Conference Agenda

The Generation IV (Gen-IV) International Forum is a framework for international co-operation and collaboration in research and development for the next generation nuclear energy systems. Within the sodium-cooled Fast Reactor (SFR) system arrangement, there are four projects; System Integration Assessment (SIA), Advanced Fuel (AF), Component Design & BOP (CD&BOP), and Safety & Operation (SO). The SFR SO project addresses the areas of safety technology and reactor operation technology developments. It aims for (1) analyses and experiments that support establishment of the safety approaches and validate the performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFRs. The tasks in the SO area are categorized by the following three work packages (WP). WP-SO-1 "Methods, Models and Codes" is for the development of tools used to evaluate the safety. WP-SO-2 "Experimental Programs and Operational Experience" is for the operation, maintenance and testing experiences in experimenta facilities and SFRs. WP-SO-3 "Studies of Innovative Design and Safety Systems" is for safety technologies of Gen-IV reactors such as active and passive safety systems and other specific design features. This paper includes recent activities of member countries and organizations within the SFR SO project.

Liquid metal cooled fast reactors (LMFR) use liquid metal as the primary coolant of the reactor core. First demonstrated in the 1950s, they were never fully deployed compared with the light water reactor technologies. However, the early 2000s saw a resurgence of interest, particularly in Sodium Fast Reactors (SFR) and Lead Fast Reactors (LFR) as Generation IV designs, due to their potential to significantly reduce the amount and toxicity of nuclear waste in a closed fuel cycle. This investigation is part of Tractebel’s effort to evaluate new tools for both SFR and LFR modeling. CTF, a subchannel thermal-hydraulic code for Light Water Reactor applications, has been used at Tractebel since 2015. It incorporates state-of-the-art models, correlations, and methods for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) modeling. Recently, new features have been developed in CTF to model SFR and LFR reactor cores. Given Tractebel’s expertise with the code, CTF is a promising candidate for developing LMFR modeling capabilities. This study focuses on validating CTF models for SFRs using data from two test facilities: TAMU 61-pin isothermal tests (Texas A&M University and SEFOR (Consortium of Southwest Atomic Energy Associates, Karlsrühe Laboratory, Euratom, General Electric). This data is obtained through participation in OECD/NEA benchmarks LMFR T/H and SFR-UAM. Key models of interest include friction factor correlations, turbulent mixing, and Nusselt correlations for heat transfer in liquid metals. This paper presents the preliminary outcomes of these investigations.

The use of Separate Effect Tests (SET) and Integral Effect Tests (IET) is a common practice in support of Sodium Fast Reactors (SFR) designs. These tests are built to analyse physical phenomena and their measured data can serve as validation database for simulation codes. In the framework of the Franco-Japanese collaboration on Research and Development for SFR thermal hydraulics, transient tests were performed in the IET named PLANDTL2 test facility in Japan. This IET’s instrumented test section is composed of an electrically heated core and a hot pool with a Dipped Heat Exchanger (DHX). The Intermediate Heat Exchanger (IHX) and the Electro-Magnetic Pump (EMP) are located in a deported primary loop. Studied transients consist in transition from forced convection to natural convection, in the pool and in the primary circuit, under various decay heat removal operations using the DHX.

It was observed that in the long term, a cliff effect occurs, meaning that the apparent steady natural convection is perturbed if a threshold is reached. Instabilities and flow rate oscillations from positive to negative values in the primary loop are observed after a period of smooth natural circulation. The unstable behaviour results from the competition between IHX and DHX cooling, the latter leading to an increase in thermal stratification in the hot pool. This paper aims to analyse this phenomenon, bring a comprehensive criterion for the onset of instable behaviours and give some general guidelines to avoid such effects for accidental transient management.

The evaluation of reactivity feedback in sodium-cooled fast reactors owing to core deformation during the power increase needs a comprehensive understanding of the interactions among neutronics, thermal-hydraulics, and structural mechanics in the core. However, conventional reactor core design evaluation methods often lack accuracy due to oversimplifications in modeling. To deal with this, JAEA has developed an evaluation method that couples several analysis codes implementing detailed models of these phenomena. In the neutronics calculation, core deformation reactivity is based on the first-order perturbation theory and GEM reactivity is determined using a function of core flow rate based on Monte Carlo calculation results of the reactivity. Other reactivities due to the Doppler effect, density reductions of fuel, cladding, coolant, and wrapper tube, and the axial thermal expansion of control rods are calculated by multiplying their temperature increases by their respective reactivity coefficients. The thermal-hydraulics inside fuel assemblies and inter-wrapper regions between neighboring assemblies are modeled as flow networks. The deformation of assemblies is modeled by FEM beam elements. These codes are coupled and synchronized depending on the time scale of each physical phenomenon’s variation to effectively simulate core transients. In this study, the evaluation method was validated by FFTF LOFWOS Test #13 analysis. Comparison between the analyses and test results revealed that the analyses had uncertainties concerning the inclination of the assembly on the core support plate, pad stiffness, and the temperature flattening effect of inter-wrapper flow, which influence deformation reactivity. These uncertainties need further investigation for accurate analysis.

Fuel failures during normal operations and transient scenarios involve rather high uncertainty due to various factors, such as defects in manufacturing, operating conditions, cladding dose, fuel cladding chemical interaction, fuel cladding mechanical interaction, plenum pressurization and cladding thermal creep. While the failure of a single fuel pin poses minimal risk by itself, the potential for pin-to-pin failure propagation (or decrease in failure margin of the neighboring pins) may exist within a fuel bundle. Numerous studies have explored the potential for cascading pin failure, but only in-pile tests with live fuel have created the sudden rupture and rapid fission gas release resulting from cladding failure. A unique burst technique has been developed at Oregon State University to replicate the depressurization of fission gas during fuel failure. This was achieved by laser-welding thin stainless-steel film, laser-etched to create defects, onto partially voided surrogate fuel pins. These pre-defected surrogate fuel pins were then inserted into a 19-pin quartz stainless-steel surrogate fuel bundle, that allowed for the visualization of gas release within typical liquid metal fast reactor (LMFR) geometry and dimensions using a matching index of refraction technique. Failures within the bundle were tested at various burst pressures, coolant flow rates, and breach sizes to characterize the gas release component of fuel failure in a controlled separate effects test. The data from these experiments will inform the design and experimental parameters for future tests in sodium flow loop and contribute to validation of models for unprotected transient events, which currently lack corresponding experimental data.

Japan Atomic Energy Agency has started developing analytical technologies to simulate disrupted core of metal-fuel sodium-cooled fast reactors. This paper reports the development of physical models implemented into the SIMMER code for metal-fuel fast reactor simulations and results of in-pile experiment analysis as a code validation.

To apply the SIMMER code to the metal-fuel fast reactor, priority is given to implementation of two feasible models to represent phenomena specific to a fuel damage accident in the reactor. One of the feasible models is a eutectic formation with a contact of fuel and steel, and the other is an in-pin behavior of molten fuel slug with low melting point. The eutectic formation is treated both in the pin and after pin failure. Furthermore, a cladding failure due to a cladding thinning by the eutectic formation and a molten fuel discharge through the cladding failure can be represented by combining the two models.

To validate the implemented models, this study performed an analysis of the TREAT experiment. The calculation shows that the eutectic formation thins cladding at a top of fuel slug and the cladding failure occurs. The molten fuel in the pin is discharged from the cladding failure to a coolant flow channel. The new models improve the pin failure and a formation of blockage by broken pin and a eutectic material which was observed when not using the models.