Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: PWR Vessel Internals
2007 Update

IAEA TECDOC No. 1557

Subject Classification: 0703-Reactor technology

English IAEA-TECDOC-1557; (ISBN:978-92-0-105107-3); 74 pp.; € 15.00; Date Published: 2007

Download PDF (1.55 MB)

This publication is an update of IAEA TECDOC Series No. 1119 and provides current ageing management guidance for pressurized water reactor (PWR) vessel internals. Directed towards all those involved in the operation and regulation of PWRs, this publication deals with age related degradation and ageing management of PWR vessel internals. It presents the requirements and methodologies utilized for the assessment and management of ageing of these components and provides the technical basis for ensuring that the required safety and operational margins are maintained throughout the remainder of plant life.

Download Order Form PDF
You might also like

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

Performance Assessment of Passive Gaseous Provisions (PGAP) ...

Read more