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Relevant bibliographies by topics / RELAP5-3D

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Academic literature on the topic 'RELAP5-3D'

Author: Grafiati

Published: 4 June 2021

Last updated: 14 March 2023

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Journal articles on the topic "RELAP5-3D"

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Forsmann, J. Hope. "RELAP5-3D User Tools." Nuclear Technology 193, no. 1 (2016): 213–17. http://dx.doi.org/10.13182/nt14-141.

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Mesina, G. L., D. L. Aumiller, F. X. Buschman, and M. R. Kyle. "Modeling Moving Systems with RELAP5-3D." Nuclear Science and Engineering 182, no. 1 (2016): 83–95. http://dx.doi.org/10.13182/nse15-3.

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R. Gonzalez, A. Petruzzi, F. D’Auria, and O. Mazzantini. "Identification of Limiting Case Between DBA and SBDBA (CL Break Area Sensitivity): A New Model for the Boron Injection System." Journal of Energy - Energija 62, no. 1-4 (2022): 81–92. http://dx.doi.org/10.37798/2013621-4220.

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Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete

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Prošek, Andrej, and Ovidiu-Adrian Berar. "BETHSY 9.1b test calculation with TRACE using 3D vessel component." Journal of Energy - Energija 62, no. 1-4 (2022): 93–103. http://dx.doi.org/10.37798/2013621-4221.

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Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not

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Sánchez, V., W. Jaeger, M. Boettcher, and B. Truong. "Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools." Science and Technology of Nuclear Installations 2010 (2010): 1–14. http://dx.doi.org/10.1155/2010/470794.

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The Institute of Neutron Physics and Reactor Technology (INR) is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 react

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Venturini, Alessandro, Marco Utili, and Nicola Forgione. "Numerical Simulations with RELAP5-3D and RELAP5/mod3.3 of the Second Experimental Campaign on In-Box LOCA Transients for HCLL TBS." Energies 14, no. 15 (2021): 4544. http://dx.doi.org/10.3390/en14154544.

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In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in Lead LIthiUM), a facility that reproduces the LiPb loop of the HCLL TBS. Two experimental campaigns were carried out by simulating the rupture of a stiffening plate and the related helium injection in the LiPb loop. The obtained experimental data were used to check the capabilities of RELAP5 system code to reproduce the pressure wave propagation

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Popov, Emilian L., Graydon L. Yoder, and Valeri Velichkov. "VVER-1000 Model Assessment Using RELAP5-3D." Nuclear Technology 149, no. 3 (2005): 304–8. http://dx.doi.org/10.13182/nt05-a3597.

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Mesina, G. L., D. L. Aumiller, and F. X. Buschman. "Extremely Accurate Sequential Verification of RELAP5-3D." Nuclear Science and Engineering 182, no. 1 (2016): 1–12. http://dx.doi.org/10.13182/nse14-151.

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Balestra, P., F. Giannetti, G. Caruso, and A. Alfonsi. "New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors." Science and Technology of Nuclear Installations 2016 (2016): 1–15. http://dx.doi.org/10.1155/2016/1687946.

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The latest versions of RELAP5-3D©code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D©water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D©d

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Uspuras, E., A. Kaliatka, and E. Bubelis. "Validation of RELAP5-3D RBMK-1500 reactor model." Nuclear Energy 43, no. 5 (2004): 257–64. http://dx.doi.org/10.1680/nuen.43.5.257.53136.

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Dissertations / Theses on the topic "RELAP5-3D"

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Anderson, Nolan Alan. "Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum." Texas A&M University, 2006. http://hdl.handle.net/1969.1/4160.

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The Very High Temperature Reactor (VHTR) system behavior should be predicted during normal operating conditions and during transient conditions. To predict the VHTR system behavior there is an urgent need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. Thi

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Vavassori, Giacomo. "Swing check valvecharacterization: 3D CFDvalidation of one dimensionalmodels used in RELAP5." Thesis, KTH, Mekanik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-214556.

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In a previous thesis work a swing check valve was studied with CFD analysisin order to nd correlations that could provide a good input for a onedimensionalmodel of the same. In this document, starting from the previousthesis results and using the model by Li and Liou as the reference work, a checkvalve is investigated and the hydraulic torque coecients identied. In this wayit becomes possible to analyze the behavior of the same valve with a 1D codecalled RELAP5.The rst part of the work was dedicated to understanding the dynamics lyingbehind the movement of swing check vale, and to the construc

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Moore, Eugene James Thomas. "Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application." Texas A&M University, 2003. http://hdl.handle.net/1969.1/3996.

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High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and economical solutions to the worldÂs energy crisis. HTGRs are capable of generating high temperatures during normal operation, introducing design challenges related to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling abilities of system analysis codes, used to develop an understanding of light water reactor phenome

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Ghrist, Melissa Renee. "Zero gravity two-phase flow regime transition modeling compared with data and relap5-3d predictions." [College Station, Tex. : Texas A&M University, 2008. http://hdl.handle.net/1969.1/ETD-TAMU-2353.

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Silva, Rodney Aparecido Busquim e. "Implications of advanced computational methods for reactivity initiated accidents in nuclear reactors." Universidade de São Paulo, 2015. http://www.teses.usp.br/teses/disponiveis/3/3139/tde-20072016-142605/.

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Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). T

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MOSCATO, Ivo. "THERMAL-HYDRAULIC STUDY IN SUPPORT OF THE DESIGN OF THE DEMO BALANCE OF PLANT." Doctoral thesis, Università degli Studi di Palermo, 2020. http://hdl.handle.net/10447/395492.

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The European Research Roadmap to the Realisation of Fusion Energy, foresees that the Demonstration Fusion Power Reactor (DEMO) is going to be the successor of ITER reactor in the pathway towards the exploitation of nuclear fusion to produce electric energy on an industrial scale. It will, hence, have to deliver hundreds MW of electric power ensuring an adequate availability and reliability of operation over a reasonable time span. EU-DEMO pre-conceptual design is being conducted by research institutions and universities from 26 countries of European Union, Switzerland and Ukraine, with the ma

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Wang, Chun-Lung, and 王駿隆. "AP1000 Large Break LOCA Analysis using RELAP5-3D." Thesis, 2011. http://ndltd.ncl.edu.tw/handle/35395439085201049011.

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Bowser, Christopher Jordan. "RELAP5-3D modeling of ADS blowdown of MASLWR facility." Thesis, 2012. http://hdl.handle.net/1957/30438.

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Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool wat

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劉相君. "Margin Assessment of AP1000 In-vessel Retention using RELAP5-3D." Thesis, 2011. http://ndltd.ncl.edu.tw/handle/40831999048528243364.

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碩士<br>國立清華大學<br>核子工程與科學研究所<br>99<br>Abstract In-vessel retention（IVR）of core melt is a key severe accident management strategy adopted by advanced light water reactors such as AP600﹐AP1000. One of the viable means for In-vessel retention is external reactor vessel cooling （ERVC）which involves flooding of the reactor cavity during severe accidents. In this study﹐RELAP5-3D system thermal hydraulic code is used to simulate the behaviors of IVR design of AP1000 Pressurized Water Reactor. Both one-dimensional and multi-dimensional models of the code are adopted in the simulation. The results from

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NARCISI, Vincenzo. "Validation of RELAP5-3D© for liquid metal reactor technologies." Doctoral thesis, 2020. http://hdl.handle.net/11573/1363375.

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The present research work set in an international and national context that includes the efforts of several universities and research centers in a strict collaboration. Within the international framework, the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of “Sapienza” University of Rome (UNIROMA1) has been recently involved in the Horizon 2020 (H2020) SESAME (thermal-hydraulic Simulations and Experiments for the Safety Assessment of MEtal-cooled reactors) project. The project aims to contribute to the liquid metal-cooled fast reactors (LMFRs) development, including the

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Book chapters on the topic "RELAP5-3D"

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S. Liang, Thomas K. "Development of an Appendix K Version of RELAP5-3D and Associated Deterministic-Realistic Hybrid Methodology for LOCA Licensing Analysis." In Nuclear Power - System Simulations and Operation. InTech, 2011. http://dx.doi.org/10.5772/17345.

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Conference papers on the topic "RELAP5-3D"

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Parrinello, Valeria, Marco Lanfredini, Alessandro Petruzzi, and Marco Cherubini. "Investigations on RELAP5-3D to RELAP5-3D Coupling Methodology by PVMEXEC." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60748.

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In the framework of a BEPU (Best Estimate Plus Uncertainty) approach within the licensing process of a nuclear power plant, the need to extend the resources of nuclear system thermal-hydraulics codes, such as RELAP5-3D, arises to allow more detailed simulations of the complex 3D reality of Nuclear Power Plants (NPPs), either under normal steady-state or during various accident scenarios. Currently, it is not possible to achieve the same degree of detail for a whole nuclear system when it is simulated with RELAP5-3D and this is due to the inherent limitations in the number of components and vol

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Mesina, George L., David L. Aumiller, and Francis X. Buschman. "Automated, Highly Accurate Verification of RELAP5-3D." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-31153.

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Computer programs that analyze light water reactor safety solve complex systems of governing, closure and special process equations to model the underlying physics. In addition, these programs incorporate many other features and are quite large. RELAP5-3D[1] has over 300,000 lines of coding for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. Verification ensures that a program is built right by checking that it meets its design specifications. Recently, there

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Fiori, F., A. Kovtonyuk, A. Petruzzi, M. Kovtonyuk, and F. D’Auria. "RELAP5-3D Assessment Using a Qualified Experimental Database." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16749.

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The independent code assessment constitutes an important part in the qualification of an analytical tool for performing licensing or NPP support activities. The consistent application of adopted modeling procedures together with qualified experimental database plays a key role in the code assessment process. It should be mentioned that a consistent code assessment supported by a qualified experimental database is an important step for developing a solid ground for the uncertainty evaluation in the frame of Best Estimate Plus Uncertainty (BEPU) approach. The assessment of RELAP5-3D code capabil

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Bubelis, Evaldas, Algirdas Kaliatka, and Eugenijus Uspuras. "RELAP5-3D Code Validation for RBMK-1500 Reactors." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49122.

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This paper deals with RELAP5-3D code validation through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A successful best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total r

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Riemke, Richard A., Cliff B. Davis, and Richard R. Schultz. "RELAP5-3D Code Includes ATHENA Features and Models." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89217.

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Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, car

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Riemke, Richard A., Cliff B. Davis, and Richard R. Schultz. "Recent Hydrodynamics Improvements to the RELAP5-3D Code." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75380.

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The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radionuclide transport model, (5) improved pump model, and (6) compressor model. These changes will be discussed.

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Mesina, George L. "Reformulation RELAP5-3D in FORTRAN 95 and Results." In ASME 2010 3rd Joint US-European Fluids Engineering Summer Meeting collocated with 8th International Conference on Nanochannels, Microchannels, and Minichannels. ASMEDC, 2010. http://dx.doi.org/10.1115/fedsm-icnmm2010-30401.

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RELAP5-3D is a nuclear power plant code used worldwide for safety analysis, design, and operator training. In keeping with ongoing developments in the computing industry, we have re-architected the code in the FORTRAN 95 language [2], the current, fully-available, ANSI standard FORTRAN language. These changes include a complete reworking of the database and conversion of the source code to take advantage of new constructs. The improvements and impacts to the code are manifold. It is a completely machine-independent code that produces machine independent fluid property and plot files and expand

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Bubelis, Evaldas, Algirdas Kaliatka, and Eugenijus Uspuras. "RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22450.

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The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed qu

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Mesina, George L., and Nolan Anderson. "Enhanced Verification for RELAP5-3D Parameter and Sensitivity Studies." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-61040.

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The RELAP5-3D1 program solves a complex system of governing, closure and special process equations to model the underlying physics of nuclear power plants. For SQA (software quality assurance), the code must be verified and validated (V&V) to ensure proper performance before release to users. The physical models are validated against data from experiments and plants and verified against specifications for the computer code. In addition to physics, programs such as RELAP5-3D perform numerous other functions and processes that should also be checked to guarantee correct results. Functions in

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Riemke, Richard A., Cliff B. Davis, and Richard R. Schultz. "Heat Transfer Boundary Conditions in the RELAP5-3D Code." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48154.

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The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general tabl

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Reports on the topic "RELAP5-3D"

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Aaron J. Pawel and Dr. George L. Mesina. Uncertainty Analysis for RELAP5-3D. Office of Scientific and Technical Information (OSTI), 2011. http://dx.doi.org/10.2172/1042350.

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Alexandra E Gertman and Dr. George L Mesina. Uncertainty Analysis of RELAP5-3D. Office of Scientific and Technical Information (OSTI), 2012. http://dx.doi.org/10.2172/1056002.

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Paul D. Bayless. RELAP5-3D Developmental Assessment: Comparison of. Office of Scientific and Technical Information (OSTI), 2013. http://dx.doi.org/10.2172/1107269.

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Fisher, James Ebberly. RELAP5-3D Code Validation for RBMK Phenomena. Office of Scientific and Technical Information (OSTI), 1999. http://dx.doi.org/10.2172/911482.

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Mesina, George L. Consistent Fluid Property Evaluation in RELAP5-3D. Office of Scientific and Technical Information (OSTI), 2016. http://dx.doi.org/10.2172/1481783.

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Mesina, George L. RELAP5-3D Restart and Backup Verification Testing. Office of Scientific and Technical Information (OSTI), 2013. http://dx.doi.org/10.2172/1120815.

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Dr. George L Mesina. RELAP5-3D Developer Guidelines and Programming Practices. Office of Scientific and Technical Information (OSTI), 2014. http://dx.doi.org/10.2172/1134413.

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Dr. George L Mesina. RELAP5-3D Developer Guidelines and Programming Pra. Office of Scientific and Technical Information (OSTI), 2013. http://dx.doi.org/10.2172/1087686.

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Dr. George L Mesina. RELAP5-3D Developer Guidelines and Programming Pra. Office of Scientific and Technical Information (OSTI), 2013. http://dx.doi.org/10.2172/1087688.

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Fisher, J. E. RELAP5-3D code validation for RBMK phenomena. Office of Scientific and Technical Information (OSTI), 1999. http://dx.doi.org/10.2172/752162.

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