The work presented in this paper applies the MELCOR code developed at Sandia National Laboratories to evaluate the source terms from potential accidents in non-reactor nuclear facilities. The present approach provides an integrated source term approach that would be well-suited for uncertainty analysis and probabilistic risk assessments. MELCOR is used to predict the thermal-hydraulic conditions during fires or explosions that includes a release of radionuclides. The radionuclides are tracked throughout the facility from the initiating event to predict the time-dependent source term to the environment for subsequent dose or consequence evaluations. In this paper, we discuss the MELCOR input model development and the evaluation of the potential source terms from the dominated fire and explosion scenarios for a spent fuel nuclear reprocessing plant.

The Steam Generator Tube Rupture (SGTR) postulated event constitutes one of the most hazardous safety issues for Gen IV pool reactors, cooled by heavy liquid metals. This accidental scenario is characterized by quick water flashing when in contact with primary coolant liquid metal, causing pressure wave propagation, cover gas pressurization in the reactor main vessel as well as possible tube rupture propagation, vapour dragged through the core, oxides precipitation and consequent slugs and plugs formation. The design phase of Gen IV MYRRHA reactor addressed the SGTR scenario issues in the framework of MAXSIMA project, supported by the European Commission. This research activity was fully executed at ENEA CR Brasimone, where a new test section was designed, assembled, instrumented and implemented in the large scale pool facility CIRCE. It was supported by the execution of preliminary and detailed pre-tests analysis performed adopting SIMME-III and -IV code, respectively. This paper details the test section main features, able to host four full scale portions (each one constituted by 31 tubes) of the MYRRHA Primary Heat eXchanger (PHX), for carrying out four independent SGTR experiments. A couple of tests investigated the tube rupture at middle position between two spacer grids of the bundle. The other two tests analysed instead the rupture near the bottom tube plate. Auxiliary systems were adopted for reaching primary (Lead Bismuth Eutectic alloy, LBE) and secondary (water) coolant initial conditions in accordance with MYRRHA design. Water was injected at 16 bar and 200°C in LBE at 350°C under an argon cover gas at about atmospheric pressure. The experimental results of the first test (middle rupture), in terms of CIRCE vessel pressurization, vapour flow path through tube bundle and tubes deformation, are presented. The post-test analysis was performed by SIMMER-IV code adopting the 3D Cartesian code version. The whole main vessel of CIRCE facility and implemented test section were modelled conserving heights and flowing areas. The experimental initial conditions were successfully matched by numerical results as well as the vessel pressurization and temperature time trends in the tube bundle following the SGTR. An important engineering feedback, for MYRRHA designer, was the evidence of rupture propagation absence. Moreover, the effectiveness of implemented safety devices, rupture disks, was evaluated and characterized for pressure relief feedbacks. A wide series of high quality measured data (pressure, temperature, strain and mass flow rate) was acquired and constitutes a database enlargement for future codes validation and possible new model development.

Due to the increase of computing efficiency and power, full-resolution two-phase flow simulations have become a practical research tool for model development and analysis of reactor flows. The expansion of state-of-the-art high performance computing (HPC) facilities allows for the use of direct numerical simulation (DNS) coupled with Interface Tracking Methods (ITM) to perform full resolution simulations. Given adequate spatial and temporal resolution, DNS can resolve all relevant turbulent scales, allowing for the extraction of high quality and detailed turbulent and two-phase flow numerical data for use in model development. While larger scale bubbly flow DNS are becoming ever more affordable, it is still computationally expensive due to the requirements of the spatial discretization. This presents the largest obstacle for future applications of DNS. For this reason, mesh adaptation techniques are sought after to reduce the computational expense of bubbly flow simulations in complex geometries. By fully resolving only the areas of specific interest, the computational costs of DNS can be reduced. Grid refinement can be based on the location of the interface between the two phases, area of greatest turbulent intensity, averaged bulk fluid velocity data, or the prediction of bubble movement. Coupled with an advanced bubble tracking algorithm, the path of individual bubbles moving through the computational domain can be predicted, and the computational mesh refined within the path area. This refinement can create tracks of greater resolution for the bubbles to move through in the domain, while keeping the bulk resolution of the mesh coarser. Through these means, the overall cost of the simulation is reduced, while high quality numerical data is still obtainable. This work outlines the enhancement of existing mesh adaptation algorithms to implement the bubble tracking refinement, and its practical applications to full resolution two-phase flow simulations.

MELCOR is a fully-integrated, system-level computer code developed by Sandia National Laboratories (SNL) for the Nuclear Regulatory Commission (NRC) with the primary objective of modeling the progression of severe accidents in light water nuclear power plants [1,2,3]. Since the project began in 1982, MELCOR has undergone continuous development to address emerging issues, process new experimental information, and create a repository of knowledge on severe accident phenomena. This paper summarizes model development specifically developed for non-LWR applications such as high temperature gas reactors (HTGR), sodium fast reactors (SFR) and molten salt reactors (MSR). Beginning in 2008, active development work began on HTGR modeling in MELCOR. Models were developed for helium gas thermodynamics, oxidation of graphite, thermal hydraulics and heat transfer for both prismatic and pebble bed designs, cavity cooling systems, fuel failure and fission product release, graphite dust generation, and aerosol transport, deposition, and resuspension. In 2013, work commenced on the development of modeling capabilities for sodium fast reactors. This development included the addition of sodium as a working fluid as well as the addition of models for simulating containment fires (both spray and pool) as well as sodium atmospheric chemistry. Validation of these new models has been completed and code-to-code comparisons with the CONTAIN/LMR code has been performed. Work continues as development of sodium concrete interaction models is now underway. In 2017, work began on adding capabilities for molten salt reactors. A new equation of state for FLIBE coolant has been successfully tested in MELCOR and is now undergoing validation against experiments. The alternate working fluid model has also been extended to permit both water and one alternate condensable working fluid in the same input model.

A full scope Probabilistic Safety Assessment (PSA) was provided as an integral part of the safety case for the UK ABWR Generic Design Assessment (GDA) and this included a Level 3 PSA. The main objectives of the Level 3 PSA for GDA are to provide a demonstration of the compliance, for a single unit UK ABWR, with numerical risk targets defined in the UK Safety Assessment Principles and to support the ALARP assessment. This paper includes: • An overview of the methodology, PSA model development and illustration of some results. • A summary of the key assumptions made during the model development for the GDA phase of the project. Compliance with the numerical risk targets has been investigated through assessments against the individual off site risk from the facility (Target 7), facility dose bands (Target 8) and off site societal risk (Target 9). • Some conclusions of a peer review against the draft ASME/ANS standard for trial use (ASME/ANS RA-S-1.3, Feb 2016 for Level 3 Probabilistic Risk Assessment (PRA)), which was a key aspect of this study.

Detailed knowledge of a coolant flow in a fuel assembly of a reactor core has always been a major factor in the design of new nuclear systems. In this regard, traditionally adopted subchannel analysis codes cannot take into account local phenomena, which are quite essential. On the other hand, Computational Fluid Dynamic (CFD) is being recognized as a valuable research tool for thermal-hydraulics phenomenon in the fuel assembly geometries. Because of the high Reynolds number and geometric complexities, the practical CFD calculations are mostly limited to pragmatic Reynolds Averaged Navier-Stokes (RANS) type modelling approaches. A good prediction of the flow and heat transport inside the fuel rod bundle is a challenge for such RANS turbulence models and these models need to be validated. Although the measurement techniques are constantly getting improved, however, the CFD-grade experiments of flow mixing and heat transfer in the subchannel scale are often impossible or quite costly to be performed. In addition, lack of experimental databases makes it impossible to validate and/or calibrate the available RANS turbulence models for certain flow situations. In that context, Direct Numerical Simulation (DNS) can serve as a reference for model development and validation. The aim of this work is to design a numerical experiment in order to generate a high quality DNS database for a tight lattice bare rod bundle, which will serve as a reference for the validation purpose. The considered geometric design is based on the well-known Hooper experiment, which contains a bare rod bundle with pitch-to-diameter ratio of P/D = 1.107. Performing a DNS computation corresponding to the Hooper experiment requires a huge computational power. Hence, a wide range of unsteady RANS (URANS) study has been performed to scale-down the Reynolds number such that it is feasible for a DNS computation and at the same time it still preserves the main flow characteristics. In addition to the flow field, a parametric study for three different passive scalars is performed to take into account the heat transfer analysis. These passive scalars correspond to the Prandtl numbers of air, water and liquid metal fluids. The heat transfer of these three fluids has been studied in combination with two different boundary conditions at the walls, i.e. a constant temperature and a constant heat flux. Finally, the obtained URANS results are used to compute the Kolmogorov and Batchelor length scales in order to estimate the overall meshing requirements for the targeted DNS.

A Probabilistic Safety Assessment (PSA) was provided as an integral part of the safety case for the United Kingdom Advanced Boiling Water Reactor (UK ABWR) Generic Design Assessment (GDA). The main objectives of PSA for GDA is to provide a demonstration of the compliance, for a single unit UK ABWR, with numerical risk targets defined in the UK Safety Assessment Principles (SAP) and to support the “As Low As Reasonably Practicable” (ALARP) assessment. This content of this paper includes: • An overview of the UK ABWR PSA • Identification of the PSA scope • PSA model development to compare with numerical risk targets and inform design and operational features • Illustration of PSA results, e.g., Core Damage Frequency (CDF), Large Release Frequency (LRF) and Large Early Release Frequency (LERF) • Peer reviews • Use of PSA in ALARP assessment

The continuous generation of graphite dust particles in the core of a High Temperature Reactor (HTR) is one of the key challenges of safety during the operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurisation accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions. On the one hand, the knowledge of the behaviour of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behaviour can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces. The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5m and a tube diameter of 0.05m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass. In particular, this test facility offers the possibility to analyse the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a 3D Laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle Image Velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation time are available. The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for CFD-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of Nuclear Safety.

Compared with conservation evaluation model, best estimate plus uncertainty (BEPU) method can obtain more realistic results and gain larger license margins with respect to the safety criteria. In view of this, a BEPU method named 4S (SNERDI Statistical Solution for Safety) has been developed, according to the basic principles of evaluation model development and assessment of RG 1.203. The characteristics of 4S method are as follows: The output uncertainty is quantified by using random sampling and propagation of input uncertainties. Global sensitivity analysis is used to support PIRT establishment. Uncertainties of model parameters are calibrated and validated by using separate effects tests considering measuring uncertainties. DAKOTA code is used for uncertainty and sensitivity analysis. An automatic BEPU analysis platform has been developed by coupling DAKOTA and different reactor safety analysis codes, and code calculations can be performed in parallel. BEPU analysis of mass and energy release and containment pressure response of CAP1400 under a postulated double-ended cold leg break loss of coolant accident (DECL LOCA) has been carried out by coupling DAKOTA, a mass and energy release analysis code and a containment analysis code. In total, 21 uncertain input parameters are considered. To make the results more stable, the sample size is 124 and the third highest peak pressure is used as the pressure upper bound (with 95%/95% probability/confidence) based on Wilks’ formula. The calculated results show that the peak pressure upper bound is obviously lower than the present conservation method used in license application, with more than 10% analysis margin. Influences of input parameter uncertainties on the containment peak pressure have been analyzed, according to the partial rank correlation coefficients calculated by DAKOTA. The results show that the input parameters mainly affecting the peak pressure are the containment condensation heat transfer multiplier, initial containment temperature, break resistance, decay heat, initial containment pressure, Core Makeup Tank (CMT) resistance multiplier and initial containment humidity.

In-Vessel Retention is a key severe accident management strategy for reactors such as AP/CAP series reactors. The IVR success evaluation criterion is whether the RPV is melted through or not at the final RPV state. Once the RPV lower head melt through, the liquid corium will flow into the reactor cavity and will lead to complex phenomena, such us steam explosion and the reaction between the corium and concrete. These will make temperature and pressure of the containment vessel rise quickly and is a threat to the integrity of the containment vessel. When the wall surface of RPV lower head heating condition exceed the critical heat flux, the temperature rises rapidly, it is generally assumed that the RPV lower head in this state will inevitably melt through. This is the so-called IVR failure. In order to study the possible failure modes and mechanism of RPV lower head under the IVR measures, an experimental facility called TRECT is built. By measuring the parameters such as temperature, flow of the test section to study the influence to CHF by the parameters such as flow velocity and angle. All of these can provide reliable basis to the effectiveness appraisal and model development on the area of severe accident mitigation measures (IVR). To be specific, the test section is rectangular channel whose section is 50 × 20 mm. The upper surface is the heat surface and using a direct current heating mode to supply heat power. The heat flux can reach 1.5MW/m 2 . We use this upper surface heated rectangular channel to simulate RPV ERVC channel. By adjust the angle of test section to simulate the different circum ferential location of RPV lower head. And the Adjusting range can be 0° to 90°. The experimental results show that flow rate was reduced by 11% in the experiments, the critical heat flux density increased by 4.5%. Inclined angle increased from 16° to 29°, CHF increased by 7.9%.

The systems computer code is a key part of the evaluation model for safety analysis of nuclear reactors. The systems code utilizes a set of governing equation that is simplified from the fundamental Navier-Stokes equations and closure models to describe the transport of mass, momentum, and energy of single phase or multiphase fluid throughout the reactor coolant systems. Following the Evaluation Model Development and Assessment Process, an assessment matrix is established where Separate Effects Tests and Integral Effects Tests are selected based on phenomena identification and ranking table. The purpose of the assessment matrix is to validate the systems code against the important phenomena for the safety analysis. The code biases and uncertainties are established and the effect of scale could then be determined. The assessment matrices of major systems codes, RELAP5/MOD3, TRACE Ver.5.0 and W COBRA/TRAC-TF2, for the reactor safety analysis are reviewed and compared in this study for the Loss of Coolant Accident (LOCA) safety analysis perspectives. The scenarios are divided into small break LOCA and large break LOCA. The phenomena bases of the separate effects tests in those assessment matrices are discussed following its PIRT. The comparison demonstrates the capability of each systems code.

Experiments of counter-current two-phase flow of upward steam flow and condensing downward film flow in a pipe were performed. The experiments were intended to examine water accumulation in steam generator U-tubes during intermediate and small break loss-of-coolant accidents of a pressurized water reactor. The inner diameter and the length of a test flow channel used in the experiments were 18 mm and 4 m, respectively. Experiments were performed at higher steam velocity a little than the velocity that was expected just after scram as the first trial. There was no water drainage form the test pipe to the lower plenum. All condensed water was entrained by steam to flow out from the top of the test pipe to the upper plenum. The test pipe was filled with the water lump and the water film, then these were blown up upward and the inner wall of the test pipe became dry. Again the test pipe was filled with the water lump and the water film, then these were blown up upward and the inner wall of the test pipe became dry. This process was iterated at short intervals. The flow state in the test pipe is highly chaotic and agitated. Condensed water flows up and down at high frequencies. It is indicated that to examine the time averaged void fraction and the two-phase pressure drop of the counter-current flow are required.

The possibility of a spent fuel severe accident has received increasing attention in the last decade, and in particular following the Fukushima accident. Several large scale experiments and also separate effect tests have been conducted to obtain a data base for model development and code validation. The outcome of the Sandia BWR Fuel Project was used to define the flow parameters adjusted for the low pressure and the increased flow resistance due to the presence of the spent fuel racks which resulted in reduced buoyancy driven natural circulation flow compared with reactor geometry. The possibility of a zirconium fire, using the flow parameters obtained from the spent fuel experiments, is investigated in the present work. The important outcome of the study is that spent fuel will degrade if temperatures above 800 K are reached. In partial loss of coolant accidents the flow through the lower bottom nozzle is blocked and because there is no cross flow possible due to the spent fuel racks the coolant flow in the upper dry part of the spent fuel is limited by the steam production in the lower still wetted part of the fuel. This accident scenario leads to the fastest heat up in a postulated spent fuel accident. The influence of different kind of spent fuel storage (hot neighbour and cold neighbour) is investigated. An important factor in these calculations is the radial heat transfer to the neighbouring fuel assemblies. In the present work limits of the spent fuel storage under accident conditions (minimum allowed water levelin the pool) and total loss of coolant (maximum coolable decay heat per fuel assembly) are shown and explained.

PSA specialists in ÚJV Řež, a. s. maintain a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP), a four-unit VVER-440 plant type, which is operated in the Czech Republic. This project has been established as a broad framework for all plant activities related to risk assessment and as a support for risk-informed decision making carried out at this plant. In addition to recommendations for design and operation measures in order to increase the plant safety, it provides a basis and platform for all PSA applications at Dukovany NPP. The Living PSA model for Dukovany NPP is an integrated model representing the complete scope of Level 1 and Level 2 PSA for all plant operational modes. It produces the unit specific outputs for any Dukovany NPP unit. The RiskSpectrum ® PSA software has been used for development and quantification of the PSA model. It is continuously updated and extensively used for various PSA applications at Dukovany NPP (e.g. risk monitoring, evaluation of plant Technical Specification changes, support for procedure development and training process, event analysis, etc.). The paper focuses on the important features of the Living PSA project for Dukovany NPP. It also discusses the broad experience gained during model development and update as well as possible future enhancements.

Hydrogen risk mitigation issues based on catalytic recombiners cannot exclude flammable clouds to be formed during the course of a severe accident in a Nuclear Power Plant. Consequences of combustion processes have to be assessed based on existing knowledge and state of the art in CFD combustion modelling. The Fukushima accidents have also revealed the need for taking into account the hydrogen explosion phenomena in risk management. Thus combustion modelling in a large-scale geometry is one of the remaining severe accident safety issues. At present day there doesn’t exist a combustion model which can accurately describe a combustion process inside a geometrical configuration typical of the Nuclear Power Plant (NPP) environment. Therefore the major attention in model development has to be paid on the adoption of existing approaches or creation of the new ones capable of reliably predicting the possibility of the flame acceleration in the geometries of that type. A set of experiments performed previously in RUT facility and Heiss Dampf Reactor (HDR) facility is used as a validation database for development of three-dimensional gas dynamic model for the simulation of hydrogen-air-steam combustion in large-scale geometries. The combustion regimes include slow deflagration, fast deflagration, and detonation. Modelling is based on Reactive Discrete Equation Method (RDEM) where flame is represented as an interface separating reactants and combustion products. The transport of the progress variable is governed by different flame surface wrinkling factors. The results of numerical simulation are presented together with the comparisons, critical discussions and conclusions.

In this work, in-depth considerations about FPGA-CPU architectures design methods are presented together with some possible plant applications. FPGAs are used because of their potentials in guaranteeing parallelism, synchronism and high reliability tasks; on the other hand CPUs are crucial to higher precision computations and to reproduce data and results on a graphical interface, thus resulting in an improvement in the human-machine-interface (HMI). Design choices are here reported to justify the benefits from the use of both CPU-FPGA approaches. Taking advantage of these features, the LabVIEW environment and its digital platform is expanded in order to develop a novel tool able to support designers during the fundamental phases of a supervision and control tool realization: from the model development to the digital hardware implementation and testing. LabVIEW is able to communicate with several languages like Matlab or Scilab for a theoretical model study and simulation, and VHDL for a digital implementation, taking the advantages from both development environments. Main results have been reported concerning studies of control systems in nuclear plant facilities, showing how a full-digital platform can be very useful in developing monitoring and control instrumentation. This new tool is thought to improve the so-called hardware in the loop (HIL) simulations and to accelerate the process of prototypes implementation and testing, using a more accurate and reliable environment in terms of performance and safety.

High temperature gas-cooled reactor (HTGR), especially the pebble-bed core type reactor, will inevitably cause the wear the graphite components and generate graphite dust in the core. The graphite dust is taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. The objective of this paper is to develop a code for calculating the behaviour of graphite dust in the primary circuit of HTGR. The paper is focused on development of models for predicting the deposition rates of the dust. The purpose of the work is to estimate the amount and distribution of deposited dust during plant life time, which was assumed to be 40 full-power years. The result will lay the foundation for further studies of fission products releasing and interaction with dust under accident conditions.

The Institute for Nuclear Waste Management and Reactor Safety (IEK-6) at the Forschungszentrum Juelich (JUELICH) investigates accident scenarios of gas-cooled High-Temperature Reactors (HTR), especially the air ingress scenario. Fluid mechanic codes and models are developed and validated using in-house experimental databases. This paper describes the development of a computational fluid dynamics (CFD) model with the porous media approach in order to investigate the air ingress scenario: “rupture of the pressure compensation line” at the HTR-Modul. Bases on a brief introduction on this chosen air ingress scenario, the used model equations are explained, as well as the necessary simplifications of the physical model. Afterwards, an overview of the used experiments for validation purposes is given. As an outlook, the assumptions and boundary conditions of the chosen scenario are shown and the setup for the air ingress calculation is presented.

Safety reports have shown that tons of solid particles would be generated as dusts in the operation of ITER facility. The dust particles include carbon, beryllium and tungsten with diameters ranging from a few to a few hundreds microns. The particles deposit downwards and mostly accumulated on the surfaces of the diverter on the bottom side of the vacuum vessel (VV). In accident scenarios, e.g., loss of vacuum accident (LOVA), the potentially combustible dust particles can be suspended by the air ingress and entrained into the whole volume of the VV, and impose a risk of dust explosions in case of unintentionally ignition to the whole ITER facility. Therefore the mechanism of particle resuspension was investigated theoretically in the work. A force balance approach and numerical fittings have been utilized to develop a semiempirical particle resuspension model based on a group of particle resuspension experimental data. The model has been applied into a three-dimensional computational fluid dynamics code, GASFLOW. The model validation has been done by comparison of the numerical predictions about particle resuspension rates in given incoming flows against the corresponding experimental data. The comparisons have proved the validity of the developed model about particle resuspension.

As an important part in both product development and analytical model validation, Westinghouse continues to develop prototype testing capabilities. Towards better understanding fuel assembly’s structural performance during a seismic event, our recent effort was to develop a new testing capability for simulating multiple fuel assembly impact that may occur during a seismic event. The test system was developed by adding a pendulum system to our existing fuel assembly mechanical test stand. Following qualification of the testing system, a prototype fuel assembly was tested using selected pendulum parameters. Fuel assembly displacement, impact loads at various contact locations, and the spacer grid deformation as a function of impact forces were obtained. The test results provided new insights and understanding of the fuel assembly’s structural performance during a simulated seismic event, and are useful for model development and validation.