Loss of coolant accidents
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This study was conducted to carry out an independent thermal-hydraulic safety analysis for LB-LOCA of VVER-1200/491 and to investigate the agreement of the SAR conservative results with requirement stated in the IAEA Specific Safety Guide by using RELAP5/MOD 3.3 computer code with both conservative and best estimate approach.
10p visnape 30-01-2023 7 4 Download
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The International Thermonuclear Experimental Reactor (ITER) Vacuum Vessel Pressure Suppression System (VVPSS) limits the Vacuum Vessel (VV) internal pressure, in case of loss of coolant (LOCA) or other pressurizing accidents from the in-vessel components, to 150 kPa (abs).
14p vironald 15-12-2022 17 3 Download
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Westinghouse type plants in the United States (US) and Europe have a manual trip procedure of the reactor coolant pumps (RCP) in the case of a small break Loss of Coolant Accident (LOCA) with the high pressure injection (HPI) system functioned. This is in response to the Three Mile Island (TMI) accident in order to maintain as much the forced core cooling as possible.
6p vironald 15-12-2022 10 3 Download
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Nuclear fuel claddings can balloon and rupture at high temperatures under internal gas pressure in case of design basis accidents like loss-of-coolant-accident (LOCA). The thermal phenomena surrounding the ballooning and cracking was investigated in a series of experiments performed using zirconium alloy cladding tubes at the Centre for Energy Research in Hungary.
9p vironald 15-12-2022 11 3 Download
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The paper presents the modeling of the Test Facility ROSA-2/LSTF in the calculation code CATHARE 2.V2.5. OECD/NEA ROSA-2 Project Test 7 was conducted with the Large Scale Test Facility on June 14, 2012. The experiment simulated the thermal-hydraulic responses during a PWR 13% cold leg Intermediate Break Loss Of Coolant Accident (IBLOCA).
8p christabelhuynh 30-05-2020 15 0 Download
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The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing.
9p christabelhuynh 30-05-2020 9 0 Download
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This paper will provide an update on results from the feasibility study and discuss the attributes of the coated Mo cladding design to meet the challenging requirements for improving fuel tolerance to severe loss of coolant accidents.
6p christabelhuynh 30-05-2020 15 2 Download
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The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.
10p christabelhuynh 31-05-2020 27 0 Download
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The paper covers the results of VVER core reflooding studies in fuel assembly (FA) mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup.
11p christabelhuynh 30-05-2020 34 1 Download
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This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200 °C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples.
8p christabelhuynh 30-05-2020 9 0 Download
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This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents.
13p christabelhuynh 29-05-2020 6 0 Download
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The management of hydrogen safety and prevention of overpressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity.
10p minhxaminhyeu5 30-06-2019 17 0 Download
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The aim of this paper is to predict the hydrogen buildup within containment resulting from severe accidents. The prediction was based on NPPs of an optimized power reactor 1,000. The increase in the hydrogen concentration in severe accidents is one of the major factors that threaten the integrity of the containment.
9p minhxaminhyeu5 30-06-2019 7 0 Download
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In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model.
14p minhxaminhyeu5 30-06-2019 15 1 Download
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In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity.
8p minhxaminhyeu5 30-06-2019 7 0 Download
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The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area.
10p minhxaminhyeu4 26-06-2019 9 0 Download
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This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms.
7p minhxaminhyeu3 12-06-2019 6 0 Download
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In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models.
12p minhxaminhyeu3 25-06-2019 22 0 Download
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In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.
13p minhxaminhyeu3 25-06-2019 14 0 Download
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The X/Q (s/m3 ) quality and the maximum dose values were calculated within an area of 40 km radius from the NPP site, where X/Q (s/m3 ) is the ratio of activity concentration to release rate. Based on the obtained results on dose distribution the necessary measures for nuclear emergency preparedness have been proposed according to the IAEA recommendations.
9p thuyliebe 09-10-2018 32 0 Download