Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor

Author:

Hu B.X.,Wu Y.W.,Tian W.X.,Su G.H.,Qiu S.Z.

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference20 articles.

1. Thermal-hydraulic design considerations for Clinch River Breeder Reactor plant intermediate heat-exchangers;Aburomia;Mechanical Engineering,1976

2. Agrawal, A.K., et al., 1978. An Advanced Thermo Hydraulic Simulation Code for Transients in LMFBRs (SSC-L Code), BNL-NUREG-50773.

3. Primary pipe rupture accident analysis for Clinch River Breeder Reactor;Albright;Nuclear Technology,1978

4. Main Features of the BN-800 Passive Shutdown Rods, Absorber Materials, Control Rods and Designs of Shutdown Systems for Advanced Liquid Metal Fast Reactors;Alexandrov,1995

5. Cahalan, J.E., Wei, T.Y.C., 1990. Modeling developments for the SAS4A and SASSYS computer codes. In: Proceedings of the 1990 international fast reactor safety meeting.

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