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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Radionuclide Inventory and Decay Heat Quantification Methodology for Severe Accident Simulations.

The MELCOR and MACCS computer codes require inputs for radionuclide inventory and decay heat in order to simulate severe accident phenomenology, source term, and consequences. Therefore, Sandia National Laboratories ( SNL ) has developed accurate and automated methods using SCALE 6 in conjunction with automation scripts to directly generate MELCOR and MACCS input records for radionuclide inventories and decay heating. Consistent information between the two codes is essential for realistic modeling of severe nuclear accidents it is the radionuclide inventory and decay power that are the principal safety problems of any severe accident that assumes successful reactor shutdown. SNL plans to conduct best-estimate calculations and uncertainty analyses using MELCOR and MACCS for several reactors, including units 1 - 3 of Fukushima Daiichi, which entail accurate and technically-scrutable information for the input models. The SNL methodology for generating radionuclide and decay power inputs directly from SCALE6 calculations is mostly presented in the context of supporting Fukushima modeling efforts. Furthermore, certain historical assumptions used by MELCOR and MACCS in the abstraction of radionuclide classes and their properties are reviewed, and these assumptions are compared to modern code predictions by SCALE6.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Poster — Thur Eve — 52: A Web-based Platform for Collaborative Document Management in Radiotherapy

We describe DepDocs, a web-based platform that we have developed to manage the committee meetings, policies, procedures and other documents within our otherwise paperless radiotherapy clinic. DepDocs is essentially a document management system based on the popular Drupal content management software. For security and confidentiality, it is hosted on a linux server internal to our hospital network such that documents are never sent to the cloud or outside of the hospital firewall. We used Drupal's in-built role-based user rights management system to assign a role, and associated document editing rights, to each user. Documents are accessed for viewing using either a simple Google-like search or by generating a list of related documents from a taxonomy of categorization terms. Our system provides document revision tracking and an document review and approval mechanism for all official policies and procedures. Committee meeting schedules, agendas and minutes are maintained by committee chairs and are restricted to committee members. DepDocs has been operational within our department for over six months and has already 45 unique users and an archive of over 1000 documents, mostly policies and procedures. Documents are easily retrievable from the system using any web browser within our hospital's network.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A Critical Review of Heat Pipe Experiments in Nuclear Energy Applications

This critical review provides heat pipe (HP) experimental data sets that contain pertinent information regarding nuclear technology that may be beneficial to researchers. Heat pipes have been shown to have a tremendously positive impact on nuclear technologies and will continue to become a more prevalent technology as more nuclear reactor concepts embrace this robust technology. Most previous reviews may focus on only a specific HP design or application, and some are backdated. Herein, this critical review extends previous efforts; integrates and summarizes previously reported HP experimental efforts; and provides updates with recently reported results in the literature for HPs in all nuclear-related applications, including space power (thermal radiators, core cooling, and electricity production), microreactors (emergency core cooling, hybrid control rods, and reactor core cooling), and HP involvement in other nuclear-related technologies (spent fuel pool cooling). The two main objectives of this critical review are (1) to facilitate the development of HP codes by outlining some of the existing experimental data sets to validate their codes and directing developers to these efforts and (2) to provide comprehensive information regarding the vast applicability of HPs used in the nuclear industry, including the theory of operation and limitations to supplement researchers in the development of new ideas for potential applications in nuclear-related technologies. The review clearly shows extensive and diverse experimental data sets for HPs developed under diverse testing conditions depending on the available nuclear application for validation purposes. Thus, this critical review is oriented to providing attention to the existing efforts rather than determining gaps in HP research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

RELAP5-3D Developmental Assessment: Comparison of Versions 4.3.4i and 4.2.1i

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.3.4i and 4.2.1i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Transport model comparison studies of intermediate-energy heavy-ion collisions

Transport models are the main method to obtain physics information on the nuclear equation of state and in-medium properties of particles from low to relativistic-energy heavy-ion collisions. The Transport Model Evaluation Project (TMEP) has been pursued to test the robustness of transport model predictions in reaching consistent conclusions from the same type of physical model. To this end, calculations under controlled conditions of physical input and set-up were performed with various participating codes. These included both calculations of nuclear matter in a box with periodic boundary conditions, which test separately selected ingredients of a transport code, and more realistic calculations of heavy-ion collisions. Over the years, six studies have been performed within this project. In this intermediate review, we summarize and discuss the present status of the project. We also provide condensed descriptions of the 26 participating codes, which contributed to some part of the project. These include the major codes in use today. After a compact description of the underlying transport approaches, we review the main results of the studies completed so far. They show, that in box calculations the differences between the codes can be well understood and a convergence of the results can be reached. These studies also highlight the systematic differences between the two families of transport codes, known under the names of Boltzmann–Uehling–Uhlenbeck (BUU) and Quantum Molecular Dynamics (QMD) type codes. However, when the codes were compared in full heavy-ion collisions using different physical models, as recently for pion production, they still yielded substantially different results. This calls for further comparisons of heavy-ion collisions with controlled models and of box comparisons of important ingredients, like momentum-dependent fields, which are currently underway. Furthermore, our evaluation studies often indicate improved strategies in performing transport simulations and thus can provide guidance to code developers. Results of transport simulations of heavy-ion collisions from a given code will have more significance if the code can be validated against benchmark calculations such as the ones summarized in this review.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Safety of Hydrogen Systems Installed in Outdoor Enclosures

The Hydrogen Safety Panel brings a broad cross-section of expertise from the industrial, government, and academic sectors to help advise the U.S. Department of Energy’s (DOE) Fuel Cell Technologies Office through its work in hydrogen safety, codes, and standards. The Panel’s initiatives in reviewing safety plans, conducting safety evaluations, identifying safety-related technical data gaps, and supporting safety knowledge tools and databases cover the gamut from research and development to demonstration and deployment. The Panel’s recent work has focused on the safe deployment of hydrogen and fuel cell systems in support of DOE efforts to accelerate fuel cell commercialization in early market applications: vehicle refueling, material handling equipment, backup power for warehouses and telecommunication sites, and portable power devices. This paper resulted from observations and considerations stemming from the Panel’s work on early market applications. This paper focuses on hydrogen system components that are installed in outdoor enclosures. These enclosures might alternatively be called “cabinets,” but for simplicity, they are all referred to as “enclosures” in this paper. These enclosures can provide a space where a flammable mixture of hydrogen and air might accumulate, creating the potential for a fire or explosion should an ignition occur. If the enclosure is large enough for a person to enter, and ventilation is inadequate, the hydrogen concentration could be high enough to asphyxiate a person who entered the space. Manufacturers, users, and government authorities rely on requirements described in codes to guide safe design and installation of such systems. Except for small enclosures used for hydrogen gas cylinders (gas cabinets), fuel cell power systems, and the enclosures that most people would describe as buildings, there are no hydrogen safety requirements for these enclosures, leaving gaps that must be addressed. This paper proposes that a technical basis be developed to enable code bodies to write requirements for the range of enclosures from the smallest to the largest.

08 HYDROGEN↗
DOE OSTI2017

Review of CTF’s Fuel Rod Modeling Using FRAPCON-4.0’s Centerline Temperature Predictions

Coolant Boiling in Rod Arrays–Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF’s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF’s fuel rod modeling to address shortcomings in CTF’s temperaturepredictions. CTF is compared to FRAPCON which is U.S. NRC’s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model.In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF’s dynamic gap conductance model.First case is chosen to show temperature is predicted correctly with CTF’s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF’s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF’s dynamic gap conductance model. Improving the CTF’s dynamic gap conductance model will allowprediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

Toptan, Aysenur↗
DOE OSTI2022

SolarAPP+ Pilot Analysis: Performance and Impact of Instant, Online Solar Permitting

The National Renewable Energy Laboratory (NREL) led a collaborative effort to develop the Solar Automated Permit Processing Plus (SolarAPP+), a no-cost solar permitting software solution to address residential solar photovoltaic (PV) permitting resource constraints and streamline solar adoption processes among authorities having jurisdiction (AHJs). The SolarAPP+ is an online portal that automates permit plan review, enabling an instant permit approval process for code-compliant residential PV systems. Based on national model building, electrical, and fire codes, the SolarAPP+ automatically performs a compliance check of permit inputs against code requirements and produces an inspection checklist that can be used to verify installation practices, workmanship, and adherence to the approved design. NREL conducted a two-phase pilot with five participating AHJs and 16 solar contractors, spanning from November 2020 through December 2021, evaluating the ability of SolarAPP+ to deliver instant permits and its impact on four critical areas within each AHJ: permit review timelines; solar adoption timelines; AHJ time saved; and inspection results. The pilot confirmed that SolarAPP+ reduced permit review times to less than one day, reduced solar adoption timelines by an average of 12 days, saved an estimated 2,067 staff hours, and had comparable inspection results to those found in traditional permitting. The pilot further identified key lessons for improving SolarAPP+, including ways to streamline SolarAPP+ adoption, development of training resources, and expansion of supported products.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2020

Simplified Performance Rating Method - Review of Existing Tools, Rulesets, and Programs. Literature and technical review

The Performance Rating Method (PRM) in ASHRAE Standard 90.1 (ASHRAE 2016) is a simulation ruleset for establishing minimum code compliance and for rating a building’s beyond-code applications. Building energy modeling (BEM) to comply with the PRM is often expensive and time consuming due to the complexity of code requirements. The goal of this project is to develop and codify a simplified PRM (S-PRM) approach for commercial buildings that will expand the use of BEM for small or simple buildings by defining a low-cost, simplified approach for creating robust and detailed models. This report summarizes a review of (i) the more commonly used existing codes, (ii) incentive programs, and (iii) the existing simplified energy modeling tools to identify which requirements are good candidates for simplification.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

Application of a computational glass model to the shock response of soda-lime glass

This article details the implementation and application of the glass-specific computational constitutive model by Holmquist and Johnson [1] to simulate the dynamic response of soda-lime glass under high rate and high pressure shock conditions. The predictive capabilities of this model are assessed through comparison of experimental data with numerical results from computations using the CTH shock physics code. The formulation of this glass model is reviewed in the context of its implementation within CTH. Using a variety of experimental data compiled from the open literature, a complete parameterization of the model describing the observed behavior of soda-lime glass is developed. Simulation results using the calibrated soda-lime glass model are compared to flyer plate and Taylor rod impact experimental data covering a range of impact and failure conditions spanning an order of magnitude in velocity and pressure. In conclusion, the complex behavior observed in the experimental testing is captured well in the computations, demonstrating the capability of the glass model within CTH.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Natural Circulation Patterns in the VHTR Air-Ingress Accident and Related Issues

Natural circulation patterns in the VHTR during a hypothetical air-ingress accident have been investigated using computational fluid dynamic (CFD) methods in order to compare results from the previous 1-D model which was developed using GAMMA code for the air-ingress analyses. The GT-MHR 600 MWt reactor was selected to be the reference design and modeled by a half symmetric 3-D geometry using FLUENT 6.3, a commercial CFD code. CFD simulations were carried out as the steady-state calculation, and the boundary conditions were either assumed or provided from the 1-D GAMMA code results. Totally, 12 different cases have been reviewed, and many notable results have been obtained through in this work. According to the simulations, natural circulation patterns in the reactor were quite different from the previous 1-D assumptions. A large re-circulation flow with thermal stratification phenomena was clearly observed in the hot-leg and the lower plenum in the 3-D model. This re-circulation flow provided about an order faster air-ingress speed (0.46 m/s in superficial velocity) than previously predicted by 1-D modeling (0.02~0.03 m/s). It indicates that the 1-D air-ingress modeling may significantly distort the air-ingress scenario and consequences. In addition, complicated natural circulation patterns are eventually expected to result in very complex graphite oxidations and corrosion behaviors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

CMS Full Simulation for Run 3

We report the status of the CMS full simulation for Run 3. During the long shutdown of the LHC a significant update has been introduced to the CMS code for simulation. The CMS geometry description is reviewed. Several important modifications were needed. CMS detector description software is migrated to the DD4Hep community developed tool. We will report on our experience obtained during the process of this migration. Geant4 10.7 is the CMS choice for Run 3 simulation productions. We will discuss arguments for this choice, the strategy of adaptation of a new Geant4 version, and will report on the physics performance of the CMS simulation. A special Geant4 Physics List configuration FTFP_BERT_EMM will be described, which provides a compromise between simulation accuracy and CPU performance. A significant fraction of time for simulation of CMS events is spent on tracking of charged particles in a magnetic field. In the CMS simulation a dynamic choice of Geant4 parameters for tracking in field is implemented. A new method is introduced into simulation of electromagnetic components of hadronic showers in the electromagnetic calorimeter of CMS. For low-energy electrons and positrons a parametrization of GFlash type is applied. Results of tests of this method will be discussed. In summary, we expect about 25% speedup of the CMS simulation production for Run 3 compared to the Run 2 simulations.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2022

Big Hole Drilling Support for Nuclear Testing, 1985-1992: An Architectural Survey of the Area 1 Subdock, Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) planned to demolish two buildings at the Area 1 Subdock at the Nevada National Security Site (NNSS) in Nye County, Nevada, to meet environmental management mission requirements. A review was conducted under Title 54 United States Code (USC) § 306101 (commonly known as Section 106 of the National Historic Preservation Act) and its implementing regulations, 36 Code of Federal Regulations (CFR) Part 800. As a result, a Memorandum of Agreement (MOA) was developed to mitigate the effects of the building demolitions. Stipulation III.B of the MOA requires an architectural survey of the Area 1 Subdock. Prior to this survey, the Subdock had not been systematically recorded. Therefore, an area of approximately 33 hectares (81 acres) was surveyed for historic properties by Desert Research Institute personnel. This effort resulted in the identification, recording, and evaluation of the potential Area 1 Subdock Historic District (SHPO Resource No. D377), including the identification of its contributing components. This district is recommended as eligible for the National Register of Historic Places (NRHP) under Criteria A and C. It contains 14 primary resources which include individual buildings, structures, storage yards, and infrastructure. Of these resources, all except one are recommended as elements that contribute to the district during its period of significance corresponding to nuclear testing from 1985 through 1992. Four of the resources (B18847, B18848, S2772, S2773) are recommended as individually eligible for the NRHP.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Single-Dose Immunogenic DNA Vaccines Coding for Live-Attenuated Alpha- and Flaviviruses

Single-dose, immunogenic DNA (iDNA) vaccines coding for whole live-attenuated viruses are reviewed. This platform, sometimes called immunization DNA, has been used for vaccine development for flavi- and alphaviruses. An iDNA vaccine uses plasmid DNA to launch live-attenuated virus vaccines in vitro or in vivo. When iDNA is injected into mammalian cells in vitro or in vivo, the RNA genome of an attenuated virus is transcribed, which starts replication of a defined, live-attenuated vaccine virus in cell culture or the cells of a vaccine recipient. In the latter case, an immune response to the live virus vaccine is elicited, which protects against the pathogenic virus. Unlike other nucleic acid vaccines, such as mRNA and standard DNA vaccines, iDNA vaccines elicit protection with a single dose, thus providing major improvement to epidemic preparedness. Still, iDNA vaccines retain the advantages of other nucleic acid vaccines. In summary, the iDNA platform combines the advantages of reverse genetics and DNA immunization with the high immunogenicity of live-attenuated vaccines, resulting in enhanced safety and immunogenicity. This vaccine platform has expanded the field of genetic DNA and RNA vaccines with a novel type of immunogenic DNA vaccines that encode entire live-attenuated viruses.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2019

Evalulation of the Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC N690-18, for Application to Nuclear Power Plants

This report describes the assessment of the recently published ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities. This work was performed for the U.S. Nuclear Regulatory Commission for potential endorsement of N690-18 for use in the design of nuclear power plants. Currently the NRC endorses the use of an earlier version: N690-1994 (R2004), and thus, a technical review of the updated N690 Specification, which contains significant changes, is warranted. The N690-18 Specification reflects advances and improvements in steel design gained from experience, analytical studies, and experimental data over the years. The N690-18 specification also contains a new Appendix N9 that addresses the design of safety-related steel-plate composite (SC) walls which are a relatively new type of structural member for which previously there has not been any U.S. code or standard that governs its design. The technical review of N690-18 was performed in order to assess the adequacy of the provisions for use by the NRC in updating its regulatory guidance for nuclear power plants. The research developed the technical basis for acceptance of the new specification and also identified areas where additional staff guidance is needed for the design of safety-related steel structures at nuclear power plants. The results of the research have determined that the N690-18 specification is a significant enhancement from the N690-1994 (R2004) edition. The new specification contains substantial upgrades in the design of structural steel and includes new provisions for design of SC wall sections, which can benefit the nuclear power industry as well as the NRC review process. All of the improvements, if endorsed by the NRC with certain regulatory positions, will allow clearer and more consistent regulatory guidance, and will avoid the need for justification by licensees and related reviews each time certain methods or approaches are used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Evaluation of the Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC N690-18, for Application to Nuclear Power Plants

This report describes the assessment of the recently published ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities. This work was performed for the U.S. Nuclear Regulatory Commission for potential endorsement of N690-18 for use in the design of nuclear power plants. Currently the NRC endorses the use of an earlier version: N690-1994 (R2004), and thus, a technical review of the updated N690 Specification, which contains significant changes, is warranted. The N690-18 Specification reflects advances and improvements in steel design gained from experience, analytical studies, and experimental data over the years. The N690-18 specification also contains a new Appendix N9 that addresses the design of safety-related steel-plate composite (SC) walls which are a relatively new type of structural member for which previously there has not been any U.S. code or standard that governs its design. The technical review of N690-18 was performed in order to assess the adequacy of the provisions for use by the NRC in updating its regulatory guidance for nuclear power plants. The research developed the technical basis for acceptance of the new specification and also identified areas where additional staff guidance is needed for the design of safety-related steel structures at nuclear power plants. The results of the research have determined that the N690-18 specification is a significant enhancement from the N690-1994 (R2004) edition. The new specification contains substantial upgrades in the design of structural steel and includes new provisions for design of SC wall sections, which can benefit the nuclear power industry as well as the NRC review process. All of the improvements, if endorsed by the NRC with certain regulatory positions, will allow clearer and more consistent regulatory guidance, and will avoid the need for justification by licensees and related reviews each time certain methods or approaches are used.

42 ENGINEERING↗