Small Modular Reactor Concept Research Articles

A risk-informed and performance-based approach to defining prevention and mitigation

The terms “prevention” and “mitigation” are often used to define safety functions in nuclear power plants and other complex engineered systems. For existing light water reactor (LWR) nuclear plants, these terms are typically used in the context of preventing core damage and mitigating the consequences of a severe accident. Balancing the reliance on prevention and mitigation is often stated as an important principle of defense-in-depth. To apply these terms to formulate design requirements for advanced reactors, it is necessary to define specifically what is to be prevented and mitigated because the concept of core damage is not generally applicable to advanced non-LWR concepts and designs. Advanced non-LWR reactor concepts use different fuels, moderators, coolants and different strategies for achieving the functional containment of radionuclides. The adoption of the small modular reactor concept also leads to plants with many small reactors. This paper discusses a risk-informed and performance-based definition of prevention and mitigation that can be applied to any reactor concept or design. It was developed as part of the Licensing Modernization Project (LMP) which has introduced a new approach to developing a safety case to support design and licensing. A new way of thinking about prevention and mitigation was needed to formulate advanced reactor design and special treatment requirements for structures, systems and components (SSCs) that could be applied to a wide diversity of designs. The LMP approach discussed in this paper includes criteria to identify safety significant SSC prevention and mitigation functions and uses SSC reliability and capability targets to inform the selection of design and special treatment requirements and to support the evaluation of defense-in-depth adequacy. These prevention and mitigation functions are defined in the context of event sequences that comprise the licensing basis events that anchor the safety case for the reactor operating license.

Application of the passive turbocharger to PRHRS in integral PWR type SMR with an sCO2 power cycle

One of the potential design concepts for small modular reactor(SMR) is coupling an integral pressurized water reactor(IPWR) with a supercritical CO2(sCO2) power cycle to reduce the plant footprint. However, the passive safety systems of this SMR concept have not been well established compared to steam power cycle based SMRs. In this study, a passive turbocharger utilizing the unique characteristics of sCO2 is proposed as a way to improve the passive safety of sCO2 power cycle. This is the concept of installing a turbine-compressor set at the connection of passive residual heat removal system(PRHRS) to effectively remove the high level of residual heat in the early stages of an accident rather than relying only on natural circulation cooling. Based on the findings, it can be concluded that proposed sCO2 turbocharger concept has the potential to enhance performance of PRHRS and reduce the height of natural circulation loop leading to smaller volume.

CESAR experiment: Sodium boiling during Transient OverPower

In the framework of nuclear revival under the constraint of global warming and in the context of the emergence of new small modular reactor concepts, sodium cooled fast breeders seem to be a good candidate for maximizing the energy yield of the fuel and burning actinides. In the case of rapid reactivity insertion, characterizing transient sodium boiling is essential, as it can lead to dry out and rupture of fuel pins, but also as it has an impact on core reactivity. As sodium is a metal with a low Prandtl number, its boiling behavior is different from that of water. In order to provide insights and data on boiling of sodium during transient overpower, past experimental results obtained in the frame of the CESAR program (Caracterisation de l’Ebullition du Sodium en Accident de Réactivité) performed at CEA Grenoble in the 70th are described and main learnings are given. The CESAR test loop and experimental procedure are described, and results of seven transient tests with different inlet flow rates, temperatures, pressures and transient powers are described and interpreted on the basis of a comparison with single-phase simulations. The phenomenon of sodium superheat at boiling onset is studied. These data, which are rare in the literature, can be used to validate simulation tools.

Multiscale analysis of the boron dilution sequence in the NuScale reactor using TRACE and SUBCHANFLOW

In the ‘High-Performance Advanced Methods and Experimental Investigations for the Safety Evaluation of Generic Small Modular Reactors’ McSAFER H2020 European project the main aim pursued is to advance the research in safety analysis methods for water cooled Small Modular Reactors (SMRs). Hence, several light-water-cooled SMR concepts were selected along with the corresponding accident scenarios to be modelled to evaluate the potential application of advanced modeling tools based on the coupling of different codes and their benefits and drawbacks. In this study, the simulation of the postulated boron dilution scenario caused by the malfunction of the Chemical and Volume Control System (CVCS) in the NuScale reactor is performed using different modeling tools and approaches. More in particular, a TRACE-1D, a TRACE-3D and a TRACE-3D/SUBCHANFLOW (SCF) models have been developed to perform this analysis. Two cases have been carried out in this work: the first case is based on the calculation of the boron concentration evolution within the Reactor Coolant System (RCS) of NuScale to compute the time to loss of shutdown margin (SDM), and secondly, a best estimate calculation is performed to evaluate the plant response considering the reactivity feedbacks and the performance of the safeguards and safety systems of NuScale according to the public literature. These investigations have demonstrated that the applied modeling approaches and tools can capture the physics of the problem providing less conservative results than those obtained using the analytical methods presented in the DCA report of NuScale. In addition, it has been demonstrated that the NuScale design is robust against the consequences of the boron dilution transient and that no relevant asymmetrical effects appears due to the singular arrangement of the Helically Coiled Steam Generators (HCSGs) surrounding the riser.

Future nuclear fission reactors – uncertainties, the effect of parameter choice and an application to small modular reactor concepts

Abstract. Today, steep declines in the generation costs of renewable energy systems, particularly solar photovoltaic and wind, combined with a recent spur in storage and flexible technologies mainly driven by batteries and renewable-energy-generated hydrogen are driving a paradigm shift in energy systems across the world, and renewable energy dominates investments in installed electricity generation systems (Ram et al., 2022). In contrast, in 2022 USD 4.8 billion (ca. 21 %) of the total spending in research and development expenditures by countries within the Organisation for Economic Co-operation and Development (OECD) was still spent on nuclear development alongside the current discussion in recent literature about the role of new nuclear developments in combating the recent energy crisis and climate change (Asuega et al., 2023; Ingersoll et.al., 2020; IEA, 2023). Since predicting future costs of technologies not yet developed is a complex exercise, we start this paper with a comparison of two different approaches from production theory and show that they have a theoretically equal structure (Steigerwald et al., 2023). This we then apply to estimate a range of potential production costs for 15 so-called small modular reactor (SMR) concepts for which sufficient data are available. These include light-water, high-temperature and fast reactors. In a third step, we apply the Monte Carlo method to benchmark the cost projections assumed by the manufacturers by varying production parameters in simulations of the net present value (NPV) and the levelized cost of electricity (LCOE). Besides this, we test whether the differences between the manufacturer estimates and ours differ between technology families of SMR concepts and apply a sensitivity analysis. This contributes to an intensifying debate in the literature on the economics and finance of SMR concepts (Steigerwald et al., 2023). In a final step we compared our results with current costs of renewable energy technologies and could conclude that none of the tested concepts is able to compete economically with existing renewable technologies, not even when taking their variability and necessary system integration costs into account. We conclude that any technology foresight has to take the case specifics into account as much as possible, including technological and institutional specifics; this also holds for SMR concepts.

Uncertainties in estimating production costs of future nuclear technologies: A model-based analysis of small modular reactors

Predicting future costs of technologies not yet developed is a complex exercise that includes many uncertain parameters and functional forms. In that context, small modular reactor (SMR) concepts that are in a rather early development stage claim to have cost advantages through learning effects, standardized design, modularization, co-siting economies, and other factors, such as better time-to-market even though they exhibit negative economies of scale in their construction costs due to their lower power output compared to conventional nuclear reactors. In this paper, we compare two different approaches from production theory and show that they have a theoretically equal structure. In the second step, we apply these approaches to estimate a range of potential construction costs for 15 SMR projects for which sufficient data is available. These include water cooled, high temperature, and fast neutron spectrum reactors. We then apply the Monte Carlo method to benchmark the cost projections assumed by the manufacturers by varying the investment costs, the weighted average cost of capital, the capacity factor, and the wholesale electricity price in simulations of the net present value (NPV) and the levelized cost of electricity (LCOE). We also test whether the differences between the manufacturer estimates and ours differ between technology families of SMR concepts and apply a sensitivity analysis. Here we contribute to an intensifying debate in the literature on the economics and finance of SMR concepts. The Monte Carlo analysis suggests a broad range of NPVs and LCOEs: Surprisingly, the lowest LCOE is calculated for a helium-cooled high-temperature reactor, whereas all of the light water reactors feature higher LCOEs. None of the tested concepts is able to compete economically with existing renewable technologies, not even when taking their variability and necessary system integration costs into account. The numerical results also confirm the importance of the choice of production theory and parameters. We conclude that any technology foresight has to take as much of the case specifics into account, including technological and institutional specifics; this also holds for SMR concepts.

Computational model for thermohydraulic analysis of an integral pressurized water reactor with mixed oxide fuel (Th, Pu)O2

The use of advanced generation III+ and IV nuclear reactors, and their applications, has become important, seen as a means capable of contributing to the global transition to more sustainable, affordable and reliable energy systems. This technology, which could be integrated into future carbon-free electric power generation systems with high proportions of different renewable energy sources, includes Small Modular Reactors (SMR). There are about 100 different proposed projects of Generation III+ and IV, of which about 50 are SMR concepts, in various stages of development and of different types of technologies. Other important issues for achieving the long-term sustainability of nuclear energy are the proper use of its fuel sources and the improvement of nuclear waste management. Therefore, fuels based on a mixture of oxides have been used successfully in several countries. In addition, the incorporation of thorium-based fuel is a current challenge for the new designs of advanced reactors. The present paper focuses on the analysis of a small modular integral pressurized water reactor (iPWR) with Thorium-Uranium Oxide (Th-U MOX) mixtures. A thermohydraulic model is developed using the Ansys CFX program, which allows the calculation of the temperature distribution in the section where the highest power is produced within the SMR IPWR core (critical section). The temperature distributions in the fuel, clad and coolant were calculated with the objective of verifying that they were within the safety limits.

Small Modular Reactors: Learning From the Past

Abstract Most emerging small modular reactor (SMR) designs resemble older reactors that were designed in the early days of the nuclear technology. Experience with operating these old reactors can contribute to the licensing of new SMRs, by showing that some safety concepts were already proven, and their viability was demonstrated. This paper shows examples of older reactors in each of the reemerging SMR concepts of integrated pressurized water reactors, high temperature gas cooled reactors, molten salt reactors, and liquid metal cooled fast reactors. The Canadian experience with the WR-1 organic cooled reactor is also discussed to examine whether it can inform the development on an organic cooled SMR.

Demonstration of a coupled computational fluid dynamics approach for modelling non-water cooled small modular reactors

This paper provides an overview of the capabilities of a new toolset that is being developed at Canadian Nuclear Laboratories for modelling advanced non-water cooled small modular reactor concepts. Throughout the study, a technology agnostic approach was adopted. The integrated toolset includes computational fluid dynamics, neutronics and system thermalhydraulics codes to handle a range of multi-physics capabilities. The toolset has been applied to two different non-water cooled SMR concepts: molten salt and prismatic gas-cooled reactors. A step-wise approach was used in which the models were first tested using standalone codes, followed by execution of coupled-code simulations. This paper demonstrates the application of the toolset to gas-cooled and molten-salt reactor concepts, including transient simulations using the CFD code STAR-CCM+ coupled with the Monte Carlo neutronics code Serpent 2 and with the system thermalhydraulics code RELAP5-3D. The scenarios considered include a step reactivity insertion and a pump trip for the molten salt reactors. Overall, the results obtained so far demonstrates the potential of the developed methodology at CNL in simulating the non-water cooled concepts.

Integral PWR-Type Small Modular Reactor Developmental Status, Design Characteristics and Passive Features: A Review

In recent years, the trend in small modular reactor (SMR) technology development has been towards the water-cooled integral pressurized water reactor (iPWR) type. The innovative and unique characteristics of iPWR-type SMRs provide an enhanced safety margin, and thus offer the potential to expand the use of safe, clean, and reliable nuclear energy to a broad range of energy applications. Currently in the world, there are about eleven (11) iPWR-type SMRs concepts and designs that are in various phases of development: under construction, licensed or in the licensing review process, the development phase, and conceptual design phase. Lack of national and/or internatonal comparative framework for safety in SMR design, as well as the proprietary nature of designs introduces non-uniformity and uncertainties in regulatory review. That said, the major primary reactor coolant system components, such as the steam generator (SG), pressurizer (PRZ), and control rod drive mechanism (CRDM) are integrated within the reactor pressure vessel (RPV) to inherently eliminate or minimize potential accident initiators, such as LB-loss of coolant accidents (LOCAs). This paper presents the design status, innovative features and characteristics of iPWR-type SMRs. We delineate the common technology trends, and highlight the key features of each design. These reactor concepts exploit natural physical laws such as gravity to achieve the safety functions with high level of margin and reliability. In fact, many SMR designs employ passive safety systems (PSS) to meet the evolving stringent regulatory requirements, and the extended consideration for severe accidents. A generic classification of PSS is provided. We constrain our discussion to the decay heat removal system, safety injection system, reactor depressurization system, and containment system. A review and comparative assessment of these passive features in each iPWR-type SMR design is considered, and we underline how it maybe more advantageous to employ passive systems in SMRs in contrast to conventional reactor designs.

Comparisons of Reduced Moderation Small Modular Reactors With Heavy Water Coolant

This paper presents the comparison of two reduced moderation small modular reactor concepts with heavy water coolant. Two reduced moderation small modular reactors, RMSMR-Th and RMSMR-MOX, are proposed for the sustainable utilization of nuclear resources. The design concepts are established on modifications of the well-experienced pressurized water reactor technology. Tightly-packed lattice and heavy water coolant are employed to yield a hard neutron spectrum, which proved advantageous for increasing the conversion ratio as well as lowering the burnup reactivity swing between beginning of cycle and end of cycle. Thorium-uranium dioxide fuel and MOX fuel are compared using the same core arrangement, and the small modular reactor concept is adopted to reduce void coefficients. Radial blanket region and axial blanket region are adopted to enhance the fissile breeding and a three-zone fuel configuration is adopted to flatten the power distribution. Core burnup calculations were carried out to investigate the available cycle length, the conversion ratio, the power peaking factors, reactivity coefficients, etc. Light-water based thermal-hydraulic models were employed to examine the safety features of the concepts. Numerical simulations indicate that both RMSMR-Th and RMSMR-MOX can sustain the power generation of 100MWe by 7 years without refueling. Compared with thorium-uranium dioxide fuel, MOX fuel is helpful in reducing the burnup reactivity swing, increasing the conversion ratio and increasing the mDNBR value. However, the positive void coefficient becomes a serious problem making RMSMR-MOX less attractive than the RMSMR-Th concept.

Development of the universe based geometry criticality search capability in RMC and analysis of NHR200-II reactor

Criticality search is one of the important aspects to perform accurate lifecycle analysis of nuclear reactor in Monte Carlo code, in which the geometry criticality search is particularly important for some kinds of reactor systems, such as BWR and so on, to use control rods to adjust the criticality and power in lifecycle. To solve the Monte Carlo geometry criticality search problem correctly and efficiently, the Universe Based Geometry Criticality Search (UBGCS) capability along with the Non-Converged Sub-Iteration (NCSI) acceleration method is newly developed and implemented in reactor physics Monte Carlo code RMC. The UBGCS capability relies on the multi-layer universe geometry description structure in RMC and could perform eigenvalue criticality search not only to single or several surfaces, but also complex cells and universes combination. By using the NCSI method, the RMC’s UBGCS capability could achieve accurate and efficient Monte Carlo criticality search for complex problems.Based on the newly developed UBGCS capability and other advanced techniques in RMC, the lifecycle analysis of a nuclear heating small modular reactor concept NHR200-II is taken out to have a better understanding of the reactor physics characteristics of this reactor. NHR200-II is a new 200 MWt commercial size heating reactor using control rods as power regulating system developed by INET, Tsinghua University according to the new energy and environment requirements in China. In this article, the first detailed Monte Carlo model of NHR200-II has been build including different types of complex fuel assemblies, cruciform control rods, full core structures and so on. The cold zero power and hot zero power state of the NHR200-II reactor are analyzed based on the detailed model. Also using the Monte Carlo reactor physics/thermal hydraulics and UBGCS capabilities in RMC, the full core pin-by-pin burnup calculation of NHR200-II are performed and analyzed.

CFD Simulations of Molten Salt Reactor Experiment Core

At present, no clear guidelines exist for modeling non-water-cooled small modular reactors (SMRs) despite the rising need for high-fidelity simulation tools to support regulators and the industry. Most SMR concepts currently under the Canadian prelicensing review adopted non-water-cooled–reactor technologies [molten salt reactor (MSR), gas-cooled reactor, and liquid metal–cooled reactor] that are new for Canada. There is a need for a modeling tool set that is broadly applicable for the assessment of advanced technologies used in SMRs. Computational fluid dynamics (CFD) can be used in performance evaluation and safety analysis of non-water-cooled SMRs for modeling three-dimensional (3-D) fluid flow and heat transfer in geometries of arbitrary complexity without resorting to geometry-specific empirical correlations. This study investigates the capabilities of existing models within a commercial CFD code to simulate the flow and heat transfer characteristics in a MSR configuration. The Oak Ridge National Laboratory (ORNL) Molten Salt Reactor Experiment (MSRE) configuration was simulated in this study using a stand-alone CFD approach, and CFD predictions were assessed with ORNL data. Intricate geometry details within the MSRE core were included in the computational model to study the associated geometric effects. The results obtained in this study showcased the ability of CFD to predict 3-D effects within the computational domain especially at the lower plenums. The predicted trends for the temperature rise in the fuel and moderator within the core were in good agreement with the ORNL data. The results presented in this paper constitute the first step in developing Canadian Nuclear Laboratories’ capability for CFD modeling of non-water SMRs.

Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EDF) Research and Development (R&D) Center, in the Pericles department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMR) with the modular accident analysis program version 5 in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a stepwise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of the lower head geometry, vessel steel ablation, and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. The results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identify modeling improvements.

TRANSPORTATION OF SMALL MODULAR REACTOR MODULES: WHAT DO THE EXPERTS SAY?

One of the key characteristics of small modular reactors (SMRs), as their name emphasised, is the modularization. Modularization implies factory production, which in turn implies transportation of large, heavy, complex and fragile modules from the factory to the site. Various vendors and organisations are developing several SMR concepts and designs, but there are extremely limited information about the crucial element of modules transportation. Conversely, in other industries (e.g. Oil & Gas), the experience on modules transportation is much greater. This paper provides a structured analysis for the knowledge transfer from the general literature (i.e. other major infrastructure) to the SMR world. Firstly, the paper provides a summary of the literature about transporting large modules. In the second part, the paper presents and discusses the results of a series of interviews with transport industry experts about large modules transportation. The third part provides a summary of the findings and the key takeaways.

Small modular high temperature reactor optimisation part 2: Reactivity control for prismatic core high temperature small modular reactor, including fixed burnable poisons, spectrum hardening and control rods

This article investigates a high temperature prismatic core small modular reactor concept based on the U-Battery. To achieve a long operational time without reloading of the fuel, a high fissile fuel load is required, to provide the required excess reactivity which needs to be compensated to allow safe operational states in all stages of the life of the core. This is achieved through the means of reactivity control methods. In this article we deploy a design methodology for controlling the reactivity in a design based on the U-Battery.The initial excess reactivity within the core was 2000 pcm, implying that the reactivity control systems were required to bring this to unity. To reduce this reactivity, we investigated three main reactivity control methods. Initially the control rod designs which is kept constant. Then a full range of fixed burnable poison designs are evaluated. The final reactivity control method investigated the process of flux spectrum hardening by changing the moderation within the core.This work managed to identify that maximum full power days without any reactivity control as 1364 days. The fixed burnable poisons deployed reduced this lifetime by 62 days due to a reduced neutron economy. However, by removing part of the central reflector the flux spectrum hardens (increases the proportion of fast neutrons) this provides additional bred plutonium provided an additional 62 full power operating days. The meant that the overall design saw no lifetime reduction due to reactivity control devices, without including the use of control rods.Due to the implemented reactivity control systems, the highest point of reactivity has been identified as 1.03, this meant that the control rod position required to reduce this reactivity is limited to the maximum of 50% insertion. Identifying the power profile at this position showed additional loading at the fuel below this point, where additional thermal distribution work will be undertaken.

Extension of the sub-channel code ANTEO+ to the mixed convection regime

Liquid Metal cooled Fast Reactors, and notably Heavy Liquid Metal cooled systems, are promising options for achieving, in a relatively short term, the various advantages brought about by Small Modular Reactors (SMRs). To increase safety, natural circulation in the primary circuit to cool the core, even at rated power, is also considered, possibly bringing the core thermal exchange dynamics inside the mixed convection regime. To address, at the outset of the design process, thermal-hydraulics aspects, suitable tools must be employed, like the sub-channel code ANTEO+. ANTEO+ was, however, specifically built for the forced convection regime, trying to maximize modeling efficiency inside that regime; to make it applicable also for these SMRs concepts, its validity domain has been extended to the mixed convection regime. In this work, such extension process is outlined, presenting the new code structure, including the solved set of equations and models in the new anticipated application domain. A thorough validation for sub-channel and clad outer temperatures is also presented, confirming the ability of ANTEO+ in reproducing experimental data, homogeneously, in its new anticipated validity domain, with a relatively high degree of accuracy when compared to a reference tool like COBRA-IV-I-MIT. The need for Heavy Liquid Metal-cooled experiments dedicated to validation for sub-channel tools has also been highlighted so to greatly enhance confidence in the final code accuracy assessment, ultimately easing its use in the design phase.

Advanced Nuclear Open Air-Brayton Cycles for Highly Efficient Power Conversion

Nuclear heated open air-Brayton cycles have been investigated both as topping cycles for combined cycle Brayton-Rankine cycles and as standalone recuperated Brayton cycles. The peak turbine inlet temperature chosen for the analysis was 933 K for a range of Generation IV molten salt reactors or lead cooled reactors. A baseline power level of 25 MW(electric) was chosen to be representative of some of the small modular reactor concepts being considered. Extensions to higher temperatures and power levels were evaluated. Thermal efficiencies in the 45% to 46% range can be achieved by both the combined cycle systems and the recuperated systems, though the combined cycle systems achieve about a 1% to 1.5% improvement over the recuperated systems.The nuclear heated open air-Brayton systems have several advantages over current light water reactor and other Generation IV systems. The analysis demonstrates that the cycle thermal efficiencies are higher than other proposed systems. The gas turbine hardware is readily available over a broad range of power levels. And both the combined cycle and recuperated systems require significantly less circulating water for waste heat rejection than any other proposed systems.

Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of historical, full scale operating experience, a general uncertainty and sensitivity analysis methodology was developed to help aid SMR designs through this process. Uncertainty was quantified through the empirical cumulative distribution function (ECDF) created from a desired data set. Linear regression techniques were applied to measure sensitivity. This methodology was demonstrated through the thermal hydraulic steady-state subchannel analysis of the MASLWR concept using RELAP5-3D Version 4.0.3 and VIPRE-01 Mod 2.2.1. Twelve uncertain input parameters were selected. System response uncertainty in the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel temperature, and maximum clad temperature was evaluated. These figures were shown to satisfy U.S. NRC regulatory requirements for steady state operation at the 95 percent probability and 95 percent confidence level under the evaluated conditions. Sensitivity studies showed input parameters affecting local power generation within the core had a large influence on MDNBR, maximum fuel temperature, and maximum clad temperature.

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 (100W) 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 keff of 385pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of keff by 13pcm (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 −33pcm from its 169pcm worth simulated in Serpent.