ASME Graphite code considerations
This summarizes the INL assessment of the ASME Code for graphite
36 MATERIALS SCIENCE↗
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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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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↗
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↗
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↗
Transient analysis of a micro-reactor using the DireWolf code suite
Transient analyses of heat pipe micro-reactors are necessary to ensure that hypothetical accident scenarios do not comprise reactor safety. Due to its small size and reliance on heat-pipes for cooling, the micro-reactor design introduced in this paper is a tightly coupled system which requires multi-physics tools to accurately model transient events. Idaho National Laboratory's DireWolf code suite based on the MOOSE framework is tailor-built to model heat-pipe reactors. This paper demonstrates DireWolf's ability to simulate the coupled thermal-neutronics transient behavior of a heat-pipe micro-reactor. The transient events presented here include an inadvertent rotation of all control drums simultaneously and a sudden complete rotation of a single control drum. A detailed description of each event is provided along with simulation results, including time dependent power and temperature distributions, and discussion. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Unstructured mesh-based multi-physics reactor transient simulation with iMC code
In this study, we present a time-dependent Monte Carlo simulation result about a complex fuel-loaded light water reactor system with thermal-hydraulic feedback using the KAIST iMC code. The intra-pin temperature distribution is calculated based on Monte Carlo-tallied detailed power distribution and the finite element heat transfer method. The temperature effect of fuel, absorber, and coolant are adequately reflected into the dynamic response by the on-the-fly Doppler broadened cross-section correction. For the time-dependent Monte Carlo transport simulation, the predictor-corrector quasi-static Monte Carlo method is used for this study. We tested the coupled analysis framework with a 17-by-17 fuel assembly loaded with the centrally-shielded burnable absorber (CSBA) fuel element. The temperature feedback effect on the material cross-section and the moderator density incurred negative reactivity in response to the external positive reactivity insertion, showing the inherent stability of the reactor system. The pin-wise power and temperature change during the system transient are presented and discussed. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NRC-INL Collaboration on Codes & Standards
This is the presentation on the NRC-INL Collaboration on Codes and Standards for presentation at the 2024 NRC Standards Forum.
99 GENERAL AND MISCELLANEOUS↗
Class B Materials Code Case
Class B Materials Code Case to include: FY25 - High temperature design methodology, overview of ASME Section III (Division 5), Design for elevated temperature service, Goal to address current gaps, evaluation of proposed rules, allowable stress development, Class A materials (existing ASME approach), variable confidence index, data extrapolation, bounding value for SEE, and next steps.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Corrigendum: Thermal dynamics in the flux-coordinate independent turbulence code GRILLIX
Here, we have discovered that one of the original articles findings, that the actual choice of the dimensionless parameter χ ∥i0 for the ion heat conductivity seemed to play no role for the simulation results, was not physical. Instead, due to a mistake in the handling of input parameters, χ ∥i0 was actually not used in the code – the dimensionless electron heat conductivity χ ∥e0 was used instead.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Modelling of electrostatic ion-scale turbulence in divertor tokamaks with the gyrokinetic code $\mathrm{COGENT}$
Here, continuum gyrokinetic simulations of electrostatic ion scale turbulence are presented for the case of a diverted (single-null) tokamak geometry. The simulation model, implemented in the finite-volume code COGENT, solves the long-wavelength limit of the full-F gyrokinetic equation for ion species coupled to a vorticity equation for electrostatic potential variations, where a fluid model is used for an electron response. The model describes the ion scale ion temperature gradient (ITG) and resistive drift modes as well as neoclassical ion physics effects. Different turbulence regimes are observed depending on the plasma profiles, and the roles of a self-consistent background electric field and an X-point geometry are explored. In particular, increasing the pedestal density gradient and the corresponding radial electric field is demonstrated to suppress the ITG turbulence, whereas the same edge plasma background can still be destabilized by the resistive modes when the plasma resistivity is increased. The effects of X-point geometry are assessed by comparing cross-separatrix simulations with counterpart calculations performed for a toroidal annulus geometry. For the simulation parameters considered, similar global behaviour is observed in both cases, whereas strong local suppression of turbulence fluctuations is demonstrated near the X-point for the case of a single-null geometry.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
A GPU ‐Accelerated 3D Unstructured Mesh Based Particle Tracking Code for Multi‐Species Impurity Transport Simulation in Fusion Tokamaks
ABSTRACT This paper presents the multi‐species global impurity transport capability developed in a GPU‐accelerated fully 3D unstructured mesh‐based code, GITRm, to simultaneously track multiple impurity species and handle interactions of these impurities with mixed‐material surfaces. Different computational approaches to model particle‐surface interaction or surface response have been developed and compared. Sheath electric field is taken into account by employing a fast distance‐to‐boundary calculation, which is carried out in parallel on distributed or partitioned meshes on multiple GPUs without the need for any inter‐process communication during the simulation. Several example cases, including two for the DIII‐D tokamak, that is, one with the SAS‐V divertor and the other with the collector probes, are used to demonstrate the utility of the current multi‐species capability. For the DIII‐D probe case, the capability of GITRm to resolve the spatial distribution of particles in localized regions, such as diagnostic probes, within non‐axisymmetric tokamak geometries is demonstrated. These simulations involve up to 320 million particles and utilize up to 48 GPUs.
Nath, Dhyanjyoti D. [Scientific Computation Resear↗