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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. [ABSTRACT FROM AUTHOR]

The WIMS (Winfrith Improved Multigroup Scheme) reactor physics code is actively being developed for whole core modelling of a range of Small Modular Reactor types including the Pressurized Water Reactor (PWR), High Temperature Reactor (HTR), and Liquid Metal Cooled Fast reactor (LMFR). These developments include the capability for whole core multiphysics modelling with neutronics and thermal hydraulic feedback, as well as methods to determine the power deposition from neutron and gamma heating. Flux solutions are obtained using a wide variety of deterministic methods including diffusion theory, SP3, and full transport with the method of characteristics and Sn discrete ordinates methods, as well as multi-group Monte Carlo methods. The SP3 method allows both steady state and time dependent transient solutions by solving the time dependent SP3 equations. A wide variety of nuclear data libraries are available with WIMS including data from the JEF3.3, ENDF/B-VII.0 and CENDL3.1 nuclear data evaluations. This paper presents validation of the latest version of the WIMS code, WIMS11, for PWR and HTR systems. Comparisons are made against physics data obtained from the OECD/NEA PWR Watts Bar multi-physics benchmark and the IAEA HTR-10 benchmark, as well as neutron and gamma heating experiments that took place on the NESSUS reactor at Winfrith in the United Kingdom. In each case, validation of WIMS has been obtained by comparison either against measured data, or results provided by other benchmark participants that have been obtained with alternative deterministic or Monte Carlo methods. [ABSTRACT FROM AUTHOR]

The WIMS/PANTHER Embedded Supercell Method (ESM) provides a significant improvement in prediction accuracy for radial power distributions for PWR reactors compared to the standard "two-step" approach, without the need for a significant increase in computational resource. Recent papers at PHYSOR conferences have outlined the details of the method and demonstrated its operation, and the accuracy improvements possible, by means of benchmarking calculations. This paper applies the method to a 4-loop PWR in the U.K, and three PWRs (3-loop and 2-loop) in Belgium. Comparisons are made against measured data from the start-of-cycle physics testing performed for each cycle, and power-shape measurements collected during the cycle using a conventional "two-step" nodal reactor solution, and with the ESM. All results will be presented with the JEF2.2 nuclear data library, for ease of comparison between the methods and previously reported results, although the effects of more modern evaluations will be commented upon. The benchmark calculations referred to above studied a challenging MOX/UO2 benchmark core akin to an SMR. The four reactors studied here include conventional UO2 only core designs and cycles with UO2/MOX mixed cores. A variety of boron-and gadolinium-based burnable absorbers are also present. The data is used to show that the method both operates successfully for real reactor problems, and delivers improvements in the prediction accuracy of measured parameters. [ABSTRACT FROM AUTHOR]