IAEA Concludes Four-year Research Project on Fast Reactor Safety Features
Block control room of BN-800's Unit 4. The BN-800 reactor is a
sodium-cooled fast breeder reactor, built at the Beloyarsk Nuclear
Power Plant. (Photo: Rosenergoatom)
Results of a four-year IAEA Coordinated Research Project
(CRP) on improving system and safety analysis of sodium cooled fast
reactors (SFRs) are now available. The recently published IAEA TECDOC, Benchmark Analysis of EBR-II Shutdown Heat Removal Tests, is a result of work done by experts from 19 organisations in 11 Member States.
The SFR technology development traces its beginnings to the Experimental Breeder Reactor I (EBR-I) at Argonne National Laboratory in the United States, which first generated useable amounts of electricity in December 1951. The succeeding decades saw construction and operation of experimental and prototype fast reactor facilities in the US, the Soviet Union, the United Kingdom, France, Germany, Japan and India. At present, a new generation of fast reactors have been introduced with the BN-800 in Russia operating since 2015, the China Experimental Fast Reactor (CEFR) operating in China since 2011 and the Prototype Fast Breeder Reactor (PFBR) under construction in India since 2004.
SFRs represent a significant advance over established and evolutionary light water reactor designs in terms of efficient resource utilization, passive safety1, reliability and management of high-level waste.
The SFR technology development traces its beginnings to the Experimental Breeder Reactor I (EBR-I) at Argonne National Laboratory in the United States, which first generated useable amounts of electricity in December 1951. The succeeding decades saw construction and operation of experimental and prototype fast reactor facilities in the US, the Soviet Union, the United Kingdom, France, Germany, Japan and India. At present, a new generation of fast reactors have been introduced with the BN-800 in Russia operating since 2015, the China Experimental Fast Reactor (CEFR) operating in China since 2011 and the Prototype Fast Breeder Reactor (PFBR) under construction in India since 2004.
SFRs represent a significant advance over established and evolutionary light water reactor designs in terms of efficient resource utilization, passive safety1, reliability and management of high-level waste.
The
Experimental Breeder Reactor-II (EBR-II) operated at Argonne-West in
Idaho from 1964 to 1994 and was the backbone of the U.S. breeder reactor
effort, as it successfully demonstrated the concept of a liquid
metallic sodium being used as a coolant in a nuclear reactor. (Photo:
Argonne National Laboratory)
In
response to growing interest, in June 2012, the IAEA initiated a
four-year Coordinated Research Project (CRP) with the objective of
improving SFR modelling and simulation tools for safety validation,
alongside training the next generation of researchers, analysts and
designers through international benchmark exercises.
The participating organisations in this CRP demonstrated, via simulation and modelling, how SFRs would stand a severe accident with no core damage, a scenario which was experimentally tested in the 1980s in the EBR-II reactor. Several meetings held during the project have provided insight into the performance and reliability of simulation codes to be used for the system design and the safety analysis of future prototype reactors, with particular emphasis on shutdown heat removal phenomena2.
“Nothing would have been possible without the outstanding and precious work performed by Argonne National Laboratory,” said Stefano Monti, Head of the IAEA Nuclear Power Technology Development Section.
“Sometimes we underestimate the intrinsic value of technical data provided by our Member States to carry out our activities, but we have to consider that this information comes from very expensive and complex experimental campaigns. This will undoubtedly allow our Member States with an active programme on fast reactors to improve their capability to perform reliable safety analysis of innovative reactor types.”
__________
1 A safety feature that does not require manual or electronic actions to safely shut down a reactor in the event of an emergency (usually overheating resulting from a loss of coolant or loss of coolant flow). Such reactors tend to rely more on engineering of components, such that their predicted behaviour, would slow, rather than accelerate, a nuclear reaction.
2 The removal of residual heat from the reactor core after shutdown, and during and after appropriate operational states and accident conditions.
The participating organisations in this CRP demonstrated, via simulation and modelling, how SFRs would stand a severe accident with no core damage, a scenario which was experimentally tested in the 1980s in the EBR-II reactor. Several meetings held during the project have provided insight into the performance and reliability of simulation codes to be used for the system design and the safety analysis of future prototype reactors, with particular emphasis on shutdown heat removal phenomena2.
“Nothing would have been possible without the outstanding and precious work performed by Argonne National Laboratory,” said Stefano Monti, Head of the IAEA Nuclear Power Technology Development Section.
“Sometimes we underestimate the intrinsic value of technical data provided by our Member States to carry out our activities, but we have to consider that this information comes from very expensive and complex experimental campaigns. This will undoubtedly allow our Member States with an active programme on fast reactors to improve their capability to perform reliable safety analysis of innovative reactor types.”
__________
1 A safety feature that does not require manual or electronic actions to safely shut down a reactor in the event of an emergency (usually overheating resulting from a loss of coolant or loss of coolant flow). Such reactors tend to rely more on engineering of components, such that their predicted behaviour, would slow, rather than accelerate, a nuclear reaction.
2 The removal of residual heat from the reactor core after shutdown, and during and after appropriate operational states and accident conditions.
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