The UK Department of Business, Energy and Industrial Strategy (BEIS) recently launched an R&D programme in Digital Reactor Design, incorporating the development of a Nuclear Virtual Engineering Capability with an integrated Modelling and Simulation programme. A key challenge of nuclear reactor design and analysis is the system complexity, which arises from a wide range of multi-physics phenomena being important across multiple length scales. This project constitutes the first step towards developing an integrated nuclear digital environment (INDE) linking together models across physical domains and incorporating real world data across all stages of the nuclear lifecycle. Simulation case studies will be developed within the INDE framework, delivering an enhanced modelling capability while ensuring the framework has immediate application. For these case studies have been specified that are relevant to design and operation phases for AGR and PWR type reactors. The AGR case considers the through-life structural performance of graphite bricks. This involves modelling of multi-scale, multi-physics phenomena in the support of reactor operations. The PWR case study is based on core multiphysics modelling, with potential relevance to operating and future PWRs, and in particular in the design of SMRs.
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.
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.
Rolls-Royce and a UK Consortium are progressing the design and development of a Small Modular Reactor (SMR) Power Station. The SMR programme is a phased design cycle, progressing through the Rolls-Royce gated review process. The project aims to deploy the first of a kind SMR in the UK by the end of the next decade. In this paper, the development methodology for the reactor core design is discussed, along with a selection of the key technical challenges that have been addressed during the concept design phase. Lessons learned from past projects have been identified, to help improve the design efficiency for the SMR. The concept design has been developed in an iterative fashion, with different analysis disciplines carefully integrated around a common set of objectives. Key economic requirements for an SMR core include maximising fuel economy, cycle length and thermal power while remaining small enough to enable a modular build approach. Top-level safety requirements include control of reactivity, control of core temperature and control of release of radioactivity/radioactive material. A set of surrogate design limits has been used alongside the true safety limits to avoid the need for detailed transient subchannel or fuel performance analysis in this phase. This has allowed the design to mature and be characterised very quickly, while also maintaining high confidence that all performance and safety requirements will be met when detailed analyses are undertaken. This paper describes the different analyses that have been undertaken to date, including a variety of reactor physics and thermal hydraulics calculations. The paper discusses the limits used, how they have been used to optimise the design solution and why they provide high confidence in the core design’s performance.
For liquid metal-cooled fast reactors (LMFRs), improved predictive modelling is desirable to facilitate reactor licensing and operation and move towards a best estimate plus uncertainty (BEPU) approach. A key source of uncertainty in fast reactor calculations arises from the underlying nuclear data. Addressing the propagation of such uncertainties through multiphysics calculations schemes is therefore of importance, and is being addressed through international projects such as the Sodium-cooled Fast Reactor Uncertainty Analysis in Modelling (SFR-UAM) benchmark. In this paper, a methodology for propagation of nuclear data uncertainties within WIMS is presented. Uncertainties on key reactor physics parameters are calculated for selected SFR-UAM benchmark exercises, with good agreement with previous results. A methodology for coupled neutronic-thermal-hydraulic calculations within WIMS is developed, where thermal feedback is introduced to the neutronic solution through coupling with the ARTHUR subchannel code within WIMS and applied to steady-state analysis of the Horizon 2020 ESFR-SMART project reference core. Finally, integration of reactor physics and fuel performance calculations is demonstrated through linking of the WIMS reactor physics code to the TRAFIC fast reactor fuel performance code, through a Fortran-C-Python (FCP) interface. Given the 3D multiphysics calculation methodology, thermal-hydraulic and fuel performance uncertainties can ultimately be sampled alongside the nuclear data uncertainties. Together, these developments are therefore an important step towards enabling propagation of uncertainties through high fidelity, multiphysics SFR calculations and hence facilitate BEPU methodologies.
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