Proceedings of the 2023 Water Reactor Fuel Performance Meeting : WRFPM2023, July 17–21, Xi’an, China / edited by Jianqiao Liu, Yongjun Jiao.

Format:

Book

Author/Creator:

Liu, Jianqiao.

Contributor:

Jiao, Yongjun.

Series:

Springer Proceedings in Physics, 1867-4941 ; 299

Language:

English

Subjects (All):

Nuclear physics.

Electric power-plants.

Thermodynamics.

Security systems.

Nuclear engineering.

Nuclear Physics.

Power Stations.

Security Science and Technology.

Nuclear Energy.

Local Subjects:

Nuclear Physics.

Power Stations.

Thermodynamics.

Security Science and Technology.

Nuclear Energy.

Physical Description:

1 online resource (384 pages)

Edition:

1st ed. 2024.

Place of Publication:

Singapore : Springer Nature Singapore : Imprint: Springer, 2024.

Summary:

The Water Reactor Fuel Performance Meeting (WRFPM) held in Asia has merged with TopFuel in Europe and LWR Fuel Performance in the United States to form the globally most influential conference in the field of nuclear fuel research. WRFPM2023 is organized by Chinese Nuclear Society (CNS) in cooperation with the Atomic Energy Society of Japan (AESJ), Korean Nuclear Society (KNS), European Nuclear Society (ENS), American Nuclear Society (ANS), the Interna-tional Atomic Energy Agency (IAEA) with the support from China Nuclear Energy In¬dustry Corporation (CNEIC) and TVEL. Conference Topics: 1. Advances in water reactor fuel technology and testing 2. Operation and experience 3. Transient and off-normal fuel behaviour and safety related issues 4. Fuel cycle, used fuel storage and transportation 5. Innovative fuel and related issues 6. Fuel modelling, analysis and methodology.

Contents:

Intro

Contents

Contributors

Numerical Investigation on the Effect of Fuel Pulvers on Axial Fuel Relocation

1 Introduction

2 Model Setup

2.1 Pellet Model

2.2 Cladding Model

2.3 Model Setting

3 Results and Discussion

3.1 Fuel Stack Reduction

3.2 Fuel Mass Fraction

3.3 Fuel Filling Ratio

4 Conclusions

References

On the Creep Collapse of the Cladding Considering the Irradiation Growth Effect

2 Methods

2.1 Constitutive Model of the Cladding

2.2 Finite Element Model

2.3 Loading History

2.4 Creep Collapse Failure Criterion

4 Conclusion

Practical Development of Accident Tolerant Fecral-Ods Fuel Claddings for BWRs in Japan

2 Experiments

2.1 Fatigue Test

2.2 Micro-Hardness and Tensile Tests of Irradiated Specimens

3 Results and Discussions

3.1 Fatigue Test

3.2 Micro-Hardness and Tensile Tests of Irradiated Specimens

Experimental Study on Pool Boiling Heat Transfer Characteristics of SiC Cladding Under Atmospheric Pressure

2 Experimental Apparatus and Procedure

2.1 Boiling Apparatus

2.2 Experimental Procedure

2.3 Data Acquisition

2.4 Uncertainty Analysis

3 Observations and Results

3.1 Boiling Heat Transfer Curve

3.2 Bubble Growth Curve, Departure Diameter, and Departure Frequency

3.3 Heat Transfer Model

4 Summary and Conclusions

Preliminary Development of a Simulation Capability for Zircaloy Clad Ballooning in LOCA

2 Material Properties and Behavior Models

3 Implicit Integration Algorithm of Creep

4 Simulations and Discussion

5 Conclusion

Updates to the IAEA Guide on Fuel Reliability and Performance

1 Introduction.

2 In-Reactor Fuel Performance Issues and Mitigation Measures

3 Fuel Design Changes to Improve Reliability and Performance

3.1 Fuel Design Changes, Verification and Validation

3.2 Quantification and Management of Fuel Design and Operating Margins

3.3 Good Practices for Fuel Design Change Verification

3.4 Qualify Assurance, Qualification and Licensing of Fuel Design

4 Good Practices to Improve Fuel Reliability During Operation

5 Conclusions

Parametric Study of Phenomena Influencing Secondary Hydriding During LOCA Transients

2 Materials and Methods

2.1 Materials

2.2 Sample Geometry and Constitutive Materials

2.3 High Temperature (HT) Oxidation

2.4 Post-Test Characterization

2.5 Test Matrix

3 Experimental Results and Discussion

3.1 Metallographic Analysis

3.2 Influence of Oxidation Duration on Secondary Hydriding

3.3 EPMA &amp

µ-LIBS Micro-Analysis

Radial Hydride Precipitation in Fuel Cladding During Back-End Cooling Transient Under Decreasing Pressure

2 Modeling the Influence of Decreasing Stress on Radial Hydride Precipitation

2.1 Constant Stress Model

2.2 Decreasing Stress Model

3 Pressurization Tests Performed on Zircaloy-4 Tubes

3.1 Tested Material

3.2 Testing Conditions

3.3 Post-Test Examinations

3.4 Test Matrix and Results

4 Test Analysis and Interpretation

Analysis and Assessment of BEO-Doped Fuel with Fuel Rod Performance Code Jasmine

2 Theoretical Model

2.1 Non-Irradiation Model

2.2 Irradiation Model

3 Verification and Evaluation

3.1 Experimental Verification

3.2 Performance Evaluation

Reassessment of FRAPTRAN's Cladding Failure Criteria in LOCA Within R2CA H2020 Project.

1 Introduction

2 FRAPTRAN Description

2.1 Cladding Mechanical Deformation Model

2.2 LOCA Failure Criteria

3 Modifications in the Code

3.1 High-Temperature Creep Deformation

3.2 Phase Transformation

3.3 IRSN New Burst Criteria

4 Results of the Validation Cases

5 Full Core Analysis Results

6 Conclusions

Numerical Calculation on Thermal Expansion Of UO2 - 3 Vol% Mo Microplate Pellet

2 Numerical Calculation Model

3 Results

Preliminary Study on the Torque Coefficient and Filtering Coefficient for Threaded Fasteners in Fuel Assembly

2 Mechanical Model

2.1 The Relationship Between the Applied Torque and Preload of Screw

2.2 The Redistribution of External Load Between Screw and Members

3 Analysis of Threaded Connection in Fuel Assembly

3.1 The Structural Description of PWR Fuel Connection

3.2 Torque Coefficient Analysis

3.3 Filtering Coefficient Analysis

ThN's Lattice-Assisted Thermal Conductivity Revisited

2 Results and Discussion

2.1 Phonon Density of States

2.2 Heat Capacity

2.3 Lattice Thermal Conductivity

3 Summary

How to Deal with the Threat of New Energy to the Safe Operation of Nuclear Fuel

1 Energy Situation

2 Analysis of the Situation of Nuclear Power Units Peak Load Regulation

3 Technical Measures to Deal with Nuclear Power Units Peak Load Regulation

3.1 Balance of the Amplitude and Frequency of Peak Load Regulation with Multi-reactor Management

3.2 Control the Rate of Change of Reactor Power

3.3 The Moving Speed of the Control Rod

3.4 Conclusion

4 Solutions in the Future

4.1 Steam Energy Supply

4.2 Pumped Storage Power Station

4.3 Self-built Energy Storage

4.4 Grid Construction.

References

Study on Thermal Hydraulic Characteristics of Rod-Type Fuel Irradiation Test Section

2 Numerical Models

2.1 Geometry

2.2 Model Settings

2.3 Mesh Sensitivity Analysis

2.4 Hydraulic Test Verification

3 Calculation Results and Discussion

3.1 Design Condition

3.2 Non-rated Flow Conditions

Modeling and Analyzing of Fuel with Missing Pellet Surface (MPS) Defect Based on Multiphysics Method

2 Modeling Details

2.1 Material Properties

2.2 Modeling Geometry and Mesh

2.3 Fuel Input Parameters

2.4 Modeling Approach

3.1 The Verification of This Code

3.2 The Analysis of Fuel with MPS

3.3 Comparisons of Different Depths of MPS

3.4 Comparisons of Different Fuels with MPS

AFA 3G Operating Experience

2 Customer Requests for Product Performance Evolution

2.1 Burn-Up Capacity and Longer Cycles

2.2 Thermal-Hydraulic Performance

2.3 Maneuverability

2.4 Reliability

3 AFA 3G Technology Improvements

3.1 Q12 for High Dimensional Stability

3.2 Optimized 718 Alloy

3.3 M5Framatome Cladding Material

Reference

Progress on Modelling the Thermo-Mechanical Performance of Accident-Tolerant Fuels

2 Cr-Doped UO2 Pellets

2.1 Background on Cr-Doped UO2 Properties

2.2 FRAPCON-4.0 Extension to Cr-Doped UO2 Simulation

2.3 Validation Under Reactor and Fast Power Ramp Conditions

3 FeCrAl Cladding

3.1 Background on FeCrAl Properties

3.2 FRAPTRAN Extension to FeCrAl Cladding Simulation

3.3 Verification Under LOCA Conditions

Additive Manufacturing Process Design For Thimble Plug Assembly

1.1 Study Conditions

1.2 Research Results.

1.3 Additive Structure Design

1.4 Additive Manufacturing Parts Design

1.5 SLM Process Design

1.6 Molding Size Optimization

1.7 Welding Process Design

1.8 Conclusions

Research and Application of Radioactive Control Methods on the Primary Circuit of PWR Fuel Cladding with Loss of Air Tightness

2 Abnormal Fuel Cladding Damage in a Domestic Nuclear Power Plant

3 Primary Circuit Radioactive Control During Unit State Transition

3.1 Control Power to Prevent the Expansion of Fuel Cladding Rupture

3.2 Primary Circuit Water Quality Control

4 Three Wastes and Ventilation Control During the Overhaul

5 Summary

Sensitivity Analyses of Thermal Hydraulic Parameters in ATWS by Rods Failure-Loss of Offsite Power of the Third Generation Nuclear Power Plant

1 Foreword

2 Objects and Methods

2.1 Introduction

2.2 Tools and Methodology

2.3 Main Proposes

3 Sensitivity Analysis

3.1 Thermal Power

3.2 Primary Pressure

3.3 Primary Temperature

3.4 Pressurizer Water Level

3.5 VDA Valve

3.6 Turbine Trip

4 Comprehensive Analysis

4.1 All Uncertainty Considered

4.2 Acceptance Criteria

Confirmation of the Design Characteristics of the TVS-K Design After Operation in the PWR Reactor at Ringhals-3 NPP

2 Design Parameters

3 Examinations of the TVS-K Skeleton and Top Nozzle

3.1 Skeleton Visual Inspection and Length Measurements

3.2 Skeleton Stiffness Test

3.3 Skeleton Bow Measurements

3.4 Top Nozzle Examinations

4 Non-destructive Examinations on Guide Tube, Instrumentation Tube and Spacer Grids

4.1 Visual Inspection with Photo and Video Fixing the External State of the GT and IT

4.2 Length Measurements of the GT and IT.

4.3 Measurements of Oxide Film Thickness on the Outer Surface of the GT and IT by Eddy Current Method.

Other Format:

Print version: Liu, Jianqiao Proceedings of the 2023 Water Reactor Fuel Performance Meeting

ISBN:

9789819971572

9819971578

OCLC:

1411349373