An OECD/NEA sub-group on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM) has been formed under the NSC/WPRS/EGUAM to check the use of best-estimate codes and data. This work comes from the desire to design reactors with improved safety performance while preserving a sustainable source of energy at a rather low cost.
There is a strong incentive to design reactors with improved safety performance while preserving a sustainable source of energy at a rather low cost. The Generation IV International Forum (GIF) has defined the key research goals for advanced Sodium-cooled Fast Reactors (SFR):
Sodium-cooled Fast Reactors offer the most promising type of reactors to achieve such Generation IV goals at a reasonable time scale given the experience accumulated over the years. However, it is recognized that new regulations and safety rules as they exist worldwide are requiring improved safety performance. In particular, one of the foremost GIF objectives is to design cores that can passively avoid core damage when the control rods fail to scram in response to postulated accident initiators (e.g., inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core.
Recently, the International Atomic Energy Agency (IAEA) produced guidance on the use of deterministic safety analysis (DSA) for the design and licensing of nuclear power plants (NPPs): ‘‘Deterministic Safety Analysis for Nuclear Power Plants Specific Safety Guide’’. Since the early days of civil nuclear power, the conservative approach has been used and is still widely used today. However, the desire to utilize current understanding of important phenomena and to maximize the economic potential of NPPs without compromising their safety has led many countries to use best-estimate codes and data together with an evaluation of the uncertainties.
The group benefits from the results of a previous Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force work which demonstrated that for the benchmark cores under study the major source of bias between participants is coming from nuclear data.
Doppler and Void coefficients were calculated as well as some important dynamic characteristics of the core. Missing in the benchmark were feedback coefficients associated with thermal expansions and hence transient studies were not performed.
The UAM-SFR working group will have to define the grace time or the margin to melting available in the different identified accidental scenarios, have to apply the Best Estimate Plus Uncertainty (BEPU) methodology, and possibly recommend some changes to the design so that it meets some safety concerns.
The work is progressive so as to avoid possible compensating errors. Two SFR cores are being studied: a large 3600MWth oxide core and a medium 1000MWth metallic core. In order to assess tools being used for studying these cores, various sub-exercises have been set up for what concerns neutronics with cell, sub-assembly, super-cell and core benchmarks under steady state conditions either at BOL conditions or at EOEC depending on the benchmark. A sub-assembly depletion benchmark is being set up before going into full-core calculations with depletion.
Since the objective is to define the grace period or the margin to melting available in the different accident scenarios and this within uncertainty margins, uncertainties of different origins (methods, neutronics, thermal-hydraulic, fuel behavior) once identified and evaluated will be propagated through. At first two simple Unprotected Transients over Power (UTOP) and Loss of Flow (ULOF) are proposed because they allow useful insights without need to model the secondary loop and the primary vessel (negligible impact).
Another benchmark on control rod withdrawal will challenge tools on a particularly difficult asymmetrical transient.
In order to ensure validity to these exercises, the sub-group incorporates some experimental validations on neutronics, thermal hydraulics, fuels and systems. This will be done with neutronic experiments from ICSBEP & IRPhEP, SEFOR, thermal hydraulic experiments from THORS and system experiments with the SUPER-PHENIX start-up transient programme.
Last reviewed: 05 April 2018Top