyofumi

doi:10.1080/18811248.2011.636538

Cited by 111 publications

(63 citation statements)

References 1 publication

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“…The event sequence by MELCOR calculation is summarized in Table 1. According to the corresponding analytical study presented by Hirano, et al, (2012), the reactor core isolation cooling system (RCIC) at Unit 2 started up automatically at 14:50 immediately after the earthquake and had been activated continuously for 3 days. The RCIC stopped at 13:25 on 14 March 2011 for unknown reason, then the water level started to decrease and the reactor pressure vessel (RPV) pressure started to increase.…”

Section: Accident Scenario and Analysis Conditionsmentioning

confidence: 99%

Estimation of Source Term Uncertainty in a Severe Accident With Correlated Variables

Zheng

Itoh

Tamaki

et al. 2014

Volume 6: Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls (I&C); Fusion Engineering;

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The quantitative evaluation of the fission product release to the environment during a severe accident is of great importance. In the present analysis, integral severe accident code MELCOR 1.8.5 has been applied to estimating uncertainty of source term for the accident at Unit 2 of the Fukushima Daiichi nuclear power plant (NPP) as an example and to discussing important models or parameters influential to the source term. Forty-two parameters associated with models for the transportation of radioactive materials were chosen and narrowed down to 18 through a set of screening analysis. These 18 parameters in addition to 9 parameters relevant to in-vessel melt progression obtained by the preceding uncertainty study were input to the subsequent sensitivity analysis by Morris method. This one-factor-at-a-time approach can preliminarily identify inputs which have important effects on an output, and 17 important parameters were selected from the total of 27 parameters through this approach. The selected parameters have been integrated into uncertainty analysis by means of Latin Hypercube Sampling technique and Iman-Conover method, taking into account correlation between parameters. Cumulative distribution functions of representative source terms were obtained through the present uncertainty analysis assuming the failure of suppression chamber. Correlation coefficients between the outputs and uncertain input parameters have been calculated to identify parameters of great influences on source terms, which include parameters related to models on core components failure, models of aerosol dynamic process and pool scrubbing.

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Section: Accident Scenario and Analysis Conditionsmentioning

confidence: 99%

Estimation of Source Term Uncertainty in a Severe Accident With Correlated Variables

Zheng

Itoh

Tamaki

et al. 2014

Volume 6: Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls (I&C); Fusion Engineering;

Self Cite

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“…2) are 600 x 300 x 400 mm concrete blocks with a Ø300 x 250 mm hemicylindrical cavity (area ~ 350 cm 2 , volume ~9 L) in which corium is poured. Several types of concrete have been studied (Table II), either representative of nuclear reactor basemats or specially designed to study one character of concrete.…”

Section: Concrete Test Sectionsmentioning

confidence: 99%

“…In current LWRs, this concrete basemat is the ultimate barrier, while in the EPR [1], the reactor pit provides temporary melt retention before spreading in the core catcher. Calculations of the Fukushima Dai-ichi accident at the time of writing suggest [2] that vessels from units 1-3 have failed and that corium has interacted with the concrete.…”

Section: Introductionmentioning

confidence: 99%

Contributions of the Vulcano Experimental Programme to the Understanding of McCi Phenomena

Journeau

Piluso²,

Correggio³

et al. 2012

Nuclear Engineering and Technology

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“…E-mail: takeda.takeshi@jaea.go.jp. accident of the Fukushima Dai-ichi reactor [1], it is important to line up various cooling measures for reactor core and containment. The PWR (pressurized water reactor) electric utilities and vender, etc.…”

Section: Introductionmentioning

confidence: 99%

RELAP5 Code Study of ROSA/LSTF Experiments on PWR Safety System Using Steam Generator Secondary-Side Depressurization

Takeda¹,

Ohnuki²,

Nishi³

2015

JEPE

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RELAP5 (reactor excursion and leak analysis program, version 5) code analyses were performed on two ROSA/LSTF (rig of safety assessment/large scale test facility) experiments on PWR (pressurized water reactor) safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG (steam generator) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by core boil-off. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. In the 4-in. break test, on the other hand, there was no core uncovery and heatup due to smaller break flow rate than in the 8-in. break test. Adjustment of Cd (break discharge coefficient) for two-phase discharge flow predicted the break flow rate reasonably well. The code well predicted the overall trend of the major thermal-hydraulic response observed in the two LSTF tests by the Cd adjustment. The code, however, overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.

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Insights from review and analysis of the Fukushima Dai-ichi accident

Cited by 111 publications

References 1 publication

Estimation of Source Term Uncertainty in a Severe Accident With Correlated Variables

Estimation of Source Term Uncertainty in a Severe Accident With Correlated Variables

Contributions of the Vulcano Experimental Programme to the Understanding of McCi Phenomena

RELAP5 Code Study of ROSA/LSTF Experiments on PWR Safety System Using Steam Generator Secondary-Side Depressurization

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