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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
NASA NTRS1999

Characterizing the Properties of a Woven SiC/SiC Composite Using W-CEMCAN Computer Code

A micromechanics based computer code to predict the thermal and mechanical properties of woven ceramic matrix composites (CMC) is developed. This computer code, W-CEMCAN (Woven CEramic Matrix Composites ANalyzer), predicts the properties of two-dimensional woven CMC at any temperature and takes into account various constituent geometries and volume fractions. This computer code is used to predict the thermal and mechanical properties of an advanced CMC composed of 0/90 five-harness (5 HS) Sylramic fiber which had been chemically vapor infiltrated (CVI) with boron nitride (BN) and SiC interphase coatings and melt-infiltrated (MI) with SiC. The predictions, based on the bulk constituent properties from the literature, are compared with measured experimental data. Based on the comparison. improved or calibrated properties for the constituent materials are then developed for use by material developers/designers. The computer code is then used to predict the properties of a composite with the same constituents but with different fiber volume fractions. The predictions are compared with measured data and a good agreement is achieved.

Murthy, Pappu L. N.↗
DOE OSTI2016

DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems

ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plant Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Non-local slicing approaches for NNLO QCD in MCFM

We present the implementation of several processes at Next-to-Next-to Leading Order (NNLO) accuracy in QCD in the parton-level Monte Carlo program MCFM. The processes treated are pp → H, W ± , Z, W ± H, ZH, W ± γ, Zγ and γγ and, for the first time in the code, W + W - , W ± Z and ZZ. Decays of the unstable bosons are fully included, resulting in a flexible fully differential Monte Carlo code. The NNLO corrections have been calculated using two non-local slicing approaches, isolating the doubly unresolved region by cutting on the zero-jettiness, $\mathscr{T}_0$, or on q T , the transverse momentum of the colour singlet final-state particles. We find that for most, but not all processes the q T slicing method leads to smaller power corrections for equal computational burden.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2022

Coupled neutronic-thermal-hydraulic simulations of the European SFR core

Within the European SFR (Sodium-cooled Fast Reactor) - Safety Measures Assessment and Research Tools (ESFR-SMART) project, steady-state coupled simulation of the ESFR core has been performed using several core analysis packages, with the objective of quantifying the coupling effect. Focus is on the fuel Doppler effect and coolant expansion effect. Standalone neutronics calculations in TRACE/PARCS (PSI), DYN3D (HZDR) and WIMS (Jacobs) showed superb agreement with the reference Serpent power distribution (root mean square - rms discrepancies of 1.3%, 1.5% and 0.7% respectively). Results for COUNTHER (CIEMAT) were also in reasonable agreement, but with a somewhat higher discrepancy of 3.7%. Temperature distributions from thermal-hydraulic calculations were also compared and are found to be in good agreement. The effect of Doppler and coolant density feedback on core power distribution was predicted by TRACE/PARCS, DYN3D and WIMS to be between 0.4% and 0.8% rms difference in assembly powers. Reactivity coefficients for perturbations in the inlet temperature, flow rate and core power were shown to be negative for these three codes, with values of roughly -0.5 pcm/C. degree, -0.3 pcm/C. degree and -3.5 pcm/C. degree respectively. A preliminary investigation of differential thermal expansion effects indicates that this may have a significant effect on core power distribution of a few %, greater than anticipated a priori and may warrant inclusion in coupled core analysis to ensure the accurate calculation of power distributions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy's Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. There is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report details a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To achieve this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. This report describes this work in full. We begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from across the NEAMS toolkit to enable BISON-PROTEUS integration. Furthermore, we show how Warthog leverages a tool known as DataTransferKit to seamlessly enable the transfer for solution data between disparate frameworks and mesh formats. To end, we demonstrate tests for the direct software coupling of BISON and PROTEUS using Warthog, and discuss current impediments and solutions to the construction of physically realistic input models for this coupled BISON-PROTEUS system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Warthog: Progress on Coupling BISON and PROTEUS

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Office of Nuclear Energy at the Department of Energy (DOE) provides a robust toolkit for modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines, with two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line (FPL) and the Reactor Product Line (RPL) provide advanced computational technologies that serve each respective field effectively. There is currently a lack of integration between the product lines, impeding future improvements of simulation solution fidelity. In order to mix and match tools across the product lines, a new application called Warthog was produced. Warthog is built on the Multi-physics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory (INL). This report details the continuing efforts to provide the Integration Product Line (IPL) with interoperability using the Warthog code. Currently, this application strives to couple the BISON fuel performance application from the FPL using the PROTEUS Core Neutronics application from the RPL. Warthog leverages as much prior work from the NEAMS program as possible, enabling interoperability between the independently developed MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. Previous work performed on Warthog allowed it to couple a pin cell between the two codes. However, as the temperature changed due to the BISON calculation, the cross sections were not recalculated, leading to errors as the temperature got further away from the initial conditions. XSProc from the SCALE code suite was used to calculate the cross sections as needed. The remainder of this report discusses the changes to Warthog to allow for the implementation of XSProc as an external code. It also discusses the changes made to Warthog to allow it to fit more cleanly into the MultiApp syntax of the MOOSE framework. The capabilities, design, and limitations of Warthog will be described, in addition to some of the test cases that were used to demonstrate the code. Future plans for Warthog will be discussed, including continuation of the modifications to the input and coupling to other SHARP codes such as Nek5000.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Plutonium Critical Mass Curve Comparison to Mass at Upper Subcritical Limit (USL) Using Whisper

Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the MCNP ® Monte Carlo radiation transport package. Standard approaches to validation rely on the selection of benchmarks based upon expert judgment. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application or set of applications being analyzed. Using these benchmarks, Whisper computes a calculational margin. Whisper attempts to quantify the margin of subcriticality (MOS) from errors in software and uncertainties in nuclear data. The combination of the Whisper-derived calculational margin and MOS comprise the baseline upper subcritical limit (USL), to which an additional margin may be applied by the nuclear criticality safety analyst as appropriate to ensure subcriticality. A series of critical mass curves for plutonium, similar to those found in Figure 31 of LA-10860-MS, have been generated using MCNP6.1.1 and the iterative parameter study software, WORM_Solver. The baseline USL for each of the data points of the curves was then computed using Whisper 1.1. The USL was then used to determine the equivalent mass for plutonium metal-water system. ANSI/ANS-8.1 states that it is acceptable to use handbook data, such as the data directly from the LA-10860-MS, as it is already considered validated (Section 4.3 4) “Use of subcritical limit data provided in ANSI/ANS standards or accepted reference publications does not require further validation.”). This paper attempts to take a novel approach to visualize traditional critical mass curves and allows comparison with the amount of mass for which the k eff is equal to the USL (calculational margin + margin of subcriticality). However, the intent is to plot the critical mass data along with USL, not to suggest that already accepted handbook data should have new and more rigorous requirements for validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Deterministic modelling of the SPERT IV reactor transients using the multi-physics capability of WIMS

The ANSWERS WIMS reactor physics code is being developed for whole core multi-physics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding. An integrated thermal hydraulic solver permits temperature and density variations to feedback to the neutronics calculation. This capability can be applied to both steady state and time dependent transient problems. Nuclear reactor design and safety case development requires assessment of a range of reactor transients, to inform both normal operation limitations and accident scenario analysis. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver for the simulation of reactor transients in whole core models. To support the validation of the multi-physics capability of WIMS, this capability has been applied to the reactivity insertion transient experiments performed in the plate fuelled SPERT-IV reactor. This study employs WIMS using the whole core solver MERLIN, which calculates the time-dependent flux distribution and magnitude, coupled to ARTHUR, which solves for the time-dependent thermal hydraulics solution, and provides thermal feedback via the cross sections generated by GEOM, which performs resonance shielding calculations and generates cross section data for the plate geometry. This spatial kinetics model, with dynamic cross section generation, allows for variations in the temperature and neutron flux profile with time.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such as MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Investigation of the OECD/NEA PWR MOX/UO{sub 2} core transient benchmark using a coupled whole core pin-by-pin route in WIMS

This paper investigates the new CAMELOT coupling route developed in WIMS, which uses prepared tabulated cross-sections combined with the flux solver MERLIN module and the integrated sub-channel thermal hydraulics solver ARTHUR module to solve for the coupled core state. This coupling route is applied to the OECD/NEA pressurized water reactor (PWR) core transient benchmark exercise, involving 3-D neutronics and thermal-hydraulics analysis of a full-sized PWR core featuring MOX fuel at various core states, as well as simulating a control rod ejection transient. Analysis of the results for each part of the benchmark demonstrate that the high-resolution pin-by-pin analysis used in WIMS closely matches the other participants in the benchmark, with differences of between 20-200 pcm for k-effective, 1-2.5% RMS difference for assembly-averaged powers, 2-8% difference in rod worth calculations, < 200 ppm for calculations of critical boron concentration, zero difference in delayed neutron fraction calculation and a very close match for power response to the rod ejection transient. This is the first application of the WIMS-CAMELOT approach for the transient analysis of a full-sized reactor and as such further improvements such as reducing computational cost through parallelization, optimisation and memory reduction, are currently in active development. This paper demonstrates the accuracy and flexibility of the CAMELOT coupling route, providing a straightforward and consistent user image that can be easily used for a variety of modelling problems involving coupled neutronics and thermal-hydraulics. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C k 's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.

42 ENGINEERING↗
DOE OSTI2018

Accurate Bandgap of Zr{sub x}Al{sub 1−x}N Using Modified Becke–Johnson (mBJ) Exchange Potential

Structural parameters and the electronic structure of wurtzite Zr{sub x}Al{sub 1−x}N (0 ≤ x ≤ 1) have been investigated by means of the full-potential linearized augmented plane wave (FP-LAPW) with the modified Becke–Johnson (mBJ) potential, which provides an improved value of the bandgap, as embodied in the WIEN2k code. We found that the calculated band structure shows that the crystal is a semiconductor with a direct bandgap for x = 0 (AlN) and metallic for x ≥ 0.25. The comparison of our results with the experimental and other theoretical calculations shows that the results obtained by mBJ are much superior to other theoretical techniques and are very close to the experimental ones.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2015

A COMPREHENSIVE PHOTOMETRIC STUDY OF THE CONTACT BINARY GN BOO WITH POSSIBLE MAGNETIC ACTIVITIES

Multi-color photometric data of GN Boo observed from 2010 to 2013 are presented. The intrinsic variations of the light curves are remarkable, and their phenomena are probable evidence of stronger magnetic activities on the surfaces of the components. Based on all CCD times of minimum light, a secular increase superimposed on a cyclic oscillation is found. The orbital period increases at a rate of dP/dt = +1.74 × 10{sup −7} days yr{sup −1}, which can be explained by mass transfer from the less massive component to the more massive one. The period and amplitude of the cyclic variation are P = 9.5632 yr and A = 0.0046 days, respectively, which correspond to the previous published results. Using the 2010 version of the W–D code, five sets of photometric solutions were derived from our new data. The results imply that the stellar spot, the degree of fill-out, and the temperature difference ΔT between the components of GN Boo are variable. It is inferred that the magnetic activities perhaps influence the outer radius of the component, causing the temperature of the component star and the level of contact to change.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2015

Response to FESAC survey, non-fusion connections to Fusion Energy Sciences. Applications of the FES-supported beam and plasma simulation code, Warp

The Fusion Energy Sciences Advisory Committee’s subcommittee on non-fusion applications (FESAC NFA) is conducting a survey to obtain information from the fusion community about non-fusion work that has resulted from their DOE-funded fusion research. The subcommittee has requested that members of the community describe recent developments connected to the activities of the DOE Office of Fusion Energy Sciences. Two questions in particular were posed by the subcommittee. This document contains the authors’ responses to those questions.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Ab initio study of structural, electronic, magnetic alloys: XTiSb (X = Co, Ni and Fe)

Structural, electronic and magnetic properties of three semi-Heusler compounds of CoTiSb, NiTiSb and FeTiSb were calculated by the method (FP-LAPW) which is based on the DFT code WIEN2k. We used the generalized gradient approximation (GGA (06)) for the term of the potential exchange and correlation (XC) to calculate structural properties, electronic properties and magnetic properties. Structural properties obtained as the lattice parameter are in good agreement with the experimental results available for the electronic and magnetic properties was that: CoTiSb is a semiconductor NiTiSb is a metal and FeTiSb is a half-metal ferromagnetic.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗