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The 7th international workshop on the integrity of nuclear components

The International journal of pressure vessels and piping, 2010, Vol.87 (1), p.1-1

Park, Youn-Won

2010

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An investigation of the degradation characteristics for casting stainless steel, CF8M, under high temperatures

Nuclear engineering and design, 2000-06, Vol.198 (3), p.227-240

2000

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Integrity Evaluation System of CANDU Reactor Pressure Tube

Journal of mechanical science and technology, 2003-07, Vol.17 (7), p.947-957

2003

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Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

International journal of pressure vessels and piping, 2010, Vol.87 (1), p.33-40

2010

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Sensitivity analysis of input parameters for pressure tube integrity evaluation

Nuclear engineering and design, 2005-08, Vol.235 (17), p.1909-1918

2005

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Optimum local failure model of steam generator tubes with multiple axial through-wall cracks

Nuclear engineering and design, 2005-08, Vol.235 (17), p.2099-2108

2005

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Investigation on constraint effect of reactor pressure vessel under pressurized thermal shock

Nuclear engineering and design, 2003-02, Vol.219 (3), p.197-206

2003

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Effects of thermal aging on the low cycle fatigue behavior of austenitic–ferritic duplex cast stainless steel

Nuclear engineering and design, 2001-05, Vol.206 (1), p.35-44

2001

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Development of an Integrity Evaluation System for Nuclear Power Plants

JSME International Journal Series C Mechanical Systems, Machine Elements and Manufacturing, 2003, Vol.46(4), pp.1464-1472

2003

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Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

Nuclear engineering and design, 2001-03, Vol.205 (1), p.1-11

2001

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Round robin analysis of pressurized thermal shock for reactor pressure vessel

Nuclear engineering and design, 2003-12, Vol.226 (2), p.141-154

2003

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Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

Nuclear engineering and design, 2002-05, Vol.214 (1), p.13-23

2002

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A study on technique to estimate impact location of loose part using Wigner-Ville distribution

Progress in nuclear energy (New series), 2003, Vol.43 (1-4), p.261-266

2003

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Optimum Global Failure Prediction Model of Steam Generator Tube With Two Parallel Axial Through-Wall Cracks

Journal of pressure vessel technology, 2005-05, Vol.127 (2), p.123-128

2005

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Pressurized thermal shock analyses of a reactor pressure vessel using critical crack depth diagrams

The International journal of pressure vessels and piping, 1999-10, Vol.76 (12), p.813-823

1999

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Leak-before-break assessment of CANDU pressure tube considering leak detection capability

Nuclear engineering and design, 1999-07, Vol.191 (2), p.205-216

1999

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Seventh International Workshop on the Integrity of Nuclear Components

International journal of pressure vessels and piping, 2010, Vol.87 (1)

PARK, Youn-Won

2010

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A Probabilistic Integrity Assessment of Flaw in Zirconium Alloy Pressure Tube Considering Delayed Hydride Cracking

International journal of modern physics. B, Condensed matter physics, statistical physics, applied physics, 2003-04, Vol.17 (8n09), p.1587-1593

2003

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Development of Web-Based Fatigue Life Evaluation System for Reactor Pressure Vessel

Solid state phenomena, 2007, Vol.120, p.25-30

2007

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Development and validation of a smartphone addiction scale (SAS)

PloS one, 2013-02, Vol.8 (2), p.e56936-e56936

2013

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Statistical Reliability Assessment of UT Round-Robin Test Data for Piping Welds

Key engineering materials, 2004-01, Vol.270-273, p.1743-1748

2004

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Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Reactor Pressure Vessel

Key engineering materials, 2004-01, Vol.270-273, p.2244-2249

2004

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Integrity Evaluation System of CANDU Reactor Pressure Tube

KSME international journal, 2003, Vol.17 (7), p.947-957

2003

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Thermal Aging and Low Cycle Fatigue Characteristics of CF8M in a Nuclear Reactor Coolant System

Key engineering materials, 2000-01, Vol.183-187, p.975-980

2000

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The Effect of Aging Temperatures and Time on Mechanical Properties of CF8M

Key engineering materials, 2000-01, Vol.183-187, p.577-582

2000

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The 7th international workshop on the integrity of nuclear components

An investigation of the degradation characteristics for casting stainless steel, CF8M, under high temperatures

Integrity Evaluation System of CANDU Reactor Pressure Tube

Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

Sensitivity analysis of input parameters for pressure tube integrity evaluation

Optimum local failure model of steam generator tubes with multiple axial through-wall cracks

Investigation on constraint effect of reactor pressure vessel under pressurized thermal shock

Effects of thermal aging on the low cycle fatigue behavior of austenitic–ferritic duplex cast stainless steel

Development of an Integrity Evaluation System for Nuclear Power Plants

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

Round robin analysis of pressurized thermal shock for reactor pressure vessel

Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

A study on technique to estimate impact location of loose part using Wigner-Ville distribution

Optimum Global Failure Prediction Model of Steam Generator Tube With Two Parallel Axial Through-Wall Cracks

Pressurized thermal shock analyses of a reactor pressure vessel using critical crack depth diagrams

Leak-before-break assessment of CANDU pressure tube considering leak detection capability

Seventh International Workshop on the Integrity of Nuclear Components

A Probabilistic Integrity Assessment of Flaw in Zirconium Alloy Pressure Tube Considering Delayed Hydride Cracking

Development of Web-Based Fatigue Life Evaluation System for Reactor Pressure Vessel

Development and validation of a smartphone addiction scale (SAS)

Statistical Reliability Assessment of UT Round-Robin Test Data for Piping Welds

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Reactor Pressure Vessel

Integrity Evaluation System of CANDU Reactor Pressure Tube

Thermal Aging and Low Cycle Fatigue Characteristics of CF8M in a Nuclear Reactor Coolant System

The Effect of Aging Temperatures and Time on Mechanical Properties of CF8M

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