When progressing from the International Thermonuclear Experimental Reactor (ITER) to the Demonstration Fusion Reactor (DEMO), a system for transferring plasma heat exhaust to a power conversion system is necessary for the so-called Balance of Plant (BOP). During the preconceptual phase of the EU-DEMO project, different BOP concepts were investigated in order to identify the main requirements and feasible architectures to achieve that goal in the most efficient way. This paper comprises the investigations performed during the DEMO preconceptual design phase (p-CDP) and compares the different variants. The main aspect was focused on the helium-cooled pebble bed (HCPB) breeding blanket (BB) concept. After all assessments were performed, the indirect coupled design (ICD) was chosen as the reference configuration for the DEMO HCPB BOP for further development and optimization. The ICD provides decoupling using a molten salt storage loop, which accumulates thermal power during plasma pulses that are released during dwell periods. The work is supported by simulations using design codes EBSILON and MATLAB/SIMULINK, providing the basis for the next design phase.

The paper presents a transient simulation phase of the new benchmark on a large sodium fast reactor (SFR). This phase of the benchmark is devoted to the modeling of selected operational transients performed during startup tests of the French SFR Superphénix. Six operational transients were selected for the analysis. The specifications of a simplified thermal hydraulic model equipped with point kinetics reactivity data and boundary conditions for the selected transients are given in the paper. The developed model contains a necessary thermal hydraulic description of the primary system components, assumptions to account for thermal expansion reactivity feedback from out-of-core structures, neutron kinetics parameters, power distribution, and reactivity coefficients. The neutronic input parameters were obtained with the help of the Monte Carlo code serpent during the first phase of the benchmark related to the static neutronic characterization of the core. In this study, the solution of the transient benchmark was obtained with three thermal hydraulic system codes, namely, TRACE, SIM-SFR, and ATHLET. The numerical results, compared to the available experimental data, exhibit a reasonable mutual agreement. Particular discrepancies between calculations and experiments could not be fully resolved. Therefore, a set of recommendations for achieving an improved agreement was proposed. In general, the proposed transient benchmark can be seen as an effective tool for validation and cross comparisons of system codes applied for safety analyses of SFRs, including approbation and comparison of different modeling features for thermal expansion of the out-of-core structures.

The European sodium fast reactor—safety measures assessment and research tools (ESFR-SMART) project (Contract number: 754501) focuses on the development of innovative safety design options for future sodium-cooled fast reactors (SFRs). The goal of this paper is to show how the objective provision tree (OPT) method, which is part of the integrated safety analysis methodology (ISAM) developed by generation IV International forum (GIF), can help in defining safety guidelines for the definition and the study of ESFR-SMART innovative design options. The objective provision tree method provides a view of the implementation of main safety functions in accordance with defense-in-depth principle. The method has been adapted to ESFR-SMART needs and guidance for a consistent and homogeneous application has been developed. The method is developed for each safety function with a specific adaptation for the confinement function. Beyond the allocation of the main equipment ensuring the safety functions to the different levels of defense-in-depth, the adapted method proposes a systematic identification of the mechanisms likely to degrade them. Then, safety features to cope with these mechanisms are investigated as well as the potential common cause failures of main equipment belonging to different levels of defense-in-depth. This adapted method allows to study and then to improve the independence between the levels of defense-in-depth as required for future reactors in Europe by Western European Nuclear Regulators’ Association (WENRA). The paper provides the outline of guidance for OPT method adaptation to ESFR-SMART project needs, examples of application, and main outcomes of the use of the method.

During the Pre-Conceptual Design Phase of the EU-DEMO, two BOP solutions for WCLL and HCPB were elaborated, as close as possible to industrial standards. Nevertheless, each solution has open issues to be investigated, analytically and experimentally, in the Conceptual Design Phase (CDP). For the HCPB, the functionality and operability of the Helium-Molten Salt Heat Exchanger, and the coupling to a helium loop with a prototypic helium blower, is of primary interest. In addition, the operation of the pulse, dwell and transitions will be investigated within the new build infrastructure, HELOKA-US (Upgrade Storage), to be erected at KIT. The design requires a certain flexibility, since the final parameters of the Primary Heat Transfer System of DEMO may vary, due to plasma optimizations during CDP. HELOKA-US benefits from the high-pressure helium loop HELOKA-HP, erected to test HCPB-Breeding Blanket and First Wall modules, as well as from the competencies of preparing, handling and testing of various molten salts used for heat transfer optimization and natural convection.

Accurate coupled neutronic-thermal-hydraulic analysis of sodium fast reactor (SFRs), including European sodium fast reactor—safety measures assessment and research tools (ESFR-SMART), requires an accurate calculation of the fuel-to-clad gap conductance. This is typically derived using a fuel performance code and is dependent on fuel rating and burn-up history, among other parameters. In this paper, the gap conductance of the mixed oxide fuel (MOX) pins in the ESFR-SMART project is investigated through predictive modeling in seven independent fuel performance codes, to provide confidence in results and to aid the understanding of the uncertainties associated with gap conductance predictions. A parameterization is then defined. A single pin model for both the inner and outer fuel was built. The fuel was burned for 2100 effective full power days, with the axial power distribution varying over time. This paper presents a comparison of the gap conductance and predicted fuel temperature distribution between codes. The values produced from the codes are then combined to produce a best estimate prediction of the gap conductance expressed as a function of nodal fuel rating and burn-up, for both inner and outer fuel for all seven codes. A 2D fit was applied to the data thus obtained. The spread between results is such that, to 95% confidence, gap conductance predictions may vary from the correlation by up to a factor of ∼4. The gap conductance results obtained show a general increase of conductance with fuel rating and burn-up, from 0.22 at 0 burn-up and 10 kW/m to 0.45 W/cm/K at 0 burn-up and 50 kW/m and to 1.00 W/cm/K at 150 GWd/t and 50 kW/m. Some spread between codes has been noted and appears to be consistent with the spread published earlier by several code developers. There is good agreement between codes at low burn-up for both the central and surface fuel temperature predictions. The spread between codes increases with burn-up due to multiple phenomena, including gap re-opening, joint oxyde gain (JOG) formation, and clad swelling.

The ESFR-SMART European project (Contract No. 754501) focuses on the development of innovative safety design options for European sodium-cooled fast reactor (ESFR). The task of Work Package 1.3 is to assess the impact of the new safety measures on the reactor behavior in the transients protected by either active or passive reactor shutdown systems. The aim of Task 4 in this Work Package is to evaluate the passive reactor shutdown system performance in the ESFR core. This paper deals with the results of this evaluation, which is based on the analysis of four transients passively protected by 12 diversified shutdown device (DSD) rods. Simulations have been done with the SIM-SFR system code and demonstrated that DSD rods are capable to shut down ESFR in a timely manner, in order to avoid the negative consequences of the analyzed transients. Although a total loss of heat sink transient is a practically eliminated event, it was included in the analysis to estimate the grace time before the core meltdown.