[en] Conclusion: The analysis of SFR safety related operational experience gained in Russia enables us to make a conclusion about industrial application of this reactor technology and, in particular, sodium coolant technology, and it also demonstrates the possibility of ensuring reliable and safe operation of sodium cooled fast reactors on the example of BN-600 power unit operation.

[en] Summary: • KAERI is concentrating on the development of key technologies for the implementation of safety design concept: - Validation of Passive function of DHRS; - MA bearing metallic fuel. • The licensing of demonstration SFR will be pursued in the current regulatory framework on the basis of deterministic approach. The safety design will be supported by probabilistic approach. • A system analysis code for SFR system has been developed. This code will be validated further with available test data worldwide. • An evaluation methodology for DBEs is under development which aims to perform the safety analysis of demonstration SFR.

[en] Summary of UTOP Analysis: • BOL (conservative): Fuel melts – With in-pin fuel motion, power stabilises at 111% nominal. • BOEC (conservative): Fuel melts – With in-pin fuel motion, reactor power reduces to 44% nominal. • TOP event will not result in CDA with in pin fuel motion reactivity feed back is present.

No abstract available

[en] Conclusions: Russian regulatory documents regulating the issues related to safety of reactor facilities and NPPs, including sodium cooled fast reactors, are developed with regard for the gained operation experience and IAEA recommendations, and they meet the up-to-date level of safety requirements.

[en] The main feature of the next stage of SFR development is simultaneous achievement of the goals of improvement of economical and safety characteristics. In our vision the main path for this is max use of inherent safety characteristics, as well as safety systems based on passive operating principles. This approach was used in optimization of characteristics of BN-1200 in the stage of its conceptual design. The main characteristics of the BN-1200 reactor taken for design development are presented.

[en] Challenge: • Extremely valuable knowledge (Phénix and less from Creys-Malville): - fuel and core; - sodium technology and steel which can contain it; -components and operation of a SFR. • Progress can still be made: - to raise safety level (= 3rdgeneration); - performances (fuel, reactor, energy conversion). • Challenge: design futures SFR / requirements ~ PWR (availability, safety, cost, financial risk) + sustainability. • Operating feedback experience → improvements / (Phénix, CreysMalville, EFR) and innovations (R&D CEA, AREVA and EDF). • Passing on the knowledge and skills acquired in the past to men and women who will develop, design, construct and operate the future SFR.

[en] Concluding Remarks: – Concept of DiD shall be applied to the safety design of advanced SFR. – Safety level can be further improved especially by enhancing prevention and mitigation features with more emphasis on passive safety features. – Through prevention, detection, and control of accident, CDA shall be excluded from DBE. – Toward a commercialization of SFR, not only prevention but also mitigation of typical severe core damage are necessary to be enhanced. – In particular, the safety approach with elimination of severe recriticality is highly desirable and will contribute to establish public acceptance of the SFR.

[en] U.S. SFR Safety Experience: • The US reactor development program has demonstrated that liquid sodium metal cooling in an SFR contributes to excellent safety performance: – Excellent heat removal and heat transport characteristics; – Natural circulation decay heat removal; – Passive reactor power reduction in beyond-design-basis accidents. • Core melting accidents have shown that safe shutdown of an SFR is possible without severe consequences: – Metallic fuel is compatible with liquid sodium; – Accident progression can be safely terminated; • Reactors were refueled and operated after the accidents. • ANS has started development of ANS Standard 54.1, “Nuclear Safety Criteria and Design Process for SodiumCooled Reactor Nuclear Power Plants”: – An update and revision of the ANS 54.1 Standard.

[en] Summary: • Innovative experimental techniques for comprehensively addressing the safety issues related to seismic integrity, decay heat removal requirements, severe accident scenarios, post accident heat removal, sodium fire mitigation, development of innovative sodium fire extinguishers, sodium aerosol dispersion studies are the priorities. • Experiments are planned to study both science and technology aspects concurrently. • High emphasis on involvement of younger scientists and engineers and students from academic institutions. • National and international collaborations are encouraged to share the facilities and joint investigations on the basis of challenges and mutual interest.