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At least 379 records · Page 21

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DOE OSTI2021

HiPACE++ v0.1

Hipace++ is the new implementation of the DESY-LBNL code Hipace (not distributed nor licensed) for plasma acceleration, relying on the quasi-static particle-in-cell (PIC) method, enabling fast and accurate simulations of plasma accelerators for problems satisfying the quasi-static approximation. Plasma acceleration is a priority in the ATAP division, and such problems represent a wide range of parameters. Hipace++ is an HPC (high-performance computing) code running on supercomputers and it provides the first efficient implementation of the quasi-static PIC method on Graphic Processing Units (GPU), showing 10x speedup and using much less resources than CPU implementations for production simulations. It currently runs on Nvidia GPUs and will ultimately be portable to more accelerated HPC platforms. The porting strategy is novel, and can be adapted to other flavors of quasi-static PIC. Besides, Hipace++ aims at using mesh refinement for higher efficiency, thus enabling simulations unfeasible today. For this purpose, Hipace++ depends on the open-source library AMReX (Adaptive Mesh Refinement at EXascale) developed at LBNL.

Vay, Jean-Luc↗

DOE OSTI2014

Insight from Fukushima Daiichi Unit 3 Investigations using MELCOR

During the emergency response period of the accidents that took place at Fukushima Daiichi in March of 2011, researchers at Oak Ridge National Laboratory (ORNL) conducted a number of studies using the MELCOR code to help understand what was occurring and what had occurred. During the post-accident period, the Department of Energy (DOE) and the US Nuclear Regulatory Commission (NRC) jointly sponsored a study of the Fukushima Daiichi accident with collaboration among Oak Ridge, Sandia, and Idaho national laboratories. The purpose of the study was to compile relevant data, reconstruct the accident progression using computer codes, assess the codes predictive capabilities, and identify future data needs. The current paper summarizes some of the early MELCOR simulations and analyses conducted at ORNL of the Fukushima Daiichi Unit 3 accident. Extended analysis and discussion of the Unit 3 accident is also presented taking into account new knowledge and modeling refinements made since the joint DOE/NRC study.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Mesoscale Modeling Framework Design: Subcontract Report

This subcontract is aimed at developing the mesoscale computational capability to help ensure success of the project “Multiscale Capability for Exploring Transport Phenomena in Battery” at the Lawrence Livermore National Laboratory (LLNL). We develop an efficient computer code that implements the smoothed boundary method for simulating charge transport and coupled microstructure evolution in battery electrode architectures.

25 ENERGY STORAGE↗

DOE OSTI2019

Improved Calculation of Two-photon opacity

This report will describe an improved computer code for two-photon opacity. The new code incorporates many recent advances and is ready to start to face the experiments. It incorporates the difficult mathematical techniques for handling free states and free-free matrix elements.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Sirepo : an open-source cloud-based software interface for X-ray source and optics simulations

Sirepo , a browser-based GUI for X-ray source and optics simulations, is presented. Such calculations can be performed using SRW ( Synchrotron Radiation Workshop ), which is a physical optics computer code, allowing simulation of entire experimental beamlines using the concept of a `virtual beamline' with accurate treatment of synchrotron radiation generation and propagation through the X-ray optical system. SRW is interfaced with Sirepo by means of a Python application programming interface. Sirepo supports most of the optical elements currently used at beamlines, including recent developments in SRW . In particular, support is provided for the simulation of state-of-the-art X-ray beamlines, exploiting the high coherence and brightness of modern light source facilities. New scientific visualization and reporting capabilities have been recently implemented within Sirepo , as well as automatic determination of electron beam and undulator parameters. Publicly available community databases can be dynamically queried for error-free access to material characteristics. These computational tools can be used for the development and commissioning of new X-ray beamlines and for testing feasibility and optimization of experiments. The same interface can guide simulation on a local computer, a remote server or a high-performance cluster. Sirepo is available online and also within the NSLS-II firewall, with a growing number of users at other light source facilities. Our open source code is available on GitHub.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2014

RADTRAN 6 Technical Manual

This Technical Manual contains descriptions of the calculation models and mathematical and numerical methods used in the RADTRAN 6 computer code for transportation risk and consequence assessment. The RADTRAN 6 code combines user-supplied input data with values from an internal library of physical and radiological data to calculate the expected radiological consequences and risks associated with the transportation of radioactive material. Radiological consequences and risks are estimated with numerical models of exposure pathways, receptor populations, package behavior in accidents, and accident severity and probability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Validation of RELAP5-3D Using HE-FUS3 Data

The RELAP5-3D computer code was validated using data from the HE-FUS3 experimental facility. HE-FUS3 is a helium-cooled, electrically-heated experimental facility that was used to generate experimental data for a benchmark exercise aimed at the validation of system transient analysis codes for very-high-temperature gas reactor applications. RELAP5-3D was validated using results from seven steady-state tests and one transient that simulated a loss-of-flow accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

MACCS Theory Manual

This report describes the models of the MACCS computer code as presented in MACCS Version 3.10.0. The purpose of MACCS is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. From a given release of radioactive material into the atmosphere, MACCS estimates the extent and magnitude of radiological contamination, offsite doses, protective actions, socioeconomic impacts and costs, and health effects. Since the weather at the time of an accident is not predictable, MACCS supports various sampling options to run a representative set of simulations to evaluate weather variability. MACCS simulates atmospheric transport with a straight-line Gaussian plume segment model. From the estimated air and ground concentrations, MACCS models dose projections through several dose exposure pathways. These exposures can be offset by protective actions during the emergency response and long-term recovery of the accident. MACCS users directly specify the evacuation and sheltering area, while other protective actions (e.g., relocation, farmland restrictions, decontamination) are based on user-specified dose or concentration limits. While protective actions help reduce dose accumulation, they also cause social and economic impacts. MACCS models the extent of displaced individuals and land contamination, and the cost of offsite property damage, economic disruptions, and various accident expenditures caused by protective actions. Finally, from the dose accumulation, MACCS estimates early and stochastic health effects according to dose-response models. The purpose of consequence analyses is to be able to understand and estimate the impact of nuclear accidents. Consequence analysis is an essential tool to inform determinations of adequate protection of the public, to understand nuclear power hazards, to measure the value of regulations, and to help us appreciate the importance of nuclear safety. As such, MACCS has a variety of regulatory uses including environmental analyses (10 CFR 51.53, 52.47), regulatory cost-benefit analyses, backfit analyses (10 CFR 50.109), consequence analysis studies such as SOARCA (NUREG-1935), Level 3 PRA studies, and risk-informing of emergency planning (10 CFR 50 App. E and 50.47). This report updates the previous MACCS theory manual (NUREG/CR-4691 Vol. 2; Chanin, Sprung, Ritchie, & Jow, 1990) and accompanies the MACCS User's Guide (SAND-2021-1588) that describes the use and input requirements of the graphical user interface of MACCS known as WinMACCS. The MACCS User's Guide is also a reference guide that describes data input file formats, describes various software components in the MACCS code suite, and provides a set of example tutorials for running WinMACCS. Also, soon to be published is a MACCS input parameter guidance report (NUREG/CR-7270) that provides technical bases for commonly used MACCS input values. This page left blank

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Application of Jacobian-free Newton–Krylov method in implicitly solving two-fluid six-equation two-phase flow problems: Implementation, validation and benchmark

This work represents a first-of-its-kind successful application to employ advanced numerical methods in solving realistic two-phase flow problems with two-fluid six-equation two-phase flow model. These advanced numerical methods include high-resolution spatial discretization scheme with staggered grids (high-order) fully implicit time integration schemes, and Jacobian-free Newton–Krylov (JFNK) method as the nonlinear solver. The computer code developed in this work has been extensively validated with existing experimental flow boiling data in vertical pipes and rod bundles, which cover wide ranges of experimental conditions, such as pressure, inlet mass flux, wall heat flux and exit void fraction. Additional code-to-code benchmark with the RELAP5-3D code further verifies the correct code implementation. The combined methods employed in this work exhibit strong robustness in solving two-phase flow problems even when phase appearance (boiling) and realistic discrete flow regimes are considered. Transitional flow regimes used in existing system analysis codes, normally introduced to overcome numerical difficulty, were completely removed in this work. As a result, this in turn provides the possibility to utilize more sophisticated flow regime maps in the future to further improve simulation accuracy.

42 ENGINEERING↗

DOE OSTI2014

Numerical and experimental investigation of melting with internal heat generation within cylindrical enclosures

There have been significant efforts by the heat transfer community to investigate the melting phenomenon of materials. These efforts have included the analytical development of equations to represent melting, numerical development of computer codes to assist in modeling the phenomena, and collection of experimental data. The understanding of the melting phenomenon has application in several areas of interest, for example, the melting of a Phase Change Material (PCM) used as a thermal storage medium as well as the melting of the fuel bundle in a nuclear power plant during an accident scenario. The objective of this research is two-fold. First a numerical investigation, using computational fluid dynamics (CFD), of melting with internal heat generation for a vertical cylindrical geometry is presented. Second, to the best of authors knowledge, there are very limited number of engineering experimental results available for the case of melting with Internal Heat Generation (IHG). An experiment was performed to produce such data using resistive, or Joule, heating as the IHG mechanism. The numerical results are compared against the experimental results and showed favorable correlation. Uncertainties in the numerical and experimental analysis are discussed. Based on the numerical and experimental analysis, recommendations are made for future work.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2013

Lawrence Livermore National Laboratory National Emission Standards for Hazardous Air Pollutants (NESHAPs) 2012 Annual Report

Lawrence Livermore National Security, LLC operates facilities at Lawrence Livermore National Laboratory (LLNL) where radionuclides are handled and stored. These facilities are subject to the U.S. Environmental Protection Agency (EPA) National Emission Standards for Hazardous Air Pollutants (NESHAPs) in Code of Federal Regulations (CFR) Title 40, Part 61, Subpart H, which regulates radionuclide emissions to air from Department of Energy (DOE) facilities. Specifically, NESHAPs limits the emission of radionuclides to the ambient air to levels resulting in an annual effective dose equivalent of 10 mrem (100 µSv) to any member of the public. Using measured and calculated emissions, and building-specific and common parameters, LLNL personnel applied the EPA-approved computer code, CAP88-PC, Version 1.0, to calculate the dose to the maximally exposed individual member of the public for the Livermore site and Site 300.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2014

Lawrence Livermore National Laboratory Radionuclide Air Emission Report for 2013 (NESHAPs 2013 Annual Report)

Lawrence Livermore National Security, LLC operates facilities at Lawrence Livermore National Laboratory (LLNL) where radionuclides are handled and stored. These facilities are subject to the U.S. Environmental Protection Agency (EPA) National Emission Standards for Hazardous Air Pollutants (NESHAPs) in Code of Federal Regulations (CFR) Title 40, Part 61, Subpart H, which regulates radionuclide emissions to air from Department of Energy (DOE) facilities. Specifically, NESHAPs limits the emission of radionuclides to the ambient air to levels resulting in an annual effective dose equivalent of 10 mrem (100 μSv) to any member of the public. Using measured and calculated emissions, and building-specific and common parameters, LLNL personnel applied the EPA-approved computer code, CAP88-PC, Version 4.0.0.570, to calculate the dose to the maximally exposed individual member of the public for the Livermore Site and Site 300.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2015

LLNL NESHAPs 2014 Annual Report

Lawrence Livermore National Security, LLC operates facilities at Lawrence Livermore National Laboratory (LLNL) where radionuclides are handled and stored. These facilities are subject to the U.S. Environmental Protection Agency (EPA) National Emission Standards for Hazardous Air Pollutants (NESHAPs) in Code of Federal Regulations (CFR) Title 40, Part 61, Subpart H, which regulates radionuclide emissions to air from Department of Energy (DOE) facilities. Specifically, NESHAPs limits the emission of radionuclides to the ambient air to levels resulting in an annual effective dose equivalent of 10 mrem (100 μSv) to any member of the public. Using measured and calculated emissions, and building-specific and common parameters, LLNL personnel applied the EPA-approved computer code, CAP88-PC, Version 4.0.1.17, to calculate the dose to the maximally exposed individual member of the public for the Livermore Site and Site 300.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

LLNL NESHAPs 2015 Annual Report - June 2016

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

NESHAPs CY19 Report: Radionuclide Air Emission Report for 2019

Lawrence Livermore National Security, LLC operates facilities at Lawrence Livermore National Laboratory (LLNL) in which radionuclides are handled and stored. These facilities are subject to the U.S. Environmental Protection Agency (EPA) National Emission Standards for Hazardous Air Pollutants (NESHAPs) in Code of Federal Regulations (CFR) Title 40, Part 61, Subpart H, which regulates radionuclide emissions to air from Department of Energy (DOE) facilities. Specifically, NESHAPs limits the emission of radionuclides to the ambient air to levels resulting in an annual effective dose equivalent of 10 mrem (100 μSv) to any member of the public. Using measured and calculated emissions, and building-specific and common parameters, LLNL personnel applied the EPA-approved computer code, CAP88-PC, Version 4.0.1.17, to calculate the dose to the maximally exposed individual member of the public for the Livermore Site and Site 300. In 2019, LLNL maintained its compliance with 40 CFR 61, Subpart H. All radioactive air emissions resulted in calculated doses far below the annual 10 mrem (100 μSv) sitewide standard. The annual doses to the site-wide maximally exposed individual member of the public at the Livermore Site and Site 300 from planned and unplanned operations in 2019 are: Livermore Site: 4.3 x 10⁻³ mrem (4.3 x 10⁻² μSv); Site 300: 9.5 x 10⁻⁸ mrem (9.5 x 10⁻⁷ μSv).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

LLNL NESHAPs 2023 Annual Report

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2016

Experimental Data Sets for CFD Calculations of Flow and Heat Transfer in Deformed Fuel Assemblies

Liquid-metal fast-breeder reactors (LMFBRs) have traditionally used fuel assemblies consisting of a hexagonal array of pins enclosed in a hexagonally shaped duct. Each fuel pin is surrounded by a wire that is wrapped in a helical configuration, which is used to separate the pins sufficiently to allow sodium coolant to flow through the assembly. The helical wire wrap also promotes mixing of the coolant. Experience with these assemblies has shown that they begin to deform with time during normal operation leading to significant changes in core thermal-hydraulics. For example, deformations of assemblies in the Experimental Breeder Reactor-2 and the Fast Flux Test Facility led to pin overheating resulting from pin-to-duct and pin-to-pin contact. This problem is well known and was investigated extensively in the 1970's and early 1980's. This was a period, however, before the development of modern Computational Fluid Dynamics (CFD) computer codes. Therefore, the experiments performed at the time did not record their results in sufficient detail to be useful for compiling validation datasets for CFD codes. Most of the experiments focused only on bulk measurements, such as the pressure drop across the bundle. When the effects of deformation of the fuel assembly were investigated, the studies performed to quantify these effects considered only bare pins without the wire wrap. With the resurgence of interest in LMFBRs in recent years, particularly in designs that leverage long-life performance of the fuel, the need has arisen for data that are suitable for the validation and benchmarking of CFD codes to predict the thermal-hydraulic behavior of flow in wire-wrapped fuel bundles. Such data must include high-resolution, transient, full-field measurements of the fluid velocity and temperature and an accurate record of the test conditions, including a precise description of the geometry. This paper describes a project to produce a validation-quality set of data that can be used to benchmark CFD codes for modeling flow and thermal performance in wire-wrapped fuel assemblies of the kind used by LMFBR designs. In addition, this effort investigates the changes to the flow and other phenomena that occur in the deformed geometries that can form later in a fuel assembly's life. This work is a joint effort of four organizations. The experimental work is being performed by AREVA and Texas A and M University. The computational work, both to support the experiments and to perform blind benchmark calculations, is being done by TerraPower, LLC, and Argonne National Laboratory (ANL). The experiments use a geometry that consists of a complete assembly of 61 wire-wrapped pins, which is surrounded by a hexagonal duct. This configuration is considered the minimum size necessary to adequately capture the important results of interest, such as the thermal gradients across the assembly and the velocity field effects in the assembly resulting from the surrounding duct. These types of effects cannot be captured by a geometry consisting of only a small group of fuel pins. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

A method for smoothing multiple yield functions

A method is presented that produces a single C 2 differentiable and convex yield function from a plastic model that contains multiple yield surfaces that are individually C 2 differentiable and convex. The method is suitable for implementation in numerical computer codes. Here, the error incurred through the smoothing procedure is quantified. The method is illustrated through a number of examples. Two simple finite-element models are solved to compare this method with existing approaches.

42 ENGINEERING↗