The Serpent Monte Carlo code has been in public distribution for 15 years, and has a large international user basis with both research and commercial applications. Serpent is currently developed as part of the Kraken multi-physics framework, which has dedicated capabilities for core-level reactor physics analyses. In Kraken, Serpent can be used either as a high-fidelity neutronics solver, or for generating homogenized group constants for the Ants nodal neutronics code. The neutron and photon transport modes in Serpent enable using the code also for various stand-alone applications beyond reactor physics, such as radiation shielding and fusion neutronics. This paper presents a review of the current status and capabilities of Serpent, corresponding to the latest release 2.2.1. The main features are introduced, with references to publications with more detailed methodological descriptions.

This work presents a high-fidelity pin-by-pin simulation approach for a NuScale-like Small Modular Reactor core during a rod ejection accident (REA). We coupled 3D Monte Carlo neutron transport (Serpent), subchannel thermal-hydraulic (SUBCHANFLOW) and fuel performance (TRANSURANUS) codes using the Interface for Code Coupling (ICoCo), which is part of the EU's Salome open source platform. To resolve fuel intra-assembly details, we simulated all the fuel rods and channels, subdividing them into axial slices and transferred calculated data between the codes using scalar fields saved in memory variables. Two different REA scenarios were modelled, and the behaviour of fresh-loaded cores with conventional UO2 fuel with Zr-4 cladding and accident tolerant fuel (ATF) materials, U3Si2 fuel with FeCrAl cladding, were analysed. In both scenarios, the control rod was ejected within 0.1 s, followed by a SCRAM after two seconds. In the first moderate scenario, the control rod ejection occurred at 75% of the nominal power, whereas in the second accident scenario, it occurred at hot zero power (HZP) conditions. In the first scenario, the power increase was around 25%, while in the HZP case it amounted up to 600% and 300% of the nominal power for the core loaded with UO2 and ATFfuel and cladding, respectively. Detailed calculations were conducted on a High-Performance Computer (HPC). The results demonstrated the robustness and flexibility of the coupled code system, providing full-core behaviour and rod-level safety parameters and predicting as needed during the safety analysis support of the licensing processes. This paper outlines the system setup, presents rod-level results and underlines the usefulness to assess the performance of SMR-cores loaded with different fuel types under various REA scenarios. In the scenarios considered, we did not observe significant fuel rod deformations, and the core loaded with ATF-fuel and cladding showed a large margin to melting.

The NEA Data Bank is an international reference centre for computer codes, nuclear and thermochemical data which has traditionally used simple file servers, and even DVDs, to deliver valuable content to end users across the globe. With the recent implementation of a self hosted GitLab system at the NEA it has enabled the Data Bank to streamline delivery, automate processes and testing, while empowering code owners and developers with a secure platform to collaborate and develop codes. In this paper we present the NEA Git- Lab system with some concrete examples of codes such as Kraken, PHITS and FISPACT-II taking advantage of many of the services and functionalities provided by GitLab. We illustrate methodologies on how to work effectively with third party software in a position of being a custodian of code, rather that a code owner, providing DevSecOps as a service.

This article contributes to the validation of the Kraken framework for VVER-1000 fuel cycle modelling. Four fuel cycles from the X2 VVER-1000 benchmark are simulated along with the start-up tests at the beginning of each cycle. Predicted boron letdown curves, power distributions and start-up test results for the four fuel cycles are compared to measured data from the benchmark as well as the results of similar reactor analysis tools. The comparisons to measured data show that Kraken performs well in the task. The accuracy of Kraken is comparable to other nodal diffusion based solvers.

The objective of this paper is to start the validation of the coupled code system of the nodal neutronics solver Ants and the system code TRACE. The process consists of two exercises of well-known VVER-1000 coolant transient benchmarks: a core-vessel simulation of a main steam line break transient of the V1000CT-2 benchmark and a full plant simulation of a main coolant pump trip of the Kalinin-3 benchmark. The Ants-TRACE results are compared against other code solutions in the V1000CT-2 benchmark and measurement data in the Kalinin-3 benchmark. The results show good agreement in both benchmarks, with the deviations ranging mostly within the measurement error or the range of deviations of other published solutions. The results confirm the correct implementation of the coupling and successful modeling of relevant phenomena in coolant transient events in a VVER-1000 reactor.

This paper defines a 3D full core neutronics benchmark which is based on the NuScale small modular reactor (SMR) concept. The paper provides a detailed description of the NuScale-like core, a list of ex-pected outputs, and a reference solution to the benchmark exercises obtained with the Monte Carlo code Serpent.The benchmark was developed in the framework of the Euratom McSAFER project and can be used for verification of computational chains dedicated to 3D full-core neutronics simulations of water cooled SMRs.The paper is supplemented with a digital data set to ease the modeling process.(c) 2023 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

The Kraken computational framework is a new modular calculation system designed for coupled core physics calculations. The development started at VTT Technical Research Centre of Finland in 2017, with the aim to replace VTT’s outdated legacy codes used for the deterministic safety analyses of Finnish power reactors. In addition to conventional large PWRs and BWRs, Kraken is intended to be used for the modeling of SMRs and emerging non-LWR technologies. The main computational modules include the Serpent Monte Carlo neutron and photon transport code, the Ants nodal neutronics solver, the FINIX fuel behavior module and the Kharon thermal hydraulics code, all developed at VTT. The core physics solution can be further coupled to system-scale simulations. In addition to development, significant effort has been devoted to verification and validation of the implemented methodologies. The reduced-order Ants code has been successfully used for steady-state, transient and burnup simulations of PWRs with rectangular and hexagonal core geometry. The Ants–Kharon–FINIX code sequence is actively used for the core design tasks in VTT’s district heating reactor project. This paper is a general overview on the background, functional description, current status and future plans for the Kraken framework. Due to the short history of development, Kraken has not yet been comprehensively validated or applied to full-scale core physics calculations. A review of previous studies is instead provided to exemplify the practical use.

The pin power reconstruction methodology for hexagonal geometry in the multigroup nodal neutronics program Ants is described and verified against full core Serpent Monte Carlo solutions in four VVER benchmarks. In order to provide the best possible reference solution, Serpent uses the same nuclear data libraries and simulation options both to generate the required group constants for Ants and to solve the full core reference solution. The root mean square of the relative differences in 2D pin powers between Ants and Serpent was between 0.83% and 1.11% in the four benchmarks, while the root mean square of the relative differences in the 2D intra-assembly peaking factor (Kk) was between 0.54% and 0.71%. Increasing the number of energy groups in the nodal solution tended improve the accuracy of the nodal solution as did the use of critical spectrum condensed group constant data instead of infinite spectrum group constants.

In this work the nodal neutronics code Ants is coupled with the subchannel code SUBCHANFLOW and the fuel behavior code SuperFINIX within VTT’s Kraken framework in order to evaluate thermal margins. The coupling is implemented using Cerberus which is the multi-physics driver of Kraken. The capabilities of the new coupled code system are demonstrated by modeling the first operating cycle of BEAVRS. Calculated and measured boron concentrations are compared and selected pin-by-pin results are presented at 3 points during the operating cycle. In addition, the modularity of the Kraken framework is highlighted by modeling the depletion of a 3D single fuel assembly with both Ants and Serpent based code systems. This capability can be used to assess the accuracy of nodal neutronics vs continuous energy Monte Carlo in the estimation of thermal margins. Finally, some possible topics for future work are introduced.

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

This work presents the results for a coupled neutronic-thermalhydraulic-thermomechanic pin-level depletion calculation of a PWR fuel assembly using Serpent2-SUBCHANFLOW-TRANSURANUS. This tool is based on a semi-implicit depletion scheme with pin-by-pin feedback, mesh-based field exchange and an object-oriented software design. The impact of including fuel-performance capabilities is analyzed, with focus on high-burnup effects. The treatment of the Doppler feedback to the neutronics is examined as well, in particular the use of radial fuel-temperature profiles or radially averaged values.

In 2019 the government of Finland made a decision to phase out of coal in energy production in a period of just ten years. The Finnish energy sector is currently looking for alternative technologies to replace coal-fired power plants, used especially in large cities for producing electricity and low-temperature heat for the local district heating network. The production of low-carbon electricity is expected to grow within the near future, along with the commissioning of the Olkiluoto 3 nuclear power plant and increasing share of wind power. The lost district heating capacity, however, is more difficult to replace. To anticipate the transition, municipal energy companies have turned their attention to clean alternatives, including nuclear energy. In an effort to meet the government climate goals, VTT Technical Research Centre of Finland has launched a project to design a small, simplified and passively safe PWR for district heating applications. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120 degrees C temperature. The design combines conventional LWR technology with an innovative containment function, capable of decay heat removal without any mechanical moving parts. The reactors can be constructed partially or fully underground, or retro-fitted into an existing boiler plant. This paper presents an overview of the pre-conceptual reactor design, together with some general background on district heating reactor technology. More detailed design and safety analyses are provided in two separate papers at this ICONE-28 conference.

The continuous improvement in nuclear industry safety standards and reactor designers’ and operators’ commercial goals represent a push for the development of highly accurate methodologies in reactor physics. This fact, combined with the availability of vast computational resources, allowed the development of a wide range of coupled state-of-the-art neutronic-thermal-hydraulic calculation tools worldwide during last decade. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes, suitable for high-fidelity calculations for PWR and VVER reactors. This work presents the results for a pin-by-pin coupled burnup calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel thermal-hydraulics code SUBCHANFLOW (SCF, developed by KIT, Germany) for two different VVER-type fuel assembly types. For such purpose, a recently refurbished master-slave coupling scheme is considered, which provides several new features such as burnup and transient calculations capabilities for square and hexagonal geometries. Main aspects of this coupling are presented for this burnup case, showing some of the capabilities now available. On top of that, the obtained global results are compared with available published data from a similar high-fidelity approach for the same FA design, showing a good agreement. Finally, a brief analysis of the main resources requirement and main bottlenecks identification are also included. The results presented here provide valuable insights and pave the way to tackle the final goals of the McSAFE project, which includes full-core pin-by-pin depletion calculation within a fully coupled MC-TH approach.