[en] Mechanical behavior of the monolithic mini-plates during a post-fabrication furnace annealing was investigated. Monolithic fuel is a proposed fuel form to accomplish higher uranium densities in the reactor core and thermal cycling is a standard performance evaluation procedure for these fuel elements. To evaluate the mechanical performance of the plate under a thermal loading, a thermo-mechanical Finite element simulation was performed. All three stages of the thermal cycling process were considered: (1) heating of a newly fabricated plate to 500 degrees C, (2) holding at a constant temperature of 500 degrees C for 60 min., and finally (3) cooling the plate to room temperature. Fabrication induced residual stress fields were implemented as the initial state for the thermal cycle model. It was shown that the fuel foil remains in the elastic regime during the entire process, while the cladding material exhibits additional plasticity. In particular, simulations have revealed the existence of a critical temperature at which the net stress fields on the fuel foils change directions. This stress reversal occurs between 400 and 450 degrees C which matches the experimental blister temperature of irradiated plates. It was shown that the fuel foil would be in fully tensile state above this transition temperature,facilitating the initiationof blisters. Long transverse edges and the regions around the corners of the fuel foil were identi?ed as possible blister locations. The results have implied that a higher post-fabrication compressive stress field of the foil yields higher threshold temperatures; however, each thermal cycle would progressively relieve compressive stresses ofthe foil. Comparison with experiments has shown agreement,thus substantiated the capability of the model.

[en] Numerical simulation tools used for the design and safety evaluation of nuclear power plants are under continuous development, improvement and validation. In this field, a key challenge is the development and consolidation of integrated simulation platforms to support safety analysis and future fission reactor designs. These simulation platforms must integrate the dynamic 3D coupling of the codes simulating the different physics of reactor problems into a common multi-physic simulation scheme in order to enhance the prediction capability of the computations used for safety demonstration of the current light water reactors and the design of innovative light water reactors. In this context, the NURESAFE European project had two major objectives: - to deliver to European stakeholders a reliable and predictive up-to-date software platform usable for safety analysis needs for LWR and to further improve the 'safety culture' by developing a high level of expertise in the proper use of the most recent and most advanced simulation tools; this software capacity is the NURESIM (Nuclear Reactor Simulation) European reference simulation platform. - to create a community that brings together the European key-players, to promote the use of an advanced scientific basis among this community and engage it in advanced and predictive simulation for light water reactors. The NURESIM software platform offers: 1) higher fidelity methods as for example pin-by-pin resolution rather than nodal methods and innovative computational fluid dynamics methods 2) an integrated multi-physics environment enabled by the SALOME open-source software (http://www.salome-platform.org/). SALOME provides a generic user-friendly interface and is designed to facilitate the coupling of computing codes in a heterogeneous distributed environment as well as to facilitate interoperation between CAD modeling and codes. 3) a toolbox for uncertainty quantification, sensitivity analysis and model calibration: the URANIE open-source software (http://sourceforge.net/projects/uranie/). The NURESIM software platform was developed from 2006 to 2015 by 22 European organizations. It incorporates from now the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling for present and future reactors together with, sensitivity and uncertainty tools as well as multi-scale and multi-physics features. It provides tools parallel to those of industry, with higher fidelity. These tools can simulate design basis accidents of light water reactors as well as normal operation. The different physics of the NURESIM platform were validated and fully integrated into the platform using SALOME, in order to provide a standardized state-of-the-art code system to support safety analysis of current and evolving light water reactors. (authors)

[en] Two-phase turbulence has been studied using a Direct Numerical Simulation (DNS) of an upward turbulent bubbly flow in a so-called plane channel. Fully deformable monodispersed bubbles are tracked by a Front-Tracking algorithm implemented in TrioCFD code on the TRUST platform. Realistic fluid properties are used to represent saturated steam and water in pressurised water reactor (PWR) conditions. The large number of bubbles creates a void fraction of 10%. The Reynolds friction number is 180. Time- and space-averaging is used to compute the main variables of the averaged scale description (e.g. void fraction, liquid and vapour velocities..) along with the Reynolds stresses and the turbulent dissipation rate tensor. Altogether, they provide reference profiles to assess and further improve Reynolds Stress models. A low-Reynolds version of the SSG model (Speziale et al., 1991) called EBRSM (Manceau and Hanjalic, 2002; Manceau, 2005) is applied in the context of two-phase flows with additional interfacial production terms. The model has been implemented and tested in the two-fluid Euler-Euler model of NEPTUNE-CFD code. The comparison with DNS demonstrates that the interfacial momentum closure plays a dominant role over the turbulent closure hypothesis in the present physical conditions. (author)

[en] The safety assessment of Sodium-cooled Fast Reactors (SFR) requires to account for hypothetical severe accidents involving the melting down of the core materials. This paper deals with the modeling of a fuel vaporization transient that might occur in a SFR in case of severe accident. After a nuclear power excursion, some fuel might be molten and vaporized. In this case, the expansion of fuel vapor might generate a mechanical stress on the reactor vessel and structures. Assessing the vessel integrity is of major importance for the reactor design. A fuel vaporization and expansion modeling, which has been simplified using a Dimensional Analysis, is presented. The modeling is implemented in a tool, called DETONa, able to perform fast calculations, of the order of one minute. The vaporized fuel's thermal exchange with the reactor liquid coolant leading to a possible coolant vaporization is simulated by DETONa. The coolant is assumed to be entrained into the fuel vapor. A droplet entrainment model based on Rayleigh-Taylor instabilities associated to their diameter's limitation using Weber stability criterion is proposed. The modeling is validated on experimental results and on code-to-code comparisons. Parametric calculations are conducted on a reactor case. The impact of the initial molten fuel mass, its initial temperature, critical Weber number and radiative heat transfer are investigated. The non-adiabatic modeling and the adiabatic modeling yield results different by 40% in certain cases. DETONa is shown to be sensitive to the fuel initial temperature, the heat transfer coefficient and the Rayleigh-Taylor wavelength, involving variations that can range to 18%. (authors)

[en] The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10 MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.

[en] The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. Another mandatory constraint is use of low enriched uranium (at or below 20 w/o). The feasibility of using this fuel is assessed by looking at two factors: cycle lengths and fuel material failure rates. Other considerations (e.g., safety parameters such as reactivity coefficients, feedback, etc.) were not considered at this stage of the study. The study includes the examination of increases in the TRISO kernel sizes without changing the thickness of any of the coating layers. In addition, cases where the buffer layer thickness is allowed to vary are also considered. The study shows that a naive use of UO2 (even up to 20 w/o enrichment) results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. The increase of fissile inventory can be accomplished through multiple means, including higher particle packing fraction, higher enrichment, larger fuel kernel sizes, and the use of higher density fuels (that contain a higher number of U atoms per unit volume). In this study, starting with the recognized highest packing fraction practically achievable (44%), combinations of the other means have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios.

[en] A transient heat balance model was developed to assess the impact of a Submersible Mixer Pump (SMP) on radioactive liquid temperature during the process of waste mixing and removal for the high-level radioactive materials stored in Savannah River Site (SRS) tanks. The model results will be mainly used to determine the SMP design impacts on the waste tank temperature during operations and to develop a specification for a new SMP design to replace existing longshaft mixer pumps used during waste removal. The present model was benchmarked against the test data obtained by the tank measurement to examine the quantitative thermal response of the tank and to establish the reference conditions of the operating variables under no SMP operation. The results showed that the model predictions agreed with the test data of the waste temperatures within about 10%

[en] For the Pressurized Thermal Shock (PTS) issue, the NEPTUNE-CFD code solves the Eulerian two-fluid model with specific models for liquid/gas interfaces, which are much larger than the computational cells size. The CFD validation database dedicated to PTS includes the TOPFLOW-PTS experiment, which represents condensation phenomena in a PWR cold leg, with the Emergency Core Cooling system (ECC) and a downcomer. The present study deals with NEPTUNE-CFD calculations of steady-state steam-water tests 3-16, 3-17, 3-18 and 3-19, which differ each other by the ECC liquid inlet flow rate (m_LECC). So the liquid turbulence, which is the main input of the condensation models, is changed from one test to the other, firstly in the ECC region. The direct and first order effect of the ECC flow on the liquid temperature is shown with sensitive two-phase flow regime transitions, which require a careful meshing. This condition being fulfilled, satisfactory NEPTUNE-CFD results mesh-independence on refinement is shown. Present CFD is able to calculate the effect of m_LECC on condensation and temperatures. CFD versions with κ-ε and R_ij-_ε SSG turbulence modeling are compared and the improvement brought by the new version with R_ij-_ε SSG is shown. (authors)

[en] The possible biological consequences of a release of cadmium due to a design basis earthquake in INL's nuclear fuel reprocessing cell are evaluated.

[en] The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility