Development and benchmarking of a mechanical model for GFR plate-type fuel

The reference fuel design currently being considered within the Generation-IV Gas-cooled Fast Reactor (GFR) project is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium-uranium carbide, while the matrix material is silicon carbide. The present paper describes the mechanical part of a thermal-mechanical model being developed for studying the transient behavior of this highly heterogeneous fuel type. Benchmarking has been carried out against detailed finite-elements modeling (FEM). The resultant thermal-mechanical model can provide reliable fuel and cladding (matrix) stress/strain conditions to evaluate temperatures and neutronic feedbacks. As such, it has been integrated into PSI's coupled code system FAST, which aims at the comprehensive safety analysis of advanced fast reactor systems. The detailed FEM analysis of the GFR fuel has been useful not only for benchmarking the new model, but also for obtaining an in-depth understanding of fuel stress/strain characteristics, which cannot be reproduced with simplified models. Thereby, the range of applicability of the new model has clearly been defined. In particular, the 3D FEM analysis has revealed a concentration of stresses at the pellet corners during pellet/matrix contact, which could lead to fuel element failure. This effect is found to be mitigated considerably, if the fuel pellets are shaped in a manner which enhances the contact area. [All rights reserved Elsevier].

Development and benchmarking of a 2D transient thermal model for GFR plate-type fuel

Rakesh Chawla, Konstantin Mikityuk, Petr Petkevich

Several new fuel designs are currently being considered within the generation-IV gas-cooled fast reactor (GFR) project. One of them, which is the GFR reference design, is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium-uranium carbide, while the matrix material is silicon carbide. A two-dimensional approach to simulation of the thermal behaviour (for both steady-state and transient analysis) of such fuel, taking into account the inner heterogeneity, has been developed and benchmarked against detailed finite-element calculations. The resultant model can provide reliable fuel and matrix temperatures to determine neutronic feedbacks, as also to evaluate fuel integrity. As such, it has been integrated into PSIs coupled code system FAST, which aims at the comprehensive safety analysis of advanced fast reactor systems.[All rights reserved Elsevier].

Elsevier Science Ltd.2007

Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

Rakesh Chawla, Gaëtan Girardin, Konstantin Mikityuk, Gérald Rimpault

The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic-ceramic (CERCER) plate-type fuel-elements with (U-Pu) carbide fuel contained within a SiC inert matrix. For the neutronic calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European fast reactor (EFR)-type pattern for the control assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal-hydraulic constraints could be appropriately accounted for. The thermal-hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins.Complementary to the design study, neutronic investigations have been performed to assess the impact of the control assemblies in the GFR core in greater detail (rod interactions, shift of the flux, peaking factors, etc.). Thus, considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual assemblies within the first bank. Large anti-shadowing effects also occur, the most prominent being that between the two CSD banks, where the total assembly worth is almost doubled in comparison to the sum of the individual values. Additional investigations have been performed and, in this context, it has been found that computation of the first-order eigenvalue and the eigenvalue separation is a robust tool to anticipate control assembly interactions in a large fast-spectrum core. One interesting finding is that the interactions are lower when one of the control assembly banks is located at a radial position corresponding to the minimum in the first-order eigenvalue (at approximately half the core radius).An important aspect of the design optimization process has been to obtain a low heterogeneity effect for the control assembly worth. This was achieved by applying a methodology based on reactivity equivalence, with the optimization goal set at minimizing the shadowing effects between the absorber pins as well as the self-shielding within the absorber pins. A relatively low reduction of ~13% was obtained as the heterogeneity effect for the control assembly worth, the corresponding value in the case of Super-phenix being significantly higher, viz. about 20%.[All rights reserved Elsevier].

Elsevier Science Ltd.2008

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

Rakesh Chawla, Aurélia Chenu, Jiri Krepel, Konstantin Mikityuk, Kaichao Sun

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.