Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment
Deterministic Evaluation for the Integrity of Reactor Pressure Vessel

IAEA TECDOC No. 1627

Subject Classification: 0703-Reactor technology

English IAEA-TECDOC-1627; (ISBN:978-92-0-111109-8); 228 pp.; € 15.00; Date Published: 2010

Download PDF (3.42 MB)

Pressurized thermal shock (PTS) analysis, which is a part of reactor pressure vessel (RPV) structural integrity assessment, is associated with large thermal shocks and, in some cases, with low temperature repressurization of the RPV after a certain time. At present, several different procedures and approaches are used for RPV integrity assessment. This publication provides benchmark calculations and analyses which compare effects of individual parameters on the final RPV integrity assessment. Based on this, the publication provides recommendations for best practice for their implementation in PTS procedures.

Download Order Form PDF
You might also like

Benchmark Analysis for Condition Monitoring Test Techniques of Aged Low Voltage Cables in Nuclear Power Plants ...

Read more

Instrumentation and Control Systems for Advanced Small Modular Reactors ...

Read more

Severe Accident Mitigation through Improvements in Filtered Containment Vent Systems and Containment Cooling Strategies ...

Read more

Preparing and Conducting Review Missions of Instrumentation and Control Systems in Nuclear Power Plants ...

Read more

Application of Field Programmable Gate Arrays in Instrumentation and Control Systems of Nuclear Power Plants ...

Read more

Technical Challenges in the Application and Licensing of Digital Instrumentation and Control Systems in Nuclear Power Pl ...

Read more

Treatment of Residual Sodium and Sodium Potassium from Fast Reactors ...

Read more

Performance Analysis Review of Thorium TRISO Coated Particles during Manufacture, Irradiation and Accident Condition Hea ...

Read more

Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel ...

Read more

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs) ...

Read more

Benchmark Analyses on the Control Rod Withdrawal Tests Performed during the PHÉNIX End-of-Life Experiments ...

Read more

Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors ...

Read more

Design Features and Operating Experience of Experimental Fast Reactors ...

Read more

Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear ...

Read more

Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors ...

Read more