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Proc. ASME. ICONE28, Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, V002T07A022, August 4–6, 2021
Paper No: ICONE28-64703
... advantages are Keywords: Thermal-hydraulics, system code, solar salt, its thermal stability, large temperature range, low cost, high MMR, V & V density, and low vapor pressure. For example, MMR is a prismatic block design which has emerged as one preferred 1. INTRODUCTION concept of gas cooled reactors (GCRs...
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T01A013, August 4–5, 2020
Paper No: ICONE2020-16599
... to eventually diminish with increasingly lower hold points, indicating a reduction in retained water. Keywords: dry cask storage, vacuum drying, residual water, water retention, dew point, thermal-hydraulics NOMENCLATURE DAQ data acquisition system DP dew point MS mass spectrometer NPT national pipe thread PV...
Proc. ASME. ICONE2020, Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors, V002T11A003, August 4–5, 2020
Paper No: ICONE2020-16196
... information in each calculation step. The validation result using FNCC provides good agreement with a result of the one-column flow distribution experiment. thermal-hydraulics high temperature gas-cooled reactor DEVELOPMENT OF A FLOW NETWORK CALCULATION CODE (FNCC) FOR HIGH TEMPERATURE GAS-COOLED...
Proc. ASME. ICONE2020, Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors, V002T11A004, August 4–5, 2020
Paper No: ICONE2020-16199
...Abstract Abstract In the thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR), unintended flows such as gap flows between columns, cross flows between column layers and gap flows between permanent reflectors should be analyzed to minimizing...
Proc. ASME. ICONE2020, Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation, V003T13A040, August 4–5, 2020
Paper No: ICONE2020-16741
...MODELING OF THERMAL HYDRAULIC CHARACTERISTICS FOR A LBE-COOLED FAST REACTOR HELICAL COILED TYPE STEAM GENERATOR Xueyou Ding1, Qinglong Wen1,2 Zhiqiang Chen1, Shenhui Ruan1, Cheng Cheng3 1Department of Nuclear Engineering, School of Energy and Power Engineering, Chongqing University, Chongqing...
Proc. ASME. ICONE21, Volume 4: Thermal Hydraulics, V004T09A103, July 29–August 2, 2013
Paper No: ICONE21-16570
... uncertainty analysis for complex computational codes (e.g., RELAP5 thermal-hydraulics) computations. Different sampling techniques, dependencies between uncertainty sources, and accurate inference on results are among the issues to be considered. The dynamic behavior of the system codes executed in each time...
Proc. ASME. ICONE21, Volume 3: Nuclear Safety and Security; Codes, Standards, Licensing and Regulatory Issues; Computational Fluid Dynamics and Coupled Codes, V003T06A037, July 29–August 2, 2013
Paper No: ICONE21-16165
... Anticipated transient PSA for operating NPP VVER ATWS safety analysis thermal-hydraulics MTC Safety analysis on operating nuclear power plant (NPP) plays an important role on nuclear energy application, especially after the severe accident of Fukushima plants. This paper focuses...
Proc. ASME. ICONE20-POWER2012, Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems, 439-447, July 30–August 3, 2012
Paper No: ICONE20-POWER2012-54871
... plant behavior during the first four phases of the accident is discussed and analyzed in comparison to available post-accident data and measurements. The calculation captures the plant response in terms of the thermal- hydraulics very well during the first two phases. However, during the reflooding...
Proc. ASME. ICONE18, 18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 401-413, May 17–21, 2010
Paper No: ICONE18-29386
... objective is to provide unique experimental data that are needed for the validation of thermal- hydraulic codes and to support safety assessments for VVER- 1000 reactors. Under OECD PSB-VVER project, five experiments were to be performed. Performance of each experiment in PSB-VVER was preceded by design...
Proc. ASME. ICONE18, 18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 893-903, May 17–21, 2010
Paper No: ICONE18-29933
... of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements. LWR PERSEO Test CATHARE Thermal-Hydraulics Decay Heat Removal 1 Copyright © 2010 by ASME Proceedings of the 18th International Conference on Nuclear Engineering...
Giacomino Bandini, Maddalena Casamirra, Francesco Castiglia, Mariarosa Giardina, Paride Meloni, Massimiliano Polidori
Proc. ASME. ICONE16, Volume 3: Thermal Hydraulics; Instrumentation and Controls, 759-768, May 11–15, 2008
Paper No: ICONE16-48720
... circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety...
Proc. ASME. ICONE10, 10th International Conference on Nuclear Engineering, Volume 3, 95-103, April 14–18, 2002
Paper No: ICONE10-22098
... models. Light Water Reactor Severe Accident Thermal-Hydraulics In-Vessel Debris Cooling 10 G FO A, fet ta- as en annu the r evalu Some gap fluxes There deve restri and avera patte fluxes We e simpl perfo Institu Proceedings of ICONE10 10th International Conference on Nuclear...
Proc. ASME. ICONE10, 10th International Conference on Nuclear Engineering, Volume 4, 69-81, April 14–18, 2002
Paper No: ICONE10-22350
... 05 03 2009 As the result of the advancing TCP/IP based inter-process communication technology, more and more legacy thermal-hydraulic codes have been coupled with neutronics codes to provide best-estimate capabilities for reactivity related reactor transient analysis. Most...
Proc. ASME. ICONE12, 12th International Conference on Nuclear Engineering, Volume 3, 795-801, April 25–29, 2004
Paper No: ICONE12-49496
... 25 11 2008 R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power company, reactor vendors, universities since 2002...