Reducing Energy Demand in Buildings Through State Energy Codes
Building Codes Project for the 2013 Building Technologies Office's Program Peer Review
2013 BTO Program Peer Review↗
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Results for “Coding”
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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Southeast Enertgy Efficiency Alliance's Building Energy Codes Project
Building Codes Project for the 2013 Building Technologies Office's Program Peer Review
2013 BTO Program Peer Review↗
EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes
The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.
Code Validation↗
National Electric Code Scope
Presentation on the issue of the limits of the scope of the National Electric Code and the likelihood that it will be resolved in the next cycle of its revision. Includes a timeline for public input.
17 WIND ENERGY↗
National Electric Code Status
This presentation provides a synopsis of changes impacting the small wind industry in the 2017 National Electric Code.
17 WIND ENERGY↗
ASME Graphite code considerations
This summarizes the INL assessment of the ASME Code for graphite
36 MATERIALS SCIENCE↗
Modeling of Fission Gas Behavior in Uranium Oxide Nuclear Fuel Applied to Engineering Fuel Performance Codes
This presentation provides an update on the I-NERI project 2017-004-E between INL and JRC-Karlsruhe (European Commission). I-NERI is a collaboration initiative between the European institutions and the US DOE which favors collaborative work. INL is involved in multiple INERI projects, principally on the experimental side but also on the modeling side with this project, which focuses on fission gas modeling. This year's I-NERI meeting is in the USA (LLNL) and G. Pastore, who is the US PI for this, will provide the update. The project has led to fission gas developments that are being applied in the Bison code and strengthened collaborations of INL in this area with JRC and POLIMI.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
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↗
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↗
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↗
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↗
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↗
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↗
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↗
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↗
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↗
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↗