Nuclear reactor safety systems
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Continued part 1, part 2 of ebook "Wind energy systems: Optimising design and construction for safe and reliable operation" provides readers with content including: wind energy systems - optimising design and construction for safe and reliable operation; solid oxide fuel cell technology - principles, performance and operations; handbook of advanced radioactive waste conditioning technologies; nuclear reactor safety systems;...
294p lytamnguyet 04-08-2023 10 5 Download
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The natural convection method is used in many advanced nuclear reactors, such as Generation IV and Small Modular Reactors to improve the safety and reliability. In order to investigate the natural circulation (NC) behavior in A NuScale Power Module Reactor, a numerical study on the NC in primary system for such kind of reactor was obtained by using RELAP5.
10p vicedric 08-02-2023 5 3 Download
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This study deals with the feasibility study of a new in-vessel core melt retention (IVCMR) strategy capable to extend the coping period in the event of adverse situations, involving the melting of the core. Since Fukushima accident, many studies have been carried out to resolve the severe accident mitigation issues related to the corium stabilization inside and outside the reactor vessel.
8p vironald 15-12-2022 14 5 Download
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SuperCritical Water Reactor(SCWR) applies water beyond the thermodynamic critical point as the coolant, which aims to achieve high efficiency around 45% compared to 33% for existing commercial light water reactors.
15p vironald 15-12-2022 11 4 Download
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In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis.
9p vironald 15-12-2022 8 4 Download
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The difficulty in predicting locally and globally the transient evolution of two-phase or multiphase flows in complex systems is well recognized in nuclear thermal-hydraulics. Large efforts involving the expenditure of huge resources during the last three decades in previous century brought to the creation of giant databases (e.g. including experimental data and results of computer code calculations) and to the perception that the safety of nuclear reactors is guaranteed notwithstanding residual areas of unawareness.
7p vironald 15-12-2022 11 3 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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A miniature integrated nuclear reactor design with gravity independent autonomous circulation (ACMIR) was newly proposed. The reactor core, energy transfer system of Stirling and linear electric motors are integrated in the reactor pressure vessel to achieve high power density and autonomous circulation capability.
8p vironald 15-12-2022 7 3 Download
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The system code package AC2 by GRS for safety analyses of nuclear reactors from normal operation to severe accidents has been updated with a new release. We briefly describe the main modules of AC2 2019: ATHLET 3.2, ATHLET-CD 3.2 and COCOSYS 3.0 and selected improvements in these codes.
16p vironald 15-12-2022 10 4 Download
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Supercritical carbon dioxide (S-CO2) Brayton cycle has many advantages including high power conversion efficiency at mediate temperature, compact configuration, high system simplicity and low efficiency loss using dry cooling, which make it well suited to nuclear reactor applications.
23p vironald 15-12-2022 9 3 Download
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Reliable mechanical valves that can withstand the corrosive and high-temperature conditions in Molten Salt Reactors (MSRs) have not yet been demonstrated. In their place, freeze valves (sometimes called freeze plugs) represent a unique nuclear design solution for isolating salt flow during operations. Additionally, in some cases, they are intended to perform safety-related functions.
17p vironald 15-12-2022 5 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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The assessment in this paper is necessary to demonstrate that sufficient decay heat conservatism is retained in the UNF-ST&DARDS bounding as-loaded spent fuel analysis methodology. This paper also demonstrates the time dependent impact of various parameters such as last cycle power on decay heat values.
25p vironald 15-12-2022 6 3 Download
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The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned.
6p christabelhuynh 29-05-2020 13 1 Download
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The Best-Estimate Plus Uncertainty (BEPU) is applied as Deterministic Approach for safety analysis of Nuclear Power Plant using the system analysis code. The system analysis code such as Relap5/Mod3.3 is required to be able to simulate the thermal-hydraulic behavior of nuclear reactor in some accident scenarios.
7p vineptune2711 04-11-2019 10 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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Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance.
12p minhxaminhyeu5 30-06-2019 7 0 Download
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This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.
12p minhxaminhyeu5 30-06-2019 8 0 Download
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This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models.
8p minhxaminhyeu5 30-06-2019 14 0 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