Liquid Metal-cooled Fast Reactors Research Articles

Evaluation of proliferation resistance of SMRs with COMPRE methodology assisted by Pu production

The proliferation resistance (PR) of several small modular reactors (SMR) is investigated in terms of material characteristic and other extrinsic parameters with COMPRE methodology developed by KINAC. In this work, plutonium (Pu) production amount and isotopic yield are newly introduced as key metrics in material characteristic. Their functionalization is performed with a linear regression method based on sparse core depletion data. Nevertheless, a clear discrepancy is observed in Pu buildup and its quality depending on SMR type and core burnup status. Some of the non-light water SMRs exhibit a higher proliferation concern and their PR score trends are clearly different compared to conventional pressurized water reactors (PWR). Other extrinsic parameters are evaluated based on development status of representative SMRs. At present, liquid metal cooled-fast reactor (LMFR) and molten salt reactor (MSR) receive low scores due to their limited technical maturity in reactor design and related I&C equipment. We conclude that distinct safeguards approaches, suitable for a wide range of SMR designs should be developed for commercial use in the near future.

Advanced Liquid-Metal-Cooled Fast Reactor Core Design Using Modern Optimization Algorithms

When performing core loading pattern design in fast spectrum reactors, it is often assumed that the larger neutron mean free path in fast reactors makes the core loading pattern less significant than in thermal reactors. Due to this assumption, the literature often includes homogeneous core designs for liquid-metal-cooled fast reactors (LMFRs). In this paper, heterogeneous loading patterns are investigated using modern LMFR multiphysics analysis. It was found that the figures of merit (FOMs) used to design cores are very sensitive to the core loading pattern, and better core designs (measured by the FOM) can be obtained from heterogeneous designs. Based on this investigation, a need for a modern LMFR core loading pattern design methodology is identified and developed. The methodology is demonstrated through the optimization of the core loading pattern of a sodium-cooled fast reactor. The reactor design used in this demonstration is based on the Super Power Reactor Innovative Small Modular (SPRISM) core, and an optimized loading pattern is obtained by searching for the fuel enrichment and locations of the driver and blanket assemblies. The objectives of the search are to reduce the fuel cost and peaking factors, while meeting the design constraints. To calculate the fuel cost, a preliminary cost model is developed and applied for transuranium fuel loading. Upon establishing the methodology, six optimization algorithms are tested for their effectiveness in solving the LMFR core loading pattern problem. Four of the algorithms are population based: Gray Wolf Optimization, Salp Swarm Optimization, Whale Optimization Algorithm, and the Moth Flame Optimization. A reinforcement learning algorithm, called the proximal policy optimization, was also selected. Finally, the differential evolution (DE) algorithm was selected as the choice of an evolutionary-based algorithm. All algorithms showed a competitive converged design. However, the DE showed more favorable performance since it was able to converge to a superior design compared to the rest of the algorithms with a reasonable number of design sample evaluations, in addition to avoiding local minima entrapment.

METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization

The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.

Nuclear microreactors: Status review and potential applications; the Mexican case

Microreactors are a class of small modular reactor with a typical power output up to 10 MWe. Their factory fabricated, transportable, and self-regulating design makes the microreactors suitable for supplying electricity to off-grid communities, emergency situations, and space and naval applications. Likewise, microreactors are appropriate for non-power applications such as hydrogen and synthetic fuels production, high and low temperature process heat, various industries (mining, cement, etc.), district heating, and water desalination. Many microreactor concepts are under development around the world with different nuclear technologies, such as the Micro Modular Reactor, a high temperature gas-cooled reactor; Aurora, a liquid metal-cooled fast reactor; and eVinci, a heat pipe-cooled reactor, among others. A recent market assessment of microreactors, conducted by the Idaho National Laboratory, identifies five potential markets for microreactors: Isolated Operations, Distributed Energy, Resilient Urban, Disaster Relief, and Marine Propulsion. In the case of Mexico, which remains dependent on fossil fuels for its energy supply (oil covers 44% of the energy source, while natural gas covers 38% and coal 3%), all potential markets are attractive to consider deploying microreactors, particularly Isolated Operations and Distributed Energy, due to its industrial sector and remote/isolated communities. Mexico’s mining cement, and iron and steel industries are the major consumers of electricity, which is generated primarily with fossil fuels. Similarly, oil accounts for 99.7% of the energy mix of the national transport sector. Furthermore, isolated regions in Mexico, such as the state of Baja California Sur, which generates its own electricity mainly with natural gas, fuel oil, and diesel, could implement microreactors, not only as a potential solution for their electricity supply, but also to implement cogeneration application as water desalination.

Fuel cycle cost comparison between lead and sodium cooled fast reactors

Sodium and lead coolants are the most recognized coolants for Liquid Metal-cooled Fast Reactor (LMFR). The two coolants have significantly different physical, thermal and chemical properties, which results in major core design differences between lead- and sodium-cooled systems. This work attempts to quantify the impact of the coolant type on the fuel cycle cost of a LMFR by comparing the Levelized Fuel Cycle Cost (LFCC).The paper is composed of two studies with the goal of comparing the advantage of the superior neutronic performance of lead versus the advantage of the tight lattice that can be achieved with sodium coolant. In the first study, the influence of the neutronic differences between lead and sodium on the fuel cycle cost are isolated and quantified. This is achieved by comparing the fuel cycle performance of a Lead-cooled Fast Reactor (LFR) core model to that of a Non-optimized SFR (NSFR) model. The two models have a similar geometry, fuel loading, fuel type, and structural materials. The only major difference between the two cores was the coolant (i.e. lead vs. sodium). Based on this comparison, it is shown that the better neutronic performance of lead coolant results in longer cycle length for the LFR core. It is also shown that this difference in cycle length can result in up to 30.8% better LFCC for a lead-cooled system compared to a sodium-cooled one. In the second study, a Sodium-cooled Fast Reactor (SFR) is designed that is optimized for the sodium coolant. The design process for the Long-life core Sodium Fast Reactor (LSFR) realizes the compatibility of sodium coolant with structural materials and allows for a more compact design. The design parameters and constraints of the design are presented, and the objective of the design is to reduce the LFCC. In this study, the design with the sodium coolant offers a 2% reduction in the LFCC, along with a smaller core size.The overall results show that if you compare sodium and lead using the same reactor geometry, the neutronic benefits of lead coolant can lead to a 30% reduction in the LFCC. However, if you optimize the geometry to take advantage of the higher power density allowed with sodium coolant, the sodium coolant can offer a 2% LFCC reduction or a smaller core size.

Validation of RANS simulations on a 61-pin wire-wrapped fast-reactor fuel assembly under the presence of localized blockages

One of the fuel assembly designs considered for liquid metal-cooled fast reactors (LMFR) uses wires helically wrapped around fuel pins as spacers. The understanding of the coolant behavior in LMFR fuels under the influence of a partially blocked flow area is required for a correct safety assessment of the Nuclear Power Plant. Accumulation of debris or cladding deformation may generate a partial or total flow blockage of coolant channels at different locations. In this paper, Unsteady Reynolds-Averaged Navier-Stokes (URANS) is applied using different turbulence models to predict the axial friction factor under the presence of blockages, placed at different positions on the bundle cross section. The simulated Reynolds number was 17,000. Results are compared with experimental friction factor data produced at the Texas A&M experimental facility. The sensitivity analysis of the RANS turbulence models concluded that the k-∊ Realizable model is the recommended one for this Reynolds number because it presented the minimum relative error with respect to the experimental data that was found below 5%.

Design, Modeling, and Analysis of a Compact-External Electromagnetic Pumping System for Pool-Type Liquid Metal-Cooled Fast Reactors

Current pool-type Liquid Metal-Cooled Fast Reactors (LMCFRs), either under development or operational, immerse main reactor components in the primary coolant, (i.e. sodium), that includes heat exchangers, shielding structures, and pumping systems. Proposed main pumping systems, for some reactors that are under development, use electromagnetic pumps (EMPs) for primary coolant circulation. Annular linear induction pumps (ALIPs) are the preferred type since they are known for their advantages over mechanical centrifugal pumps (MCPs) when used to circulate liquid metals. This is due to ALIPs’ absence of moving parts such as shafts and impellers, seals and bearings, auxiliary lubrication systems, and simplicity of flow and pressure control mechanism. The immersion of reactor components in the primary sodium in the reactor vessel minimizes the likelihood of radioactive coolant leakage and loss of coolant accidents. However, since there is only one access point to reactor components, the immersion of ALIPs prevents additional potential advantages. Online pump inspection and replacement, reduction of negative effects on pump components due to the high temperature and radiation environment, an additional heat removal mechanism for self-cooled ALIPs, and simple decommissioning procedures are some possible advantages. This paper discusses a study conducted to investigate the possibility of using large EMPs, ALIP type, that are located outside the reactor vessel and connected in parallel, instead of in vessel sodium immersed ones for pool-type LMCFRs. The large, outside-vessel EMP idea is tested on a liquid metal-cooled test reactor design using an experimentally validated multiphysics finite element analysis tool. Specifically, the steady state reactor’s primary circuit cooling and pressure requirements are used to design two, in-vessel ALIPs and then two outside-vessel ALIPs. The two pumping systems are compared in terms of their geometries, performance characteristics, and impact on overall reactor design. It is found that the outside-vessel EMP design provides the same performance requirements and offers additional advantages compared to the in-vessel pump with only minimal reactor vessel modification and a slight drop in efficiency.

SIMMER modelling of accident initiation phase in sodium fast reactors

The SIMMER-III code was developed mainly for simulation of hypothetical reactor accidents after core melting, but can be applied also for accident initiation phase simulations in sodium and other liquid–metal-cooled fast reactors. New thermal hydraulic and neutronic simulation approaches and models have been developed for treatment of the initiation phase. Two examples of simulation of unprotected loss of coolant flow (ULOF) transients, in ESFR-SMART and FFTF reactors, are presented. It can be concluded that the SIMMER code has a large potential for application to initiation phase analyses.

Applicability of RANS models and pressure drop in edge subchannels for 19-pin wire-wrapped fuel bundle channel in CiADS

The accelerator-driven subcritical system has a strong transmutation ability and high inherent safety, and it is internationally recognized as the most promising long-life nuclear waste disposal device. This study involves the construction of a Visual Hydraulic ExperimentaL Platform (VHELP) for the purpose of evaluating the applicability of Reynolds-averaged Navier–Stokes (RANS) models and analyzing the pressure distribution within the fuel bundle channel of China initiative accelerator-driven system (CiADS). Measurements of thirty differential pressures in edge subchannels within a 19-pin wire-wrapped fuel bundle channel were obtained under different conditions using deionized water. The pressure distribution in the fuel bundle channel at Reynolds numbers of 5000, 7500, 10,000, 12,500, and 15,000 was simulated using Fluent. The results show that RANS models obtained accurate results, and the shear stress transport k-ω model provided the most accurate prediction of the pressure distribution. The difference between the results of the Shear stress transport (SST) k-ω model and experimental data was the smallest, and the maximum difference was ±5.57%. Moreover, the error between the experimental data and numerical results of the axial differential pressure was smaller than that of the transverse differential pressure. The pressure periodicity in axial and transverse directions (one pitch) and a relatively three-dimensional pressure measurements were studied. The static pressure fluctuated and decreased periodically as the z-axis coordinate increased. These results can facilitate research on the cross-flow characteristics of liquid metal-cooled fast reactors.

A novel friction factor model for wire-wrapped rod bundles

Wire-wrapped rod bundles are often considered for fast spectrum nuclear reactors, especially liquid metal cooled fast reactors (LMFRs). Accurate calculation of the pressure drops of these bundles is key for proper thermal–hydraulic design of the reactor. Existing models, while able to accurately predict bundle friction factor, are complex and can be difficult to use without utilizing a numerical implementation. In this paper, a new model for bundle friction factor is proposed based on exact laminar solutions of the Navier-Stokes equations for a simplified rod bundle geometry. These exact solutions can then be adjusted via geometric and empirical parameters to the prototypic geometry. The proposed model is able to calculate the bundle friction factor with an average absolute percent difference of 10.8 % compared to the available experimental data. This compares favourably to existing correlations, offering similar performance while minimizing the number of empirical coefficients and the overall effort required. The model is also valuable for determining a theoretical lower bound for bundle friction factor for a given geometry, with some data found to be lower (>10 %) than this bound. The overall methodology can also in principal be applied to other geometries where similar simplifications could be made.

Numerical study on inter-wrapper flow and heat transfer characteristics in liquid metal-cooled fast reactors

Inter-wrapper flow (IWF) is a typical thermal hydraulic phenomenon that exists in narrow gaps among the adjacent assemblies of a liquid metal fast reactor (LMFR) that can dissipate heat from assemblies and transfer heat among adjacent assemblies. Moreover, when blocked or reduced flow accidents occur, the IWF has a significant effect on reactor safety, especially reducing the peak cladding temperature. In this study, a computational fluid dynamics (CFD) model of the Karlsruhe liquid metal laboratory IWF experimental setup (called MYRRHA design) is built. The model contains three fuel assemblies with seven rods using lead–bismuth eutectic as working fluid. The new mesh generation method, correlation of turbulence model, and fluid–solid coupling heat transfer were considered in the CFD model. The established model was initially validated, and the simulation results were found to be in good agreement with experimental data. The influence of IWF on the thermal hydraulic characteristics within the rod bundle was revealed. The results indicated that the IWF rate can affect the heat transfer capability of the bundle under. The amount of heat removed increased by up to 10% with the IWF rate in a single bundle. The occurrence of IWF reduces the wall surface temperature and appropriately balances the temperature of adjacent bundles; however, it cannot effectively reduce the peak temperature of the cladding. The complex heat transfer characteristics and flow mechanism under the influence of IWF and wire mixing were revealed. The work presented in this paper represents an important step toward the IWF research on blocked flow scenarios.

Flow-Induced Vibration of Solitary Wire-Wrapped Cylinders in Axial Flow

Vibration of nuclear power plant components can cause fretting wear and fatigue that can eventually lead to component failure. Flexible, high-aspect-ratio components under flow, such as the wire-wrapped cylindrical fuel elements in a liquid metal-cooled fast reactor core, are particularly susceptible to vibration due to their low natural frequencies. The flow-induced vibrations experienced by such components tend to be random and of low amplitude and frequency; however, at critical flow velocities these components can experience self-excited, fluid-elastic instabilities that can lead to immediate failure. Such failures of critical reactor components, particularly those that act as fission product barriers, can lead to prolonged shutdowns of nuclear power plants and even to their permanent closure. Thus, a better understanding of the vibration response of wire-wrapped cylinders in axial flow is needed. This study details the development of a theoretical model that incorporates the effects of a helical wire wrap along a cylinder to understand its impact on the dynamic response of the cylinder under flow. This theoretical model is compared against experimental vibration data of varying geometries of solitary wire-wrapped cylinders in confined axial flow. The results of this study provide an improved knowledge of how a helical wire wrap can affect the dynamic response of a cylinder under flow.

Upscaling and downscaling the heat transfer process coupled with neutronic reflected core for sodium-cooled fast nuclear reactor

The heat transfer phenomena are crucial in nuclear reactors' design and safety analysis due to the feedback effects with the neutronic processes for power generation. Nuclear reactors are heterogeneous systems containing hundreds of thousands of fuel pins that exhibit power distribution through the space, and the temperatures between the coolant fluid and pins are different. This work analyzes the heat transfer process in liquid metal-cooled fast nuclear reactors with two upscaled energy equations based on the volume averaging method, which is widely applied for the analysis of transport in multiphase systems. The upscaled heat transfer model is coupled with a neutronic reflected core model including feedback effects from the nuclear fuel and liquid metal temperatures. The coupled mathematical models are employed within a called downscaling procedure, which represents a novel methodology with scopes beyond conventional problems in heat transport processes in nuclear reactors. The downscaling process allows us to increase the degree of resolution in the reactor core since this considers the scale of the fuel assembly and that of individual pins at the smallest scale, i.e. a single fuel pin surrounded by liquid metal. This point is crucial for analyzing hot spots in the reactor core.

Pressure drop and flow characteristics in partially blocked wire wrapped rod bundles

One of the fuel assembly designs considered for liquid metal-cooled fast reactors (LMFR) uses wires helically wrapped around fuel pins as spacers. The understanding of the coolant behavior in LMFR fuels under the influence of a partially blocked flow area is required for a correct safety assessment. Accumulation of debris or cladding deformation may generate a partial or total flow blockage of coolant channels at different locations. Unsteady Reynolds-Averaged Navier-Stokes (URANS) is applied using different turbulence models to predict the axial friction factor under the presence of blockages. Results are compared with experiments. The different turbulence models presented a variable performance for the blockage scenarios under study, finding k-∊ Realizable the recommended model for turbulent regime and k-ω SST the recommended model at transition regime. A Large Eddy Simulation was also implemented to simulate the case with the largest blocked area finding its results in reasonable agreement with the experiments.

A multiphysics simulation suite for liquid metal-cooled fast reactors

A new multiphysics simulation suite has been created to model Liquid Metal-cooled Fast Reactors (LMFRs). LUPINE: “LMFR Utility for Physics Informed Nuclear Engineering” is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified PN (SPN) neutron transport equations and multiphysics models to simulate thermal hydraulics and thermal expansion. Results of validation benchmarks are presented using the SPN solver and runtime results indicate that the SPN solver scales well for increasing SPN truncation order. To demonstrate the multiphysics coupling capabilities of LUPINE, the Advanced Burner Reactor (ABR) MET-1000 benchmark is modeled using coupled neutronics, thermal hydraulics, and thermal expansion models. It is shown that the SP3 method is sufficient to model the SPN effects in the ABR. The multiphysics models are showcased by calculating several multiphysics reactivity coefficients including: power defect, thermal expansion coefficient, Doppler coefficient, and Coolant Temperature Coefficient (CTC).

Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

Development of a simple model for estimating the design limit of core void reactivity to prevent re-criticality of MOX-fueled cores in liquid metal-cooled fast reactors

The mixed oxide (MOX)-fueled core for liquid metal-cooled fast reactors has such high core performance and favorable neutron economy as to achieve an average fuel burnup of over 90 GWD/t with high power density. However, traditional MOX-fueled cores which are the ones without specific design measures to reduce the positive void reactivity worth has the potential to exceed prompt criticality during hypothetical core disruptive accidents (CDAs). Should the prompt criticality be exceeded, the core materials would heat up until their temperatures would exceed their boiling points. As a result, a release of significant mechanical energy cannot be ruled out, imposing serious damage on the reactor vessel. The released mechanical energy, termed “energetics,” is dominated by a rapid insertion of positive reactivity due to a fuel–coolant interaction (FCI) during the initiating phase of an unprotected loss-of-flow (ULOF) event. Considering the mechanism of energetics generation, a simple model was developed to estimate the design limit of positive component of core void reactivity against a ULOF event for traditional MOX-fueled cores. This simple model is constructed to evaluate the design limit based on the relation between the allowable positive reactivity insertion rate defined by the core characteristics and that assumed by the effect of FCI phenomena. The validity of the model is discussed by comparison with SAS4A which is the CDA analysis code devoted to calculate the initiating phase.

VALIDATION OF WIMS11 FOR SMALL MODULAR REACTOR ANALYSIS

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.

FAST REACTOR MULTIPHYSICS AND UNCERTAINTY PROPAGATION WITHIN WIMS

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.

Upscaled heat transfer coefficients for a liquid metal-cooled fast nuclear reactor

In this paper, we calculate the heat transfer effective coefficients for a Liquid Metal-Cooled Fast Reactor. We upscaled the local energy equation of a fuel pin and its surrounding liquid metal (cooler fluid) to the whole fuel assembly representing the entire nuclear reactor core. Such an upscaling procedure was carried out within the volume averaging framework. The upscaled model contains two coupled energy equations, and jointly three closure problems are presented which assist to numerically compute the effective coefficients. The closure problems consist of boundary-value problems, whose solution must be carried out in a representative geometry of fuel assembly. The closure problems were solved to obtain the closure variables, with which the effective coefficients related to conductivity and convection heat transfer were obtained. The upscaled model results were compared with the more-intensive numerical solution of the pin-scale mathematical model. An acceptable agreement was found validating the mathematical structure of the upscaled model, the closure problems, and the definitions of effective parameters.