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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 1,783 records · Page 99
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Grid Code Essentials and Streamlining Process for Interconnections

This presentation was part of a webinar entitled "A Practitioner's Guide to Implementing Solar Rooftop Programs and Navigating Net-Metering Policies," hosted by USAID Clean Power Asia and included speakers the Hawaii Natural Energy Institute as well as NREL. This presentation outlines grid codes in general, and IEEE Std 1547-2018 specifically, and discusses their adoption into interconnection requirements. Key capabilities, now enabled by IEEE Std 1547-2018, such as voltage regulation and frequency ride-through are described.

DER↗
DOE OSTI2012

Duct Insulation and Sealing Requirements in Commercial Buildings - Code Notes

Duct insulation and sealing, especially insulated supply ducts delivering conditioned air within a building, save energy. The intent of energy-efficiency codes, as related to duct insulation and sealing, is to keep mechanically warmed or cooled air as close to a constant, desired temperature as possible and prevent the conditioned air from escaping the duct system while it is being moved to spaces where it is needed. If reduced heat transfer through insulated ducts is accounted for in the heating, ventilating, and air conditioning (HVAC) load calculations, it may even be possible to reduce the size of HVAC equipment.

Duct systems, thermal envelope, conditioned air, H↗
DOE OSTI2021

Integrated FY21 Elevated Temperature Mechanical Testing Results for Alloy 709 Code Case

This report provides the status of creep, fatigue, and creep-fatigue testing that transpired in Fiscal Year 2021 at Argonne National Laboratory, Idaho National Laboratory and Oak Ridge National Laboratory. This testing is being conducted to develop the data package to qualify Alloy 709 in Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. This would permit the use of Alloy 709 for elevated-temperature nuclear construction. Preliminary results continue to demonstrate the improved creep and fatigue resistance of Alloy 709 compared to 316H stainless steel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Consumer Guide to Recycling Codes

Learn about the recycling symbols that indicate whether products or packaging can be recycled. This fact sheet from Energy Saver describes the recycling codes for plastic, glass, and paper that indicate how materials should be recycled or disposed of.

recycling, recycling codes, Energy Saver↗
DOE OSTI2022

Session 4: ASME Code Rules

General review of code rules What do they cover. What they don’t. Underlying assumptions Probability of failure Material property parameters Stresses and loads Degradation What is NonmetallicsWork Group (NWG) doing? Failure in components Redefining failure Degradation rules Oxidation degradation Irradiation degradation Molten Salt Issues Abrasion/Erosion

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A709 Code Case Design Parameters

Joint ART Materials/AMMT Program Review to be given at DOE Headquarters on June 5-8, 2023. A709 Code Case Design Parameters regarding preliminary efforts have been started to develop design parameters from test data. Thus far, most of the data evaluated were from the first two commercial heats. Data from these commercial heats continue to support the creep strength advantage of A709 over 316H. Design parameters will be refined as more data are available. Work on the development of other design parameters will begin as data become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Comparison of direct and quasi-static methods for neutron kinetic calculations with the EDF R and D COCAGNE code

Quasi-Static (QS) methods are quite popular in the reactor physics community and they exhibit two main advantages. First, these methods overcome both the limits of the Point Kinetic (PK) approach and the issues of the computational effort related to the direct discretization of the time-dependent neutron transport equation. Second, QS methods can be implemented in such a way that they can be easily coupled to very different external spatial solvers. In this paper, the results of the coupling between the QS methods developed by Politecnico di Torino and the EDF R and D core code COCAGNE are presented. The goal of these activities is to evaluate the performances of QS methods (in term of computational cost and precision) with respect to the direct kinetic solver (e.g. {theta}-scheme) already available in COCAGNE. Additionally, they allow to perform an extensive cross-validation of different kinetic models (QS and direct methods). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

A Newton-Krylov solution to the porous medium equations in the agree code

In order to improve the convergence of the AGREE code for porous medium, a Newton-Krylov solver was developed for steady state problems. The current three-equation system was expanded and then coupled using Newton's Method. Theoretical behavior predicts second order convergence, while actual behavior was highly nonlinear. The discontinuous derivatives found in both closure and empirical relationships prevented true second order convergence. Agreement between the current solution and new Exact Newton solution was well below the convergence criteria. While convergence time did not dramatically decrease, the required number of outer iterations was reduced by approximately an order of magnitude. GMRES was also used to solve problem, where ILU without fill-in was used to precondition the iterative solver, and the performance was slightly slower than the direct solution. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

An overview of the ENEA activities in the field of coupled codes NPP simulation

In the framework of the nuclear research activities in the fields of safety, training and education, ENEA (the Italian National Agency for New Technologies, Energy and the Sustainable Development) is in charge of defining and pursuing all the necessary steps for the development of a NPP engineering simulator at the 'Casaccia' Research Center near Rome. A summary of the activities in the field of the nuclear power plants simulation by coupled codes is here presented with the long term strategy for the engineering simulator development. Specifically, results from the participation in international benchmarking activities like the OECD/NEA 'Kalinin-3' benchmark and the 'AER-DYN-002' benchmark, together with simulations of relevant events like the Fukushima accident, are here reported. The ultimate goal of such activities performed using state-of-the-art technology is the re-establishment of top level competencies in the NPP simulation field in order to facilitate the development of Enhanced Engineering Simulators and to upgrade competencies for supporting national energy strategy decisions, the nuclear national safety authority, and the R and D activities on NPP designs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Assessment of the 2D MOC solver in MPACT: Michigan parallel characteristics transport code

MPACT (Michigan Parallel Characteristics Transport Code) is a new reactor analysis tool being developed by researchers at the University of Michigan as an advanced pin-resolved transport capability within VERA (Virtual Environment for Reactor Analysis). VERA is the end-user reactor simulation tool being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The MPACT development project is itself unique for the way it is changing how students perform research to achieve the instructional and research goals of an academic institution, while providing immediate value to the industry. One of the major computational pieces in MPACT is the 2D MOC solver. It is critical that the 2D MOC solver provide an efficient, accurate, and robust solution over a broad range of reactor operating conditions. The C5G7 benchmark is first used to test the accuracy of the method with a fixed set of cross-sections. The VERA Core Physics Progression Problems are then used to compare the accuracy of both the 2D transport solver and also the cross-section treatments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Mathematical framework for the analysis of dynamic stochastic systems with the RAVEN code

RAVEN (Reactor Analysis and Virtual control Environment) is a software code under development at Idaho National Laboratory aimed at performing probabilistic risk assessment and uncertainty quantification using RELAP-7, for which it acts also as a simulation controller. In this paper we will present the equations characterizing a dynamic stochastic system and we will then discuss the behavior of each stochastic term and how it is accounted for in the RAVEN software design. Moreover we will present preliminary results of the implementation. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Verification of the three-dimensional tetrahedral grid S{sub N} code THOR

In this work current capabilities implemented in the novel, arbitrary-order, tetrahedral-grid short characteristics S{sub N} radiation transport code THOR are verified based on four benchmark problems: (1) A one-group Method of Manufactured Solution (MMS) problem on a cuboidal domain, (2) an infinite medium eigenvalue problem with up-scattering, (3) a homogeneous torus and (4) a bare cube eigenvalue problem with anisotropic scattering up to order three. The first benchmark problem exercises the various spatial discretization options available in THOR: The short characteristics method in conjunction with polynomial expansions of the source and face fluxes either using the complete or Lagrange family of arbitrary orders. Using the numerical solution's order of convergence test in the framework of a mesh refinement study, correct implementation of a selection of spatial expansion orders is demonstrated for two meshes with tetrahedral aspect ratios close to unity and 50. The second benchmark problem exercises the implementation of angular fluxes on reflective boundary faces that are implicit within a mesh sweep, and up-scattering. The third benchmark problem comprises cyclic dependencies within the mesh sweep thus exercising the algorithm devised for 'breaking' the cyclic dependencies. Finally, the fourth benchmark problem, a simple bare cube, is used to test correct implementation of the anisotropic scattering capability. For all test problems THOR obtains solutions that converge to the reference/exact solution with the expected rate thereby contributing to our confidence in the correctness of its tested features in the present implementation. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Validation and bias quantification of criticality safety codes for SRNS operations

The validation and bias quantification methods employed by the Criticality Safety Engineering group at Savannah River Nuclear Solutions (SRNS) are discussed. Both SCALE and MCNP are validated against benchmark experiments, where these benchmark models have been divided into subgroups representative of the types of fissile material operations at SRNS. A quantification of the bias is made for each subgroup for each code and the three methods employed for quantification are discussed. The bias information is employed directly in the establishment of the Upper Safety Limit (USL) value of the neutron multiplication factor when using standardized documented SRNS methods. A simple example of validation and bias quantification is given and the procedure for setting the USL using the derived bias information is also presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Qualification of Numerical and Semi-Analytical Modeling Codes for Thermal Design of HLW Repository - 18455

Disposal in a geological repository is a viable option for safely isolating high-level radioactive waste. Thermal processes induced by decay heat from the radioactive waste may impact the integrity of the engineered barrier system, and may lead to fluid flow and transport; and the associated migration of radionuclides. This study looks at an analysis of thermal conduction for the emplacement of high-level waste in a repository located in bedded salt. The simulations used analytical and numerical methods to study conductive heat transport in a repository. This study concentrated on heat flow in a central waste package with contributions from adjacent waste packages and drifts. Of particular importance to this work is comparison of predictions of peak temperature at the drift wall and at the waste package surface using different solution methods. Simulation results of analytical and numerical modeling methods are compared. The results show that predictions of the codes were comparable under identical initial and boundary conditions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

RISMC Industry Application no. 1 (ECCS/LOCA) - Core Characterization Automation: Lattice Codes Interface for PHISICS/RELAP5-3D

The U.S. NRC is currently proposing rulemaking '10 CFR 50.46c' to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that the cladding performance cannot be evaluated in isolation anymore. Both cladding and ECCS performance need to be considered in a coupled manner. This also implies that models for cladding performance as well as LOCA methodologies will need to be updated for regulatory purposes. Furthermore, efforts need to be made to find the limiting case within the regulatory input space. The Risk-Informed Safety Margin Characterization (RISMC) Pathway, as part of the DOE Light Water Reactor Sustainability (LWRS) research and development program initialed several programs at the Idaho National Laboratory (INL) to support the industry in the transition to the new rule and to offer potential solutions to this LOCA/ECCS analysis. One of these programs is the Industry Application no. 1 (IA1). Its purpose is to provide to the plant owner/operator a vehicle to inform decisions to manage margins related to compliance with the proposed 10 CFR 50.46c rule. IA1's goal is to develop an Integrated Evaluation Model (IEM) to revisit how uncertainties are propagated across the physical disciplines and data involved, as well as well as how risks are evaluated in a LOCA under 10 CFR 50.46c. This IEM is called LOCA Tool-kit for the US (LOTUS) and it connects five major disciplines involved in LOCA analysis. The focus of LOTUS is to establish the automation interfaces among the five disciplines. LOTUS will utilize in a first step existing computer codes. The risk-informed margins management approach for IA1 will provide a means of quantifying the impact on the key LOCA analysis figures-of-merit like peak cladding temperature (PCT), equivalent cladding reacted (ECR), etc. of a change in LOCA analysis inputs. The information that the risk analysis and associated tools provide can then be used for decision-making and margin management. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Analysis of the VERA Core Physics Benchmark Problems using cosRMC Code

In reactor physics, two kinds of numerical methods are generally employed to solve the neutron transport equations, one of which is the deterministic method and the other one is the Monte Carlo method. Monte Carlo is widely used to provide reference solution because of its ability to deal with continuous energy cross sections and complex geometry. Just like any statistical process, the Monte Carlo method requires repetition to achieve a small relative uncertainty, and therefore may inevitably have larger simulation times than deterministic method. Fortunately, the great advancements of computer technology make it possible to apply Monte Carlo method in reactor core analysis. To promote nuclear power plant system design and simulation software self-reliance development, a Monte Carlo neutron/photon transport code cosRMC has been developed jointly by State Nuclear Power Software Development Center (SNPSDC) and Tsinghua University. The cosRMC is aimed at achieving full core calculations and analysis with high fidelity and efficiency by means of advanced methodologies and algorithms as well as high performance computing techniques. The cosRMC has many functions as criticality calculation, fixed-source calculation, burnup calculation, etc. In order to ensure its reliability, several types of verification and validation (V and V) work have been and are being carried out for cosRMC. The simulation results of cosRMC based on Problem 1 to 3 of VERA Core Physics Benchmark showed its good capability of 2D lattice and 3D assembly modeling. In this paper, some progressive verification results of multi-assembly and full core problems are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗