SIMULATION OF BREED AND BURN FUEL CYCLE OPERATION OF MOLTEN SALT REACTOR IN BATCH-WISE REFUELING MODE

The Molten Salt Reactor (MSR) is one of the most revolutionary Gen-IV reactors and it can be operated, especially with chloride salts, in the so-called breed and burn fuel cycle. In this type of fuel cycle the fissile isotopes from spent fuel do not need to be reprocessed, because the excess bred fuel covers the losses. The liquid phase of the MSR fuel assures its instant homogenization, and the reactor can be operated with batch-wise refueling thus reaching an equilibrium state. At the same time, the active core of the chloride fast MSR needs to be bulky to limit neutron leakage. In this study, the code Serpent 2 was coupled to the Python script BBP to simulate batch-wise operation of the breed and burn MSR fuel cycle. The script, previously developed for solid assemblies shuffling, was modified to simulate fuel homogenization after fertile material addition. Several fuel salts and fission products removal strategies were simulated and their impact was analyzed. Similarly, the influence of blanket volume was assessed in a two-fluid core layout. The results showed that the reactivity initially grows during the irradiation period and later decreases. The blanket has a large impact on the performance and it can be used to further increase the fuel burnup or to shrink the active core size. The breed and burn fuel cycle in MSR can reach high fuel utilization without fuel reprocessing and a multi-fluid layout can help to decrease the core size

MAPPING OF SODIUM VOID EFFECT AND DOPPLER CONSTANT IN ESFR-SMART CORE WITH MONTE CARLO CODE SERPENT AND DETERMINISTIC CODE ERANOS

The Sodium Fast Reactor is one of the most technologically developed Gen-IV reactors, which can close the nuclear fuel cycle. Its criticality safety directly depends on the sodium void effect and Doppler constant. Hence the knowledge of their local distribution is important. These coefficients can be mapped by deterministic or Monte Carlo codes, where the latter provide higher modeling accuracy, but are also strongly computer demanding and subject to stochastic noise issues. In this study, the void effect and Doppler constant have been enumerated for the ESFR core by Serpent2 and ERANOS2 codes, preserving a six-batch operation scheme. The Serpent code was coupled to the Python script BBP to simulate batch-wise operation in a radially infinite inner core configuration; the ERANOS code was applied to the whole core geometry and the batch-wise operation was simulated by the EQL3D routine. Sodium void effect and Doppler constant spatial maps with different levels of refinement were produced, as well as the time evolution of the integral coefficients during the transition from initial cycle to equilibrium cycle. Both codes indicate deterioration of these coefficients during the transition. The equilibrium cycle performance of the inner core zone from the ERANOS calculation was compared with Serpent results and they showed reasonable agreement. For very fine mapping, the Monte Carlo method employed was computationally very demanding and the enumerated effect was lower than the stochastic noise. In general, the Serpent model practically excludes modeling assumptions and produces reliable results for reasonably sized maps, which can be combined if needed with the high spatial resolution results obtained by ERANOS simulations

The EQL0D fuel cycle procedure and its application to the transition to equilibrium of selected molten salt reactor designs

In recent years, Molten Salt Reactors (MSRs), one of the Generation IV concepts, have noticeably gained in interest. Several molten salt reactor types use fuel in liquid form, bringing various advantages, notably in terms of fuel cycle. However, simulating the evolution of liquid fuel under irradiation necessitates specific tools and methods, for example to simulate the removal of insoluble fission products from the salt mixture or continuous refueling of the reactor. For this purpose, the Serpent 2-based procedure EQL0D has been developed. It is designed for both equilibrium and finite-time burn-up calculations in liquidfueled molten salt reactors. This paper first details the most important methods implemented in the procedure to simulate liquid-fuel systems, after which the adequacy of the procedure is verified by comparing benchmark results with a comparable code. Examples of applications of this tool to the start-up and transition to closed fuel cycle of the historical single- and two-fluid Molten Salt Breeder Reactors as well as the more recent Molten Salt Fast Reactor, a fast-spectrum concept, are then presented. The obtained results show the challenge faced by thorium-cycle breeder MSRs to transition from available nuclear fuels to a closed thorium cycle, as many candidate fuels prove to be unusable for the transition to a closed cycle. (C) 2020 Elsevier Ltd. All rights reserved

EQL3D: ERANOS based equilibrium fuel cycle procedure for fast reactors

The advanced fast reactors of the fourth generation should be capable to breed their own fuel from poorly fissile 238U and to recycle the actinides from their own spent fuel. However, this recycling or actually the closure of fuel cycle has negative impact on the safety parameters. The goal of this work is to develop a numerical tool, which can simulate and confirm the capability of these reactors to operate with closed fuel cycle, and which can evaluate their safety parameters. The tool is named equilibrium fuel cycle procedure for fast reactors (EQL3D) and is based on the ERANOS 2.1 code platform. Equilibrium cycle or virtually equilibrium method for considering the homogeneous recycling of actinides is a known approach; however, in EQL3D the equilibrium method is newly applied for hexagonal-z 3D and r-z 2D core geometries and typically 33 energy-group neutron-flux calculations. The utilization of hexagonal-z 3D geometry enables to characterize the equilibrium cycle for complex reloading patterns within a multi-batch scheme. Furthermore, EQL3D enables comparison of the advanced fast reactors on a common basis of their equilibrium cycle reactivity swing, fuel composition, breeding gain and safety-related parameters. The Gas-cooled Fast Reactor (GFR) was selected for verification and optimization of the EQL3D procedure. The GFR geometry was based on an international neutronics benchmark with a simple setup and potential for latter upgrade. It was used to show the impact of several EQL3D options e.g. different isotope evolution models, geometry selection, or cross-section recalculation frequency, on the equilibrium parameters. The results demonstrate the capability of the procedure to calculate the equilibrium fuel cycle for advanced fast reactors. Among others, also the ability of GFR benchmark core to be operated with closed fuel cycle is shown

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Coupled 3-D Neutronics/Thermal-Hydraulics Optimization Study For Improving The Response Of A 3600 Mw(Thermal) Sfr Core To An Unprotected Loss-Of-Flow Accident

The sodium-cooled fast reactor (SFR), as a fast neutron spectrum system, is characterized by several performance advantages. In particular, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. However, the SFR has one dominating neutronics drawback, namely, there is generally a positive reactivity effect when there is voiding of the sodium coolant in the core. Furthermore, this effect becomes even stronger in the equilibrium closed fuel cycle. Considering that in a hypothetical SFR unprotected loss-of-flow (ULOF) accident scenario, i.e., flow rundown without SCRAM, sodium boiling can be anticipated to occur, it is crucial to assess the corresponding impact of the positive sodium void effect. An optimization study for improving the safety characteristics of a large [3600-MW(thermal)] SFR has currently been conducted in the above context. The dynamic core response to a reference ULOF scenario is investigated with the use of a coupled three-dimensional neutronics/thermal-hydraulics PARCS/TRACE model. The starting point of the study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the sodium void effect, the core has been modified by introducing an upper sodium plenum, along with a boron layer above it. Furthermore, the original core height-to-diameter ratio is reduced. In comparison to the reference ESFR core behavior, certain improvements are achieved, thanks to the static neutronics optimization carried out. However, these changes are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting during the considered ULOF event. Thermal-hydraulics optimization has thus been considered necessary, in order to (a) prevent sodium flow blockage in the fuel channel and (b) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. The corresponding measures taken are (a) the introduction of an innovative wrapper design, which features small openings in each side surface of the fuel assembly, and (b) replacement of the original upper sodium plenum by an extended fission gas plenum. Following implementation of these thermal-hydraulics-related design changes, one arrives at a final configuration of the SFR core, in which, for the selected accident scenario, a new "steady state" involving stable sodium boiling is found to be achievable, with melting of neither cladding nor fuel. Such a satisfactory behavior has been confirmed not only for the beginning-of-life state of the core but also for the equilibrium closed fuel cycle