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At least 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Total Residual Radiation Source Term Produced by the Hiroshima Detonation

Residual radiation from a nuclear detonation consists of fission products and activation products produced by the excess neutrons released during the explosion. The ability to accurately predict nuclear fallout begins with a calculation of the residual radiation source term. The intent of this paper is to define the residual radiation source term for the Hiroshima detonation. The calculation was performed using the Livermore Weapon Activation Code (LWAC), which is a simplified three-region model that estimates the residual radiation associated with the fission products, the unburnt fuel, and activation products produced in the weapon components, the surrounding air, and the ground in the vicinity of ground zero. For the Hiroshima detonation, ~1200 radionuclides were produced. A time-dependent solution for the total activity of each of the major residual radiation source terms is provided in the Appendices. In addition, the timedependent exposure rates produced by the activation rings are included in the report.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2016

limiting rods selection in large break LOCA analysis with the RELAP5-3D code

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10 CFR 50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently more fuel rods need to be analyzed under LOCA conditions to maintain the safety margin, compared to the current practice in which only one hot rod needs to be analyzed. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods under LOCA conditions. In this work, the best-estimate plus uncertainty analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. With the new analysis method, the limiting transient case and the limiting rods can be easily identified in response to the proposed new rulemaking. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

Miao, Yinbin (ORCID:0000000231284275)↗
DOE OSTI2019

Initial Neutronics and Thermal-Hydraulic Coupling for Spent Nuclear Fuel Canister

This report documents work performed supporting the US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.05, “Direct Disposal of Dual Purpose Canisters.” In particular, this appendix fulfills the M3 milestone, M3SF-19OR010305015, “Multiphysics criticality consequence analysis capability development status report,” as Revision 1 to M3SF-19OR010305016, “Initial neutronic and thermal hydraulic coupling for waste package” within work package SF-19OR01030501, “Direct Disposal of Dual Purpose Canisters–ORNL.” This report presents the initial development status of a multiphysics criticality consequence simulation framework, Terrenus. Terrenus currently couples two physics codes: the Monte Carlo radiation transport code Shift and the thermal-hydraulics code COBRA-SFS. The coupling capability has been demonstrated using a simplified 3 x 3 fuel pin lattice. Terrenus will be developed further to include (1) general geometry package support, allowing modeling of a spent nuclear fuel (SNF) cask with all structural details, (2) a depletion solver to determine the change in nuclide composition at the end of each time step, and (3) a mechanics code to determine any structural impact due to a criticality event. Future research and development works include identifying/developing a modern two-phase thermal-hydraulics code and developing an approach to solve fast neutronic transients. The goal of this multiphysics framework is to determine the feasibility of direct disposal of currently loaded SNF canisters by including or excluding a criticality event from a repository performance analysis framework in terms of consequences. The revision 1 to this report includes extending Terrenus to a 17 x 17 pressurized water reactor (PWR) fuel assembly. We also investigated the phenomena necessary to quench a critical reaction, and investigated the water level necessary to achieve criticality in a typical canister. Finally, we describe the initial implementation of a critical power search capability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of isotopic measurements and burn-up value of sample M11 of rebus project - 14404

Estimation of the burn-up value and the corresponding isotopic composition of the nuclear fuel are the most important subjects for criticality analysis and source term estimation in the framework of spent fuel management. The only practical way to estimate the fuel composition and burn-up value is the isotopic depletion calculation with well qualified codes and nuclear data. Such qualification requires the availability of well evaluated experimental measurements to be used as benchmarks. The purpose of this study is evaluation of the isotopic measurements of M11 sample in REBUS project in order to judge their adequacy for qualification of the codes and libraries. Additionally the sensitivity analyses to manufacturing and operation parameters are performed in order to obtain the uncertainty due to both types of parameters. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

PeleMP: The Multiphysics Solver for the Combustion Pele Adaptive Mesh Refinement Code Suite

Combustion encompasses multiscale, multiphase reacting flow physics spanning a wide range of scales from the molecular scales, where chemical reactions occur, to the device scales, where the turbulent flow is affected by the geometry of the combustor. This scale disparity and the limited measurement capabilities from experiments make modeling combustion a significant challenge. Recent advancements in high-performance computing (HPC), particularly with the Department of Energy's Exascale Computing Project (ECP), have enabled high-fidelity simulations of practical applications to be performed. The major physics submodels, including chemical reactions, turbulence, sprays, soot, and thermal radiation, exhibit distinctive computational characteristics that need to be examined separately to ensure efficient utilization of computational resources. This paper presents the multiphysics solver for the Pele code suite, called PeleMP, which consists of models for spray, soot, and thermal radiation. Here, the mathematical and algorithmic aspects of the model implementations are described in detail as well as the verification process. The computational performance of these models is benchmarked on multiple supercomputers, including Frontier, an exascale machine. Results are presented from production simulations of a turbulent sooting ethylene flame and a bluff-body swirl stabilized spray flame with sustainable aviation fuels to demonstrate the capability of the Pele codes for modeling practical combustion problems with multiphysics. This work is an important step toward the exascale computing era for high-fidelity combustion simulations providing physical insights and data for predictive modeling of real-world devices.

42 ENGINEERING↗
DOE OSTI2020

Multiphysics for nuclear energy applications using a cohesive computational framework

With the recent development of advanced numerical algorithms, software design, and low-cost high-performance computer hardware, reliance on coupled multiphysics to predict the behavior of complex physical systems is beginning to become standard practice. This is especially true in nuclear energy applications where strong nonlinear interdependencies exist between reactor physics, radiation transport, multi-scale nuclear fuels performance, thermal fluids, etc. Resolving these nonlinear dependencies requires choices in multiphysics software approaches. Two main multiphysics modeling and simulation approaches have emerged. The first is based upon "code coupling" where disparate physics codes of different software design, code languages, and spatial and temporal integration schemes are coupled together with relatively complex data passing interfaces. The second multiphysics software approach is to employ a "cohesive" framework where all physics applications are developed with a common software design, i.e., data structures, syntax, input format, integrated spatial and temporal discretization schemes, etc. In this paper we present the Multiphysics Object-Oriented Simulation Environment (MOOSE) development and runtime framework and describe the framework's cohesive modeling and simulation multiphysics approach. Then, a "cohesive-like" extension of the MOOSE framework is presented where MOOSE-based physics software applications are efficiently coupled to non-MOOSE (external) physics codes to form multiphysics applications using MOOSE's unique interface capabilities. Finally, several examples of MOOSE's cohesive and cohesive-like multiphysics applications will be demonstrated. These multiphysics demonstrations will incorporate both MOOSE-based applications and external codes, including Nek5000, RELAP-7, TRACE, BISON, and Pronghorn.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Implementation of a Spacer Grid Rod Thermal-Hydraulic Reconstruction (ROTHCON) Capability into the Thermal-Hydraulic Subchannel Code CTF

The Consortium for Advanced Simulation of Light Water Reactors is developing a core simulator capability known as the Virtual Environment for Reactor Applications (VERA) to address nuclear industry challenge problems such as crud-induced power shift (CIPS). The CTF thermal-hydraulic (T/H) subchannel code provides thermal feedback in the coupled neutronics, T/H, crud chemistry simulation that VERA performs. It has been discovered that the coarse meshing approach used by CTF (in which fuel rods are discretized into four azimuthal segments) can be a source of error in predicting crud growth and boron distribution in VERA CIPS calculations. Spacer grid effects lead to complex rod-to-fluid heat transfer behavior that, when not resolved, can lead to error in the prediction of crud growth and boron deposition. A higher-fidelity computational fluid dynamics approach can be used instead of CTF, but this leads to excessive simulation times. This paper presents an approach for using high-fidelity computational fluid dynamics data to create shape functions that are used in CTF to reconstruct rod surface heat transfer behavior as a function of spacer grid geometry. The approach is demonstrated for a 5 × 5 rod bundle facility with five mixing vane grids under a range of operating conditions encountered in nominal pressurized water reactor conditions. Furthermore, it is demonstrated that the grid heat transfer maps are successful at introducing a higher-fidelity heat transfer modeling capability into CTF.

CFD-informed↗
DOE OSTI2013

Compilation of Quality Assurance Documentation for Analyses Performed for the Resumption of Transient Testing Environmental Assessment

This is a companion document to the analyses performed in support of the environmental assessment for the Resumption of Transient Fuels and Materials Testing. It is provided to allow transparency of the supporting calculations. It provides computer code input and output. The basis for the calculations is documented separately in INL (2013) and is referenced, as appropriate. Spreadsheets used to manipulate the code output are not provided.

99 GENERAL AND MISCELLANEOUS↗
NASA NTRS1979

Computerized systems analysis and optimization of aircraft engine performance, weight, and life cycle costs

The computational techniques utilized to determine the optimum propulsion systems for future aircraft applications and to identify system tradeoffs and technology requirements are described. The characteristics and use of the following computer codes are discussed: (1) NNEP - a very general cycle analysis code that can assemble an arbitrary matrix fans, turbines, ducts, shafts, etc., into a complete gas turbine engine and compute on- and off-design thermodynamic performance; (2) WATE - a preliminary design procedure for calculating engine weight using the component characteristics determined by NNEP; (3) POD DRG - a table look-up program to calculate wave and friction drag of nacelles; (4) LIFCYC - a computer code developed to calculate life cycle costs of engines based on the output from WATE; and (5) INSTAL - a computer code developed to calculate installation effects, inlet performance and inlet weight. Examples are given to illustrate how these computer techniques can be applied to analyze and optimize propulsion system fuel consumption, weight, and cost for representative types of aircraft and missions.

Fishbach, L. H.↗
NASA NTRS1990

RAMSCRAM: A flexible ramjet/scramjet engine simulation program

With the resurgence of interest in high supersonic and hypersonic flight there is a need to simulate airbreathing engines which may be used in this flight regime. To meet this requirement the RAMSCRAM code was developed. The code calculates 1-D flow properties at each component interface and the overall performance of the engine. It uses equilibrium thermodynamics which accounts for dissociation and allows for any fuel or combination of fuels. The program can simulate ramjet, scramjet, rocket, and ducted rocket engines.

Burkardt, Leo A.↗
NASA NTRS2019

Risk Analysis Associated with Loss of Toxic Gases During Orion Landing and Recovery Operations

Mission, landing and recovery operations for the Orion crew module involve reentry into the Earth's atmosphere and the deployment of three Nomex parachutes to slow the descent before landing along the west coast of the United States. Orion may have residual fuel (hydrazine, N2H4) or coolant (ammonia, NH3) on board which are both highly toxic to crew in the event of exposure. These risks were evaluated using a first principles analysis approach through fluid dynamics modeling. Plume calculations were first performed with the ANSYS Fluent computational fluid dynamics code. Data were then extracted at locations relevant to crew safety such as the snorkel fan inlet and the egress hatch. Mixing calculations were performed to quantify exposure concentrations within the crew bay before and during egress and departure. Finally, results included herein were used to inform the Orion post-landing Concept of Operations (ConOps) so that strategies could be formulated to maintain crew safety in the event of the loss of fuel or coolant.

Orion↗
DOE OSTI2015

Effectiveness of actinide management in fast reactors over a finite recycling period - 15123

Long-lived nuclear waste primarily consists of transuranic (TRU) isotopes. Full recycling of TRU isotopes can, in theory, lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited time frame, the radiotoxicity of the 'final' core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In this paper, the fuel cycle code ORION is used to model the recycle of light water reactor (LWR)-produced TRUs in sodium-cooled fast reactors (SFRs). With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. Break-even SFRs are much less effective from a point of view of reducing waste radiotoxicity. Thorium (Th)-burner fuel cycles are found to result in slightly lower radiotoxicity than U-burner fuel cycles for the same number of generations. Closed break-even Th fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years, provided that Pa is recycled. While this is a very long time frame, it is substantially better than the time frame required by other strategies. The integrated decay heat over the scenario time frame is substantially higher for break-even Th-SFRs than break-even U-SFRs as a result of much higher {sup 90}Sr production. As decay heat affects repository sizing, this weakens the argument for the Th cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1987

Chemical reacting flows

Future aerospace propulsion concepts involve the combination of liquid or gaseous fuels in a highly turbulent internal air stream. Accurate predictive computer codes which can simulate the fluid mechanics, chemistry, and turbulence combustion interaction of these chemical reacting flows will be a new tool that is needed in the design of these future propulsion concepts. Experimental and code development research is being performed at Lewis to better understand chemical reacting flows with the long term goal of establishing these reliable computer codes. The approach to understanding chemical reacting flows is to look at separate simple parts of this complex phenomena as well as to study the full turbulent reacting flow process. As a result research on the fluid mechanics associated with chemical reacting flows was initiated. The chemistry of fuel-air combustion is also being studied. Finally, the phenomena of turbulence-combustion interaction is being investigated. This presentation will highlight research, both experimental and analytical, in each of these three major areas.

Mularz, Edward J.↗
DOE OSTI2025

High-Fidelity Modeling of Fuel-To-Coolant Thermomechanical Transport Behaviors Under Transient Conditions

This report summarizes the work completed under NEUP project number 21-24006. The objectives of this project are to advance the high-fidelity modeling capabilities and important phenomena that is important for high-burnup UO 2 and accident tolerant fuels (ATF) during transient conditions. Accurate modeling of the time-dependent phenomena that impact material performance must be used to determine the figures of merit and safety margin. Phenomena such as fuel fragmentation, cladding oxidation, pellet-clad interaction, clad ballooning, and clad rupture are examples that pose challenges to modeling during these transients. This project focused on the development of high-fidelity tightly coupled multiphysics models that can capture the time-dependent material response and associated thermal hydraulic conditions during these events. These models can then be validated against existing separate effects tests and in-pile integral experiments and will be used to model Transient Reactor Test facility (TREAT) loss-of-coolant accidents (LOCA) experiments. To achieve the project objective, we used a combination of NEAMS and NRC codes to model various LOCA test sets for the separate effects and in-pile integral experiments. BlueCRAB tool set, which can accurately predict material response at a sub-fuel pin level, as well as modeling the entire reactor system response to these events. Fuel performance was modeled using BISON (various versions) and FAST (version 1.2.1). BISON and FAST can model on a sub-fuel pin level the fuel performance under transient conditions. Both have simplified thermal hydraulic models that are capable of providing basic coolant boundary conditions. To better capture the thermomechanical interaction between the fuel, clad, and coolant, more sophisticated thermal hydraulic models are necessary.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Monte Carlo/CFD Coupling for Accurate Modeling of the Delayed Neutron Precursors and Compressibility Effects in Molten Salt Reactors

Multiphysics reactor analysis has become one of the leading activities of most nuclear engineering research groups. Coupled neutronics and thermal/hydraulics (T/H) simulations have traditionally been performed by adopting deterministic transport tools and reactor system T/H codes. The main reason for avoiding higher fidelity numerical methods like continuous-energy Monte Carlo analysis has been the high computational cost of coupled, full-scale simulations. Nowadays, high fidelity multiphysics simulations represent a hot research topic, with several applications involving advanced nuclear systems. Liquid fuel Molten Salt Reactors (MSRs) feature peculiar physical phenomena that are not present in solid fuel reactors. Some of these phenomena are not correctly resolved by legacy reactor physics tools, and require specific treatments. This work focuses in particular on the analysis of the effect of the delayed neutron precursors drift and the impact of liquid fluid thermal expansion and compressibility in postulated criticality excursion transients in fast-spectrum MSRs. A new modeling tool based on a Monte Carlo approach for neutronics (Serpent code) coupled to a Computational Fluid Dynamic (CFD) analysis (OpenFOAM) was developed to obtain accurate modeling and simulation of these phenomena. The development of the Serpent Monte Carlo code [1] at VTT in the past few years has made this code particularly attractive for performing multiphysics studies. The availability of a flexible multiphysics interface, on-the-fly Doppler broadening and density-based collision rejection, and Delta-tracking, proved of paramount importance for the computationally efficient performance of coupled Monte Carlo/CFD calculations [2]. OpenFOAM is a C++ multiphysics toolkit featured by automatic matrix construction and solution capabilities for scalar and vector partial differential equations (PDEs), adopting several possible differencing and interpolation schemes, based on the Finite-Volume method. The ease of implementing the constitutive equations of different physical phenomena makes OpenFOAM particularly suitable for the modeling of complex coupled phenomena. In the past, several research activities proposed the adoption of both Serpent and OpenFOAM for the modeling and simulation of MSRs and the associated fuel cycles. In this work, these tools have been internally coupled for the steady-state and transient analysis of fast-spectrum systems, accounting for the effect of delayed neutron precursors transport and the salt thermal expansion and compressibility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

VERA 3.3 Release Notes

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics-fuel performance problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. Two neutronics capabilities are provided in this distribution including MPACT (from the University of Michigan) that supports 2D lattices and Insilico (from Oak Ridge National Laboratory) that supports 2D lattices, 2D core planes, and 3D core geometries. Both codes include integrated cross section capabilities that provide problem-specific cross sections for the problems defined. A thermal-hydraulics capability is provided with CTF (an updated version of the COBRA-TF code developed by the Pennsylvania State University) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes an initial coupled neutronics/thermal-hydraulics capability to allow calculations of single assemblies using Insilico or MPACT coupled with CTF. The VERA fuel rod performance subcomponent BISON-CASL calculates, on a 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Specifically targeted at VERA’s core simulator functionality (VERA-CS), the existing VERA subchannel-neutronics capability based on CTF and Insilico has been extended to use the axisymmetric 2D R-Z fuel rod modeling functionality of BISON-CASL to produce a three-way coupled neutronics, thermal-hydraulics and fuel performance capability known as Tiamat. This distribution is intended to be used for testing and evaluation of the released components of VERA. Testing within CASL has focused specifically on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Conceptual System Design of a Supercritical CO{sub 2} cooled Micro Modular Reactor - 18492

Small modular reactor (SMR) concepts proposed in the past have compact size and a longer life reactor core than the conventional nuclear power plants. However, previous concepts based on the water-cooled technologies with a steam Rankine cycle have shown that the complete modularization is still technically challenging. This study suggests an innovative concept of a reactor directly cooled by supercritical carbon dioxide (S-CO{sub 2}) with S-CO{sub 2} Brayton cycle as a power conversion system, namely KAIST MMR (Micro Modular Reactor). To enhance neutron economy and maximize the fuel volume fraction of the core, a uranium carbide fuel controlled by drum type control rods was used in the fast-spectrum MMR. The core has long life (20 years) without refueling or reshuffling as well as inherent safety features. The reactor can be used as a distributed power source, and replace outdated fossil fuel power plants for small cities. Moreover, we propose a S-CO{sub 2} Brayton cycle as a power conversion system to achieve small, high performance and lightweight module. Thanks to the compact core and power conversion system, the entire system can be contained in a single module. The target of the system is to achieve small and compact system that can be transported via ground transportation of a single module containing the core and the power conversion system altogether. In order to meet this target single module s total weight is minimized in the order of 150 tons. The external size of a module is less than 7 m in length and 4 m in diameter. It produces 12 MWe electricity from 36 MWt reactor core. The S-CO{sub 2} Brayton cycle was optimized, and the cycle components were designed while maximizing the performance and minimizing the weight. A passive decay heat removal system utilizing shell and tube heat exchangers was also developed, and examined its performance with transient analysis code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗