The Application of the Combined Fission Matrix Theory in Fast Reactors

The combined fission matrix theory is a recently-developed hybrid neutron transport method. It features high efficiency, fidelity, and resolution whole-core transport calculation. The theory is based on the assumption that the fission matrix element ai,j is dominated by the property of the destination cell i. This assumption can be well explained in thermal reactors, and the combined fission matrix method has been validated in a series of thermal neutron system benchmarks. This work examines the feasibility of the combined fission matrix theory in fast reactors. The European Sodium Fast Reactor is used as the numerical benchmark. Compared to the Monte Carlo method, the combined fission matrix theory reports a 64 pcm keff difference and 8.3% 2D RMS error. The error is much larger than that in thermal reactors, and the correction ratio cannot significantly reduce the material discontinuity error in fast reactors. Overall, the combined fission matrix theory is more suited for thermal reactor transport calculations. Its application in fast reactors needs further developments.

Word from the Horizon-2020 EU ESFR-SMART Project Coordinator

This Special Issue of the ASME Journal of Nuclear Engineering and Radiation Science includes 2 editorials and 25 technical papers presenting main achievements of the Horizon-2020 European Union ESFR-SMART project (European Sodium Fast Reactor Safety Measures Assessment and Research Tools) supported by the EURATOM grant and launched in September 2017. ESFR-SMART gathers a consortium of 19 organizations (see Figure 1): Research centres, industries, universities, Technical and Scientific Support Organizations as well as Small to Medium Enterprise aiming at enhancing further the safety of Generation-IV Sodium Fast Reactors (SFRs) and, in particular, of the commercial-size European Sodium Fast Reactor (ESFR) in accordance with the European Sustainable Nuclear Industrial Initiative (ESNII) roadmap.

ESFR-SMART Core Safety Measures and Their Preliminary Assessment

A large 3600 MW-thermal European Sodium Fast Reactor (ESFR) concept has been studied in Horizon-2020 ESFR-SMART (ESFR Safety Measures Assessment and Research Tools) project since September 2017, following an earlier EURATOM project, CP-ESFR. In the paper, we describe new ESFR core safety measures focused on prevention and mitigation of severe accidents. In particular, we propose a new core configuration for reducing the sodium void effect, introduce passive shutdown systems, and implement special paths in the core for facilitation of molten fuel discharge in order to avoid re-criticalities after a hypothetical severe accident. We describe and assess the control and shutdown system, and consider options for burning minor actinides.

Optimization of the European Sodium Fast Reactor Secondary Sodium Loop as Part of the ESFR-SMART Project

Based on feedback from existing reactors and current projects, the European Sodium Fast Reactor Safety Measures Assessment and Research Tools (ESFR SMART) project proposes an optimization of the secondary circuit with the main aim of improving safety. Besides, the optimization also leads to a simplification of the circuits and therefore to a reduction of the cost of the reactor. For the implementation of the proposed new design option, some points require further R&D to validate their feasibility.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

As part of the ESFR-SMART project, safety assessments are being conducted on the updated European Sodium Fast Reactor (ESFR) design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In the current paper, the analyzed accidental scenario is the so called Protected Station Blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and SIM-SFR system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRS). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project have been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

Alternative Absorber Materials for Control Rods in ALFRED

The next-generation reactors require improved safety performance and longer cycle length, which initiate the research on alternative absorber materials. In this context, potential absorber materials including borides (B4C, HfB2, and ZrB2), rare earth oxides (Eu2O3, Gd2O3, Sm2O3, and Dy2TiO5), metals/alloys (Hf and AIC), and metal hydride (HfHx) were compared in a large sodium fast reactor. The design of control rods for Generation-IV fast reactors strongly depends on the core characteristics. In this paper, some alternative absorbers are assessed in a lead fast reactor ALFRED using depletion capability in the Monte-Carlo particle transport code OpenMC. Results show that the ALFRED reference control rod design with B4C largely satisfies the shutdown and operation requirements. 60% 10B enriched HfB2 and HfH1.18 can replace the operation part of the reference design. In the future, the safe operating life of B4C and HfB2 should be assessed taking into account the irradiation-induced swelling, temperature margin, and gas release. HfH1.18 has a limited and local influence on the core power distribution. Eu2O3 has little loss on the absorption ability after 5 cycle irradiation. This oxide absorber satisfies the shutdown function even with only half control rod insertion, while its critical insertion depth at beginning of the cycle should be increased to realize reactivity compensation function.

SPX Benchmark Part II: Transient Results

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 modelling of selected operational transients performed during start-up 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 necessary thermal hydraulic description of the primary system components, assumptions to account for thermal expansion reactivity feedbacks 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 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 modelling features for thermal expansion of the out-of-core structures.

Decay Heat Characterization for the European Sodium Fast Reactor

In this work a detailed assessment of the decay heat power for the commercial-size European Sodium-cooled Fast Reactor (ESFR) at the end of its equilibrium cycle has been performed. The summation method has been used to compute very accurate spatial- and time-dependent decay heat by employing state-of-the-art coupled transport-depletion computational codes and nuclear data. This detailed map provides basic information for subsequent transient calculations of the ESFR. A comprehensive analysis of the decay heat has been carried out and interdependencies among decay heat and different parameters characterizing the core state prior to shutdown, such as discharge burnup or type of fuel material, have been identified. That analysis has served as a basis to develop analytic functions to reconstruct the spatial-dependent decay heat power for the ESFR for cooling times within the first day after shutdown.