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

IAEA TECDOC No. 1733

Subject Classification: 0703-Reactor technology

English IAEA-TECDOC-1733; (ISBN:978-92-0-100314-0); 392 pp.; € 18.00; Date Published: 2013

Download PDF (28.74 MB)

The IAEA organizes International Collaborative Standard Problems (ICSPs) to facilitate the development and validation of computer codes for design and safety analysis of nuclear power plants. The implementation of an ICSP usually includes an experimental investigation of interesting phenomena and simulation of the experiment with computer codes. This publication presents the outcome of an ICSP assessing the capability of system thermohydraulic computer codes for integral reactor system design and safety analysis. It took place at the experimental facility at the Oregon State University in the United States of America. The publication details the ICSP tests and results from all participants, provides a description of the computer codes used and the models developed, and includes some discussion by each participant concerning the results that were achieved during blind and open phases of the calculation for their individual models. The publication concludes with lessons learned and recommendations for the future.

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

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

Nuclear Reactor Technology Assessment for Near Term Deployment ...

Read more