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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

SmAHTR-CTC Neutronic Design

Building on prior experience for the 2010 initial SmAHTR neutronic design and on 2012 neutronic design for the advanced high temperature reactor (AHTR), this paper presents the main results of the neutronic design effort for the newly re-purposed SmAHTR-CTC reactor concept. The results are obtained based on full-core simulations performed with SCALE6.1. The dimensionality of the SmAHTR design space is reduced by using constraints originating in material fabricability, fuel licensing, molten salt chemistry, thermal-hydraulic and mechanical considerations. The new design represents in many regards a substantial improvement from the neutronic performance standpoint over the 2010 SmAHTR concept. Among other results, it is shown that fuel cycle length of over 2 years or discharged fuel burnup of 40GWd/MTHM are possible with a low, 8% fuel enrichment in a once-through fuel cycle, while 8-year once-through fuel cycle lengths are possible at higher fuel enrichments.

Ilas, Dan↗
DOE OSTI2015

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff ) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades are inserted in various locations and at varying degrees during BWR operation based on the reload design. The presence of control blades during depletion hardens the neutron spectrum locally due to both moderator displacement and introduction of a thermal neutron absorber. The reactivity impact of control blade presence is investigated herein, as well as the effect of multiple (continuous and intermittent) exposure periods. The coupled effects of control blade presence on power density, void profile, or burnup profile have not been considered to date but will be addressed in future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Subplane-based Control Rod Decusping Techniques for the 2D/1D Method in MPACT

The MPACT transport code is being jointly developed by Oak Ridge National Laboratory and the University of Michigan to serve as the primary neutron transport code for the Virtual Environment for Reactor Applications Core Simulator. MPACT uses the 2D/1D method to solve the transport equation by decomposing the reactor model into a stack of 2D planes. A fine mesh flux distribution is calculated in each 2D plane using the Method of Characteristics (MOC), then the planes are coupled axially through a 1D NEM-P$_3$ calculation. This iterative calculation is then accelerated using the Coarse Mesh Finite Difference method. One problem that arises frequently when using the 2D/1D method is that of control rod cusping. This occurs when the tip of a control rod falls between the boundaries of an MOC plane, requiring that the rodded and unrodded regions be axially homogenized for the 2D MOC calculations. Performing a volume homogenization does not properly preserve the reaction rates, causing an error known as cusping. The most straightforward way of resolving this problem is by refining the axial mesh, but this can significantly increase the computational expense of the calculation. The other way of resolving the partially inserted rod is through the use of a decusping method. This paper presents new decusping methods implemented in MPACT that can dynamically correct the rod cusping behavior for a variety of problems.

Graham, Aaron M↗
DOE OSTI2018

SCALE 6.2 REACTOR PHYSICS CODE ACCURACY ASSESMENT FOR LIGHT WATER REACTOR FUEL

Verification and validation results of the SCALE 6.2 lattice physics codes TRITON/NEWT and Polaris are presented. The SCALE/KENO continuous energy Monte-Carlo code is used as a reference to compare results of more than 2,000 combinations of pressurized water reactor and boiling water reactor lattice designs and state points. Radiochemical assay measurements were also used to assess the accuracy of depletion calculations. Results show that Polaris and TRITON/NEWT using ENDF/B_VII.1 produce accurate lattice physics calculations and cross section data for core simulators.

Mertyurek, Ugur↗
DOE OSTI2017

Uncertainty and Sensitivity Analysis of TREAT Transient Test #15

The U.S. Department of Energy (DOE) is preparing for the resumption of transient testing using the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL). In order to support the analysis of TREAT operation and experiment planning, the MAMMOTH reactor physics team has recently begun development of uncertainty and sensitivity analysis for the TREAT core simulated via MAMMOTH. MAMMOTH is a high-fidelity three-dimensional transient multi-physics code developed at INL under the DOE-NE Advanced Modeling and Simulation program (NEAMS) [1]. This research is part of the development presenting some initial uncertainty and sensitivity analysis of TREAT Transient Test #15 [2]. In this summary, we employ the Sobol sampling strategy implemented in RAVEN [3] to efficiently perform the sensitivity and uncertainty analysis. The idea is using Sobol functional decomposition with generalized polynomial chaos (gPC) expansions to estimate the responses of interest [4]. The gPC coefficients are then computed via the spectral projection method with sparse grid approach. In addition, this strategy includes evaluation of the second or higher order sensitivity indices, allowing the study of interactions between several input variables. The principal focus of this analysis is to quantify the uncertainties in the total power and maximum fuel temperatures with respect to uncertainties in the initial transient conditions, such as rated power, fuel temperature, graphite fuel density and control rod boron concentration. In addition, both the first order and second order time-dependent sensitivity indices are computed and analyzed. The results show that the interactions between initial control rod boron concentration and the rated power will contribute at most 1% to the total uncertainty of the power peak.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

NOBLE METAL MASS TRANSPORT MODEL FOR MOLTEN SALT REACTOR ANALYSIS IN VERA-CS

Molten Salt Reactor designs require robust multiphysics modelling capabilities to further the development of this next generation technology. Due to the unique liquid fuel design, traditional reactor physics analysis codes do not capture all of the important physics of these reactors. This work builds upon a general species transport model implemented into the multiphysics core simulator suite VERA-CS, by adding source and sink models for non-soluble fission products, specifically the noble metals. Noble metal fission products are born in the fuel-salt, but do not form stable fluorides, and instead plate out on various surfaces in the reactor fuel loop. A concentration gradient diffusion-convection based mass transfer model is implemented into the general species transport model within CTF-the subchannel thermal hydraulics code in VERA-CS-and verified. Two noble metal decay chains are analyzed, and their steady state bulk and surface concentrations are calculated for a simple flow loop in CTF that is representative of the Oak Ridge National Laboratory Molten Salt Reactor Experiment (MSRE). Non-soluble species 99Mo to 99Tc captures wall plate out of decaying species, and non-soluble 135Te to soluble 135I and gaseous 135Xe captures wall loss of decaying species. Results are compared with MSRE data and the possible multiphysics effects of this phenomena are identified. Future work to enhance the robustness of the physical model and couple it with neutronics to evaluate these effects of Noble Metal mass transport is discussed.

Walker, Samuel↗
DOE OSTI2012

Application of fully ceramic microencapsulated fuels in light water reactors

This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Benchmarking of calculation schemes in APOLLO2 and COBAYA3 for WER lattices

This paper presents solutions of the NURISP WER lattice benchmark using APOLLO2, TRIPOLI4 and COBAYA3 pin-by-pin. The main objective is to validate MOC based calculation schemes for pin-by-pin cross-section generation with APOLLO2 against TRIPOLI4 reference results. A specific objective is to test the APOLLO2 generated cross-sections and interface discontinuity factors in COBAYA3 pin-by-pin calculations with unstructured mesh. The VVER-1000 core consists of large hexagonal assemblies with 2 mm inter-assembly water gaps which require the use of unstructured meshes in the pin-by-pin core simulators. The considered 2D benchmark problems include 19-pin clusters, fuel assemblies and 7-assembly clusters. APOLLO2 calculation schemes with the step characteristic method (MOC) and the higher-order Linear Surface MOC have been tested. The comparison of APOLLO2 vs. TRIPOLI4 results shows a very close agreement. The 3D lattice solver in COBAYA3 uses transport corrected multi-group diffusion approximation with interface discontinuity factors of Generalized Equivalence Theory (GET) or Black Box Homogenization (BBH) type. The COBAYA3 pin-by-pin results in 2, 4 and 8 energy groups are close to the reference solutions when using side-dependent interface discontinuity factors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Development status of the lattice physics code in COSINE project

LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Application of power time-projection on the operator-splitting coupling scheme of the TRACE/S3K coupled code

This paper reports refinement studies on the temporal coupling scheme and time-stepping management of TRACE/S3K, a dynamically coupled code version of the thermal-hydraulics system code TRACE and the 3D core simulator Simulate-3K. The studies were carried out for two test cases, namely a PWR rod ejection accident and the Peach Bottom 2 Turbine Trip Test 2. The solution of the coupled calculation, especially the power peak, proves to be very sensitive to the time-step size with the currently employed conventional operator-splitting. Furthermore, a very small time-step size is necessary to achieve decent accuracy. This degrades the trade-off between accuracy and performance. A simple and computationally cheap implementation of time-projection of power has been shown to be able to improve the convergence of the coupled calculation. This scheme is able to achieve a prescribed accuracy with a larger time-step size. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Quantification of reactivity margin in BWR spent fuel pools

Boiling water reactor (BWR) spent fuel pool (SFP) criticality safety analyses typically use the in-core k{sub inf} criterion methodology to demonstrate margin to safety. This method models all the fuel in the SFP with a limiting lattice design at a peak reactivity just above a specific in-core k{sub inf} limit. To bound all the fuel in the pool and future fuel with reasonable margin, these limits are often set conservatively high compared to the peak reactivities of actual BWR bundle designs. When current regulatory expectations are met for the quantification of bias and uncertainty adders to the criticality analyses, some results are quantified that are close to the 0.95 in-rack reactivity limit required by 10CFR50.68. The small margins being calculated in the criticality safety analyses have resulted in concerns that SFPs have less margin to safety than was present in the past; however, inherent in the peak reactivity in-core k{sub inf} criterion methodology used to assess these margins is a very conservative model of the fuel. While it is known that all the fuel in a BWR SFP is typically burned well past its peak reactivity exposure, it has been unknown how much reactivity margin could be quantified if bundles were modeled more realistically by accounting for their actual discharge exposure profiles, operational void fractions, control histories, and depletion fuel temperatures. In this study, fuel bundles were depleted to their explicit nodal discharge exposure while accounting for actual operational conditions defined by GEH/GNF's core simulator code, PANACEA. The use of these depleted isotopics in the criticality safety analysis is commonly referred to as taking 'burn-up credit' in the analysis. This paper provides a quantification of the reactivity margin available from burn-up credit in two currently operating BWR spent fuel pools. The evaluation demonstrates that many of the bundles in these storage configurations are significantly below the in-rack reactivity limit, even with the necessary bias and uncertainty adders included. The utilization of BWR burnup credit appears to be an attractive solution to consider when evaluating options for reduced neutron absorber credit or for quantifying additional margin to the regulatory in-rack k{sub eff} limit. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Verification of the pin power reconstruction method in NESTLE

NESTLE is a 3D nodal core simulator that solves the few group neutron diffusion equation using the nodal expansion method (NEM) for the steady state (eigenvalue or fixed-source), adjoint (eigenvalue), and transient (eigenvalue or fixed-source) problems. NESTLE has been in use for over 25 years and has been continuously improved and developed. NESTLE can be used to model either Cartesian geometries (for use with typical PWR, BWR fuel assemblies) or hexagonal geometries (for use with VVER and other triangularly pitched fuel assemblies). As part of NESTLE's ongoing verification and validation process, several computational comparisons have been made to quantify the discrepancies in NESTLE's hexagonal pin power reconstruction method. In this paper, we present the results of these code-to-code comparisons and find that the root mean squared (RMS) error in pin power is comparable to other nodal codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of an advanced multi-scale subchannel model for CRUD risk assessments

AREVA NP has developed a process, based upon standard licensing methods and specialized applications, for the assessment of the risk of crud induced corrosion and crud induced axial off-set. This streamlined process allows for the evaluation of the crud risk for full core subchannel by subchannel models, by applying local conditions on each fuel rod/subchannel, with axial nodes spanning several centimeters. AREVA NP has applied these tools successfully to units from various NSSS vendors. To date, there have been no plant performance issues that have been outside the performance as predicted by the AREVA NP risk assessment tools. The study presented in this paper utilized the core simulator and thermal-hydraulic packages of this tool set to evaluate the effects of changing various fuel cycle design (e.g. radial zoning patterns) and fuel assembly design parameters (e.g. fuel rod diameter, spacer grid placement) on thermal-hydraulic key decision variables (KDVs) of interest. The study may be valuable as a first step towards optimizing the fuel cycle design, optimizing the fuel lattice design, or evaluating potential operating changes such as a power up-rate or modification of the cycle length for a given PWR plant. The various geometric inputs to the AREVA NP risk assessment process were tested by individually adjusting a single parameter and observing the effect. Multi-parameter studies are also included to assess the impact of optimized fuel cycle designs. The sensitivity of the model, in terms of thermal-hydraulic results, is presented. Provided a relationship between the thermal-hydraulic results to both crud deposition thickness and lithium uptake in the zirconium cladding oxide layer, approximate impacts to the risk of crud induced corrosion and crud induced axial off-set can be estimated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Uncertainty and Sensitivity Analysis of TREAT Transient Test no. 15

The U.S. Department of Energy (DOE) is preparing for the resumption of transient testing using the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL). In order to support the analysis of TREAT operation and experiment planning, the MAMMOTH reactor physics team has recently begun development of uncertainty and sensitivity analysis for the TREAT core simulated via MAMMOTH. MAMMOTH is a high-fidelity three-dimensional transient multi-physics code developed at INL under the DOE-NE Advanced Modeling and Simulation program (NEAMS). This research is part of the development presenting some initial uncertainty and sensitivity analysis of TREAT Transient Test no. 15. In this summary, we employ the Sobol sampling strategy implemented in RAVEN to efficiently perform the sensitivity and uncertainty analysis. The idea is using Sobol functional decomposition with generalized polynomial chaos (gPC) expansions to estimate the responses of interest. The gPC coefficients are then computed via the sparse grid spectral projection method. In addition, this strategy includes evaluation of the second or higher order sensitivity indices, allowing the study of interactions between several input variables. The principal focus of this analysis is to quantify the uncertainties in the total power and maximum fuel temperatures with respect to uncertainties in the initial transient conditions, such as rated power, fuel temperature, graphite fuel density and control rod boron concentration. In addition, both the first order and second order time-dependent sensitivity indices are computed and analyzed. The results show that the interactions between initial control rod boron concentration and the rated power will contribute at most 1% to the total uncertainty of the power peak.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Generalization of NESTLE into a Multi-Energy N-Group Formulation

The NESTLE code is a few-group, diffusion-based core simulation code that employs the Nodal Expansion Method (NEM). NESTLE has the capability to solve both steady-state and transient problems, using either Cartesian or hexagonal geometry, and to apply thermal-hydraulic feedback including single and two-phase flow. A modern version of the NESTLE nodal-diffusion simulator is currently maintained and developed at the University of Tennessee. The original version of the code was limited to only two or four neutron energy groups, therefore, this article describes recent work carried out to expand NESTLE into a general multi-group solver. Generalization of a fixed group finite-difference diffusion theory into arbitrary N groups is a well-known problem [3], however, this paper describes the intricacies associated with the generalization of the Nodal Expansion Method, and the methodical and computational challenges encountered and managed along the way.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

An improved multipole approximation for self-gravity and its importance for core-collapse supernova simulations

Self-gravity computation by multipole expansion is a common approach in problems such as core-collapse and Type Ia supernovae, where single large condensations of mass must be treated. The standard formulation of multipole self-gravity in arbitrary coordinate systems suffers from two significant sources of error, which we correct in the formulation presented in this article. The first source of error is due to the numerical approximation that effectively places grid cell mass at the central point of the cell, then computes the gravitational potential at that point, resulting in a convergence failure of the multipole expansion. We describe a new scheme that avoids this problem by computing gravitational potential at cell faces. The second source of error is due to sub-optimal choice of location for the expansion center, which results in angular power at high multipole l values in the gravitational field, requiring a high—and expensive—value of multipole cutoff l {sub max}. By introducing a global measure of angular power in the gravitational field, we show that the optimal coordinate for the expansion is the square-density-weighted mean location. We subject our new multipole self-gravity algorithm, implemented in the FLASH simulation framework, to two rigorous test problems: MacLaurin spheroids for which exact analytic solutions are known, and core-collapse supernovae. We show that key observables of the core-collapse simulations, particularly shock expansion, proto-neutron star motion, and momentum conservation, are extremely sensitive to the accuracy of the multipole gravity, and the accuracy of their computation is greatly improved by our reformulated solver.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

Neutrino-driven Winds in Three-dimensional Core-collapse Supernova Simulations

Abstract In this paper, we analyze the neutrino-driven winds that emerge in 12 unprecedentedly long-duration 3D core-collapse supernova simulations done using the code F ornax . The 12 models cover progenitors with zero-age main-sequence mass between 9 and 60 solar masses. In all our models, we see transonic outflows that are at least 2 times as fast as the surrounding ejecta and that originate generically from a proto−neutron star surface atmosphere that is turbulent and rotating. We find that winds are common features of 3D simulations, even if there is anisotropic early infall. We find that the basic dynamical properties of 3D winds behave qualitatively similarly to those inferred in the past using simpler 1D models, but that the shape of the emergent wind can be deformed, very aspherical, and channeled by its environment. The thermal properties of winds for less massive progenitors very approximately recapitulate the 1D stationary solutions, while for more massive progenitors they deviate significantly owing to aspherical accretion. The Y e temporal evolution in winds is stochastic, and there can be some neutron-rich phases. Though no strong r -process is seen in any model, a weak r -process can be produced, and isotopes up to 90 Zr are synthesized in some models. Finally, we find that there is at most a few percent of a solar mass in the integrated wind component, while the energy carried by the wind itself can be as much as 10%–20% of the total explosion energy.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2019

Towards an understanding of the resolution dependence of Core-Collapse Supernova simulations

ABSTRACT Using our new state-of-the-art core-collapse supernova (CCSN) code Fornax, we explore the dependence upon spatial resolution of the outcome and character of three-dimensional (3D) supernova simulations. For the same 19 M⊙ progenitor star, energy and radial binning, neutrino microphysics, and nuclear equation of state, changing only the number of angular bins in the θ and ϕ directions, we witness that our lowest resolution 3D simulation does not explode. However, when jumping progressively up in resolution by factors of two in each angular direction on our spherical-polar grid, models then explode, and explode slightly more vigorously with increasing resolution. This suggests that there can be a qualitative dependence of the outcome of 3D CCSN simulations upon spatial resolution. The critical aspect of higher spatial resolution is the adequate capturing of the physics of neutrino-driven turbulence, in particular its Reynolds stress. The greater numerical viscosity of lower resolution simulations results in greater drag on the turbulent eddies that embody turbulent stress, and, hence, in a diminution of their vigor. Turbulent stress not only pushes the temporarily stalled shock further out, but bootstraps a concomitant increase in the deposited neutrino power. Both effects together lie at the core of the resolution dependence we observe.

79 ASTRONOMY AND ASTROPHYSICS↗