In recent years, there has been renewed interest in molten salt reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, SPECTRA and SAM are two system analysis codes that have been enhanced to include MSR-specific modeling capabilities, including delayed neutron precursor drift and modified point kinetics equations.This paper discusses the efforts taken to verify and validate these features. A standard MSR system test problem was developed to verify and demonstrate the capability of SPECTRA and SAM on the MSR transient simulation. Sixteen transients were simulated. The results obtained from SPECTRA and SAM show good agreement. The Molten Salt Reactor Experiment transient experiments were reviewed and selected to validate the SPECTRA and SAM codes. The experiments included pump startup and coastdown tests at zero power, reactivity insertion tests at different power levels, frequency tests, and a natural convection test. The simulation results from SPECTRA and SAM show good agreement with the experimental data.

High Temperature Gas-cooled Reactor (HTGR) is recognized by the international nuclear power industry as a safe type of reactor. The modular pebble-bed high temperature gas-cooled reactor uses reactor modules with relatively small thermal power between 200-600MWt, utilizing the high-temperature resistance that can be achieved by coated fuel particles. Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed a commercialscale 200 MWe High Temperature gas-cooled Reactor Pebblebed Module project (HTR-PM). The decay heat can be removed by natural mechanism, such as thermal radiation, heat conductivity, natural convection, thereby guaranteeing the safety of reactor under accident and eliminating the possibility of core melting. A passive Reactor Cavity Cooling System (RCCS) has been designed for HTR-PM to ensure the safety of the Reactor Pressure Vessel (RPV) and reactor cavity under normal operation and accident conditions. RCCS transfers heat from pressure vessels to the atmosphere through thermal radiation, heat conduction, and natural convection, and the system utilizes passive mechanism with high safety and reliability. The RCCS of HTR-PM has been designed as three independent sets. According to the preliminary design, two sets work can guarantee the safety of the PRV and reactor cavity. In this paper, the working performance of RCCS under depressurized loss of forced cooling (DLOFC) accident has been analyzed by INET with TIN-CAVCO code, and also by NRG with SPECTRA code. The analysis results of these two codes are in good agreement, proving the applicability of both codes in analysis of the HTGR as well as its RCCS. During the accidents, the maximal fuel temperature keeps below the limitation of 1620 degrees C. The results also prove the excellent performance of the RCCS. The RCCS meets the design requirements, that is the normal operation of two sets can effectively ensure the safety of the pressure vessel and the reactor cavity during the accident. Even if only one set works normally, the decay heat still can be carried out effectively and the RPV temperature will not exceed the limitation of 425 degrees C.

With the increase in computational power and capacity, and the advancements being made in numerical modeling, it has become possible to model the various physical phenomena that take place in nuclear reactors with more and more detail and accuracy. This includes phenomena related to structural mechanics, fluid dynamics and reactor physics amongst others. Additionally, there has been an increased interest to simulate these combined, interacting phenomena simultaneously by coupling various numerical tools. This coupling of different codes is currently a topic high on the research and development agenda of the international nuclear community. It is also a focus point within the research done at NRG in the national PIONEER research program funded by the Dutch ministry of economic affairs. This focus has resulted in two branches of research at NRG: multi-scale modeling of the complete primary system of a nuclear reactor by coupling a 3D Computational Fluid Dynamics (CFD) code with a 1D System Thermal Hydraulics (STH) code, and multi-physics modelling through the coupling of dedicated and advanced thermal-dynamics, structural mechanics and reactor physics solvers. The paper presents simulation results of both branches applied to fast reactors. As both fields of research require the coupling of codes, it has led to the creation of an independent, external, Fortran-based coupling tool named myMUSCLE (MultiphYsics MUltiscale Simulation CoupLing Environment) that arranges the efficient and robust coupling of the different codes. The paper presents the proof-of-principle and first validation of the myMUSCLE tool under development. Additionally, results obtained with MUSCLE-Foam will be presented, which is a new multi-physics code being developed at NRG that includes several reactor physics as well as structural mechanics solvers.

The multi-year research program carried out by NRG and funded by the Dutch ministry of economic affairs and climate is called ‘Program for Innovation and cOmpetence development for NuclEar infrastructurE and Research’ (PIONEER). The program comprises seven themes, i.e. long term operation, nuclear modelling and simulation, nuclear safety and compliance, fuels & materials, radioactive waste management, radiation protection, and innovative nuclear systems. One of the pillars in the theme of innovative nuclear systems is fast reactor research, particularly in the field of thermal hydraulics. This paper provides an overview of all fast reactor thermal hydraulics activities in the program, covering development and validation of System Thermal Hydraulics (STH) and 3D (engineering as well as high-fidelity) Computational Fluid Dynamics (CFD) codes and simulation approaches. Applications range from fundamental turbulent heat transport to core, pool and system thermal hydraulics. With the recent improvements in computational infrastructure and power, also further developments of multi-scale and multi-physics computational approaches are being integrated in the PIONEER program. A generic coupling tool ‘myMuscle’ is under development which is introduced in this paper. Recent results and current developments are presented together with an outlook for the results to be expected at the end of the current multi-year program and beyond.

In the framework of the OECD/NEA/CSNI/WGAMA, an activity on the "Status report on thermal-hydraulic passive systems design and safety assessment" has been conducted. Within this activity, a benchmark exercise, based on the experimental data developed in the full scale PERSEO (in-Pool Energy Removal System for Emergency Operation) component separate effect test facility, built at SIET (Piacenza, Italy), has been proposed and carried out. An "OPEN" benchmark exercise, hosted by ENEA, has been conducted. Twelve results from eleven Organizations were submitted. PERSEO is a full-scale separate effect test facility designed to study a new passive decay heat removal system operating in natural circulation. Test 7 is a full pressure test (7 MPa) and investigates the system stability and the system operation. The accuracy of the calculated results has been evaluated both qualitatively and quantitatively. The latter has been conducted adopting the Fast Fourier Transform Based Method. The present paper summarizes the main features of the PERSEO facility and Test 7 and discusses the main results and outcome of the benchmark exercise.

Increasing the computational power enables the nuclear community to combine existing knowledge on the variety of different physical phenomena that take place in reactors and to develop tools that can simulate these combined, interacting phenomena simultaneously. This includes phenomena related to structural mechanics, fluid dynamics, and reactor physics among others. Coupling different codes developed specifically for the analysis of separate phenomenon is currently a topic high on the research and development agenda of the international community.Based on the experience of successfully computing the dissymmetric benchmark in the Phenix reactor by coupling the system thermal-hydraulic (STH) code SPECTRA to the computational fluid dynamics (CFD) code CFX in the H2020 SESAME project, the Nuclear Research and Consultancy Group (NRG) is currently developing the code-coupling tool myMUSCLE: MultiphYsics MUltiscale Simulation CoupLing Environment. MyMUSCLE is an independent, external, Fortran-based code that arranges the efficient and robust coupling of different codes. It aims at being flexible with respect to the codes being coupled, i.e., commercial and open-source codes, while having a single coupling tool that enhances quality assurance. It is currently set up to couple SPECTRA as a STH code to CFX, Fluent, STAR-CCM+, or OpenFOAM as a CFD code. This paper presents the proof of principle and first verification of the myMUSCLE tool under development by applying it to multiscale thermal-hydraulic applications.First, a flow through a pipe is modeled as proof of principle for explicit coupling at a single coupling interface. Second, in preparation for modeling liquid-metal-cooled fast reactors, a piping system with a pool with natural convection is modeled. The results of the multiscale calculations show good agreement among the different coupled CFD codes. Finally, the preparations for simulating the TALL-3D experiment, used for generating data for validation of simulation tools for liquid-metal pools, are presented.

The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI+ project. The GEMINI+ reactor is a prismatic block-type High -Temperature Gas-Cooled Reactor (HTGR). Within tasks devoted to safety, an analysis of air ingress scenario was performed by NRG with the thermal-hydraulic system code SPECTRA, while UJV and NCBJ worked with the integral system code MELCOR. Two air ingress scenarios were analyzed: Design Basis Accident (DBA) scenario. The air ingress scenario selected as DBA is a 65 mm break of a Helium purification line on top of the Steam Generator, Large air ingress scenario. The air ingress scenario selected is guillotine break of the coaxial gas duct. The DBA scenario is characterized by a very long period with practically stagnant gas in the primary system after the initial depressurization. However, very small gas flows through the break due to counter flow and diffusion in the break region (gas velocities of 10-2 m/s) are difficult to model with system codes. Therefore the break models in the system code were calibrated by performing CFD simulations. Furthermore, a model of gas mixing by diffusion was introduced in the NRG analysis for the practically stationary gas in the primary system (gas velocities of 10-4 m/s). The large air ingress scenario, with relatively large flow through the break, easier to model from this standpoint. The NRG SPECTRA results showed that the amount of air that can reach the core is extremely small in the DBA accident (0.5 kg of graphite consumed after 100 h). The maximum depth of oxidation was 2 mu m. In the large air ingress scenario the air ingress is significantly larger. At 100 h, approximately 105 kg of graphite was consumed. Even then, the maximum depth of oxidation was very small, approximately 0.2 mm. MELCOR results, performed at NCBJ and UJV, basically confirmed that the amount of air ingress is very small. The numbers were somewhat higher, mainly due to heavy flow oscillations in the break that could not be mitigated in MELCOR calculations.

In the Netherlands, NRG has an extensive experimental irradiation program on molten salt technology. In support of this program, also system thermal hydraulic modeling capabilities are being developed in the SPECTRA code. This paper describes the (fueled) molten salt capabilities developed and presently available in the code. These include flexible input of properties and (heat) transport correlations, delayed neutron precursor drift, fission product behavior, noble gas behavior, noble metal behavior, noble metal extraction, and chromium leaching and deposition. The paper introduces the models developed and where possible their application to the Molten Salt Reactor Experiment (MSRE) data. In general, satisfactory agreements were found between experimental data and SPECTRA results, especially taking into account all the uncertainties involved in the data retrieved from MSRE which is a large integrated system test. In the next development stages, the models will be validated using separate effect tests, where clear test definition and relatively few phenomena involved will allow to quantify the accuracy of models.

In support of extensive experimental irradiation program on molten salt technology at NRG in the Netherlands, system thermal hydraulic modeling capabilities are being developed in the SPECTRA code. This paper describes the (fueled) molten salt capabilities developed and presently available in the code. These include obviously flexible input of properties and (heat) transport correlations, delayed neutron precursor drift, fission product behavior, noble gas behavior, noble metal behavior, noble metal extraction, and chromium leaching and deposition. The paper will introduce the models developed and where possible their application to the Molten Salt Reactor Experiment (MSRE) data and simulation of the Mk1-PB-FHR.

Graphite dust that will be generated in a multi-pass pebble-bed HTR (high temperature reactor), for example the Chinese HTR-PM and the South African PBMR, during normal reactor operation will be deposited inside the primary system and will become radioactive due to sorption of fission products. A significant amount of radioactive dust may be resuspended and released from the reactor cooling system in case of a depressurization accident. Therefore, accurate particle resuspension models are required for HTR/PBMR safety analyses. A review of available resuspension models applicable for monolayer and multilayer deposits is presented in this paper. It is demonstrated that for both multilayer and monolayer deposits, the main problem is the lack of data on adhesion forces and particle-to-particle contact forces. For monolayer deposits, a simple resuspension model, based on a moment balance, referred to here as KS-MB, is proposed and compared with several available resuspension models and available experimental data. It is concluded that a key factor in successful resuspension predictions is a good knowledge of the adhesion force distribution for dust particles deposited on rough surfaces. We demonstrate that relatively simple, quasi-static models, such as the KS-MB model, are as useful as the more complicated dynamic models for resuspension calculations in lack of precise data concerning adhesion forces. For multilayer deposits, resuspension modelling is even more complex. Several models exist, but there is no sufficiently extensive validation. Furthermore, the models may even give contradicting trends of the resuspension rates. The KS-MB resuspension model applicable for multilayer deposits is proposed here and validated against available experimental data. It is concluded that important factors are deposit structure as well as adhesion forces and particle-to-particle contact forces. Furthermore, it is not possible to positively identify the trend of resuspension rates in multilayer deposits. The effect of the multiple layers is overwhelmed by uncertainties in the adhesion force definitions. The main recommendation from the current work is that further measurements of adhesion forces, preferably done for the actual materials and conditions (temperatures, pressures) of the analyzed system (for example HTR-PM) are crucial for development of models and accurate prediction of resuspension. (C) 2018 Elsevier Ltd. All rights reserved.

•Benchmark on sodium-cooled fast reactor, ASTRID-like core.•NRG SPECTRA code simulation of the ASTRID ULOF scenario.•Effect of dynamic gap expansion on ASTRID-like core temperatures during ULOF.

Spent fuel pools (SFPs) are large structures equipped with storage racks designed to temporarily store irradiated nuclear fuel removed from the reactor. SFP severe accidents have long been considered as highly improbable since the accident progression is slow (in comparison with reactor core accidents) and let time to corrective operator actions. However, the accident at the Fukushima Dai-ichi Nuclear Power Plants has highlighted the vulnerability of nuclear fuels that are stored in SFPs in case of prolonged loss-of-cooling accidents and consequently renewed international interest in the safety of SFPs. In this context, the AIR-SFP project, funded by the Euratom 7th FP in the frame of the NUGENIA+ project, was launched in May 2015 with 15 participants. One of the objectives was to assess the applicability of Severe Accident (SA) codes, which were initially developed for reactor applications, to the calculation of transients in SFPs. To reach this objective, a benchmark, including a criticality risk assessment, was carried out. The degradation progression was computed by 14 participants with 6 different SA codes and 5 have participated to the criticality risk assessment. Main results are presented as well as conclusions that have been drawn concerning SA codes readiness to address these "beyond-scope" scenarios. (C) 2018 Elsevier Ltd. All rights reserved.

Validation and verification (V&V) is important and receives significant attention during the design and analysis of the nuclear power plant. As the first commercial modular high temperature gas-cooled reactor (HTR) demonstration plant, during the regulatory review of the preliminary safety analysis report (PSAR) and the final safety analysis report (FSAR) of the 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM), the authorities also pay high attention to the code validation and verification. Within the cooperation between the Nuclear Research Group (NRG) of Netherland and Institute of Nuclear and new Energy Technology (INET), Tsinghua University of China, a code to code verification was carried out between the TINTE code, a thermal-hydraulic design and accident analysis tool for the Pebble-bed High Temperature Gas-cooled Reactor (HTGR), and the SPECTRA code, a thermal-hydraulic analysis code developed at the NRG. Three typical accidents scenarios, two depressurized loss of forced cooling (DLOFC) accidents scenarios and one pressurized loss of forced cooling (PLOFC) accident scenario, have been analyzed both with the TINTE code and the SPECTRA code. The simulation results show good agreement between TINTE code and SPECTRA code. The results also indicate that, due to the inherent safety feature of the HTR-PM design, the maximal fuel temperature during above accidents would never exceed the limitation of 1620 degrees C, below which there is no additional damage for the TRISO particles, and the SiC layer of each particle can guarantee the fission-product retention capacity. This work will be helpful to the further study of the HTGR, and more code to code verification is planned. (C) 2018 Elsevier Ltd. All rights reserved.