Search Results 1-20 of 294

  • INFLUENCE OF ARTIFICIAL RADIONUCLIDE DEPOSITED ON A MONITORING POST ON MEASURED VALUE OF AMBIENT DOSE RATE

    Hiraoka H. , Hokama T. , Munakata M.

    … When it exceeds 20 μSv/h at an accidental release of radionuclides, it is stipulated that inhabitants should temporarily relocate (OIL 2) within one week. … Even if time elapsed from an accident, 𝑅𝑅 did not change. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2303, 2019

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  • EFFECT OF SUPPRESSION POOL CONDITIONS ON RISING BUBBLE PARAMETERS DURING WETWELL VENTING

    Zablackaite Giedre , Nagasaka Hideo , Takahashi Hideharu , Kikura Hiroshige

    … The severe accident in Fukushima Daiichi NPP increased the need for better understanding of an accident progression. … The events after the main earthquake followed by tsunami brought the attention to the flaws in the understanding and analysis of severe accident and made it clear which key points require more studies. … In Fukushima Daiichi Unit 1 and 3 wetwell venting was utilized as a way to minimize the release of Fission Products (FPs) to the environment. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2300, 2019

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  • Fundamental study on the filtered containment venting system for decommissioning of the Fukushima Nuclear Power Plant  [in Japanese]

    TRAN Tri Vien , TAKAHASHI Hideharu , NARABAYASHI Tadashi , KIKURA Hiroshige

    … During the decommissioning of the Fukushima Nuclear Power Plant after severe accident, the process of cutting the core debris (using a laser cutter, wire cutter or drilling machine) will generate a large amount of radioactive aerosol. … Therefore, the air cleaning system should be needed to prevent the release of radioactive materials into the environment. …

    The Proceedings of the National Symposium on Power and Energy Systems 2019.24(0), B143, 2019

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  • DEVELOPMENT OF INTEGRATED SYSTEM FOR ACCIDENT CONSEQUENCE EVALUATION USING LEVEL 2 AND LEVEL 3 PRA CODES

    Kimura Masanori , Ishikawa Jun , Oguri Tomomi , Munakata Masahiro

    … For the result of a state-of-the-art source term analysis with the THALES2, the authors implemented a preliminary analysis using the OSCAAR in order to evaluate the difference in setting of release fraction rates for the radionuclide groups. … The effective dose was calculated for some cases by changing of release fraction rates divided into 1h (case 1), 3h (case 2), 5h (case 3), none (case 4) and arbitrary hours (case 5). …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2319, 2019

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  • ADSORPTION BEHAVIOR OF CESIUM ON CALCIUM SILICATE INSULATION OF PRIMARY PIPING SYSTEM IN FUKUSHIMA DAIICHI NPP UNIT 2

    Rizaal Muhammad , Saito Takumi , Erkan Nejdet , Okamoto Koji , Osaka Masahiko , Nakajima Kunihisa , Nishioka Shunichiro

    … The pedestal area, in regard to the accident progression, should exhibit a higher dose rate than those of any other locations at the same reference level because pedestal area was the most affected area after the accident. … The impact of high temperature steam during its release on thermal insulation could be highly contaminated by fission products such as cesium. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2035, 2019

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  • SURROGATE MODEL DEVELOPMENT FOR PREDICTION OF VESSEL FAILURE MODE AND MELT RELEASE CONDITIONS IN NORDIC BWR BASED ON MELCOR CODE

    Galushin Sergey , Grishchenko Dmitry , Kudinov Pavel

    … Effectiveness of severe accident management strategy in Nordic BWR reactors depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in the risk of containment failure in Nordic BWRs. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2172, 2019

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  • THE EFFECT OF VESSELFAILURE AND MELT RELEASE ON RISK OF CONTAINMENT FAILURE DUE TO EX-VESSEL STEAM EXPLOSION IN NORDIC BWR

    Galushin Sergey , Grishchenko Dmitry , Kudinov Pavel

    … Effectiveness of the severe accident management strategy in Nordic BWR reactors depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in the ROAAM+ Framework. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2174, 2019

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  • COMPARISON OF VESSELFAILURE MODE AND MELT RELEASE CONDITIONS IN UNMITIGATED AND MITIGATED STATION BLACKOUT SCENARIOS IN NORDIC BWR USING MELCOR CODE

    Galushin Sergey , Kudinov Pavel

    … Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. … Effectiveness of this strategy depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in ROAAM+ Framework. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2173, 2019

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  • THE RESEARCH OF STARTUP SIGNAL DESIGN AND OPERATING CHARACTERISTIC FOR PASSIVE CONTAINMENT HEAT REMOVAL SYSTEM

    Xiaochuan Ding , Bin Lu , Yuanye Li , Ke Yi

    … PCS startup signal should be designed to start PCS correctly and timely to remove energy release from containment. … The design of PCS startup signal is mainly base on containment design criteria, accident analysis, and function distribution of PCS during severe accident. … The starting sequence of PCS, containment temperature and pressure response after accident are also analyzed to optimize severe accident management. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1726, 2019

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  • THE FUEL BUILDING RADIATION ENVIRONMENT OF 1000MW PWR DURING SEVERE ACCIDENT

    Zhu Zhigang , You Wei , Xie Siyang , Shi Xueyao , Zhang Liying , Wang Xiaoxia , Mi Aijun , Mao Yawei

    … When spent fuel pool (SFP) loss of cooling the spent fuel assemblies in there will eventually melt and release a number of radioactive nuclides. … Results including water level change in pent fuel pool, fission products release time during severe accident, fraction of fission products in different state and dose rate in the different positon including the place installed monitoring instrument. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1597, 2019

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  • RISK ANALYSIS ON CONTAINMENT BYPASS ACCIDENTS IN M310 NUCLEAR POWER PLANT

    Jian Yang , Jin Yan Du , Chu Ran Feng

    … For these containment bypass accidents, the large early release frequency(LERF) is 5.35E-7/r.y. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1604, 2019

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  • ANALYSIS OF AIRBORNE RADIOACTIVITY IN CONTAINMENT AND ADJACENT BUILDINGS UNDER PWR ACCIDENT CONDITIONS

    Zhang Liying , You Wei , Zhu Zhigang , Wang Xiaoxia , Mi Aijun , Mao Yawei

    … Based on the main aerosol behavior, a multi-compartment post-accident airborne radioactivity calculation model is built in this paper, which can be used in post-accident radiation protection design and sensitivity analysis to establish foundation of design improvement. … A post-accident airborne radioactivity calculation model for PWR with doublecontainment design is built, which contains 6 compartments: containment, annulus and 4 adjacent buildings. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1569, 2019

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  • PROBABILISTIC ASSESSMENT OF CREEP INDUCED SGTR IN SEVERE ACCIDENT

    Shipeng Niu , Cong Wang , Gaopeng Wang , Wei Wei , Wentao Zhu

    … The creep induced SGTR may arise in severe accident, which may bypass the containment and lead to the release of fission products. … Therefore, it is essential for severe accident analysis and especially for level 2 probabilistic safety analyses (PSA) to evaluate the probability of creep induced SGTR in severe accident. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1591, 2019

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  • ANALYSIS OF HYDROGEN AND SOURCE TERM UNDER SEVERE ACCIDENT CONDITIONS FOR THOUSAND MEGAWATT PWR

    Xiao Yubai , Wu Junmei

    … This paper uses the integrated severe accident analysis computer code, MIDAC, to simulate the large break loss of coolant accident (LB-LOCA), medium break LOCA (MBLOCA) and small break loss of coolant accident (SB-LOCA) of the thousand megawatt PWR nuclear power plant, respectively. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1539, 2019

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  • SIMULATION STUDY ON THE EFFECTS OF AIRBORNE RADIOACTIVE RELEASE ON SURROUNDING WATER BODIES IN INLAND SITE UNDER ACCIDENT CONDITIONS

    Lin Hongtao , Wang Mengxi , Zhang Jiemin , Liu Xinjian , Xue Na , Qiu Lin

    … It is necessary to consider the impact of these factors on the calculation of airborne radioactivity release to the air and then deposited on the surrounding water bodies, which is crucial to the site selection of inland nuclear power plants and the accurate assessment of the feasibility of emergency planning. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1332, 2019

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  • VALIDATION STUDY OF INITIATING PHASE EVALUATION METHOD FOR THE CORE DISRUPTIVE ACCIDENT IN AN SFR

    Ishida Shinya , Kawada Ken-ichi , Fukano Yoshitaka

    … Core Disruptive Accident (CDA) has been considered as one of the important safety issues in the severe accident evaluation of Sodium-cooled Fast Reactor (SFR), since SFR core is not in the most reactive configuration. … The average fuel temperature which corresponds to the energy release is selected as the FOM, since the energy release due to power excursion is important to evaluate whether the influence of CDA can be confined in the vessel. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1341, 2019

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  • EFFECT OF NEW DRYWELL COOLER APPLICATION FOR VENTING START TIME DURING SEVERE ACCIDENTS

    Ishida Naoyuki , Nagata Yasunori , Kimura Ryusuke , Ando Koji

    … Considering the gas release from these venting systems, it is desirable to delay venting start time or to avoid operation of the systems by ensuring sufficient PCV cooling (in view of the time needed to evacuate residents). … Here, to estimate plant performance of an advanced boiling water reactor (ABWR) applying the DWC_SA, we carried out a severe accident analysis using the MAAP code. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1324, 2019

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  • COMPUTATIONAL CAPABILITY TO STUDY AIRBORNE RELEASE OF SOLIDS AND CONTAINER BREACH DUE TO MECHANICAL INSULTS

    L.Y. Louie David , Bignell John , Le San , Dingreville Remi , N. Gilkey Lindsay , Gordon Natalie

    … Engineers performing safety analyses throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK3010, <i>Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities,</i> … to determine radionuclide source terms from postulated accident scenarios. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1438, 2019

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  • MECHANISM OF AEROSOL DECONTAMINATRION FROM A BUBBLE DURING POOL SCRUBBING

    Fujiwara Kota , Kikuchi Wataru , Nakamura Yuki , Yuasa Tomohisa , Kaneko Akiko , Abe Yutaka

    … In severe accidents (SAs) of nuclear power plants, release of gas containing fission products (FPs) from the reactor vessel is thought to be a major issue. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1197, 2019

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  • PARAMETRIC SENSITIVITY STUDY OF THE DIRECT CONTAINMENT HEATING RISK ASSESSMENTS FOR PWR

    Yu LIU , Shipeng NIU , Gaopeng WANG , Wei WEI

    … Direct Containment Heating (DCH) is one of the most serious accidents in PWR, which may lead to the early uncontrollable release of radioactive materials. … In case of a high pressure core meltdown accident in PWR, the liquid corium may be rapidly ejected into the reactor pit, finely fragmented and eventually transported into the upper space of containment. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1299, 2019

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