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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 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Statistical Analyses of Post Irradiation Experiments of the BWR Forsmark

Computer codes which are to be used in nuclear safety research need to be validated prior to their application. For depletion codes, this validation can be done using data from post irradiation experiments (PIE). These experiments consist of a radiochemical analysis of small parts of a fuel rod (FR), which has been irradiated in a nuclear power plant as part of a fuel assembly (FA). This analysis allows the deduction of the isotopic composition of nuclides important for the desired application case. In addition the operation history at the sample location needs to be known and reported with the analysis results. In our work, we perform uncertainty and sensitivity analyses of a model based on a sample from the boiling water reactor (BWR) Forsmark-3 in order to deduce the influence of uncertain system parameters on the nuclide composition of the depleted fuel. In a second step we compare the resulting uncertainties of the nuclide densities with the deviation between their calculated and experimental values. The goal is to obtain quantitative uncertainty data which can be accounted for in a validation procedure. We use the Monte Carlo sampling approach, i.e. repeating the model calculation using statistically varied sets of the uncertain input parameters and analyzing the distributions of the result values statistically. BWR depletion calculations are especially demanding due to the increased complexity and heterogeneity compared to pressurized water reactors (PWR): extensively variable enrichments, complex FA geometries, partial length FRs, strong axially varying moderator density and the common use of gadolinium as burnable absorber in some of the FRs. However, the precise and correct calculation of BWR-depletion is extremely important in view of a possible application in burnup credit for BWR FAs. Therein credit is taken for the reduced reactivity of spent fuel due to the depletion of {sup 235}U for transport and intermediate or final storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Implementation of a generalized coherent elastic scattering formulation for thermal neutron scattering analysis - 14487

The scattering interactions of thermal neutrons in matter are classified into inelastic and elastic components. In crystalline materials, coherent elastic scattering represents a dominant form of interaction. Traditionally, the treatment of this interaction in standard computer codes (e.g., the LEAPR module of the NJOY code system) was performed within the cubic approximation and restricted the analyst to a few materials that were hardwired into the code. In general, these restrictions can hinder the production of ENDF style thermal neutron scattering libraries for new materials outside the current portfolio of treated materials. In this work, a general formulation based on the exact theory of coherent elastic scattering was developed. This formulation is capable of processing the crystalline structure of any material. The cubic approximation is relaxed by accounting for the partial forces from asymmetric crystal structures in the calculation. The atom sites approximation is relaxed by utilizing a site-wise Debye-Waller factor. Specifically, the Debye-Waller factor is no longer regarded as a uniform constant across the crystal but treated as a matrix tensor that uses the integral of the partial phonon density of states. Therefore, the calculation of the coherent elastic scattering cross section is executed by applying the matrix-form of the Debye-Waller factor on each reciprocal space point and summing them up. Furthermore, the developed algorithm supports compounds with any number of elements and of any ratio to each other. This general formulation was successfully implemented and contributed to the production of ENDF style thermal neutron scattering libraries for silicon dioxide (α-SiO{sub 2}) and silicon carbide (3C-SiC). The needed input for the analysis was generated using density functional theory (DFT) atomistic simulations and dynamical matrix calculations for both materials. (authors)

42 ENGINEERING↗
DOE OSTI2022

Activation of lead coolant in Westinghouse lead fast reactor

The Westinghouse Lead-cooled Fast Reactor (LFR) is a medium-sized, passively safe, scalable reactor being developed by Westinghouse in collaboration with domestic and international organizations with the primary goals of reducing front-end capital cost and generating flexible and cost-competitive electricity while satisfying the highest standards in terms of safety and sustainability. An important aspect of new plant design is an assessment of activation level of its components so that an appropriate nuclear waste management plan can be established. As lead coolant is a distinguishing design feature of the LFR, its activation due to irradiation and the implications on establishing an appropriate waste disposal approach are important for engagement of project stakeholders, customers and nuclear safety regulators. This paper provides a brief description of the Westinghouse LFR design, a background on lead activation phenomenology, and then describes the methodology used for lead activation analysis performed with the computer code Serpent 2. It was determined that impurities naturally occurring in lead have the biggest contribution to lead activation. Corrosion and erosion products from the components immersed in lead can also be important contributors to lead activity and their impact can be limited by appropriate selection of materials and operating conditions. Polonium isotopes, resulting from bismuth generated from lead irradiation and its impurities, bring a notable contribution to the inhalation and ingestion doses during the plant operation and in the first years after the plant shutdown. However, it is not by far as important as for lead-bismuth coolant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Status of the CERBERUS Evaluation for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook

Modeling & Simulation (M&S) tools are used to analyze advanced reactor designs and the safety of current nuclear operations. As computers continue to improve, we are able to enhance resolution in our calculations. Therefore, the limitations of simulation capability are in the quality of data that is being used, including our ability to quantify the uncertainty and sensitivity of that data. In order to model systems of interest with increasing accuracy, the industry must improve key nuclear data measurements. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) compiles and evaluates experiment data in a handbook that can be used by criticality safety engineers and others to validate computer codes and cross section libraries at nuclear facilities. Both critical and subcritical experiments are included in the handbook. These experiments, along with differential measurements, can help improve the quality of nuclear data. Concerns regarding the accuracy of Cu nuclear data have been published. The large values and trend of C-E for the Zeus intermediate energy benchmark, being one of the primary examples. Furthermore, very few experiments have been designed to be sensitive to Cu (as shown in Figure 1), so an integral, critical experiment is needed to help resolve these differences.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 4, October-December 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 477 Report on the International Symposium on the Use of Probabilistic Safety Assessment for Operational Safety—PSA '91, S. Chakraborty and M. Khatib-Rahbar; 488 Good Relationships Are Pivotal in Nuclear Data Bases, A. S. Heger and B. V. Koen; 494 Technical Note: The Interagency Nuclear Safety Review Panel's Evaluation of the Ulysses Space Mission, J. A. Sholtis, Jr., D. A. Huff, L. B. Gray, N. P. Klug, and R. O. Winchester; ACCIDENT ANALYSIS: 502 The Severe Accident Analysis Program for the Savannah River Nuclear Production Reactors, M. L. Hyder; CONTROL AND INSTRUMENTATION: 511 A Framework for Selecting Suitable Control Technologies for Nuclear Power Plant Systems, R. A. Kisner; DESIGN FEATURES: 521 Containments for Gas-Cooled Power Reactors History and Status, P.W. Williams; ENVIRONMENTAL EFFECTS: 537 Indoor Radon: A Natural Risk, N. H. Harley and J.H. Harley; On the Importance of the Atmospheric Parameters in the Fission Products Distribution of a Severe Reactor Accident, M. C. Barla and A. R. Bayulken; Book Review: Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, C. R. Richmond; WASTE AND SPENT FUEL MANAGEMENT: 555 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver OPERATING EXPERIENCES: 567 Effects of Component Aging on the Westhinghouse Control Rod Drive System, K. Sullivan and W. Gunther; 577 Reactor Shutdown Experience, Compiled by J. W. Cletcher, 580 Selected Safety-Related Events, Compiled by G. A. Murphy; 582 Operating U 8 Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver, RECENT DEVELOPMENTS: 596 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 610 Reports, Standards, and Safety Guides, D. S. Queener; 614 Proposed Rule Changes as of June 30, 1991; ANNOUNCEMENTS: 520 Workshop on PC PRAISE: A Probabilistic Fracture Mechanics Personal Computer Code for Nuclear Power Plant Piping Reliability Assessment; 536 Fifth Workshop on Nuclear Power Plant Containment Integrity; 624 New OECD "International Information System on Occupational Exposure (ISOE)"; 625 Harvard School of Public Health Announces Short Courses; 625 Eighth Power Plant Dynamics, Control and Testing Symposium; 626 Second Training Course on Off-Site Emergency Planning and Response for Nuclear Accidents; 618 The Authors; 622 Reviewers of Nuclear Safety, Vol 32.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

THE DESIGN OF RADIAL HONEYCOMB LATTICES FOR IMPACT ENERGY ABSORPTION IN RADIOACTIVE MATERIALS PACKAGES

In this research we present a variation on the corrugation technique of honeycomb lattices, for cylindrical honeycombs, making them much easier to design for impact energy absorption in radioactive materials packages. This variation, termed radial honeycomb lattices, eliminates the residual strain and saddle effect. The use of honeycomb lattices provides advantages over the typically used foams. While foams are effective at absorbing impact energy, they can burn, their material properties are difficult to control, they can degrade over time, and procurement of raw materials can be dependent on timing of manufacturer batch runs. While the weaknesses of foam are strengths for honeycomb lattices, lattices have a different set of problems. Typically, when cylindrical honeycomb lattices are manufactured, they are manufactured flat, wrapped around a mandrel of the desired radius, and then brazed. This approach introduces residual strains, resulting in the saddle effect, which limits both the radial thickness and cylinder length. The radial honeycomb lattice approach presented here makes the design of thicker and longer cylinder honeycombs possible. To address these issues, we propose a honeycomb lattice which changes in cross-section from the inner to the outer radius of the cylinder. This causes the lattice to automatically wrap into a cylinder as it exits the corrugating gears. The theoretically bounding case, of a square cross-section at the inner radius, transitioning through a hexagon, to a diamond cross-section at the outer radius, results in a maximum thickness of approximately 41% of the inner radius. Full mathematical derivations, implemented in computer code, allow for the design of an entire radial honeycomb lattice, including the corrugating gears. To accomplish this only the radial thickness, inner cross-section shape, cell size, and nominal gear radius need to be specified, making the design of these lattices very efficient. Radial honeycomb lattice prototypes have demonstrated that the honeycomb does indeed wrap into a cylinder as intended, without the saddle effect, and can therefore be used to create thick-walled cylinders of any length. These design improvements make cylindrical honeycomb lattices much more accessible as a design element for radioactive materials packages

Johnson, William R. [Savannah River National Labor↗
DOE OSTI2025

TRISO-form HALEU-fueled Experiment for Transport Applications (THETA) Serpent 2 Benchmarking

A number of new advanced reactor designs use TRI-structural ISO-topic (TRISO) fuel. The uranium enrichment for the TRISO fuel is in the 5- to 19.75-weight percent range, which is considered High-Assay Low Enriched Uranium (HALEU). The Nuclear Regulatory Commission and vendors both have a strong interest in benchmarking these advanced reactor designs and having confidence in the predictive capabilities of the computer codes used to model them. In the recent years, Serpent 2 gained several capabilities focused on the reactor applications making it attractive for many reactor vendors.

Critical Experiment↗
DOE OSTI2019

As-Run ATR Physics Evaluation for the UCF-1 HSIS Experiment

This Engineering Calculations and Analysis Report (ECAR) documents the Advanced Test Reactor (ATR) detailed MCNP full core model physics analysis performed to support the irradiation of the University of Central Florida (UCF-1) Hydraulic Shuttle Irradiation System Experiment in the ATR B-7 position. The purpose of this analysis is to calculate: • The as-run neutron and gamma heating rates for the experiment components (W/g); • The as-run DPA for the experiment specimens after 20 (test train 1) and 2 (test train 2) days of irradiation in the ATR B-7 position (neutrons/cm2-sec); • The radionuclide source term after 2 days and 20 days of irradiation in the ATR B-7 position; • The decay heat associated with the UCF-1 radionuclide source term. These calculations were performed using the computer codes MCNP and ORIGEN2. The neutronics model, analysis details, and results are presented in the attached report.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Processing unit performance projection using dynamic hardware behaviors

Methods for estimating accelerator performance for dynamic hardware behaviors are disclosed. Computer program code to be executed on a first processing unit is received, and an execution of the computer code on the first processing unit is monitored to determine a plurality of performance characteristics. A plurality of dynamic hardware behaviors is determined by applying a clustering algorithm to the performance characteristics, and an equivalent accelerator portion of computer code to be executed on a second processing unit is generated by translating a set of instructions in a first portion of computer code corresponding to a first one of the plurality of dynamic hardware behaviors to an equivalent set of instructions to be executed on the second processing unit. An estimated measure of performance for executing the equivalent accelerator portion on the second processing unit is determined for the first one of the plurality of dynamic hardware behaviors.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Integration of the Nek5000 Computational Fluid Dynamics Code with the NEAMS Workbench

The mission of the US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program is to develop, apply, deploy, and support state-of-the-art predictive modeling and simulation tools for the design and analysis of current and future nuclear energy systems using computing architectures ranging from laptops to leadership-class facilities. NEAMS is organized into three product lines: the Fuels Product Line (FPL),the Reactors Product Line (RPL), and the Integration Product Line (IPL). The FPL and RPL focus on the development of high-fidelity tools such as the MOOSE-based applications BISON and MARMOT [3, 9] at Idaho National Laboratory, and the PROTEUS neutronics code and the NEK5000 computational fluid dynamics code from the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) framework [10]. The NEAMS IPL will enable users to apply high-fidelity simulations to inform lower order models for the design, analysis, and licensing of advanced nuclear systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

A tool and a methodology to use macros for abstracting variations in code for different computational demands

Scientific software used on high-performance computing platforms is in a phase of transformation because of the combined increase in the heterogeneity and complexity of models and hardware platforms. Having separate implementations for different platforms can easily lead to combinatorial explosions; therefore, the computational science community has been looking for mechanisms to express code through abstractions that can be specialized for different platforms. Most existing approaches use template meta-programming in C++, and are, therefore language specific. Here, we have developed a tool that uses customized expansion of macros to mimic some of C++ behavior in other languages. It enables unification of any code variants that may be necessary to run efficiently on different target architectures and different computational environments through use of macros with multiple alternative definitions and ability to arbitrate on definition selection for expansion. Combined with two other tools, a custom runtime, and a user specified recipe translator, our custom macroprocessor becomes a part of an overall performance portability solution that does not depend on any specific programming language. We also use macros as code-shorthand that lets code snippets become building blocks that allow variations in control flow to explore performance options. We demonstrate use of macros in Flash-X, a multiphysics multicomponent code with many Fortran legacy components derived from an earlier community code FLASH.

Heterogenous computing↗
DOE OSTI2014

The design and fabrication of two portal vein flow phantoms by different methods

Purpose: This study outlines the design and fabrication techniques for two portal vein flow phantoms. Methods: A materials study was performed as a precursor to this phantom fabrication effort and the desired material properties are restated for continuity. A three-dimensional portal vein pattern was created from the Visual Human database. The portal vein pattern was used to fabricate two flow phantoms by different methods with identical interior surface geometry using computer aided design software tools and rapid prototyping techniques. One portal flow phantom was fabricated within a solid block of clear silicone for use on a table with Ultrasound or within medical imaging systems such as MRI, CT, PET, or SPECT. The other portal flow phantom was fabricated as a thin walled tubular latex structure for use in water tanks with Ultrasound imaging. Both phantoms were evaluated for usability and durability. Results: Both phantoms were fabricated successfully and passed durability criteria for flow testing in the next project phase. Conclusions: The fabrication methods and materials employed for the study yielded durable portal vein phantoms.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2018

APPLICATION OF THE FAVOR-OCI FRACTURE MECHANICS COMPUTER PROGRAM TO ASME CODE SECTION XI, IWB-3610 FLAW ACCEPTANCE CRITERIA EVALUATIONS

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Appli-cations of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is main-tained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all prev-iously-existing probabilistic capabilities of FAVOR.The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws.The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These com-pute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected fracture toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV. Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Dickson, Terry L.↗
DOE OSTI2018

Performance Optimization of a Beam Dynamics PIC Code On Hybrid Computer Architectures

The self-consistent multi-particle tracking based on particle-in-cell method (PIC) has been widely used in particle accelerator beam dynamics study. However, the PIC simulation is time-consuming and needs to use modern parallel computers for high resolution applications. In this paper, we implemented and optimized a parallel beam dynamics PIC code on two types of hybrid parallel computer architectures: one is the GPU and GPU cluster, while the other is the "Knight Landing" CPU cluster.

Liu, Zhicong↗
DOE OSTI2021

Novel features and GPU performance analysis for EM particle transport in the Celeritas code

Celeritas is a new computational transport code designed for high-performance simulation of high-energy physics detectors. This work describes some of its current capabilities and the design choices that enable the rapid development of efficient on-device physics. The abstractions that underpin the code design facilitate low-level performance tweaks that require no changes to the higher-level physics code. We evaluate a set of independent changes that together yield an almost 40% speedup over the original GPU code for a net performance increase of 220× for a single GPU over a single CPU running 8.4M tracks on a small demonstration physics app.

Johnson, Seth R.↗
DOE OSTI2024

Towards reverse mode automatic differentiation of Kokkos-based codes

Derivative computation is a key component of optimization, sensitivity analysis, uncertainty quantification, and the solving of nonlinear problems. Automatic differentiation (AD) is a powerful technique for evaluating such derivatives, and in recent years, has been integrated into programming environments such as Jax, PyTorch, and TensorFlow to support derivative computations needed for training of machine learning models, facilitating wide-spread use of these technologies. The C++ language has become the de facto standard for scientific computing due to numerous factors, yet language complexity has made the wide-spread adoption of AD technologies for C++ difficult, hampering the incorporation of powerful differentiable programming approaches into C++ scientific simulations. This is exacerbated by the increasing emergence of architectures, such as GPUs, with limited memory capabilities and requiring massive thread-level concurrency. C++ AD tools must effectively use these environments to bring novel scientific simulations to next-generation DOE experimental and observational facilities. In this project, we investigated source transformation-based automatic differentiation using LLVM compiler infrastructure to automatically generate portable and efficient gradient computations of Kokkos-based code. We have demonstrated that our proposed strategy is feasible by investigating the usage of a prototype LLVM-based source transformation tool to generate gradients of simple functions made of sequences of simple Kokkos parallel regions. Speedups of up to 500x compared to Sacado were observed on NVIDIA V100 GPU.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Communication: Automatic code generation enables nuclear gradient computations for fully internally contracted multireference theory

Analytical nuclear gradients for fully internally contracted complete active space second-order perturbation theory (CASPT2) are reported. This implementation has been realized by an automated code generator that can handle spin-free formulas for the CASPT2 energy and its derivatives with respect to variations of molecular orbitals and reference coefficients. The underlying complete active space self-consistent field and the so-called Z-vector equations are solved using density fitting. The implementation has been applied to the vertical and adiabatic ionization potentials of the porphin molecule to illustrate its capability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Improving the Capabilities and Computational Efficiency of the RTE+RRTMGP Radiation Code (Final Report)

This report details progress on the RTE+RRTMGP radiation codes made during the period of performance. RTE+RRTMGP is a set of codes for computing radiative fluxes in planetary atmospheres. RRTMGP uses a k-distribution to provide an optical description (absorption and possibly Rayleigh optical depth) of the gaseous atmosphere, along with the relevant source functions, on a pre-determined spectral grid given temperatures, pressures, and gas concentration. RTE computes fluxes given spectrally-resolved optical descriptions and source functions. Spectrally-resolved fluxes are summarized (“reduced”) via a user extensible class. The initial release of the code and the design choices are described in Pincus et al. 2019; the codes are available on Github. Although RRTMGP was based on current (at the time) empirical spectroscopic data, RTE and RRTMGP were developed in large part to modernize software practices. The design focused on flexibility broadly interpreted: by separating code from data and allowing data to drive computation; in coupling to the host model (e.g. the coupling of clouds to radiative fluxes is user-controlled); with respect to programming languages (computational tasks are accessed via widely-compatible C interfaces); and with respect to hardware (the codes run on a range of CPU and GPU architectures). The code also puts an emphasis on modularity and clarity. RTE+RRTMGP v1.0 was released in September 20219. This award supported the evolution of the RTE+RRTMGP code base to support greater flexibility, accuracy, and efficiency.

54 ENVIRONMENTAL SCIENCES↗