Mechanical Design of Prototype Gen-IV Sodium Cooled Fast Reactor

The conceptual design of structure, system, and component for PGSFR NSSS and auxiliary system are carried out, and structural design requirements are established. The computer program for structural integrity evaluation of high temperature structure is developed. And, the test requirements and test facilities to obtain the material constants used in inelastic constitutive equation for SFR high temperature structural material are established. Preliminary structure sizing and structural evaluation for structure, system, and component for PGSFR NSSS and auxiliary system are preliminarily performed. The test specimens made of Type SS316H are prepared and preliminary cyclic test in high temperature are conducted to research the material behavior in high temperature condition. The stress analyses and structural integrity evaluation for steady state condition of reactor and auxiliary systems are performed by using the ASME BPV Division 5 Code. The leak before break(LBB) evaluation procedure for IHTS hot leg piping is developed and the stability in view of fracture mechanics of IHTS hot leg piping with through wal defect are evaluated. The cyclic hardening and ratchetting features, DSA corelation are investigated through the material tests with Type SS316 specimens. The arrangement requirement and design of the reactor NSSS are improved in accordance with the design change of reactor structure, system, and components. The functional test and drop performance tests for control rod drive-line are performed. And, In-service inspection sensors for reactor vessel and steam generator, and the ranging sensors for reactor internal are designed and performance tests are completed.