scholarly journals Rapid preliminary modeling of transport reactor cores

2021 ◽  
Vol 7 (1) ◽  
pp. 41-47
Author(s):  
Vladimir I. Korolev
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Dimensionless Parameter ◽  
Preliminary Calculation ◽  
Higher Power ◽  
Nuclear Power Unit ◽  
Design Characteristics ◽  
Under Construction ◽  
Floating Objects ◽  
Rapid Modeling

At the present time, JSC Baltiskiy zavod has built and transported to the deployment site at Pevek Akademik Lomonosov, a floating nuclear power unit (FNPU), project 20870. There are also three multi-purpose nuclear icebreakers of project 22220 (Arktika, Sibir, Ural) under construction at Baltiskiy being at different readiness stages. A decision has been made to build a nuclear icebreaker, Lider, of even a higher power. Integral reactors developed by JSC OKBM Afrikantov are installed in the nuclear icebreakers using new assembly-type cores which have not been used earlier in floating facilities. A great deal of preliminary calculation is required to give these cores as advantageous characteristics as possible. The paper proposes a procedure for rapid modeling of floating cores with varied operating and design characteristics. This procedure can be used as part of preliminary modeling. The procedure is based on using a combined dimensionless parameter proposed by the author in (Korolev 2009). A chart is presented to model the key performance of cores for floating objects with a nuclear reactor NPPs. Eight assembly-type core options, which can be installed in transport reactors of a modular or integral design, are analyzed.

An Assessment of Mechanical Properties of A508-3 Steel Used in Chinese Nuclear Reactor Vessels

2014 ◽  
Author(s):  
Meifang Yu ◽  
Zhen Luo ◽  
Y. J. Chao
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Fossil Fuels ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Construction Materials ◽  
Under Construction ◽  
Made In China ◽  
Phase Out

China has very ambitious goals of expanding its commercial nuclear power by 30 Giga-Watts within the decade and wishes to phase out fossil fuels emissions by 40–45% by 2020 (from 2005 levels). With over 50 new nuclear power plants under construction or planned and a design life of 60 years, any discussions on structural integrity become very timely. Although China adopted its nuclear technology from France or US at present time, e.g. AP1000 of Westinghouse, the construction materials are primarily “Made in China”. Among all issues, both the accumulation of the knowledge base of the materials and structures used for the power plant and the technical capability of engineering personnel are imminent. This paper attempts to compile and assess the mechanical properties, Charpy V-notch impact energy, and fracture toughness of A508-3 steel used in Chinese nuclear reactor vessels. All data are collected from open literature and by no means complete. However, it provides a glimpse into how this domestically produced steel compares with western reactor vessel steels such as US A533B and Euro 20MnMoNi55.

Reliability and redundancy allocation analysis applied to a nuclear protection system

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Alexander Lucas Busse ◽  
João Manoel Losada Moreira
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Block Diagram ◽  
Thermal Power ◽  
Life Time ◽  
Spare Parts ◽  
Protection System ◽  
Under Construction

Brazil is constructing with national technology two small nuclear reactors for propulsion and for radioisotope production with thermal power levels between 20 and 50 MW. These nuclear reactors fit more in the class of small modular reactors (SMR) than in the class of large nuclear power plants. In this article we apply the design approach of SMRs to propose an architecture of reactor protection systems for the small reactor under construction in the country. To do that the probabilistic analysis of the architecture of a nuclear reactor protection system is evaluated to determine the sensitivity of the components through an Reliability Block Diagram modeling. It was evaluated the modification of the architecture and the addition of redundancies when using components with lower life time than the components usually used for this purpose. The results showed that after one year of operation, the reference RPS system presents a failure probability of 0.17 %. The modified system, with components with lower life time, presents a point reliability value only 0.070 % lower than the reference one, but this difference grows exponentially over time, and in 10 years of operation it can reach values above 95%. Using equipment with lower life time characteristics implies a greater number of redundancies and, additionally, a greater number of maintenance procedures and spare parts. Therefore, this technical feasibility analysis should consider a RAM simulation as well.

Development of Best-Estimate ECCS Evaluation Methodology for APR1400

2009 ◽  
Author(s):  
Seok-Ho Lee ◽  
Mun-Soo Kim ◽  
Han-Gon Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Evaluation Methodology ◽  
Preliminary Calculation ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
Best Estimate ◽  
Design Characteristics ◽  
Core Cooling

Advanced Power Reactor 1400 (APR1400) is an evolutionary Pressurized Water Reactor (PWR) equipped with such advanced features as the Direct Vessel Injection (DVI), the Fluidic Device (FD) in the Safety Injection Tank (SIT), and the In-containment Refueling Water Storage Tank (IRWST) in the Emergency Core Cooling System (ECCS). To verify the performance of these advanced features, more realistic performance evaluation methodology is desired since existing methodologies use too conservative assumptions which cause negative biases to these features. In this study, therefore, a best estimate evaluation methodology for the APR1400 ECCS under large break loss of cooling accident (LBLOCA) is developed targeting operating license of the Shin Kori 3&4 nuclear power plants (SKN 3&4), the first commercial APR1400 plants. On this purpose, a variety of existing best estimate evaluation methodologies previously used are reviewed. As a result of this review, a methodology named KREM is selected for this study. The KREM is based on RELAP5/MOD3.1K and has been used for Korean operating plants since 2002 when it was first approved by Korean regulation. For this study, RELAP5/MOD3.3 (Patch 3), the latest version of RELAP series is selected since it could appropriately simulate the multi-dimensional phenomena for the APR1400 design characteristics. To quantify the code accuracy, analyses covering experimental data have been performed for 36 kinds of separated effect tests (SETs) and integral effect tests (IETs). The uncertainty in the peak cladding temperature (PCT) of the APR1400 is evaluated preliminarily. Based on the preliminary calculation, final uncertainty quantification and bias evaluation are performed to obtain the licensing PCT for Shin-Kori 3&4 plants and the result shows that the LBLOCA licensing acceptance criteria are well satisfied.

An Assessment of Mechanical Properties of A508-3 Steel Used in Chinese Nuclear Reactor Pressure Vessels

2015 ◽  
Vol 137 (3) ◽  
Author(s):  
Meifang Yu ◽  
Y. J. Chao ◽  
Zhen Luo
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Fossil Fuels ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Construction Materials ◽  
Pressure Vessels ◽  
Under Construction ◽  
Reactor Pressure

China has very ambitious goals of expanding its commercial nuclear power by 30 GW within the decade and wishes to phase out fossil fuels emissions by 40–45% by 2020 (from 2005 levels). With over 50 new nuclear power plants under construction or planned and a design life of 60 years, any discussions on structural integrity become very timely. Although China adopted its nuclear technology from France or USA at present time, e.g., AP1000 of Westinghouse, the construction materials are primarily “Made in China.” Among all issues, both the accumulation of the knowledge base of the materials and structures used for the power plant and the technical capability of engineering personnel are imminent. This paper attempts to compile and assess the mechanical properties, Charpy V-notch impact energy, and fracture toughness of A508-3 steel used in Chinese nuclear reactor pressure vessels (RPVs). All data are collected from open literature and by no means complete. However, it provides a glimpse into how this domestically produced steel compares with western RPV steels such as USA A533B and Euro 20MnMoNi55.

CREATION AND VISUALIZATION OF AN OPERATIONAL PARAMETERS ARCHIVE

2021 ◽  
Vol 9 (2) ◽  
pp. 101-110
Author(s):  
A.M. Zagrebayev ◽  
◽  
S.V. Ten
Keyword(s):  
Power Plant ◽  
Power Unit ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Operational Parameters ◽  
High Quality ◽  
Nuclear Power Unit ◽  
Visualization Of Data ◽  
Over Time ◽  
Param Eters

The developed methodologies allows visualizing the behavior of limiting param- eters over time and identifying patterns in the dynamics of changes in the parameters of a nuclear power unit, as well as determining if any of the parameters intersects the safe opera- tion settings. The relevance of the work is due to the importance of high-quality processing and visualization of data from the nuclear reactor archive using modern effective methodolo- gies. Also, the development of new approaches to the analysis of archived data can improve the efficiency of power plant operators.

FUEL BURNUP AND FUEL POOL SHIELDING ANALYSIS FOR BUSHEHR NUCLEAR REACTOR VVER-1000

2006 ◽  
Vol 15 (04) ◽  
pp. 925-938 ◽  
Author(s):  
KAMAL HADAD ◽  
NAVID AYOBIAN
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Exposure Dose ◽  
Spent Fuel ◽  
Concrete Wall ◽  
Transport Code ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Γ Rays ◽  
Under Construction

Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

scholarly journals Switched Capacitor DC-DC Converters: A Survey on the Main Topologies, Design Characteristics, and Applications

Energies ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 2231
Author(s):  
Alencar Franco de Souza ◽  
Fernando Lessa Tofoli ◽  
Enio Roberto Ribeiro
Keyword(s):  
Voltage Regulation ◽  
Inherent Difficulty ◽  
Practical Applications ◽  
Dc Power ◽  
Higher Power ◽  
Voltage Doubler ◽  
Converter Design ◽  
Design Characteristics ◽  
And Performance ◽  
Full Charge

This work presents a review of the main topologies of switched capacitors (SCs) used in DC-DC power conversion. Initially, the basic configurations are analyzed, that is, voltage doubler, series-parallel, Dickson, Fibonacci, and ladder. Some aspects regarding the choice of semiconductors and capacitors used in the circuits are addressed, as well their impact on the converter behavior. The operation of the structures in terms of full charge, partial charge, and no charge conditions is investigated. It is worth mentioning that these aspects directly influence the converter design and performance in terms of efficiency. Since voltage regulation is an inherent difficulty with SC converters, some control methods are presented for this purpose. Finally, some practical applications and the possibility of designing DC-DC converters for higher power levels are analyzed.

scholarly journals Computational Fluid Dynamics Modeling of Steam Condensation on Nuclear Containment Wall Surfaces Based on Semiempirical Generalized Correlations

2012 ◽  
Vol 2012 ◽  
pp. 1-7 ◽  
Author(s):  
Pavan K. Sharma ◽  
B. Gera ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Pressure Rise ◽  
Dynamics Modeling ◽  
Steam Condensation ◽  
Noncondensable Gases ◽  
Nuclear Containment ◽  
Design Basis ◽  
Lumped Parameter ◽  
Transport Calculation

In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.

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