Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS' core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor's small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of k eff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of k eff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of $8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients produced by Serpent. A comparison of the radial power profiles between the best PARCS model and full-core Serpent model showed largest relative differences in terms of power prediction at the core periphery, which is believed to be the product of the geometry simplifications made, the diffusion coefficients produced by Serpent, and the two-group energy structure used. The worth of a single control rod reproduced in PARCS showed a difference of À33 pcm from its 169 pcm worth simulated in Serpent.

VERIFICATION OF A REACTOR PHYSICS CALCULATION SCHEME FOR THE CROCUS REACTOR

Gaëtan Girardin, Mathieu Hursin, Andreas Pautz, Adolfo Rais, Daniel Jerôme Siefman

CROCUS is a zero power (100 W) reactor of the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology, Lausanne (EPFL). It is used for teaching and research purposes. Its modeling has relied so far on diffusion theory and point kinetics for the neutronic analysis and simplified thermal hydraulics models for accident analysis. Recently, an effort has started within the LRS to improve its modeling capabilities, the long term goal being to update the CROCUS Safety Analysis Report (SAR) for improved operational flexibility. The present work is focused on the static neutron analysis of CROCUS through the development and preliminary verification of a 3D nodal simulator (e.g. PARCS) model of the reactor, a methodology typically used in the industry for modeling of Light Water Reactors (LWR). The set of homogenized macroscopic cross-sections needed by the core simulator, referred in this work as nuclear data library, is generated by a Monte Carlo based code (e.g. Serpent). The quantities of interest for the verification of the model are the keff, and the control rod worths. An innovative homogenization approach to generate the nuclear data library is considered due to the irregular radial geometry of the CROCUS reactor. The reference solution is provided by another Monte Carlo code, MCNP5. The uncertainty due to the nuclear data in the keff prediction of Serpent is also investigated and amounts to about 500pcm which covers the deviation from unity of keff prediction by MCNP5 and Serpent for a critical CROCUS configuration. PARCS keff predictions are within 400 pcm of the Serpent results.

2014

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools. There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations. For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art. Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered. In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients. Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants. Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity. Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied. By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty. For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

EPFL2021

Investigation and Validation of Unstructured Mesh Methodologies for Modeling Experimental Reactors

Carlo Fiorina, Mathieu Hursin, Tom Mager, Andreas Pautz

This paper summarizes a methodology developed at Ecole Polytechnique Federale de Lausanne for the neutronic modeling of the CROCUS experimental reactor and proposes solutions to the challenges one may face while modeling a research reactor with a complex geometry. Indeed, the double-lattice configuration of CROCUS makes it difficult to use codes for neutron diffusion and transport relying on a structured mesh description. For this reason, and based on the available in-house competences, we decided to make use of the neutronic capabilities of the GeN-Foam multiphysics solver, which takes advantage of general finite volume methodologies on unstructured meshes to provide sufficient flexibility for the study of unconventional reactor designs. In this work, GeN-Foam is used to build a first SP3 model of CROCUS based on an unstructured mesh to have an explicit modeling of the double lattice and the water gap between the two lattices. Form functions are then used to reconstruct the intra-pin fission rates for validation against measured distributions. We also discuss the limitations of the SP3 approximation of neutron transport in regions with steep neutron flux gradients and the planned future developments.

MDPI2022

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

EPFL2021