Volume 3 Issue 2 | ASME J. of Nuclear Rad Sci.

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Guest Editorial

Special Section: Pressurized Heavy Water Reactors Safety

S. Banerjee, A. K. Nayak, T. (Nithy) Nitheanandan

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020301. doi: https://doi.org/10.1115/1.4035852

Extract

View articletitled, Special Section: Pressurized Heavy Water Reactors Safety

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Topics: Design , Heavy water reactors , Safety , Accidents , Water , Nuclear engineering , Thermal hydraulics

Special Section Papers

Pressurized Heavy Water Reactor Technology: Its Relevance Today

A. K. Nayak, S. Banerjee

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020901. doi: https://doi.org/10.1115/1.4035856

Abstract

View articletitled, Pressurized Heavy Water Reactor Technology: Its Relevance Today

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Topics: Heavy water reactors , Fuels , Uranium

Development of Technologies and Safety Systems for Pressurized Heavy Water Reactors in India

S. Banerjee, H. P. Gupta

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020902. doi: https://doi.org/10.1115/1.4035435

Abstract

View articletitled, Development of Technologies and Safety Systems for Pressurized Heavy Water Reactors in India

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Topics: Heavy water reactors , Safety , Fuels , Design

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

Thambiayah Nitheanandan, X. Cao, J.-H. Choi, D. Dupleac, D.-H. Kim, H. G. Lele, A. K. Nayak, H. P. Rammohan

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020903. doi: https://doi.org/10.1115/1.4035726

Abstract

View articletitled, Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

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Topics: Accidents , Heat , Vessels , Water , Heavy water reactors , Computers , Pressure , Fuels , Containment , Boilers

Analysis of Late Phase Severe Accident Phenomena in CANDU Plant

D. Dupleac

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020904. doi: https://doi.org/10.1115/1.4035416

Abstract

View articletitled, Analysis of Late Phase Severe Accident Phenomena in CANDU Plant

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Topics: Accidents , Vessels , Water , Zirconium , Steam , Failure , Heat transfer coefficients , Temperature

Thermomechanical Behavior of Coolant Channel Assembly in Heavy Water Reactor Under Severe Plant Condition

A. K. Dureja, P. Seshu, D. N. Pawaskar, R. K. Sinha

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020905. doi: https://doi.org/10.1115/1.4035784

Abstract

View articletitled, Thermomechanical Behavior of Coolant Channel Assembly in Heavy Water Reactor Under Severe Plant Condition

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Topics: Contact resistance , Coolants , Deformation , Heat , Heavy water reactors , Manufacturing , Pressure , Temperature , Thermomechanics , Flow (Dynamics)

Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

Andrea Alfonsi, George L. Mesina, Angelo Zoino, Nolan Anderson, Cristian Rabiti

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020906. doi: https://doi.org/10.1115/1.4035851

Abstract

View articletitled, Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

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Topics: Accidents , Cladding systems (Building) , Coolants , Cycles , Design , Flow (Dynamics) , Fuels , Nuclear fuel cycle , Oxidation , Safety

Effects of Filtered Containment Venting on Fission Product Releases During CANDU Reactor Severe Accidents

A. C. Morreale, L. S. Lebel, M. J. Brown

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020907. doi: https://doi.org/10.1115/1.4035434

Abstract

View articletitled, Effects of Filtered Containment Venting on Fission Product Releases During CANDU Reactor Severe Accidents

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Topics: Accidents , Containment , Nuclear fission , Failure , Pressure , Vessels

Collapse Criterion for the CANDU 6 Calandria Vault Floor During a Severe Accident

Robert David

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020908. doi: https://doi.org/10.1115/1.4035336

Abstract

View articletitled, Collapse Criterion for the CANDU 6 Calandria Vault Floor During a Severe Accident

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Topics: Accidents , Collapse , Concretes , Finite element model , Stress , Temperature , Thermal expansion , Failure

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

Sumit V. Prasad, A. K. Nayak

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020909. doi: https://doi.org/10.1115/1.4035691

Abstract

View articletitled, In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

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Topics: Accidents , Heat , Heat transfer , Heavy water reactors , Vessels , Water , Temperature , Cooling

Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

Rajee Guptan, A. K. Vijaya, Vibha Hari, Rajeev Nama

ASME J of Nuclear Rad Sci. April 2017, 3(2): 020910. doi: https://doi.org/10.1115/1.4035783

Abstract

View articletitled, Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

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Topics: Accidents , Containment , Failure , Safety , Probabilistic safety assessment , Accident management , Radioisotopes

Research Papers

Statistical Modeling of Stratified Two-Phase Flow

M. Benz, T. Schulenberg

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021001. doi: https://doi.org/10.1115/1.4035564

Abstract

View articletitled, Statistical Modeling of Stratified Two-Phase Flow

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Topics: Flow (Dynamics) , Turbulence , Waves , Kinetic energy , Spectra (Spectroscopy) , Two-phase flow

Experimental Study of Aqueous Chemical Trisodium Phosphate-Buffered Environment Under Post-LOCA Conditions With Head Loss Measurements

Amir Ali, Daniel LaBrier, Kerry J. Howe, Edward D. Blandford

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021002. doi: https://doi.org/10.1115/1.4034572

Abstract

View articletitled, Experimental Study of Aqueous Chemical Trisodium Phosphate-Buffered Environment Under Post-LOCA Conditions With Head Loss Measurements

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Topics: Aluminum , Corrosion , Epoxy adhesives , Epoxy resins , Fibers , Flow (Dynamics) , Metals , Temperature , Testing , Water

A Feasibility Study of Noncontact Ultrasonic Sensor for Nuclear Power Plant Inspection

Akinori Tamura, Chenghuan Zhong, Anthony J. Croxford, Paul D. Wilcox

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021003. doi: https://doi.org/10.1115/1.4035466

Abstract

View articletitled, A Feasibility Study of Noncontact Ultrasonic Sensor for Nuclear Power Plant Inspection

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Topics: Inspection , Nuclear power stations , Separation (Technology) , Thickness measurement , Ultrasonic transducers , Transducers , Pipes , Radiation (Physics) , Signals

Validation of Constant Load C-Ring Apex Stresses for Stress Corrosion Cracking Testing in Supercritical Water

R. Swift, W. Cook, C. Bradley, R.C. Newman

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021004. doi: https://doi.org/10.1115/1.4034567

Abstract

View articletitled, Validation of Constant Load C-Ring Apex Stresses for Stress Corrosion Cracking Testing in Supercritical Water

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Topics: Deflection , Springs , Stress , Stress corrosion cracking , Testing , Temperature , Thermal expansion , Water , Metals

Development of SNTA Code System for SCWR Core Steady-State Analysis

Lianjie Wang, Wenbo Zhao, Ping Yang, Yongqiang Ma, Di Lu

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021005. doi: https://doi.org/10.1115/1.4035334

Abstract

View articletitled, Development of SNTA Code System for SCWR Core Steady-State Analysis

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Topics: Coolants , Steady state , Supercritical water reactors , Thermal hydraulics , Flow (Dynamics) , Water , Density , Relaxation (Physics) , Design , Finite difference methods

Adhesion of Oxides Grown in Supercritical Water on Selected Austenitic and Ferritic/Martensitic Alloys

D. Artymowicz, C. Bradley, B. Xing, R. C. Newman

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021006. doi: https://doi.org/10.1115/1.4035331

Abstract

View articletitled, Adhesion of Oxides Grown in Supercritical Water on Selected Austenitic and Ferritic/Martensitic Alloys

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Topics: Alloys , Delamination , Adhesion , Descaling (Metals)

A Review of the Turbine Cooling Fraction for Very High Turbine Entry Temperature Helium Gas Turbine Cycles for Generation IV Reactor Power Plants

A. Gad-Briggs, P. Pilidis, T. Nikolaidis

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021007. doi: https://doi.org/10.1115/1.4035332

Abstract

View articletitled, A Review of the Turbine Cooling Fraction for Very High Turbine Entry Temperature Helium Gas Turbine Cycles for Generation IV Reactor Power Plants

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Topics: Cooling , Cycles , Temperature , Turbines , Flow (Dynamics) , Pressure , Blades

Results of Air Barbotage Experiments Simulating Two-Phase Flow in a CANDU End Shield During In-Vessel Retention

Justin H. Spencer

ASME J of Nuclear Rad Sci. April 2017, 3(2): 021008. doi: https://doi.org/10.1115/1.4035566

Abstract

View articletitled, Results of Air Barbotage Experiments Simulating Two-Phase Flow in a CANDU End Shield During In-Vessel Retention

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Topics: Boiling , Cavities , Flow visualization , Heat , Heat flux , Vessels , Water , Accidents , Air flow , Porous materials

Technical Brief

Current Severe Accident Research and Development Topics for Wolsong PHWR Safety in Korea

Y. M. Song, D. H. Kim, S. Y. Park, J. H. Song

ASME J of Nuclear Rad Sci. April 2017, 3(2): 024501. doi: https://doi.org/10.1115/1.4035690

Abstract

View articletitled, Current Severe Accident Research and Development Topics for Wolsong PHWR Safety in Korea

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Topics: Accidents , Containment , Heavy water reactors , Overpressure protection , Safety , Heat , Design , Computers

Thermal-Hydraulic Model and Program Development for Helium-Cooled Accelerator-Driven System

Tianji Peng, Zhiwei Zhou

ASME J of Nuclear Rad Sci. April 2017, 3(2): 024502. doi: https://doi.org/10.1115/1.4035333

Abstract

View articletitled, Thermal-Hydraulic Model and Program Development for Helium-Cooled Accelerator-Driven System

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Topics: Heat , Heat transfer , Helium , Transients (Dynamics) , Cooling systems , Mathematics , Fluids

Book Review

Handbook of Generation-IV Nuclear Reactors

Igor Pioro

ASME J of Nuclear Rad Sci. April 2017, 3(2): 026501. doi: https://doi.org/10.1115/1.4035327

Extract

View articletitled, Handbook of Generation-IV Nuclear Reactors

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Topics: Nuclear reactors , Small modular reactors , Nuclear engineering

Guest Editorial

Special Section: Pressurized Heavy Water Reactors Safety

Special Section Papers

Pressurized Heavy Water Reactor Technology: Its Relevance Today

Development of Technologies and Safety Systems for Pressurized Heavy Water Reactors in India

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

Analysis of Late Phase Severe Accident Phenomena in CANDU Plant

Thermomechanical Behavior of Coolant Channel Assembly in Heavy Water Reactor Under Severe Plant Condition

Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

Effects of Filtered Containment Venting on Fission Product Releases During CANDU Reactor Severe Accidents

Collapse Criterion for the CANDU 6 Calandria Vault Floor During a Severe Accident

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

Research Papers

Statistical Modeling of Stratified Two-Phase Flow

Experimental Study of Aqueous Chemical Trisodium Phosphate-Buffered Environment Under Post-LOCA Conditions With Head Loss Measurements

A Feasibility Study of Noncontact Ultrasonic Sensor for Nuclear Power Plant Inspection

Validation of Constant Load C-Ring Apex Stresses for Stress Corrosion Cracking Testing in Supercritical Water

Development of SNTA Code System for SCWR Core Steady-State Analysis

Adhesion of Oxides Grown in Supercritical Water on Selected Austenitic and Ferritic/Martensitic Alloys

A Review of the Turbine Cooling Fraction for Very High Turbine Entry Temperature Helium Gas Turbine Cycles for Generation IV Reactor Power Plants

Results of Air Barbotage Experiments Simulating Two-Phase Flow in a CANDU End Shield During In-Vessel Retention

Technical Brief

Current Severe Accident Research and Development Topics for Wolsong PHWR Safety in Korea

Thermal-Hydraulic Model and Program Development for Helium-Cooled Accelerator-Driven System

Book Review

Handbook of Generation-IV Nuclear Reactors

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