Volume 5 Issue 2 | ASME J. of Nuclear Rad Sci.

Guest Editorial

Special Section: Selected Papers From the International Youth Nuclear Congress 2018 - 26th WiN Global Annual Conference

Co-chairs of IYNCWIN18, Gabriele Voigt, Denis Janin, Melina Belinco, Cristian Vega

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020301. doi: https://doi.org/10.1115/1.4042797

View articletitled, Special Section: Selected Papers From the International Youth Nuclear Congress 2018 - 26th WiN Global Annual Conference

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Topics: Accidents , Climate change , Nuclear power

Special Section: Selected Papers from the International Youth Nuclear Congress 2018 - 26th WiN Global Annual Conference

Pyrolysis and High Performance Plasma Treatment Applied to Spent Ion Exchange Resins

Hernán Ariel Castro, Raúl Ariel Rodríguez, Vittorio Luca, Hugo Luis Bianchi

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020901. doi: https://doi.org/10.1115/1.4042193

Abstract

View articletitled, Pyrolysis and High Performance Plasma Treatment Applied to Spent Ion Exchange Resins

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Topics: Plasmas (Ionized gases) , Pyrolysis , Resins , Ion exchange , Temperature

Thermal–Hydraulic Design in the Transition Laminar–to–Turbulent Flow Regime

J. Lupiano Contreras, A. S. Doval, P. A. Alberto

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020902. doi: https://doi.org/10.1115/1.4042363

Abstract

View articletitled, Thermal–Hydraulic Design in the Transition Laminar–to–Turbulent Flow Regime

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Topics: Design , Flow (Dynamics) , Uncertainty , Cooling , Heat transfer coefficients

Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

Martín Lemes, Alicia Denis, Alejandro Soba

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020903. doi: https://doi.org/10.1115/1.4042705

Abstract

View articletitled, Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

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Topics: Accidents , Cladding systems (Building) , Coolants , Fuel rods , Fuels , High temperature , Hydrogen , Simulation , Temperature , Water

Implementation of Hydrogen Solid Solubility Data and Precipitation Threshold Stresses in the Fuel Rod Code TESPA-ROD

Felix Boldt

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020904. doi: https://doi.org/10.1115/1.4042118

Abstract

View articletitled, Implementation of Hydrogen Solid Solubility Data and Precipitation Threshold Stresses in the Fuel Rod Code TESPA-ROD

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Topics: Cladding systems (Building) , Fuel rods , Heat , Hydrogen , Precipitation , Storage , Stress , Temperature , Transients (Dynamics) , Simulation

Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

Lena Andriolo, Clément Meriot, Nikolai Bakouta

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020905. doi: https://doi.org/10.1115/1.4042360

Abstract

View articletitled, Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

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Topics: Vessels

Nuclear Communication Campaign in Spain: Basic Nuclear Fusion Course

Kevin Fernández-Cosials, Alfonso Barbas, José García Laruelo

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020906. doi: https://doi.org/10.1115/1.4042192

Abstract

View articletitled, Nuclear Communication Campaign in Spain: Basic Nuclear Fusion Course

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Topics: Nuclear fusion , Nuclear power

Glasses for Nuclear Waste Immobilization. Effect of Lutetium Addition on Yttrium Aluminosilicate Glass Crystallization Kinetics

D. C. Lago, M. B. Bortot, M. O. Prado

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020907. doi: https://doi.org/10.1115/1.4042497

Abstract

View articletitled, Glasses for Nuclear Waste Immobilization. Effect of Lutetium Addition on Yttrium Aluminosilicate Glass Crystallization Kinetics

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Topics: Glass , Yttrium , Crystallization

Validation of Selected Cesar Friction Models of the ASTECV21 Code Based on Moby Dick Experiments

I. Gómez-García-Toraño, L. Laborde

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020908. doi: https://doi.org/10.1115/1.4042119

Abstract

View articletitled, Validation of Selected Cesar Friction Models of the ASTECV21 Code Based on Moby Dick Experiments

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Topics: Flow (Dynamics) , Friction , Momentum , Pressure , Pressure drop

Sensitivity Analysis of the Thermal Diffusion Coefficient Effect on the Departure From Nucleate Boiling Ratio With the VIPRE Code

Rosario Delgado-Tardáguila, Marisol Corisco, Antonio Espejo, Daniel Navarro, Javier Riverola

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020909. doi: https://doi.org/10.1115/1.4042358

Abstract

View articletitled, Sensitivity Analysis of the Thermal Diffusion Coefficient Effect on the Departure From Nucleate Boiling Ratio With the VIPRE Code

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Topics: Design , Fuels , Nucleate boiling , Thermal diffusion , Turbulence , Enthalpy , Momentum , Heat flux , Flow (Dynamics) , Sensitivity analysis

Preparation of UF₄ by Carbochlorination of U₃O₈ and Solid-State Halogen Exchange Reaction

Cristina N. Guibaldo, Georgina De Micco, Ana E. Bohé

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020910. doi: https://doi.org/10.1115/1.4042362

Abstract

View articletitled, Preparation of UF4 by Carbochlorination of U3O8 and Solid-State Halogen Exchange Reaction

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Topics: Uranium , Carbon , Quartz , X-ray diffraction

Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

Waleed Ishaque

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020911. doi: https://doi.org/10.1115/1.4042498

Abstract

View articletitled, Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

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Topics: American Petroleum Institute , Fracture (Materials) , Fracture toughness , Junctions , Nozzles , Stress , Vessels , Design , Fitness-for-service , Steel

Evaluation of the Steam Generator Clogging Phenomena Kinetics by γ-Ray Counting

Patricia Schindler, Pierre Guy Allinei, Laurent Loubet, Veronique Pointeau

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020912. doi: https://doi.org/10.1115/1.4042851

Extract

View articletitled, Evaluation of the Steam Generator Clogging Phenomena Kinetics by γ-Ray Counting

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Topics: Boilers , Modeling , Noise (Sound) , Sensors , Fluids , Control systems , Crystals , Signals

Technical Communication During Nuclear and Radiological Emergencies With the Tools to Support International Assessment and Prognosis

Margarita Tzivaki, Edward J. Waller

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020913. doi: https://doi.org/10.1115/1.4042906

Abstract

View articletitled, Technical Communication During Nuclear and Radiological Emergencies With the Tools to Support International Assessment and Prognosis

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Topics: Emergencies , Emergency response

Micro-Calorimetric Study on the Thermal Stability of N, N-Dimethylhydroxylamine in the APOR Process

Cao Zhi, Yan Taihong, Bian Xiaoyan, Li Chuanbo, Zuo Chen, Zheng Weifang

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020914. doi: https://doi.org/10.1115/1.4042499

Abstract

View articletitled, Micro-Calorimetric Study on the Thermal Stability of N, N-Dimethylhydroxylamine in the APOR Process

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Topics: Enthalpy , Heat , Temperature , Thermal stability , Metals

Electrical Power System Modeling of Atucha II Nuclear Power Plant Using etap

Nicolás Alejandro Malinovsky

ASME J of Nuclear Rad Sci. April 2019, 5(2): 020915. doi: https://doi.org/10.1115/1.4042361

Abstract

View articletitled, Electrical Power System Modeling of Atucha II Nuclear Power Plant Using etap

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Topics: Circuits , Electricity (Physics) , Nuclear power stations , Stress , Power systems (Machinery) , Modeling , Computer software , Flow (Dynamics) , Electronic systems

Research Papers

Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka, Hiroyuki Ohshima

ASME J of Nuclear Rad Sci. April 2019, 5(2): 021001. doi: https://doi.org/10.1115/1.4042191

Abstract

View articletitled, Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

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Topics: Ducts , Flow (Dynamics) , Fuels , Manufacturing , Sodium , Temperature distribution , Temperature , Sodium fast reactors , Wire , Thermal hydraulics

Laminar Natural Convection Heat Transfer From Vertical 7 × 7 Rod Bundles in Liquid Sodium

Koichi Hata, Katsuya Fukuda, Tohru Mizuuchi

ASME J of Nuclear Rad Sci. April 2019, 5(2): 021002. doi: https://doi.org/10.1115/1.4042356

Abstract

View articletitled, Laminar Natural Convection Heat Transfer From Vertical 7 × 7 Rod Bundles in Liquid Sodium

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Topics: Cylinders , Fuel rods , Heat transfer , Natural convection , Sodium , Temperature

Experimental and Numerical Analysis of Mixing Process of Two Component Gases in a Vertical Fluid Layer in a VHTR

Tetsuaki Takeda

ASME J of Nuclear Rad Sci. April 2019, 5(2): 021003. doi: https://doi.org/10.1115/1.4041690

Abstract

View articletitled, Experimental and Numerical Analysis of Mixing Process of Two Component Gases in a Vertical Fluid Layer in a VHTR

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Topics: Gases , High temperature , Natural convection , Numerical analysis , Temperature , Diffusion (Physics) , Flow (Dynamics) , Very high temperature reactors , Low temperature , Fluids

Thermomechanical Safety Analyses for a ²³⁸Pu Production Target at the HFIR

Christopher J. Hurt, James D. Freels, Prashant K. Jain, G. Ivan Maldonado

ASME J of Nuclear Rad Sci. April 2019, 5(2): 021004. doi: https://doi.org/10.1115/1.4041295

Abstract

View articletitled, Thermomechanical Safety Analyses for a 238Pu Production Target at the HFIR

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Topics: Cycles , Irradiation (Radiation exposure) , Safety , Temperature , Stress , Heat transfer , Electrical conductance , Steady state , Heat , Cladding systems (Building)

Technology Review

Current Status and Future Developments in Nuclear-Power Industry of the World

I. Pioro, R. B. Duffey, P. L. Kirillov, R. Pioro, A. Zvorykin, R. Machrafi

ASME J of Nuclear Rad Sci. April 2019, 5(2): 024001. doi: https://doi.org/10.1115/1.4042194

Abstract

View articletitled, Current Status and Future Developments in Nuclear-Power Industry of the World

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Topics: China , Coal , Electric power generation , Nuclear power , Nuclear power stations , Nuclear reactors , Power stations , Wind , Fuels , Energy resources

Technical Brief

Study on Thallium Bromide Radioactive Detector

He Li-xia, Hao Xiao-yong, He Gao-kui

ASME J of Nuclear Rad Sci. April 2019, 5(2): 024501. doi: https://doi.org/10.1115/1.4042357

Abstract

View articletitled, Study on Thallium Bromide Radioactive Detector

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Topics: Sensors , Signals , Semiconductors (Materials) , Engineering prototypes , Gamma rays

Discussion

The Necessity to Reexamine Key Component Material With Spectrometry Analysis in the Nuclear Island

Chongzhi Wu, Jindong Zhang, Zongkui Wang, Zengkui Zhang, Shenqing Fu

ASME J of Nuclear Rad Sci. April 2019, 5(2): 025501. doi: https://doi.org/10.1115/1.4042221

Abstract

View articletitled, The Necessity to Reexamine Key Component Material With Spectrometry Analysis in the Nuclear Island

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Topics: Metals , Nuclear power , Spectroscopy , Welding , Pumps , Nozzles , Construction , Pipes , Heat

Guest Editorial

Special Section: Selected Papers From the International Youth Nuclear Congress 2018 - 26th WiN Global Annual Conference

Special Section: Selected Papers from the International Youth Nuclear Congress 2018 - 26th WiN Global Annual Conference

Pyrolysis and High Performance Plasma Treatment Applied to Spent Ion Exchange Resins

Thermal–Hydraulic Design in the Transition Laminar–to–Turbulent Flow Regime

Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

Implementation of Hydrogen Solid Solubility Data and Precipitation Threshold Stresses in the Fuel Rod Code TESPA-ROD

Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

Nuclear Communication Campaign in Spain: Basic Nuclear Fusion Course

Glasses for Nuclear Waste Immobilization. Effect of Lutetium Addition on Yttrium Aluminosilicate Glass Crystallization Kinetics

Validation of Selected Cesar Friction Models of the ASTECV21 Code Based on Moby Dick Experiments

Sensitivity Analysis of the Thermal Diffusion Coefficient Effect on the Departure From Nucleate Boiling Ratio With the VIPRE Code

Preparation of UF₄ by Carbochlorination of U₃O₈ and Solid-State Halogen Exchange Reaction

Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

Evaluation of the Steam Generator Clogging Phenomena Kinetics by γ-Ray Counting

Technical Communication During Nuclear and Radiological Emergencies With the Tools to Support International Assessment and Prognosis

Micro-Calorimetric Study on the Thermal Stability of N, N-Dimethylhydroxylamine in the APOR Process

Electrical Power System Modeling of Atucha II Nuclear Power Plant Using etap

Research Papers

Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

Laminar Natural Convection Heat Transfer From Vertical 7 × 7 Rod Bundles in Liquid Sodium

Experimental and Numerical Analysis of Mixing Process of Two Component Gases in a Vertical Fluid Layer in a VHTR

Thermomechanical Safety Analyses for a ²³⁸Pu Production Target at the HFIR

Technology Review

Current Status and Future Developments in Nuclear-Power Industry of the World

Technical Brief

Study on Thallium Bromide Radioactive Detector

Discussion

The Necessity to Reexamine Key Component Material With Spectrometry Analysis in the Nuclear Island

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Guest Editorial

Special Section: Selected Papers from the International Youth Nuclear Congress 2018 - 26th WiN Global Annual Conference

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Technology Review

Technical Brief

Discussion

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