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At least 793 records · Page 44

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DOE OSTI2013

A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Modelling of LOCA Tests with the BISON Fuel Performance Code

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

BISON code↗

DOE OSTI2015

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rod undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.

safety codes↗

DOE OSTI2013

Self-shielding models of MICROX-2 code

The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. In the previous study, a new fine-group cross section library of the MICROX-2 was generated and tested against reference calculations and measurement data. In this study, existing physics models of the MICROX-2 are reviewed and updated to improve the physics calculation performance of the MICROX-2 code, including the resonance self-shielding model and spatial self-shielding factor. The updated self-shielding models have been verified through a series of benchmark calculations against the Monte Carlo code, using homogeneous and pin cell models selected for this study. The results have shown that the updates of the self-shielding factor calculation model are correct and improve the physics calculation accuracy even though the magnitude of error reduction is relatively small. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by approximately 0.1 % and 0.2% for the homogeneous and pin cell models, respectively, considered in this study. (authors)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach

Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Development of a new flux map processing code for moveable detector system in PWR

This paper presents an introduction to the development of the flux map processing code MAPLE developed by China Nuclear Power Technology Research Institute (CNPPJ), China Guangdong Nuclear Power Group (CGN). The method to get the three-dimensional 'measured' power distribution according to measurement signal has also been described. Three methods, namely, Weight Coefficient Method (WCM), Polynomial Expand Method (PEM) and Thin Plane Spline (TPS) method, have been applied to fit the deviation between measured and predicted results for two-dimensional radial plane. The measured flux map data of the LINGAO nuclear power plant (NPP) is processed using MAPLE as a test case to compare the effectiveness of the three methods, combined with a 3D neutronics code COCO. Assembly power distribution results show that MAPLE results are reasonable and satisfied. More verification and validation of the MAPLE code will be carried out in future. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2013

Monte Carlo burnup code acceleration with the correlated sampling method. Preliminary test on an UOX cell with TRIPOLI-4{sup R}

For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Validation Study in SAS4A Code in Simulated Mild TOP Condition

Model improvements and validation efforts for SAS4A code, which is the most advanced computer code for simulating the the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors, have been performed utilizing the experimental data from the CABRI programs by JAEA (Japan Atomic Energy Agency), IRSN (Institut de Radioprotection et de Surete Nucleaire), CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) and KIT (Karlsruhe Institute of Technology). In order to confirm validity of improved models of SAS4A, a systematic and comprehensive validation plan was defined. It was based on the PIRT (Phenomena Identification and Ranking Table) result for CDA of MOX-fueled fast reactors. The fuel failure position and time are the most important parameters, which dominate the analytical results in a reactor calculation. The failure position will affect the reactivity feedback due to the fuel distribution after failure and consequently would have an effect on the energy release in the initiating phase. This paper describes the preliminary result of validation study for SAS4A code in simulated Transient Over-Power (TOP) condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Validation of continuous energy sensitivity coefficients calculations in the MORET 5 code - 14269

A continuous-energy sensitivity coefficients calculation capability has been developed in the MORET 5 Monte Carlo code recently. As discussed in this paper, it relies on the differential operator method, an adjoint weighting model (to avoid calculating the fission source derivatives) and a constrained sensitivity coefficient model for sensitivity profiles to distributions such as, for instance, the fission spectrum. For now, the coefficients are provided per medium, per isotope, per reaction on an energy binning defined by the user. The sensitivity coefficients are currently available for the following quantities: total, elastic scattering, total inelastic scattering, fission, total capture and radiative capture cross sections, average neutron produced by fission and fission spectrum. This paper deals with the validation of this new feature in the MORET code on several benchmarks and applications with various energy spectra and geometrical configurations. This validation is performed through the UACSA expert group and the Jezebel ICSBEP benchmarks, but also thanks to calculations performed in the frame of experimental program design in the STACY facility. Comparisons are made using the MCNP6.1 code and the SCALE6.1 package as references. Preliminary results show a pretty good agreement between the different software. (authors)

42 ENGINEERING↗

DOE OSTI2022

Application of Monte Carlo code SHIFT for MSBR dose rate calculations

This paper presents a study that applied the Monte Carlo code SHIFT to calculate the radiological environment within a detailed Molten Salt Breeder Reactor (MSBR) model. It represents one of the first applications of the SHIFT code in stand-alone simulations for non-light water reactors to help demonstrate the code's potential uses in the design, licensing, and operation of advanced reactors. The radiological conditions of the MSBR were modeled when the reactor is at two different operation modes: normal full power and a drained state. The Forward Weighted-Consistent Adjoint Driven Importance Sampling (FW-CADIS) hybrid method in SHIFT was applied successfully to calculate the ex-core neutron and gamma dose rates for the MSBR at full power. Neutron and gamma dose rates within the drain cell were also calculated for the MSBR at the drained state by integrating the source terms obtained from an ORIGEN-S depletion calculation into the SHIFT simulation. The results indicate that in the drained state, the delayed gammas from the depleted fuel salt are the main contributors to the dose rates. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Buffer-IPyC separation process in TRISO fuel particles simulated with Bison code

During High Temperature Gas-cooled Reactor (HTGR) operation, tristructural isotropic (TRISO) coated-particle fuel undergoes irradiation-induced changes in morphology and thermomechanical properties. Experimental results from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program show, among other things, the mechanism of gap formation between the buffer and inner pyrolytic carbon (IPyC) layers, which could be explored further via computational simulations using the Bison code. Two simulation models were developed, the debonding restricted model, where no gap formation between buffer and IPyC layers is permitted, and the debonding enabled model, where the gap between those layers is created. The inputs of the simulated models are based on the irradiation conditions from the AGR-1 experiment. The research included simulations on spherical and aspherical fuel types. Under the specific temperatures and fluences of the AGR-1 irradiation experiment, and the Bison simulations, it was concluded that the most common scenario is a gap formation along the buffer-IPyC interface, while the least possible scenario is the situation where there is no gap formation at the buffer-IPyC junction. The computational results confirmed that the sphericity of the fuel influences the thickness of the gap that occurs at the buffer-IPyC junction, in a way that with increasing aspect ratio the gap thickness increases. The results obtained for spherical and aspherical fuel are nearly identical. Finally, performed simulations match conclusions observed from the AGR-1 experiment, which as such shows that the Bison code is a good computational method for simulating the TRISO fuel. Future simulations will include the validation of performed research and comparison of the results between Bison and PARFUME codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Replacement of Legacy Analytical Codes at the Advanced Test Reactor

For each operating cycle of the Advanced Test Reactor (ATR) at Idaho National Laboratory, a Core Safety Assurance Package (CSAP) is necessary to demonstrate compliance with the safety basis approved by the United State Department of Energy (DOE). Certain computer codes are used in CSAP development, most of them developed in-house. This work describes replacement of a large set of these codes and updates previous work. Replacement of legacy codes is necessary due to computer hardware failure but also has generally improved user-friendliness.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Replacement of Legacy Analytical Codes at the Advanced Test Reactor

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Reactivity initiated accident uncertainty quantification for fuel assembly with subchannel code

UNIST CORE lab has developed a multiphysics coupling framework (MPCORE) consisting of Neutronics, Thermal Hydraulics and Fuel Performance modules. It can accommodate one-dimensional as well as sub-channel code Thermal Hydraulics (TH) module. Generally, running a transient requires more computational power due to the convergence of modules with each other. The difference between one-dimensional and sub-channel TH modules is studied in this research for a Reactivity Initiated Accident (RIA). Both TH modules are compared for a RIA uncertainty propagation in a single VERA fuel assembly with 2.11% enrichment. MPCORE is capable of analyzing the transient at any burnup point but for current work, only fresh fuel has been considered. The results have been obtained using dynamic gap heat conductance in FRAPTRAN. Peak centerline temperature, fuel enthalpy and DNBR are compared for both the approaches. The results indicate that the use of sub-channel code lead to greater safety margin for critical parameters. Computation time comparison is also presented for both the cases. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

On the Representation of the Porosity-Pressure Relationship in General Subsurface Flow Codes

The governing equations for subsurface flow codes in a deformable porous media are derived from the balance of fluid mass and Darcy's equation. One class of these codes, which we call general subsurface flow codes (GSFs), allow for more general constitutive relations for material properties such as porosity, permeability and density. Examples of GSFs include PFLOTRAN, FEHM, TOUGH2, STOMP, and some reservoir simulators such as BOAST. Depending on the constitutive relations used in GSFs, an inconsistency arises between the standard groundwater flow equation and the governing equation of GSFs, and we clarify that the reason for this inconsistency is because the Darcy's equation used in the GSFs should account for the velocity of fluid with respect to solid. Due to lack of awareness of this inconsistency, users of the GSFs tend to use a porosity-pressure relationship that comes from the standard groundwater flow equation and assumes that the relative velocity is already accounted for. For the Theis problem, we show that using this traditional relationship in the GSFs leads to significantly large errors. In this work, we propose an alternate porosity-pressure relationship that is consistent with the derivation of the governing equations in the GSFs where the solid velocity is not tracked, and show that, with this relationship, the results are more accurate for the Theis problem. The purpose of this note is to make the users and developers of these GSFs aware of this inconsistency and to advocate that the alternate porosity model derived here should be incorporated in GSFs.

58 GEOSCIENCES↗

DOE OSTI2014

Progress with the COGENT Edge Kinetic Code: Implementing the Fokker-Plank Collision Operator

Here, COGENT is a continuum gyrokinetic code for edge plasma simulations being developed by the Edge Simulation Laboratory collaboration. The code is distinguished by application of a fourth-order finite-volume (conservative) discretization, and mapped multiblock grid technology to handle the geometric complexity of the tokamak edge. The distribution function F is discretized in v∥ – μ (parallel velocity – magnetic moment) velocity coordinates, and the code presently solves an axisymmetric full-f gyro-kinetic equation coupled to the long-wavelength limit of the gyro-Poisson equation. COGENT capabilities are extended by implementing the fully nonlinear Fokker-Plank operator to model Coulomb collisions in magnetized edge plasmas. The corresponding Rosenbluth potentials are computed by making use of a finite-difference scheme and multipole-expansion boundary conditions. Details of the numerical algorithms and results of the initial verification studies are discussed. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

A new coupling of a GPU-resident large-eddy simulation code with a multiphysics wind turbine simulation tool

The development of new wind farm control strategies can benefit from combined analysis of flow dynamics in the farm and the behavior of individual turbines within one simulation environment. In this work, we present such an environment by developing a new coupling between the large-eddy simulation (LES) code GRASP and the multiphysics wind turbine simulation tool OpenFAST via an actuator line model (ALM). In addition, the implementation of the recently proposed filtered actuator line model (FALM) within the coupling is described. The new ALM implementation is cross-verified with results from four other commonly used research LES codes. The results for the blade loads and the near wake obtained with the new coupling are consistent with the other codes. Deviations are observed in the far wake. The results further indicate that the FALM is able to reduce the lift and power overprediction from which the traditional ALM suffers on coarse LES grids. This new simulation environment paves the way for future wind farm simulations under realistic weather conditions by leveraging GRASP's ability to impose data from large-scale meteorological models as boundary conditions.

17 WIND ENERGY↗