Enhanced Simmer Neutronics Tool for Studying Fast Reactor Distorted Core Configurations

International audienceCore disruptive accidents in fast reactors need to bemonitored carefully since they may lead to possible criticalityconfigurations. However, the worst-case scenariomay have small probability occurrences, but the proof ofit requires multidisciplinary studies. Even with the upgradein computer performance, calculations would requireseveral months on several parallel computers.Accurate calculations with short running times are thusrequired. Updating the neutronics module of SIMMERset up in the 1970s was therefore carried out with thehelp of routines able to handle probability tables forgenerating broad group libraries. The use of such librariestogether with new SIMMER options is now able toproduce reliable results in all sorts of situations whilemaintaining reduced calculation times.Indeed, until now, neutronics calculations from SIMMERgave results quite far from ERANOS ones (differencesin reactivity larger than 1.5 $). The discrepancieswere mainly due to the libraries used. As a consequence,in 2000, an ERANOS module (BISIM) was created togenerate SIMMER nuclear data libraries ( for both crosssections and self-shielding factors) from the ERANOSnuclear data file, thereby reducing the major source ofinconsistencies. Other improvements were added by theJapan Atomic Energy Agency, on the way of calculatingthe transport cross section and on the library group schemeso as to better calculate the k-effective within a reasonabletime frame, but also at the Commissariat à l’EnergieAtomique et aux Energies Alternatives on the b-effectivecalculation. A new option (using the Keepin data) wasimplemented in 2010 in SIMMER.Once all these optimizations were carried out, a comparisonbetween the SIMMER (III for two dimensionsand IV for three dimensions) and ERANOS results wasperformed for a series of disruptive and representativeconfigurations. While the computation time has notchanged significantly, the differences on k-effective betweenERANOS reference route results and SIMMER 16energy-group calculations were drastically reduced by;0.8$

Challenges Associated with the Mitigation of Unprotected Loss of Flow in Sodium-Cooled Fast Reactor Cores

International audienceReactivity effects associated with the mitigation ofunprotected loss of flow in sodium fast reactors are beingstudied to find ways to reduce the potential release ofmechanical energy.The studies performed with ERANOS illustrate theimportance of cladding removal as well as radial leakagechanges during the core slumpdown. Possible arrangementsand dispositions are envisioned that wouldavoid recriticality and hence the possibility of going intosevere power excursions.Challenges to be faced by safety studies that wouldascertain that no cliff-edge effects occur are then listed

Status of severe accident studies at the end of the conceptual design of astrid feedback on mitigation features

International audienceThe ASTRID reactor developed by the CEA with its industrial partners, will be used for demonstration of the safety and operability, at the industrial scale, of sodium fast reactors of the 4th generation. Among the goals assigned to ASTRID, one is to improve the safety and the reliability of such reactor (compared to previous built sodium-cooled fast reactors). Regarding the innovations promoted in the ASTRID design, a low sodium void worth core concept (CFV core) has been developed. By means of various design provisions enhancing the neutron leakage in case of sodium draining, the overall sodium void effect of the ASTRID core is near zero and could even be negative. Additionally, mitigation devices should be implemented into the core in order to limit the calorific energy released in the fuel during the secondary phase of the severe accident. This paper deals with a synthesis of severe accident studies performed during the second period of the pre-conceptual design stage of the ASTRID project (2013-2015). The main insights of the studies in term of mitigation strategy and of mitigation device design are highlighted in the paper. The CFV core transient behavior has been investigated in case of generalized core melting situations initiated by postulated reactivity insertion ramps (UTOP) and unprotected loss of flow (ULOF). In case of UTOP transients, the mechanical energy released by molten fuel vapor expansion does not exceed several tenths of megajoule. Simulated ULOF transients do not lead to energetic power excursions thanks to the mitigation provisions and to the core design. Regarding ULOF transients, early boiling phase leads to core power decrease and the primary phase of the accident is not governed by a power excursion. The paper deals with the approach and the presentation of preliminary findings regarding mitigation provisions. Those provisions are investigated by considering a postulated core degraded state representative of the end of the transition phase. The possible scenario evolutions from this degraded state provide the following parameters mass and temperature of molten materials, mass and flow rates of materials relocated on the core catcher and possible ejected material mass above the core Those parameters are used for the determination of approximate loadings for the primary vessel and for the core catcher

Description of initial reference design and identification of safety aspects

A reference design of the fluid fuel circuit and the emergency draining system has to be defined and optimized in the frame of the SAMOFAR project, together with the interaction phenomena between the reactor and the chemical plant. The present deliverable is divided in two main parts dedicated to the fuel salt circuit and the draining system. Each part comprises three sections: - a first section named ‘Feedback from EVOL and previous studies: list of identified constraints’ explains the constraints leading to some design options on the basis of past studies (typically from the EVOL project and some previous MSFR parametric studies); - a second section named ‘Proposition of initial reference design’ proposes a preliminary design, based on the constraints identified, to be used as a basis for the optimization studies foreseen in the SAMOFAR project. This design is not fixed: it will be optimized and will evolve all along the project as will this deliverable itself; - a third section named ‘Developments and studies in the frame of SAMOFAR’ in which all the partners involved in the WP1 will list the studies that are undertaken during the project. Deliverable 1.1 presents the initial reference design and operation procedures of the MSFR proposed by CNRS and the other partners of WP1 at the beginning of the project. The fuel circuit is enclosed in an internal vessel which serves as the container for the fuel salt, with 16 cooling sectors arranged circumferentially in the vessel, inserted from the top. This design results from previous physical and safety preliminary studies and aims at minimizing the fuel leakage occurrences and optimizing the use of the liquid salt both as fuel and coolant. The proposed design for the emergency draining system is based on the same idea, with a vessel containing the drained fuel salt with cooling rods to control the salt temperature on the short to long term. Such rods contain an inert salt used for its thermal inertia. The draining tank geometry has been defined in order to ensure a subcritical configuration in any situation. During the project, this description will be continuously updated by including all improvements and recommendations from the other work packages. This deliverable will thus be a living document, to follow this evolution all along the project

Multi-physics tools development

For the assessment of the safety of the MSFR during various transient scenarios, multi-physics tools are necessary to capture the appropriate physics of the reactor system. Classical codes used in reactor physics cannot be used as they do not allow for the key features of the MSFR. The special features of the MSFR include the movement of precursors with the moving fuel, the strong coupling between the neutronics and the thermal-hydraulics due to the use of liquid fuel, the internal heat generation and the shape of the core having no fuel pins as a repeated structure. These features cause a variety of phenomena occurring during transients that are particular to the MSFR. Dedicated tools have been developed before and during this project specifically for this purpose. The present document represents a basic description of the tools used in the SAMOFAR project. The capabilities of these codes include detailed computational fluid dynamics analysis, (2D, 3D, geometrically flexible) detailed neutronics (diffusion, transport), and incorporation of complex physics such as gas bubbling, melting and solidification. Specifically the code systems of CNRS, KIT/EdF, TUD, PoliMi and PSI are described

System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions

[EN] This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.The authors acknowledge the European Commission for funding the ESNII+ project in its 7th Framework Program (grant FP7605172).Bubelis, E.; Tosello, A.; Pfrang, W.; Schikorr, M.; Mikityuk, K.; Panadero, A.; Martorell Alsina, SS.... (2017). System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions. Nuclear Engineering and Design. 320:325-345. doi:10.1016/j.nucengdes.2017.06.015S32534532