Volume 8 Issue 4 | ASME J. of Nuclear Rad Sci.

Special Issue: NENE

Research Papers

Thermal Hydraulics and CFD

Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

Filip Osuský, Branislav Vrban, Štefan Čerba, Jakub Lüley, Vladimír Nečas

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041401. doi: https://doi.org/10.1115/1.4051515

View articletitled, Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

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Topics: Coolants , Design , Fast neutron reactors , Fuels , Heat , Temperature , Transients (Dynamics) , Temperature distribution , Simulation , Cladding systems (Building)

Flow and Heat Transfer Simulation in a Complete Pressurized Water Reactor Fuel Assembly Using Wall-Modeled RANS

B. Mikuž, F. Roelofs

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041402. doi: https://doi.org/10.1115/1.4051446

Abstract

View articletitled, Flow and Heat Transfer Simulation in a Complete Pressurized Water Reactor Fuel Assembly Using Wall-Modeled RANS

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Topics: Flow (Dynamics) , Fuel rods , Fuels , Manufacturing , Pressure drop , Pressurized water reactors , Temperature , Simulation , Reynolds-averaged Navier–Stokes equations , Turbulence

Reactor Physics

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Marjan Kromar, Bojan Kurinčič

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041501. doi: https://doi.org/10.1115/1.4050991

Abstract

View articletitled, Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

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Topics: Boron , Cycles , Fuels , Manufacturing , Neutrons , Nuclear design , Nuclear power stations , Nuclear fission , Cross section (Physics) , Rods

Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation

Branislav Vrban, Štefan Čerba, Jakub Lüley, Filip Osuský, Vladimír Nečas

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041502. doi: https://doi.org/10.1115/1.4051613

Abstract

View articletitled, Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation

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Topics: Fuels , Nuclear fission , Uncertainty , Water , Nuclides , Neutrons

On the Self-Shielding in the Unresolved Resonance Range

Daniel Lopez Aldama, Andrej Trkov, Dermott E. Cullen

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041503. doi: https://doi.org/10.1115/1.4052377

Abstract

View articletitled, On the Self-Shielding in the Unresolved Resonance Range

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Topics: Cross section (Physics) , Resonance , Spectra (Spectroscopy) , Probability , Leakage

Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

Radojko Jaćimović, Maria Ângela de Barros Correia Menezes

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041504. doi: https://doi.org/10.1115/1.4051249

Abstract

View articletitled, Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

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Topics: Flux (Metallurgy) , Irradiation (Radiation exposure) , Neutrons , Neutron activation analysis , Uncertainty

Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

Eva Vilímová, Tomáš Peltan, Jana Jiřičková

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041505. doi: https://doi.org/10.1115/1.4050990

Abstract

View articletitled, Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

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Topics: Graphite , Neutron flux , Neutrons , Sensors , Spent nuclear fuels , Water , Fuels

Nuclear Fuels/Cycles and Materials

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Jakub Lüley, Branislav Vrban, Štefan Čerba, Filip Osuský, Vladimír Nečas

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041601. doi: https://doi.org/10.1115/1.4052519

Abstract

View articletitled, Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

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Topics: Fuels , Grain size , Irradiation (Radiation exposure) , Modeling , Nuclear fission , Rods , Pressure , Cladding systems (Building) , Stress

Coupling of ASTEC V2.1 and RASCAL 4.3 Codes to Evaluate the Source Term and the Radiological Consequences of a Loss-of-Cooling Accident at a Spent Fuel Pool

Antonio Guglielmelli, Stefano Ederli, Fulvio Mascari, Federico Rocchi, Pietro Maccari

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041602. doi: https://doi.org/10.1115/1.4054514

Abstract

View articletitled, Coupling of ASTEC V2.1 and RASCAL 4.3 Codes to Evaluate the Source Term and the Radiological Consequences of a Loss-of-Cooling Accident at a Spent Fuel Pool

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Topics: Accidents , Cooling , Emissions , Fukushima nuclear disaster, Japan, 2011 , Meteorology , Nuclear power stations , Simulation , Spent nuclear fuels , Wind , Wind velocity

The Heat Capacity of PuO₂ at High Temperature: Molecular Dynamics Calculations

Rolando Calabrese

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041603. doi: https://doi.org/10.1115/1.4054941

Abstract

View articletitled, The Heat Capacity of PuO2 at High Temperature: Molecular Dynamics Calculations

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Topics: Heat capacity , High temperature , Oxygen , Temperature , Molecular dynamics , Enthalpy , Modeling

Nuclear Safety/Security and Beyond Design Basis Events

Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

Matjaž Žvar, Tomaž Žagar

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041701. doi: https://doi.org/10.1115/1.4052013

Abstract

View articletitled, Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

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Topics: Accidents , Design , Drills (Tools) , Education , Emergency preparedness , Nuclear power stations , Safety , Optimization , Computers , Spent nuclear fuels

Experimental Investigation on Debris Bed Quenching With Additional Non-Condensable Gas Injection

Markus Petroff, Rudi Kulenovic, Jörg Starflinger

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041702. doi: https://doi.org/10.1115/1.4051876

Abstract

View articletitled, Experimental Investigation on Debris Bed Quenching With Additional Non-Condensable Gas Injection

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Topics: Floods , Particulate matter , Quenching (Metalworking) , Temperature , Test facilities , Water , Flow (Dynamics) , Concretes

RAW Management and Decommissioning

Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

Silja Häkkinen

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041901. doi: https://doi.org/10.1115/1.4051444

Abstract

View articletitled, Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

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Topics: Approximation , Coolants , Cycles , Irradiation (Radiation exposure) , Manufacturing , Nuclides , Spent nuclear fuels , Heat , Boron , Photons

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

Davor Grgić, Mario Matijević, Paulina Dučkić, Radomir Ječmenica

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041902. doi: https://doi.org/10.1115/1.4051447

Abstract

View articletitled, Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

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Topics: Accidents , Fuels , Hardware , Neutrons , Nuclear power stations , Particulate matter , Spent nuclear fuels , Storage , Uncertainty , Manufacturing

Radiation Science

serpent 2 Validation for Radiation Shielding Applications

Silja Häkkinen

ASME J of Nuclear Rad Sci. October 2022, 8(4): 042001. doi: https://doi.org/10.1115/1.4051614

Abstract

View articletitled, serpent 2 Validation for Radiation Shielding Applications

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Topics: Iron , Neutrons , Photons , Spectra (Spectroscopy)

Fusion Engineering

Deposition of ITER Vacuum Vessel Dust Inside the Pressure Suppression System During a Loss of Coolant Accident: Experimental and Numerical Analyses

Miriam Ibba, Alessio Pesetti, Michele Raucci, Flavio Parozzi, Roberta Lazzeri, Donato Aquaro

ASME J of Nuclear Rad Sci. October 2022, 8(4): 042301. doi: https://doi.org/10.1115/1.4050768

Abstract

View articletitled, Deposition of ITER Vacuum Vessel Dust Inside the Pressure Suppression System During a Loss of Coolant Accident: Experimental and Numerical Analyses

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Topics: Dust , Water , Containers , Vessels , Pipes

Technical Briefs

Nuclear Power and Nuclear Technologies Can Benefit From Regional Implementation of Multinational Approaches

Charles McCombie

ASME J of Nuclear Rad Sci. October 2022, 8(4): 044501. doi: https://doi.org/10.1115/1.4049756

Abstract

View articletitled, Nuclear Power and Nuclear Technologies Can Benefit From Regional Implementation of Multinational Approaches

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Topics: Nuclear power , Radioactive wastes , Economics , Waste management

Experimental Investigation on the Retention of Soluble Particles by Pool Scrubbing

R. Vennemann, M. Klauck, H.-J. Allelein

ASME J of Nuclear Rad Sci. October 2022, 8(4): 044502. doi: https://doi.org/10.1115/1.4051250

Abstract

View articletitled, Experimental Investigation on the Retention of Soluble Particles by Pool Scrubbing

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Topics: Aerosols , Particulate matter , Water , Accidents , Bubbles

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Riku Tuominen, Ville Valtavirta

ASME J of Nuclear Rad Sci. October 2022, 8(4): 044503. doi: https://doi.org/10.1115/1.4051445

Abstract

View articletitled, The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

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Topics: Boiling water reactors , Emissions , Heat , Manufacturing , Nuclear fission , Photons , Probability , Resonance , Spent nuclear fuels , Nuclides

Validating the serpent-ants Calculation Chain Using BEAVRS Fresh Core Hot Zero Power Data

Ville Valtavirta, Antti Rintala, Unna Lauranto

ASME J of Nuclear Rad Sci. October 2022, 8(4): 044504. doi: https://doi.org/10.1115/1.4052731

Abstract

View articletitled, Validating the serpent-ants Calculation Chain Using BEAVRS Fresh Core Hot Zero Power Data

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Topics: Boron , Chain , Manufacturing , Errors , Temperature , Fuels , Physics

Special Issue: NENE

Research Papers

Thermal Hydraulics and CFD

Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

Flow and Heat Transfer Simulation in a Complete Pressurized Water Reactor Fuel Assembly Using Wall-Modeled RANS

Reactor Physics

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation

On the Self-Shielding in the Unresolved Resonance Range

Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

Nuclear Fuels/Cycles and Materials

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Coupling of ASTEC V2.1 and RASCAL 4.3 Codes to Evaluate the Source Term and the Radiological Consequences of a Loss-of-Cooling Accident at a Spent Fuel Pool

The Heat Capacity of PuO₂ at High Temperature: Molecular Dynamics Calculations

Nuclear Safety/Security and Beyond Design Basis Events

Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

Experimental Investigation on Debris Bed Quenching With Additional Non-Condensable Gas Injection

RAW Management and Decommissioning

Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

Radiation Science

serpent 2 Validation for Radiation Shielding Applications

Fusion Engineering

Deposition of ITER Vacuum Vessel Dust Inside the Pressure Suppression System During a Loss of Coolant Accident: Experimental and Numerical Analyses

Technical Briefs

Nuclear Power and Nuclear Technologies Can Benefit From Regional Implementation of Multinational Approaches

Experimental Investigation on the Retention of Soluble Particles by Pool Scrubbing

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Validating the serpent-ants Calculation Chain Using BEAVRS Fresh Core Hot Zero Power Data

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Issues

Special Issue: NENE

Research Papers

Thermal Hydraulics and CFD

Reactor Physics

Nuclear Fuels/Cycles and Materials

Nuclear Safety/Security and Beyond Design Basis Events

RAW Management and Decommissioning

Radiation Science

Fusion Engineering

Technical Briefs

Email alerts

RSS Feeds

ASME Journals

ASME Conference Proceedings

ASME eBooks

Resources

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