Current status of the reactor physics code WIMS and recent developments

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 between decay heat and different parameters characterizing the core state prior to shutdowns, 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.

Evaluation of the ESFR End of Equilibrium Cycle State: Spatial Distributions of Reactivity Coefficients

The Horizon 2020 European Sodium-cooled Fast Reactor Safety Measures Assessment and Research Tools (ESFR-SMART) project investigates the behavior of the commercial-size ESFR throughout its lifetime. This paper reports work focused on the end of equilibrium cycle (EOEC) loading of the ESFR, including neutronic analysis, core- and zone-wise reactivity coefficients, and more detailed local mapping of important safety-relevant parameters. Sensitivity and uncertainty analysis on these parameters have also been performed, and a detailed investigation into decay heat mapping was carried out. Due to the scope of this work, the results have been split into three papers. The nominal operating conditions and both zone-wise and local mapping of reactivity coefficients are considered in this paper. The work was performed across four institutions using both continuous-energy Monte Carlo (MC) and deterministic reactor physics codes. A good agreement is observed between the methods, verifying the suitability of these codes for simulation of large, complicated reactor configurations and giving confidence in the results for the most limiting ESFR EOEC core state for safety analysis. The results from this work will serve as the basis for the transient calculations planned for the next stage of work on the ESFR, allowing for more in-depth studies to be performed on the multiphysics behavior of the reactor.

Reduced-Order Modelling with Domain Decomposition Applied to Multi-Group Neutron Transport

Solving the neutron transport equations is a demanding computational challenge. This paper combines reduced-order modelling with domain decomposition to develop an approach that can tackle such problems. The idea is to decompose the domain of a reactor, form basis functions locally in each sub-domain and construct a reduced-order model from this. Several different ways of constructing the basis functions for local sub-domains are proposed, and a comparison is given with a reduced-order model that is formed globally. A relatively simple one-dimensional slab reactor provides a test case with which to investigate the capabilities of the proposed methods. The results show that domain decomposition reduced-order model methods perform comparably with the global reduced-order model when the total number of reduced variables in the system is the same with the potential for the offline computational cost to be significantly less expensive.

EVALUATION OF THE ESFR END OF CYCLE STATE AND DETAILED ANALYSIS OF SPATIAL DISTRIBUTIONS OF REACTIVITY COEFFICIENTS

The ESFR-SMART project is the latest iteration of research into the behaviour of a commercial-size SFR core throughout its lifetime. As part of this project the ESFR core has been modelled by a range of different reactor physics simulation codes at its end of cycle state, and the important safety relevant parameters evaluated. These parameters are found to agree well between the different codes, giving good confidence in the results.

A detailed mapping of the local sodium void worth is also performed due to the problems associated with the positive void coefficient seen in large SFR designs. The local void worth maps show that the use of zone-wise coefficients replicates the important reactivity feedbacks to a high degree, indicating their suitability for use in SFR simulations.

WHOLE CORE COUPLING METHODOLOGIES WITHIN WIMS

The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation.

This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions.

The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code.

APPLICATION OF TWO STAGE METHOD OF CHARACTERISTICS / SP3 METHODOLOGY TO TRISO-FUELLED LEU SPACE REACTOR IN WIMS 11

In order to minimise the mass of a 1MWe LEU space fission power system design, a rapid neutronics analysis tool is sought. A two-stage deterministic analysis routine has been constructed using a core-plane method of characteristics calculation followed by a full-core SP3 calculation, within the ANSWERS© code WIMS11. This is compared to a faster route that skips the core-plane calculation and also the Monte Carlo code Serpent. Results suggest sufficiently good agreement for the WIMS-based methods to be useful in a full system mass-minimising optimisation routine.

ZERO POWER CRITICALITY BENCHMARK EVALUATION OF THE MSRE IN WIMS

This paper discusses work done to benchmark the deterministic code WIMS [1] against the Monte Carlo code Serpent [2] and experiment. Comparison is made against the Molten Salt Reactor Experiment at Oak Ridge National Laboratory as well as a Serpent model produced at the University of California, Berkeley. Producing a model for an MSR is possible thanks to the development of the next version of WIMS, WIMS11. The structure of the WIMS model built is discussed, and the final predicted criticality value for the MSR is given. This compares favourably with the Serpent model; however, both codes predict values considerably different to those expected. Potential reasons for this are suggested. However, it is concluded that WIMS has successfully been benchmarked against the current state of the art. This provides confirmation that this is a valid approach for molten salt reactor research analysis.

Neutronic investigation of alternative & composite burnable poisons for the soluble-boron-free and long life civil marine small modular reactor cores

Concerns about the effects of global warming provide a strong case to consider how best nuclear power could be applied to marine propulsion. Currently, there are persistent efforts worldwide to combat global warming, and that also includes the commercial freight shipping sector. In an effort to decarbonize the marine sector, there are growing interests in replacing the contemporary, traditional propulsion systems with nuclear propulsion systems. The latter system allows freight ships to have longer intervals before refueling; subsequently, lower fuel costs, and minimal carbon emissions. Nonetheless, nuclear propulsion systems have remained largely confined to military vessels. It is highly desirable that a civil marine core not use soluble boron for reactivity control, but it is then a challenge to achieve an adequate shutdown margin throughout the core life while maintaining reactivity control and acceptable power distributions in the core. High-thickness ZrB₂ 150 μm Integral Fuel Burnable Absorber (IFBA) is an excellent burnable poison (BP) candidate for long life soluble-boron-free core. However, in this study, we want to minimize the use of 150 μm IFBA since B-10 undergoes an (n, α) capture reaction, and the resulting helium raises the pressure within the plenum and in the cladding. Therefore, we have considered several alternative and novel burnable BP design strategies to minimize the use of IFBA for reactivity control in this study: (Case 1) a composite BP: gadolinia (Gd₂O₃) or erbia (Er₂O₃) with 150 μm thickness ZrB₂ IFBA; (Case 2) Pu-240 or Am-241 mixed homogeneously with the fuel; and (Case 3) another composite BP: Pu-240 or Am-241 with 150 μm thickness ZrB₂ IFBA. The results are compared against those for a high-thickness 150 μm 25 IFBA pins design from a previous study. The high-thickness 150 μm 25 IFBA pins design is termed the “IFBA-only” BP design throughout this study. We arrive at a design using 15% U-235 fuel loaded into 13 × 13 assemblies with Case 3 BPs (IFBA+Pu-240 or IFBA+Am-241) for reactivity control while reducing 20% IFBA use. This design exhibits lower assembly reactivity swing and minimal burnup penalty due to the self-shielding effect. Case 3 provides ~10% more initial (beginning-of-life) reactivity suppression with ~70% less reactivity swing compared to the IFBA-only design for UO₂ fuel while achieving almost the same core lifetime. Finally, optimized Case 3 assemblies were loaded in 3D nodal diffusion and reactor model code. The results obtained from the 3D reactor model confirmed that the designed core with the proposed Case 3 BPs can achieve the target lifetime of 15 years while contributing to ~10% higher BOL reactivity suppression, ~70% lower reactivity swings, ~30% lower radial form factor and ~28% lower total peaking factor compared to the IFBA-only core.

On a Roadmap for Future Industrial Nuclear Reactor Core Simulation in the U.K. to Support the Nuclear Renaissance

The U.K. has initiated the nuclear renaissance by contracting for the first two new plants and announcing further new build projects. The U.K. government has recently started to support this development with the announcement of a national programme of nuclear innovation. The aim of this programme with respect to modelling and simulation is foreseen to fulfil the demand in education and the build-up of a reasonably qualified workforce, as well as the development and application of a new state-of-the-art software environment for improved economics and safety. This document supports the ambition to define a new approach to the structured development of nuclear reactor core simulation that is based on oversight instead of looking at detail problems and the development of single tools for these specific detail problems. It is based on studying the industrial demand to bridge the gap in technical innovation that can be derived from basic research in order to create a tailored industry solution to set the new standard for reactor core modelling and simulation for the U.K. However, finally, a technical requirements specification has to be developed alongside the strategic approach to give code developers a functional specification that they can use to develop the tools for the future. Key points for a culture change to the application of modern technologies are identified in the use of DevOps in a double-strata approach to academic and industrial code development. The document provides a novel, strategic approach to achieve the most promising final product for industry, and to identify the most important points for improvement.