78 FR 37848 - ASME Code Cases Not Approved for Use

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-24

...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment draft regulatory guide (DG), DG-1233, ``ASME Code Cases not Approved for Use.'' This regulatory guide lists the American Society of Mechanical Engineers (ASME) Code Cases that the NRC has determined not to be acceptable for use on a generic basis.

A Comparison of Fatigue Design Methods

DTIC Science & Technology

2001-04-05

Boiler and Pressure Vessel Code does not...Engineers, "ASME Boiler and Pressure Vessel Code ," ASME, 3 Park Ave., New York, NY 10016-5990. [4] Langer, B. F., "Design of Pressure Vessels Involving... and Pressure Vessel Code [3] presents these methods and has expanded the procedures to other pressure vessels besides nuclear pressure vessels. B.

Kinetic: A system code for analyzing nuclear thermal propulsion rocket engine transients

NASA Astrophysics Data System (ADS)

Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

The topics are presented in viewgraph form and include the following: outline of kinetic code; a kinetic information flow diagram; kinetic neutronic equations; turbopump/nozzle algorithm; kinetic heat transfer equations per node; and test problem diagram.

Overview of codes and tools for nuclear engineering education

NASA Astrophysics Data System (ADS)

Yakovlev, D.; Pryakhin, A.; Medvedeva, L.

2017-01-01

The recent world trends in nuclear education have been developed in the direction of social education, networking, virtual tools and codes. MEPhI as a global leader on the world education market implements new advanced technologies for the distance and online learning and for student research work. MEPhI produced special codes, tools and web resources based on the internet platform to support education in the field of nuclear technology. At the same time, MEPhI actively uses codes and tools from the third parties. Several types of the tools are considered: calculation codes, nuclear data visualization tools, virtual labs, PC-based educational simulators for nuclear power plants (NPP), CLP4NET, education web-platforms, distance courses (MOOCs and controlled and managed content systems). The university pays special attention to integrated products such as CLP4NET, which is not a learning course, but serves to automate the process of learning through distance technologies. CLP4NET organizes all tools in the same information space. Up to now, MEPhI has achieved significant results in the field of distance education and online system implementation.

Recent Progress in the Development of a Multi-Layer Green's Function Code for Ion Beam Transport

NASA Technical Reports Server (NTRS)

Tweed, John; Walker, Steven A.; Wilson, John W.; Tripathi, Ram K.

2008-01-01

To meet the challenge of future deep space programs, an accurate and efficient engineering code for analyzing the shielding requirements against high-energy galactic heavy radiation is needed. To address this need, a new Green's function code capable of simulating high charge and energy ions with either laboratory or space boundary conditions is currently under development. The computational model consists of combinations of physical perturbation expansions based on the scales of atomic interaction, multiple scattering, and nuclear reactive processes with use of the Neumann-asymptotic expansions with non-perturbative corrections. The code contains energy loss due to straggling, nuclear attenuation, nuclear fragmentation with energy dispersion and downshifts. Previous reports show that the new code accurately models the transport of ion beams through a single slab of material. Current research efforts are focused on enabling the code to handle multiple layers of material and the present paper reports on progress made towards that end.

Potential advantages associated with implementing a risk-based inspection program by a nuclear facility

NASA Astrophysics Data System (ADS)

McNeill, Alexander, III; Balkey, Kenneth R.

1995-05-01

The current inservice inspection activities at a U.S. nuclear facility are based upon the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. The Code selects examination locations based upon a sampling criteria which includes component geometry, stress, and usage among other criteria. This can result in a significant number of required examinations. As a result of regulatory action each nuclear facility has conducted probabilistic risk assessments (PRA) or individual plant examinations (IPE), producing plant specific risk-based information. Several initiatives have been introduced to apply this new plant risk information. Among these initiatives is risk-based inservice inspection. A code case has been introduced for piping inspections based upon this new risk- based technology. This effort brought forward to the ASME Section XI Code committee, has been initiated and championed by the ASME Research Task Force on Risk-Based Inspection Guidelines -- LWR Nuclear Power Plant Application. Preliminary assessments associated with the code case have revealed that potential advantages exist in a risk-based inservice inspection program with regard to a number of exams, risk, personnel exposure, and cost.

78 FR 37885 - Approval of American Society of Mechanical Engineers' Code Cases

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-24

...), standard design certifications, standard design approvals and manufacturing licenses, to use the Code Cases... by the ASME. The three RGs that would be incorporated by reference are RG 1.84, ``Design, Fabrication... nuclear power plant licensees, and applicants for CPs, OLs, COLs, standard design certifications, standard...

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

NASA Radiation Protection Research for Exploration Missions

NASA Technical Reports Server (NTRS)

Wilson, John W.; Cucinotta, Francis A.; Tripathi, Ram K.; Heinbockel, John H.; Tweed, John; Mertens, Christopher J.; Walker, Steve A.; Blattnig, Steven R.; Zeitlin, Cary J.

2006-01-01

The HZETRN code was used in recent trade studies for renewed lunar exploration and currently used in engineering development of the next generation of space vehicles, habitats, and EVA equipment. A new version of the HZETRN code capable of simulating high charge and energy (HZE) ions, light-ions and neutrons with either laboratory or space boundary conditions with enhanced neutron and light-ion propagation is under development. Atomic and nuclear model requirements to support that development will be discussed. Such engineering design codes require establishing validation processes using laboratory ion beams and space flight measurements in realistic geometries. We discuss limitations of code validation due to the currently available data and recommend priorities for new data sets.

Knowledge management: Role of the the Radiation Safety Information Computational Center (RSICC)

NASA Astrophysics Data System (ADS)

Valentine, Timothy

2017-09-01

The Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL) is an information analysis center that collects, archives, evaluates, synthesizes and distributes information, data and codes that are used in various nuclear technology applications. RSICC retains more than 2,000 software packages that have been provided by code developers from various federal and international agencies. RSICC's customers (scientists, engineers, and students from around the world) obtain access to such computing codes (source and/or executable versions) and processed nuclear data files to promote on-going research, to ensure nuclear and radiological safety, and to advance nuclear technology. The role of such information analysis centers is critical for supporting and sustaining nuclear education and training programs both domestically and internationally, as the majority of RSICC's customers are students attending U.S. universities. Additionally, RSICC operates a secure CLOUD computing system to provide access to sensitive export-controlled modeling and simulation (M&S) tools that support both domestic and international activities. This presentation will provide a general review of RSICC's activities, services, and systems that support knowledge management and education and training in the nuclear field.

An overview of the ENEA activities in the field of coupled codes NPP simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parisi, C.; Negrenti, E.; Sepielli, M.

2012-07-01

In the framework of the nuclear research activities in the fields of safety, training and education, ENEA (the Italian National Agency for New Technologies, Energy and the Sustainable Development) is in charge of defining and pursuing all the necessary steps for the development of a NPP engineering simulator at the 'Casaccia' Research Center near Rome. A summary of the activities in the field of the nuclear power plants simulation by coupled codes is here presented with the long term strategy for the engineering simulator development. Specifically, results from the participation in international benchmarking activities like the OECD/NEA 'Kalinin-3' benchmark andmore » the 'AER-DYN-002' benchmark, together with simulations of relevant events like the Fukushima accident, are here reported. The ultimate goal of such activities performed using state-of-the-art technology is the re-establishment of top level competencies in the NPP simulation field in order to facilitate the development of Enhanced Engineering Simulators and to upgrade competencies for supporting national energy strategy decisions, the nuclear national safety authority, and the R and D activities on NPP designs. (authors)« less

Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. D. Bess; J. B. Briggs; A. S. Garcia

2011-09-01

One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along withmore » summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.« less

Nonperturbative methods in HZE ion transport

NASA Technical Reports Server (NTRS)

Wilson, John W.; Badavi, Francis F.; Costen, Robert C.; Shinn, Judy L.

1993-01-01

A nonperturbative analytic solution of the high charge and energy (HZE) Green's function is used to implement a computer code for laboratory ion beam transport. The code is established to operate on the Langley Research Center nuclear fragmentation model used in engineering applications. Computational procedures are established to generate linear energy transfer (LET) distributions for a specified ion beam and target for comparison with experimental measurements. The code is highly efficient and compares well with the perturbation approximations.

Asymptotic Expansion Homogenization for Multiscale Nuclear Fuel Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, J. D.; Tonks, M. R.; Chockalingam, K.

2015-03-01

Engineering scale nuclear fuel performance simulations can benefit by utilizing high-fidelity models running at a lower length scale. Lower length-scale models provide a detailed view of the material behavior that is used to determine the average material response at the macroscale. These lower length-scale calculations may provide insight into material behavior where experimental data is sparse or nonexistent. This multiscale approach is especially useful in the nuclear field, since irradiation experiments are difficult and expensive to conduct. The lower length-scale models complement the experiments by influencing the types of experiments required and by reducing the total number of experiments needed.more » This multiscale modeling approach is a central motivation in the development of the BISON-MARMOT fuel performance codes at Idaho National Laboratory. These codes seek to provide more accurate and predictive solutions for nuclear fuel behavior. One critical aspect of multiscale modeling is the ability to extract the relevant information from the lower length-scale sim- ulations. One approach, the asymptotic expansion homogenization (AEH) technique, has proven to be an effective method for determining homogenized material parameters. The AEH technique prescribes a system of equations to solve at the microscale that are used to compute homogenized material constants for use at the engineering scale. In this work, we employ AEH to explore the effect of evolving microstructural thermal conductivity and elastic constants on nuclear fuel performance. We show that the AEH approach fits cleanly into the BISON and MARMOT codes and provides a natural, multidimensional homogenization capability.« less

Approximate Green's function methods for HZE transport in multilayered materials

NASA Technical Reports Server (NTRS)

Wilson, John W.; Badavi, Francis F.; Shinn, Judy L.; Costen, Robert C.

1993-01-01

A nonperturbative analytic solution of the high charge and energy (HZE) Green's function is used to implement a computer code for laboratory ion beam transport in multilayered materials. The code is established to operate on the Langley nuclear fragmentation model used in engineering applications. Computational procedures are established to generate linear energy transfer (LET) distributions for a specified ion beam and target for comparison with experimental measurements. The code was found to be highly efficient and compared well with the perturbation approximation.

Effects of Debris Entrainment and Multi-Phase Flow on Plug Loading in an MX Trench.

DTIC Science & Technology

1978-09-15

gas stream of density (pg) and velocity (Vg) is: -., * -) - * 2~ TD FD Pg (V P V) Vp-Vg I CD( TD ) (A.1) 4 where the drag coefficient (CD) is defined by...ATTN: FCPR ATTN: Code L53 , J. Forrest Field Command Naval Facilities Engineering Command Defense Nuclear Agency ATTN: Code 09M22C Livermore Division

Rocket engine system reliability analyses using probabilistic and fuzzy logic techniques

NASA Technical Reports Server (NTRS)

Hardy, Terry L.; Rapp, Douglas C.

1994-01-01

The reliability of rocket engine systems was analyzed by using probabilistic and fuzzy logic techniques. Fault trees were developed for integrated modular engine (IME) and discrete engine systems, and then were used with the two techniques to quantify reliability. The IRRAS (Integrated Reliability and Risk Analysis System) computer code, developed for the U.S. Nuclear Regulatory Commission, was used for the probabilistic analyses, and FUZZYFTA (Fuzzy Fault Tree Analysis), a code developed at NASA Lewis Research Center, was used for the fuzzy logic analyses. Although both techniques provided estimates of the reliability of the IME and discrete systems, probabilistic techniques emphasized uncertainty resulting from randomness in the system whereas fuzzy logic techniques emphasized uncertainty resulting from vagueness in the system. Because uncertainty can have both random and vague components, both techniques were found to be useful tools in the analysis of rocket engine system reliability.

Uncrackable code for nuclear weapons

ScienceCinema

Hart, Mark

2018-05-11

Mark Hart, a scientist and engineer in Lawrence Livermore National Laboratory's (LLNL) Defense Technologies Division, has developed a new approach for ensuring nuclear weapons and their components can't fall prey to unauthorized use. The beauty of his approach: Let the weapon protect itself. "Using the random process of nuclear radioactive decay is the gold standard of random number generators," said Mark Hart. "You’d have a better chance of winning both Mega Millions and Powerball on the same day than getting control of IUC-protected components."

Uncrackable code for nuclear weapons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, Mark

Mark Hart, a scientist and engineer in Lawrence Livermore National Laboratory's (LLNL) Defense Technologies Division, has developed a new approach for ensuring nuclear weapons and their components can't fall prey to unauthorized use. The beauty of his approach: Let the weapon protect itself. "Using the random process of nuclear radioactive decay is the gold standard of random number generators," said Mark Hart. "You’d have a better chance of winning both Mega Millions and Powerball on the same day than getting control of IUC-protected components."

Evaluation of the finite element fuel rod analysis code (FRANCO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

A model for neutrino emission from nuclear accretion disks

NASA Astrophysics Data System (ADS)

Deaton, Michael

2015-04-01

Compact object mergers involving at least one neutron star can produce short-lived black hole accretion engines. Over tens to hundreds of milliseconds such an engine consumes a disk of hot, nuclear-density fluid, and drives changes to its surrounding environment through luminous emission of neutrinos. The neutrino emission may drive an ultrarelativistic jet, may peel off the disk's outer layers as a wind, may irradiate those winds or other forms of ejecta and thereby change their composition, may change the composition and thermodynamic state of the disk itself, and may oscillate in its flavor content. We present the full spatial-, angular-, and energy-dependence of the neutrino distribution function around a realistic model of a nuclear accretion disk, to inform future explorations of these types of behaviors. Spectral Einstein Code (SpEC).

Integrated nuclear data utilisation system for innovative reactors.

PubMed

Yamano, N; Hasegawa, A; Kato, K; Igashira, M

2005-01-01

A five-year research and development project on an integrated nuclear data utilisation system was initiated in 2002, for developing innovative nuclear energy systems such as accelerator-driven systems. The integrated nuclear data utilisation system will be constructed as a modular code system, which consists of two sub-systems: the nuclear data search and plotting sub-system, and the nuclear data processing and utilisation sub-system. The system will be operated with a graphical user interface in order to enable easy utilisation through the Internet by both nuclear design engineers and nuclear data evaluators. This paper presents an overview of the integrated nuclear data utilisation system, describes the development of a prototype system to examine the operability of the user interface and discusses specifications of the two sub-systems.

Globalization of ASME Nuclear Codes and Standards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swayne, Rick; Erler, Bryan A.

2006-07-01

With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countriesmore » around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. (authors)« less

RADTRAD: A simplified model for RADionuclide Transport and Removal And Dose estimation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humphreys, S.L.; Miller, L.A.; Monroe, D.K.

1998-04-01

This report documents the RADTRAD computer code developed for the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation (NRR) to estimate transport and removal of radionuclides and dose at selected receptors. The document includes a users` guide to the code, a description of the technical basis for the code, the quality assurance and code acceptance testing documentation, and a programmers` guide. The RADTRAD code can be used to estimate the containment release using either the NRC TID-14844 or NUREG-1465 source terms and assumptions, or a user-specified table. In addition, the code can account for a reduction in themore » quantity of radioactive material due to containment sprays, natural deposition, filters, and other natural and engineered safety features. The RADTRAD code uses a combination of tables and/or numerical models of source term reduction phenomena to determine the time-dependent dose at user-specified locations for a given accident scenario. The code system also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The RADTRAD code can be used to assess occupational radiation exposures, typically in the control room; to estimate site boundary doses; and to estimate dose attenuation due to modification of a facility or accident sequence.« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

NASA Technical Reports Server (NTRS)

Chen, Shu-cheng S.; Veres, Joseph P.; Fittje, James E.

2006-01-01

A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps and turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance characteristics of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed.

A New Capability for Nuclear Thermal Propulsion Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

MELCOR Peer Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.; Dhir, V.K.; Gieseke, J.A.

1992-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. The newest version of MELCOR is Version 1.8.1, July 1991. MELCOR development has reached the point that the United States Nuclear Regulatory Commission sponsored a broad technical review by recognized experts to determine or confirm the technical adequacy of the code for the serious and complex analyses it is expected to perform. For this purpose, an eight-member MELCOR Peer Review Committee was organized. The Committee has completed its review of the MELCOR code: the review process and findingsmore » of the MELCOR Peer Review Committee are documented in this report. The Committee has determined that recommendations in five areas are appropriate: (1) MELCOR numerics, (2) models missing from MELCOR Version 1.8.1, (3) existing MELCOR models needing revision, (4) the need for expanded MELCOR assessment, and (5) documentation.« less

PBF Reactor Building (PER620) Cubicle 13. Plan, section, details. Note ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620) Cubicle 13. Plan, section, details. Note "quality assurance" code at bottom of drawing. Aerojet Nuclear Company. Date: May 1976. INEEL index no. 761-0620-00-400-195279 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

NASA Astrophysics Data System (ADS)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Security Hardened Cyber Components for Nuclear Power Plants: Phase I SBIR Final Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franusich, Michael D.

SpiralGen, Inc. built a proof-of-concept toolkit for enhancing the cyber security of nuclear power plants and other critical infrastructure with high-assurance instrumentation and control code. The toolkit is based on technology from the DARPA High-Assurance Cyber Military Systems (HACMS) program, which has focused on applying the science of formal methods to the formidable set of problems involved in securing cyber physical systems. The primary challenges beyond HACMS in developing this toolkit were to make the new technology usable by control system engineers and compatible with the regulatory and commercial constraints of the nuclear power industry. The toolkit, packaged as amore » Simulink add-on, allows a system designer to assemble a high-assurance component from formally specified and proven blocks and generate provably correct control and monitor code for that subsystem.« less

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

SINGER: A Computer Code for General Analysis of Two-Dimensional Reinforced Concrete Structures. Volume 1. Solution Process

DTIC Science & Technology

1975-05-01

Conference on Earthquake Engineering, Santiago de Chile, 13-18 January 1969, Vol. I , Session B2, Chilean Association oil Seismology and Earth- quake...Nuclear Agency May 1975 DISTRIBUTED BY: KJ National Technical Information Service U. S. DEPARTMENT OF COMMERCE ^804J AFWL-TR-74-228, Vol. I ...CM o / i ’•fu.r ) V V AFWL-TR- 74-228 Vol. I SINGER: A COMPUTER CODE FOR GENERAL ANALYSIS OF TWO-DIMENSIONAL CONCRETE STRUCTURES Volum« I

Software engineering and automatic continuous verification of scientific software

NASA Astrophysics Data System (ADS)

Piggott, M. D.; Hill, J.; Farrell, P. E.; Kramer, S. C.; Wilson, C. R.; Ham, D.; Gorman, G. J.; Bond, T.

2011-12-01

Software engineering of scientific code is challenging for a number of reasons including pressure to publish and a lack of awareness of the pitfalls of software engineering by scientists. The Applied Modelling and Computation Group at Imperial College is a diverse group of researchers that employ best practice software engineering methods whilst developing open source scientific software. Our main code is Fluidity - a multi-purpose computational fluid dynamics (CFD) code that can be used for a wide range of scientific applications from earth-scale mantle convection, through basin-scale ocean dynamics, to laboratory-scale classic CFD problems, and is coupled to a number of other codes including nuclear radiation and solid modelling. Our software development infrastructure consists of a number of free tools that could be employed by any group that develops scientific code and has been developed over a number of years with many lessons learnt. A single code base is developed by over 30 people for which we use bazaar for revision control, making good use of the strong branching and merging capabilities. Using features of Canonical's Launchpad platform, such as code review, blueprints for designing features and bug reporting gives the group, partners and other Fluidity uers an easy-to-use platform to collaborate and allows the induction of new members of the group into an environment where software development forms a central part of their work. The code repositoriy are coupled to an automated test and verification system which performs over 20,000 tests, including unit tests, short regression tests, code verification and large parallel tests. Included in these tests are build tests on HPC systems, including local and UK National HPC services. The testing of code in this manner leads to a continuous verification process; not a discrete event performed once development has ceased. Much of the code verification is done via the "gold standard" of comparisons to analytical solutions via the method of manufactured solutions. By developing and verifying code in tandem we avoid a number of pitfalls in scientific software development and advocate similar procedures for other scientific code applications.

Proceedings of the symposium on inservice testing of pumps and valves

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1990-10-01

The 1990 Symposium on Inservice Testing of Pumps and Valves, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provided a forum for the discussion of current programs and methods for inservice testing at nuclear power plants. The symposium also provided an opportunity to discuss the need to improve inservice testing in order to ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants resulted in the discussion of a broad spectrum of ideas and perspectives regarding the improvement ofmore » inservice testing of pumps and valves at nuclear power plants.« less

Optical Studies of the Flow Start-Up Processes in Four Convergent-Divergent Nozzles

DTIC Science & Technology

1991-03-01

Evaluation Naval Civil Engineering Laboratory Command ATTN: Code L51, J. Tancreto Nuclear Effects Laboratory Port Hueneme, CA 93043-5003 ATTN: STEWS ...Levine Edwards AFB, CA 93523-5000 104 No. of No. of Copies Organization Copies Organization ALITSTL (Tech. Lib) 2 Director ATTN: J. Lamb Los Alamos

The Advanced Software Development and Commercialization Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallopoulos, E.; Canfield, T.R.; Minkoff, M.

1990-09-01

This is the first of a series of reports pertaining to progress in the Advanced Software Development and Commercialization Project, a joint collaborative effort between the Center for Supercomputing Research and Development of the University of Illinois and the Computing and Telecommunications Division of Argonne National Laboratory. The purpose of this work is to apply techniques of parallel computing that were pioneered by University of Illinois researchers to mature computational fluid dynamics (CFD) and structural dynamics (SD) computer codes developed at Argonne. The collaboration in this project will bring this unique combination of expertise to bear, for the first time,more » on industrially important problems. By so doing, it will expose the strengths and weaknesses of existing techniques for parallelizing programs and will identify those problems that need to be solved in order to enable wide spread production use of parallel computers. Secondly, the increased efficiency of the CFD and SD codes themselves will enable the simulation of larger, more accurate engineering models that involve fluid and structural dynamics. In order to realize the above two goals, we are considering two production codes that have been developed at ANL and are widely used by both industry and Universities. These are COMMIX and WHAMS-3D. The first is a computational fluid dynamics code that is used for both nuclear reactor design and safety and as a design tool for the casting industry. The second is a three-dimensional structural dynamics code used in nuclear reactor safety as well as crashworthiness studies. These codes are currently available for both sequential and vector computers only. Our main goal is to port and optimize these two codes on shared memory multiprocessors. In so doing, we shall establish a process that can be followed in optimizing other sequential or vector engineering codes for parallel processors.« less

TEMPEST: A computer code for three-dimensional analysis of transient fluid dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fort, J.A.

TEMPEST (Transient Energy Momentum and Pressure Equations Solutions in Three dimensions) is a powerful tool for solving engineering problems in nuclear energy, waste processing, chemical processing, and environmental restoration because it analyzes and illustrates 3-D time-dependent computational fluid dynamics and heat transfer analysis. It is a family of codes with two primary versions, a N- Version (available to public) and a T-Version (not currently available to public). This handout discusses its capabilities, applications, numerical algorithms, development status, and availability and assistance.

Validation of a multi-layer Green's function code for ion beam transport

NASA Astrophysics Data System (ADS)

Walker, Steven; Tweed, John; Tripathi, Ram; Badavi, Francis F.; Miller, Jack; Zeitlin, Cary; Heilbronn, Lawrence

To meet the challenge of future deep space programs, an accurate and efficient engineering code for analyzing the shielding requirements against high-energy galactic heavy radiations is needed. In consequence, a new version of the HZETRN code capable of simulating high charge and energy (HZE) ions with either laboratory or space boundary conditions is currently under development. The new code, GRNTRN, is based on a Green's function approach to the solution of Boltzmann's transport equation and like its predecessor is deterministic in nature. The computational model consists of the lowest order asymptotic approximation followed by a Neumann series expansion with non-perturbative corrections. The physical description includes energy loss with straggling, nuclear attenuation, nuclear fragmentation with energy dispersion and down shift. Code validation in the laboratory environment is addressed by showing that GRNTRN accurately predicts energy loss spectra as measured by solid-state detectors in ion beam experiments with multi-layer targets. In order to validate the code with space boundary conditions, measured particle fluences are propagated through several thicknesses of shielding using both GRNTRN and the current version of HZETRN. The excellent agreement obtained indicates that GRNTRN accurately models the propagation of HZE ions in the space environment as well as in laboratory settings and also provides verification of the HZETRN propagator.

138. ARAII Building ARA606 floor plan for remodel as Inel ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

138. ARA-II Building ARA-606 floor plan for remodel as Inel Welding Laboratory. Shows room divisions and welding stations to be installed. Aerojet Nuclear Company 1375-ARA-II-606-E-2. Date: June 1976. Ineel index code no. 070-0606-10-400-156552. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Nuclear Engineering Computer Modules: Reactor Dynamics, RD-1 and RD-2.

ERIC Educational Resources Information Center

Onega, Ronald J.

The objective of the Reactor Dynamics Module, RD-1, is to obtain the kinetics equation without feedback and solve the kinetics equations numerically for one to six delayed neutron groups for time varying reactivity insertions. The computer code FUMOKI (Fundamental Mode Kinetics) will calculate the power as a function of time for either uranium or…

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Kula, K.R.; Paik, I.K.; Chung, D.Y.

1996-12-31

Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, includingmore » source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper.« less

Advances in Geologic Disposal System Modeling and Application to Crystalline Rock

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariner, Paul E.; Stein, Emily R.; Frederick, Jennifer M.

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of used nuclear fuel (UNF) and high-level nuclear waste (HLW). Two of the high priorities for UFDC disposal R&D are design concept development and disposal system modeling (DOE 2011). These priorities are directly addressed in the UFDC Generic Disposal Systems Analysis (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic mediamore » (e.g., salt, granite, clay, and deep borehole disposal). This report describes specific GDSA activities in fiscal year 2016 (FY 2016) toward the development of the enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. The GDSA framework employs the PFLOTRAN thermal-hydrologic-chemical multi-physics code and the Dakota uncertainty sampling and propagation code. Each code is designed for massively-parallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.

2011-02-01

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs,more » and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.« less

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

A&M. TAN607 second floor plan for cold assembly area. Metallurgical ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

A&M. TAN-607 second floor plan for cold assembly area. Metallurgical lab, chemistry lab, nuclear instrument lab, equipment rooms. Ralph M. Parsons 902-ANP-607-A 102. Date: December 1952. Approved by INEEL Classification Office for public release. INEEL index code no. 034-0607-693-106754 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Acceleration of MCNP calculations for small pipes configurations by using Weigth Windows Importance cards created by the SN-3D ATTILA

NASA Astrophysics Data System (ADS)

Castanier, Eric; Paterne, Loic; Louis, Céline

2017-09-01

In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.

SER assistant: An expert system for safety evaluation reports

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeChaine, M.D.; Levine, S.H.; Feltus, M.A.

1993-01-01

The SER Assistant is an expert system that assists engineers to write safety evaluation reports (SERs). Section 50.59 of the Code of Federal Regulations allows modifications to be made to nuclear power plants without prior US Nuclear Regulatory Commission approval if two conditions are satisfied. First, the change must not affect the technical specifications of the plant. Second, the modification must not affect a part of the plant described in the final safety analysis report, or if it does, it must not create an unreviewed safety question. The purpose of an SER is to ensure that these conditions are satisfiedmore » for the proposed modification. The SER Assistant aids this process by providing relevant, but directed, questions and information as well as giving engineers an organized environment to document their thought processes.« less

The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

NASA Astrophysics Data System (ADS)

Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

2014-10-01

The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

Assessment of MSIV full closure for Santa Maria de Garona Nuclear Power Plant using TRAC-BF1 (G1J1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crespo, J.L.; Fernandez, R.A.

1993-06-01

This document presents a spurious Main Steam Isolation Value (MSIV) closure analysis for Santa Maria de Garorta Nuclear Power Plan describing the problems found when comparing calculated and real data. The plant is a General Electric Boiling Water Reactor 3, containment type Mark 1. It is operated by NUCLENOR, S.A. and was connected to the grid in 1971. The analysis has been performed by the Apphed Physics Department from the University of Cantabria and the Analysis and Operation Section from NUCLENOR, S.A. as a part of an agreement for developing an engineering simulator of operational transients and accidents for Santamore » Maria de Gamma Power Plant. The analysis was performed using the frozen version of TRAC-BFI (GlJl) code and is the second of two NUCLENOR contributions to the International Code Applications and Assessment Program (ICAP). The code was run in a Cyber 932 with operating system NOS/VE, property of NUCLENOR, S.A.. A programming effort was carried out in order to provide suitable graphics from the output file.« less

SAM Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

Guide to Using Sierra

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaw, Ryan Phillip; Agelastos, Anthony Michael; Miller, Joel D.

2015-03-01

Sierra is an engineering mechanics simulation code suite supporting the Nation's Nuclear Weapons mission as well as other customers. It has explicit ties to Sandia National Labs' workfow, including geometry and meshing, design and optimization, and visualization. Dis- tinguishing strengths include "application aware" development, scalability, SQA and V&V, multiple scales, and multi-physics coupling. This document is intended to help new and existing users of Sierra as a user manual and troubleshooting guide.

Guide to Using Sierra

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaw, Ryan Phillip; Agelastos, Anthony Michael; Miller, Joel D.

2017-04-01

Sierra is an engineering mechanics simulation code suite supporting the Nation's Nuclear Weapons mission as well as other customers. It has explicit ties to Sandia National Labs' workfow, including geometry and meshing, design and optimization, and visualization. Dis- tinguishing strengths include "application aware" development, scalability, SQA and V&V, multiple scales, and multi-physics coupling. This document is intended to help new and existing users of Sierra as a user manual and troubleshooting guide.

120. ARAI Expansion of ARA627 shop and maintenance building for ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

120. ARA-I Expansion of ARA-627 shop and maintenance building for new use as materials and metallurgy laboratory. Shows ground floor plan addition of gas analyzer room, fatigue testing room, microscope room, and offices. Idaho Nuclear Corporation 1230-ARA-627-A-5. Date: June 1970. Ineel index code no. 068-0627-00-400-154062. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

PRA and Risk Informed Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernsen, Sidney A.; Simonen, Fredric A.; Balkey, Kenneth R.

2006-01-01

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codesmore » and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. The approach allows calculations of conditional core damage probabilities both for component failures that cause initiating events and failures in standby systems that decrease the availability of these systems. The paper relates the explicit risk based methods of the new Section XI code cases to the implicit consideration of risk used in the development of Section XI. Other topics include the needed interactions of ISI engineers, plant operating staff, PRA specialists, and members of expert panels that review the risk based programs.« less

TOOKUIL: A case study in user interface development for safety code application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gray, D.L.; Harkins, C.K.; Hoole, J.G.

1997-07-01

Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today`s safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interfacemore » named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL.« less

NEAMS Update. Quarterly Report for October - December 2011.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley, K.

2012-02-16

The Advanced Modeling and Simulation Office within the DOE Office of Nuclear Energy (NE) has been charged with revolutionizing the design tools used to build nuclear power plants during the next 10 years. To accomplish this, the DOE has brought together the national laboratories, U.S. universities, and the nuclear energy industry to establish the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program. The mission of NEAMS is to modernize computer modeling of nuclear energy systems and improve the fidelity and validity of modeling results using contemporary software environments and high-performance computers. NEAMS will create a set of engineering-level codes aimedmore » at designing and analyzing the performance and safety of nuclear power plants and reactor fuels. The truly predictive nature of these codes will be achieved by modeling the governing phenomena at the spatial and temporal scales that dominate the behavior. These codes will be executed within a simulation environment that orchestrates code integration with respect to spatial meshing, computational resources, and execution to give the user a common 'look and feel' for setting up problems and displaying results. NEAMS is building upon a suite of existing simulation tools, including those developed by the federal Scientific Discovery through Advanced Computing and Advanced Simulation and Computing programs. NEAMS also draws upon existing simulation tools for materials and nuclear systems, although many of these are limited in terms of scale, applicability, and portability (their ability to be integrated into contemporary software and hardware architectures). NEAMS investments have directly and indirectly supported additional NE research and development programs, including those devoted to waste repositories, safeguarded separations systems, and long-term storage of used nuclear fuel. NEAMS is organized into two broad efforts, each comprising four elements. The quarterly highlights October-December 2011 are: (1) Version 1.0 of AMP, the fuel assembly performance code, was tested on the JAGUAR supercomputer and released on November 1, 2011, a detailed discussion of this new simulation tool is given; (2) A coolant sub-channel model and a preliminary UO{sub 2} smeared-cracking model were implemented in BISON, the single-pin fuel code, more information on how these models were developed and benchmarked is given; (3) The Object Kinetic Monte Carlo model was implemented to account for nucleation events in meso-scale simulations and a discussion of the significance of this advance is given; (4) The SHARP neutronics module, PROTEUS, was expanded to be applicable to all types of reactors, and a discussion of the importance of PROTEUS is given; (5) A plan has been finalized for integrating the high-fidelity, three-dimensional reactor code SHARP with both the systems-level code RELAP7 and the fuel assembly code AMP. This is a new initiative; (6) Work began to evaluate the applicability of AMP to the problem of dry storage of used fuel and to define a relevant problem to test the applicability; (7) A code to obtain phonon spectra from the force-constant matrix for a crystalline lattice has been completed. This important bridge between subcontinuum and continuum phenomena is discussed; (8) Benchmarking was begun on the meso-scale, finite-element fuels code MARMOT to validate its new variable splitting algorithm; (9) A very computationally demanding simulation of diffusion-driven nucleation of new microstructural features has been completed. An explanation of the difficulty of this simulation is given; (10) Experiments were conducted with deformed steel to validate a crystal plasticity finite-element code for bodycentered cubic iron; (11) The Capability Transfer Roadmap was completed and published as an internal laboratory technical report; (12) The AMP fuel assembly code input generator was integrated into the NEAMS Integrated Computational Environment (NiCE). More details on the planned NEAMS computing environment is given; and (13) The NEAMS program website (neams.energy.gov) is nearly ready to launch.« less

Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-07-01

The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regardingmore » the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately.« less

Comparative Studies on UO2 Fueled HTTR Several Nuclear Data Libraries

NASA Astrophysics Data System (ADS)

Hidayati, Anni N.; Prastyo, Puguh A.; Waris, Abdul; Irwanto, Dwi

2017-07-01

HTTR (High Temperature Engineering Test Reactor) is one of Generation IV nuclear reactors that has been developed by JAERI (former name of JAEA, JAPAN). HTTR uses graphite moderator, helium gas coolant with UO2 fuel and outlet coolant temperature of 900°C or higher than that. Several studies regarding HTTR have been performed by employing JENDL 3.2 nuclear data libraries. In this paper, comparative evaluation of HTTR with several nuclear data libraries (JENDL 3.3, JENDL 4.0, and JEF 3.1) have been conducted.. The 3-D calculation was performed by using CITATION module of SRAC 2006 code. The result shows some differences between those nuclear data libraries result. K-eff or core effective multiplication factor results are about 1.17, 1,18 and 1,19 (JENDL 3.3, JENDL 4.0, and JEF 3.1) at Begin of Life, also at the End of Life (after two years operation) are 1.16, 1.17 and 1.17 for each nuclear data libraries. There are some different result of K-eff but for neutron spectra results, those nuclear data libraries show the same result.

Grizzly Usage and Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, B. W.; Backman, M.; Chakraborty, P.

2016-03-01

Grizzly is a multiphysics simulation code for characterizing the behavior of nuclear power plant (NPP) structures, systems and components (SSCs) subjected to a variety of age-related aging mechanisms. Grizzly simulates both the progression of aging processes, as well as the capacity of aged components to safely perform. This initial beta release of Grizzly includes capabilities for engineering-scale thermo-mechanical analysis of reactor pressure vessels (RPVs). Grizzly will ultimately include capabilities for a wide range of components and materials. Grizzly is in a state of constant development, and future releases will broaden the capabilities of this code for RPV analysis, as wellmore » as expand it to address degradation in other critical NPP components.« less

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

Closed-Cycle Engine Program Used to Study Brayton Power Conversion

NASA Technical Reports Server (NTRS)

Johnson, Paul K.

2005-01-01

One form of power conversion under consideration in NASA Glenn Research Center's Thermal Energy Conversion Branch is the closed-Brayton-cycle engine. In the tens-of-kilowatts to multimegawatt class, the Brayton engine lends itself to potential space nuclear power applications such as electric propulsion or surface power. The Thermal Energy Conversion Branch has most recently concentrated its Brayton studies on electric propulsion for Prometheus. One piece of software used for evaluating such designs over a limited tradeoff space has been the Closed Cycle Engine Program (CCEP). The CCEP originated in the mid-1980s from a Fortran aircraft engine code known as the Navy/NASA Engine Program (NNEP). Components such as a solar collector, heat exchangers, ducting, a pumped-loop radiator, a nuclear heat source, and radial turbomachinery were added to NNEP, transforming it into a high-fidelity design and performance tool for closed-Brayton-cycle power conversion and heat rejection. CCEP was used in the 1990s in conjunction with the Solar Dynamic Ground Test Demonstration conducted at Glenn. Over the past year, updates were made to CCEP to adapt it for an electric propulsion application. The pumped-loop radiator coolant can now be n-heptane, water, or sodium-potassium (NaK); liquid-metal pump design tables were added to accommodate the NaK fluid. For the reactor and shield, a user can now elect to calculate a higher fidelity mass estimate. In addition, helium-xenon working-fluid properties were recalculated and updated.

Status of nuclear Class 1 component requalification: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cooper, W.E.

1986-12-01

Qualification relates to assurance of acceptability of the component with respect to structural integrity, operability and functional capability. Requalification is required if existing qualification is lost because of: expiration of the qualified service life (life extension); reactivation of a cancelled or suspended plant; failure to conform with certain requirements of the Technical Specifications, or revision to the applicable Regulatory requirements. The alternatives to requalification are replacement or removal from service. The choice between requalification, replacement and removal from service is governed by economics. The purpose of requalification standards is to ensure the acceptability of the requalification process. A previous EPRImore » Report prepared by Teledyne Engineering Services (TES) (NP-1921) developed a rationale for, and a draft of, a generic requalification standard for Class 1 Pressure Boundary Components presently considered by the Boiler and Pressure Vessel Code published by The American Society of Mechanical Engineers (ASME/BPVC). International Energy Associates Limited (IEA) prepared another report for EPRI shortly thereafter (NP-2418), which reviewed the economic and technologies factors of nuclear plant life extension, and concluded that NP-1921 makes a strong case that the nuclear industry will benefit from the development of the proposed standard.« less

Interactive-graphic flowpath plotting for turbine engines

NASA Technical Reports Server (NTRS)

Corban, R. R.

1981-01-01

An engine cycle program capable of simulating the design and off-design performance of arbitrary turbine engines, and a computer code which, when used in conjunction with the cycle code, can predict the weight of the engines are described. A graphics subroutine was added to the code to enable the engineer to visualize the designed engine with more clarity by producing an overall view of the designed engine for output on a graphics device using IBM-370 graphics subroutines. In addition, with the engine drawn on a graphics screen, the program allows for the interactive user to make changes to the inputs to the code for the engine to be redrawn and reweighed. These improvements allow better use of the code in conjunction with the engine program.

Development of Fuel Shuffling Module for PHISICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allan Mabe; Andrea Alfonsi; Cristian Rabiti

2013-06-01

PHISICS (Parallel and Highly Innovative Simulation for the INL Code System) [4] code toolkit has been in development at the Idaho National Laboratory. This package is intended to provide a modern analysis tool for reactor physics investigation. It is designed with the mindset to maximize accuracy for a given availability of computational resources and to give state of the art tools to the modern nuclear engineer. This is obtained by implementing several different algorithms and meshing approaches among which the user will be able to choose, in order to optimize his computational resources and accuracy needs. The software is completelymore » modular in order to simplify the independent development of modules by different teams and future maintenance. The package is coupled with the thermo-hydraulic code RELAP5-3D [3]. In the following the structure of the different PHISICS modules is briefly recalled, focusing on the new shuffling module (SHUFFLE), object of this paper.« less

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Berry, Jan; Ferrada, Juan J; Curd, Warren

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predictedmore » to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance. This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.« less

Engine Cycle Analysis for a Particle Bed Reactor Nuclear Rocket

DTIC Science & Technology

1991-03-01

0 DTIC USERS UNCLASSIFIED 22a. NAME OF RESPONSIBLE INDIVIDUAL ZZb. TELEPHONE (Include Area Code) 22c. OFFICE SYMBOL Lt Timothy J . Lawrence 805-275...Cycle with 2000 MW PBR and Uncooled Nozzle J : Output for Bleed Cycle with 2000 MW PBR and Cooled Nozzle K: Output for Expander Cycle with 2000 MW PBR L...Mars with carbon dioxide, the primary component of the Martian atmosphere. Carbon dioxide would delivera smaller ! j , but its use would eliminate the

Engineering Ethics Education Having Reflected Various Values and a Global Code of Ethics

NASA Astrophysics Data System (ADS)

Kanemitsu, Hidekazu

At the present day, a movement trying to establish a global code of ethics for science and engineering is in activity. The author overviews the context of this movement, and examines the possibility of engineering ethics education which uses global code of ethics. In this paper, the engineering ethics education which uses code of ethics in general will be considered, and an expected function of global code of ethics will be also. Engineering ethics education in the new century should be aimed to share the values among different countries and cultures. To use global code of ethics as a tool for such education, the code should include various values, especially Asian values which engineering ethics has paid little attention to.

Design-Load Basis for LANL Structures, Systems, and Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

I. Cuesta

2004-09-01

This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loadsmore » not related to natural phenomena hazards, and (3) the design loads on structures during construction.« less

EVALUATION OF ULTRASONIC PHASED-ARRAY FOR DETECTION OF PLANAR FLAWS IN HIGH-DENSITY POLYETHYLENE (HDPE) BUTT-FUSION JOINTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prowant, Matthew S.; Denslow, Kayte M.; Moran, Traci L.

2016-09-21

The desire to use high-density polyethylene (HDPE) piping in buried Class 3 service and cooling water systems in nuclear power plants is primarily motivated by the material’s high resistance to corrosion relative to that of steel and metal alloys. The rules for construction of Class 3 HDPE pressure piping systems were originally published in Code Case N-755 and were recently incorporated into the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC) Section III as Mandatory Appendix XXVI (2015 Edition). The requirements for HDPE examination are guided by criteria developed for metal pipe and are based onmore » industry-led HDPE research or conservative calculations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

LaSalle, F.R.; Golbeg, P.R.; Chenault, D.M.

For reactor and nuclear facilities, both Title 10, Code of Federal Regulations, Part 50, and US Department of Energy Order 6430.1A require assessments of the interaction of non-Safety Class 1 piping and equipment with Safety Class 1 piping and equipment during a seismic event to maintain the safety function. The safety class systems of nuclear reactors or nuclear facilities are designed to the applicable American Society of Mechanical Engineers standards and Seismic Category 1 criteria that require rigorous analysis, construction, and quality assurance. Because non-safety class systems are generally designed to lesser standards and seismic criteria, they may become missilesmore » during a safe shutdown earthquake. The resistance of piping, tubing, and equipment to seismically generated missiles is addressed in the paper. Gross plastic and local penetration failures are considered with applicable test verification. Missile types and seismic zones of influence are discussed. Field qualification data are also developed for missile evaluation.« less

Analysis of Material Sample Heated by Impinging Hot Hydrogen Jet in a Non-Nuclear Tester

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Foote, John; Litchford, Ron

2006-01-01

A computational conjugate heat transfer methodology was developed and anchored with data obtained from a hot-hydrogen jet heated, non-nuclear materials tester, as a first step towards developing an efficient and accurate multiphysics, thermo-fluid computational methodology to predict environments for hypothetical solid-core, nuclear thermal engine thrust chamber. The computational methodology is based on a multidimensional, finite-volume, turbulent, chemically reacting, thermally radiating, unstructured-grid, and pressure-based formulation. The multiphysics invoked in this study include hydrogen dissociation kinetics and thermodynamics, turbulent flow, convective and thermal radiative, and conjugate heat transfers. Predicted hot hydrogen jet and material surface temperatures were compared with those of measurement. Predicted solid temperatures were compared with those obtained with a standard heat transfer code. The interrogation of physics revealed that reactions of hydrogen dissociation and recombination are highly correlated with local temperature and are necessary for accurate prediction of the hot-hydrogen jet temperature.

Encoded physics knowledge in checking codes for nuclear cross section libraries at Los Alamos

NASA Astrophysics Data System (ADS)

Parsons, D. Kent

2017-09-01

Checking procedures for processed nuclear data at Los Alamos are described. Both continuous energy and multi-group nuclear data are verified by locally developed checking codes which use basic physics knowledge and common-sense rules. A list of nuclear data problems which have been identified with help of these checking codes is also given.

Development of Yield and Tensile Strength Design Curves for Alloy 617

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nancy Lybeck; T. -L. Sham

2013-10-01

The U.S. Department of Energy Very High Temperature Reactor Program is acquiring data in preparation for developing an Alloy 617 Code Case for inclusion in the nuclear section of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. A draft code case was previously developed, but effort was suspended before acceptance by ASME. As part of the draft code case effort, a database was compiled of yield and tensile strength data from tests performed in air. Yield strength and tensile strength at temperature are used to set time independent allowable stress for construction materials in B&PVmore » Code, Section III, Subsection NH. The yield and tensile strength data used for the draft code case has been augmented with additional data generated by Idaho National Laboratory and Oak Ridge National Laboratory in the U.S. and CEA in France. The standard ASME Section II procedure for generating yield and tensile strength at temperature is presented, along with alternate methods that accommodate the change in temperature trends seen at high temperatures, resulting in a more consistent design margin over the temperature range of interest.« less

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

Comprehensive Report For Proposed Elevated Temperature Elastic Perfectly Plastic (EPP) Code Cases Representative Example Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hollinger, Greg L.

Background: The current rules in the nuclear section of the ASME Boiler and Pressure Vessel (B&PV) Code , Section III, Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200F (650C)1. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (E-PP) analysis methods and which are expected to be applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature 2,more » 3, and have been recently revised to incorporate comments and simplify their application. The revised code cases have been developed. Task Objectives: The goal of the Sample Problem task is to exercise these code cases through example problems to demonstrate their feasibility and, also, to identify potential corrections and improvements should problems be encountered. This will provide input to the development of technical background documents for consideration by the applicable B&PV committees considering these code cases for approval. This task has been performed by Hollinger and Pease of Becht Engineering Co., Inc., Nuclear Services Division and a report detailing the results of the E-PP analyses conducted on example problems per the procedures of the E-PP strain limits and creep-fatigue draft code cases is enclosed as Enclosure 1. Conclusions: The feasibility of the application of the E-PP code cases has been demonstrated through example problems that consist of realistic geometry (a nozzle attached to a semi-hemispheric shell with a circumferential weld) and load (pressure; pipe reaction load applied at the end of the nozzle, including axial and shear forces, bending and torsional moments; through-wall transient temperature gradient) and design and operating conditions (Levels A, B and C).« less

Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adrian Miron; Joshua Valentine; John Christenson

2009-10-01

The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFCmore » codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.« less

Nuclear Fuels & Materials Spotlight Volume 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schultz, Peter Andrew

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomicmore » scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.« less

Environmental Ethics and Civil Engineering.

ERIC Educational Resources Information Center

Vesilind, P. Aarne

1987-01-01

Traces the development of the civil engineering code of ethics. Points out that the code does have an enforceable provision that addresses the engineer's responsibility toward the environment. Suggests revisions to the code to accommodate the environmental impacts of civil engineering. (TW)

The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

NASA Astrophysics Data System (ADS)

Iwamoto, O.; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

2016-01-01

A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

Two-phase flow in the cooling circuit of a cryogenic rocket engine

NASA Astrophysics Data System (ADS)

Preclik, D.

1992-07-01

Transient two-phase flow was investigated for the hydrogen cooling circuit of the HM7 rocket engine. The nuclear reactor code ATHLET/THESEUS was adapted to cryogenics and applied to both principal and prototype experiments for validation and simulation purposes. The cooling circuit two-phase flow simulation focused on the hydrogen prechilling and pump transient phase prior to ignition. Both a single- and a multichannel model were designed and employed for a valve leakage flow, a nominal prechilling flow, and a prechilling with a subsequent pump-transient flow. The latter case was performed in order to evaluate the difference between a nominal and a delayed turbo-pump start-up. It was found that an extension of the nominal prechilling sequence in the order of 1 second is sufficient to finally provide for liquid injection conditions of hydrogen which, as commonly known, is undesirable for smooth ignition and engine starting transients.

EMPIRE: A code for nuclear astrophysics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Palumbo, A.

The nuclear reaction code EMPIRE is presented as a useful tool for nuclear astrophysics. EMPIRE combines a variety of the reaction models with a comprehensive library of input parameters providing a diversity of options for the user. With exclusion of the directsemidirect capture all reaction mechanisms relevant to the nuclear astrophysics energy range of interest are implemented in the code. Comparison to experimental data show consistent agreement for all relevant channels.

Proceedings of the 21st DOE/NRC Nuclear Air Cleaning Conference; Sessions 1--8

DOE Office of Scientific and Technical Information (OSTI.GOV)

First, M.W.

1991-02-01

Separate abstracts have been prepared for the papers presented at the meeting on nuclear facility air cleaning technology in the following specific areas of interest: air cleaning technologies for the management and disposal of radioactive wastes; Canadian waste management program; radiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off-gas cleaning; incineration and vitrification; adsorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; and nuclear air cleaning programs around the world. (MB)

SOC-DS computer code provides tool for design evaluation of homogeneous two-material nuclear shield

NASA Technical Reports Server (NTRS)

Disney, R. K.; Ricks, L. O.

1967-01-01

SOC-DS Code /Shield Optimization Code-Direc Search/, selects a nuclear shield material of optimum volume, weight, or cost to meet the requirments of a given radiation dose rate or energy transmission constraint. It is applicable to evaluating neutron and gamma ray shields for all nuclear reactors.

Performance Assessments of Generic Nuclear Waste Repositories in Shale

NASA Astrophysics Data System (ADS)

Stein, E. R.; Sevougian, S. D.; Mariner, P. E.; Hammond, G. E.; Frederick, J.

2017-12-01

Simulations of deep geologic disposal of nuclear waste in a generic shale formation showcase Geologic Disposal Safety Assessment (GDSA) Framework, a toolkit for repository performance assessment (PA) whose capabilities include domain discretization (Cubit), multiphysics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). GDSA Framework is used to conduct PAs of two generic repositories in shale. The first considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel. The second considers disposal of defense-related spent nuclear fuel and high level waste. Each PA accounts for the thermal load and radionuclide inventory of applicable waste types, components of the engineered barrier system, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Model domains are half-symmetry, gridded with Cubit, and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high performance computing environment on as many as 2048 processes. Equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation, waste form dissolution, radioactive decay and ingrowth, sorption, solubility, advection, dispersion, and diffusion. Simulations are run to 106 y, and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA-0003525. SAND2017- 8305 A

Final Report for Project 13-4791: New Mechanistic Models of Creep-Fatigue Crack Growth Interactions for Advanced High Temperature Reactor Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kruzic, Jamie J; Siegmund, Thomas; Tomar, Vikas

This project developed and validated a novel, multi-scale, mechanism-based model to quantitatively predict creep-fatigue crack growth and failure for Ni-based Alloy 617 at 800°C. Alloy 617 is a target material for intermediate heat exchangers in Generation IV very high temperature reactor designs, and it is envisioned that this model will aid in the design of safe, long lasting nuclear power plants. The technical effectiveness of the model was shown by demonstrating that experimentally observed crack growth rates can be predicted under both steady state and overload crack growth conditions. Feasibility was considered by incorporating our model into a commercially availablemore » finite element method code, ABAQUS, that is commonly used by design engineers. While the focus of the project was specifically on an alloy targeted for Generation IV nuclear reactors, the benefits to the public are expected to be wide reaching. Indeed, creep-fatigue failure is a design consideration for a wide range of high temperature mechanical systems that rely on Ni-based alloys, including industrial gas power turbines, advanced ultra-super critical steam turbines, and aerospace turbine engines. It is envisioned that this new model can be adapted to a wide range of engineering applications.« less

Job Prospects for Nuclear Engineers.

ERIC Educational Resources Information Center

Basta, Nicholas

1985-01-01

As the debate over nuclear safety continues, the job market remains healthy for nuclear engineers. The average salary offered to new nuclear engineers with bachelor's degrees is $27,400. Salary averages and increases compare favorably with other engineering disciplines. Various job sources in the field are noted. (JN)

Mean Line Pump Flow Model in Rocket Engine System Simulation

NASA Technical Reports Server (NTRS)

Veres, Joseph P.; Lavelle, Thomas M.

2000-01-01

A mean line pump flow modeling method has been developed to provide a fast capability for modeling turbopumps of rocket engines. Based on this method, a mean line pump flow code PUMPA has been written that can predict the performance of pumps at off-design operating conditions, given the loss of the diffusion system at the design point. The pump code can model axial flow inducers, mixed-flow and centrifugal pumps. The code can model multistage pumps in series. The code features rapid input setup and computer run time, and is an effective analysis and conceptual design tool. The map generation capability of the code provides the map information needed for interfacing with a rocket engine system modeling code. The off-design and multistage modeling capabilities of the code permit parametric design space exploration of candidate pump configurations and provide pump performance data for engine system evaluation. The PUMPA code has been integrated with the Numerical Propulsion System Simulation (NPSS) code and an expander rocket engine system has been simulated. The mean line pump flow code runs as an integral part of the NPSS rocket engine system simulation and provides key pump performance information directly to the system model at all operating conditions.

Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavisca, M.; Ghaderi, M.; Khatib-Rahbar, M.

2006-07-01

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U.S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents themore » results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990's. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models. (authors)« less

An Object-Oriented Computer Code for Aircraft Engine Weight Estimation

NASA Technical Reports Server (NTRS)

Tong, Michael T.; Naylor, Bret A.

2009-01-01

Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. At NASA Glenn Research Center (GRC), the Weight Analysis of Turbine Engines (WATE) computer code, originally developed by Boeing Aircraft, has been used to estimate the engine weight of various conceptual engine designs. The code, written in FORTRAN, was originally developed for NASA in 1979. Since then, substantial improvements have been made to the code to improve the weight calculations for most of the engine components. Most recently, to improve the maintainability and extensibility of WATE, the FORTRAN code has been converted into an object-oriented version. The conversion was done within the NASA's NPSS (Numerical Propulsion System Simulation) framework. This enables WATE to interact seamlessly with the thermodynamic cycle model which provides component flow data such as airflows, temperatures, and pressures, etc., that are required for sizing the components and weight calculations. The tighter integration between the NPSS and WATE would greatly enhance system-level analysis and optimization capabilities. It also would facilitate the enhancement of the WATE code for next-generation aircraft and space propulsion systems. In this paper, the architecture of the object-oriented WATE code (or WATE++) is described. Both the FORTRAN and object-oriented versions of the code are employed to compute the dimensions and weight of a 300-passenger aircraft engine (GE90 class). Both versions of the code produce essentially identical results as should be the case.

An Object-oriented Computer Code for Aircraft Engine Weight Estimation

NASA Technical Reports Server (NTRS)

Tong, Michael T.; Naylor, Bret A.

2008-01-01

Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. At NASA Glenn (GRC), the Weight Analysis of Turbine Engines (WATE) computer code, originally developed by Boeing Aircraft, has been used to estimate the engine weight of various conceptual engine designs. The code, written in FORTRAN, was originally developed for NASA in 1979. Since then, substantial improvements have been made to the code to improve the weight calculations for most of the engine components. Most recently, to improve the maintainability and extensibility of WATE, the FORTRAN code has been converted into an object-oriented version. The conversion was done within the NASA s NPSS (Numerical Propulsion System Simulation) framework. This enables WATE to interact seamlessly with the thermodynamic cycle model which provides component flow data such as airflows, temperatures, and pressures, etc. that are required for sizing the components and weight calculations. The tighter integration between the NPSS and WATE would greatly enhance system-level analysis and optimization capabilities. It also would facilitate the enhancement of the WATE code for next-generation aircraft and space propulsion systems. In this paper, the architecture of the object-oriented WATE code (or WATE++) is described. Both the FORTRAN and object-oriented versions of the code are employed to compute the dimensions and weight of a 300- passenger aircraft engine (GE90 class). Both versions of the code produce essentially identical results as should be the case. Keywords: NASA, aircraft engine, weight, object-oriented

Design and Analysis of Boiler Pressure Vessels based on IBR codes

NASA Astrophysics Data System (ADS)

Balakrishnan, B.; Kanimozhi, B.

2017-05-01

Pressure vessels components are widely used in the thermal and nuclear power plants for generating steam using the philosophy of heat transfer. In Thermal power plant, Coal is burnt inside the boiler furnace for generating the heat. The amount of heat produced through the combustion of pulverized coal is used in changing the phase transfer (i.e. Water into Super-Heated Steam) in the Pressure Parts Component. Pressure vessels are designed as per the Standards and Codes of the country, where the boiler is to be installed. One of the Standards followed in designing Pressure Parts is ASME (American Society of Mechanical Engineers). The mandatory requirements of ASME code must be satisfied by the manufacturer. In our project case, A Shell/pipe which has been manufactured using ASME code has an issue during the drilling of hole. The Actual Size of the drilled holes must be, as per the drawing, but due to error, the size has been differentiate from approved design calculation (i.e. the diameter size has been exceeded). In order to rectify this error, we have included an additional reinforcement pad to the drilled and modified the design of header in accordance with the code requirements.

Code of Ethics for Electrical Engineers

NASA Astrophysics Data System (ADS)

Matsuki, Junya

The Institute of Electrical Engineers of Japan (IEEJ) has established the rules of practice for its members recently, based on its code of ethics enacted in 1998. In this paper, first, the characteristics of the IEEJ 1998 ethical code are explained in detail compared to the other ethical codes for other fields of engineering. Secondly, the contents which shall be included in the modern code of ethics for electrical engineers are discussed. Thirdly, the newly-established rules of practice and the modified code of ethics are presented. Finally, results of questionnaires on the new ethical code and rules which were answered on May 23, 2007, by 51 electrical and electronic students of the University of Fukui are shown.

A surface code quantum computer in silicon

PubMed Central

Hill, Charles D.; Peretz, Eldad; Hile, Samuel J.; House, Matthew G.; Fuechsle, Martin; Rogge, Sven; Simmons, Michelle Y.; Hollenberg, Lloyd C. L.

2015-01-01

The exceptionally long quantum coherence times of phosphorus donor nuclear spin qubits in silicon, coupled with the proven scalability of silicon-based nano-electronics, make them attractive candidates for large-scale quantum computing. However, the high threshold of topological quantum error correction can only be captured in a two-dimensional array of qubits operating synchronously and in parallel—posing formidable fabrication and control challenges. We present an architecture that addresses these problems through a novel shared-control paradigm that is particularly suited to the natural uniformity of the phosphorus donor nuclear spin qubit states and electronic confinement. The architecture comprises a two-dimensional lattice of donor qubits sandwiched between two vertically separated control layers forming a mutually perpendicular crisscross gate array. Shared-control lines facilitate loading/unloading of single electrons to specific donors, thereby activating multiple qubits in parallel across the array on which the required operations for surface code quantum error correction are carried out by global spin control. The complexities of independent qubit control, wave function engineering, and ad hoc quantum interconnects are explicitly avoided. With many of the basic elements of fabrication and control based on demonstrated techniques and with simulated quantum operation below the surface code error threshold, the architecture represents a new pathway for large-scale quantum information processing in silicon and potentially in other qubit systems where uniformity can be exploited. PMID:26601310

A surface code quantum computer in silicon.

PubMed

Hill, Charles D; Peretz, Eldad; Hile, Samuel J; House, Matthew G; Fuechsle, Martin; Rogge, Sven; Simmons, Michelle Y; Hollenberg, Lloyd C L

2015-10-01

The exceptionally long quantum coherence times of phosphorus donor nuclear spin qubits in silicon, coupled with the proven scalability of silicon-based nano-electronics, make them attractive candidates for large-scale quantum computing. However, the high threshold of topological quantum error correction can only be captured in a two-dimensional array of qubits operating synchronously and in parallel-posing formidable fabrication and control challenges. We present an architecture that addresses these problems through a novel shared-control paradigm that is particularly suited to the natural uniformity of the phosphorus donor nuclear spin qubit states and electronic confinement. The architecture comprises a two-dimensional lattice of donor qubits sandwiched between two vertically separated control layers forming a mutually perpendicular crisscross gate array. Shared-control lines facilitate loading/unloading of single electrons to specific donors, thereby activating multiple qubits in parallel across the array on which the required operations for surface code quantum error correction are carried out by global spin control. The complexities of independent qubit control, wave function engineering, and ad hoc quantum interconnects are explicitly avoided. With many of the basic elements of fabrication and control based on demonstrated techniques and with simulated quantum operation below the surface code error threshold, the architecture represents a new pathway for large-scale quantum information processing in silicon and potentially in other qubit systems where uniformity can be exploited.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE PAGES

Hu, Jianwei; Gauld, Ian C.

2014-12-01

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Gauld, Ian C.

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Evaluating Open-Source Full-Text Search Engines for Matching ICD-10 Codes.

PubMed

Jurcău, Daniel-Alexandru; Stoicu-Tivadar, Vasile

2016-01-01

This research presents the results of evaluating multiple free, open-source engines on matching ICD-10 diagnostic codes via full-text searches. The study investigates what it takes to get an accurate match when searching for a specific diagnostic code. For each code the evaluation starts by extracting the words that make up its text and continues with building full-text search queries from the combinations of these words. The queries are then run against all the ICD-10 codes until a match indicates the code in question as a match with the highest relative score. This method identifies the minimum number of words that must be provided in order for the search engines choose the desired entry. The engines analyzed include a popular Java-based full-text search engine, a lightweight engine written in JavaScript which can even execute on the user's browser, and two popular open-source relational database management systems.

Current and anticipated uses of thermal-hydraulic codes in Spain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pelayo, F.; Reventos, F.

1997-07-01

Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to themore » different uses of the codes, the applications mainly come from the authority, industry, universities and research institutions. The main conclusion derived from this paper establishes that further code development is justified if the following requisites are considered: (1) Safety relevance of scenarios not presently covered is established. (2) A substantial gain in margins or the capability to use realistic assumptions is obtained. (3) A general consensus on the licensability and methodology for application is reached. The role of Regulatory Body is stressed, as the most relevant outcome of the project may be related to the evolution of the licensing frame.« less

Composite materials. Volume 3 - Engineering applications of composites. Volume 4 - Metallic matrix composites. Volume 8 - Structural design and analysis, Part 2

NASA Technical Reports Server (NTRS)

Noton, B. R. (Editor); Kreider, K. G.; Chamis, C. C.

1974-01-01

This volume discusses a vaety of applications of both low- and high-cost composite materials in a number of selected engineering fields. The text stresses the use of fiber-reinforced composites, along with interesting material systems used in the electrical and nuclear industries. As to technology transfer, a similarity is noted between many of the reasons responsible for the utilization of composites and those problems requiring urgent solution, such as mechanized fabrication processes and design for production. Features topics include road transportation, rail transportation, civil aircraft, space vehicles, builing industry, chemical plants, and appliances and equipment. The laminate orientation code devised by Air Force materials laboratory is included. Individual items are announced in this issue.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed bymore » Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.« less

TALYS/TENDL verification and validation processes: Outcomes and recommendations

NASA Astrophysics Data System (ADS)

Fleming, Michael; Sublet, Jean-Christophe; Gilbert, Mark R.; Koning, Arjan; Rochman, Dimitri

2017-09-01

The TALYS-generated Evaluated Nuclear Data Libraries (TENDL) provide truly general-purpose nuclear data files assembled from the outputs of the T6 nuclear model codes system for direct use in both basic physics and engineering applications. The most recent TENDL-2015 version is based on both default and adjusted parameters of the most recent TALYS, TAFIS, TANES, TARES, TEFAL, TASMAN codes wrapped into a Total Monte Carlo loop for uncertainty quantification. TENDL-2015 contains complete neutron-incident evaluations for all target nuclides with Z ≤116 with half-life longer than 1 second (2809 isotopes with 544 isomeric states), up to 200 MeV, with covariances and all reaction daughter products including isomers of half-life greater than 100 milliseconds. With the added High Fidelity Resonance (HFR) approach, all resonances are unique, following statistical rules. The validation of the TENDL-2014/2015 libraries against standard, evaluated, microscopic and integral cross sections has been performed against a newly compiled UKAEA database of thermal, resonance integral, Maxwellian averages, 14 MeV and various accelerator-driven neutron source spectra. This has been assembled using the most up-to-date, internationally-recognised data sources including the Atlas of Resonances, CRC, evaluated EXFOR, activation databases, fusion, fission and MACS. Excellent agreement was found with a small set of errors within the reference databases and TENDL-2014 predictions.

Generalized Nuclear Data: A New Structure (with Supporting Infrastructure) for Handling Nuclear Data

NASA Astrophysics Data System (ADS)

Mattoon, C. M.; Beck, B. R.; Patel, N. R.; Summers, N. C.; Hedstrom, G. W.; Brown, D. A.

2012-12-01

The Evaluated Nuclear Data File (ENDF) format was designed in the 1960s to accommodate neutron reaction data to support nuclear engineering applications in power, national security and criticality safety. Over the years, the scope of the format has been extended to handle many other kinds of data including charged particle, decay, atomic, photo-nuclear and thermal neutron scattering. Although ENDF has wide acceptance and support for many data types, its limited support for correlated particle emission, limited numeric precision, and general lack of extensibility mean that the nuclear data community cannot take advantage of many emerging opportunities. More generally, the ENDF format provides an unfriendly environment that makes it difficult for new data evaluators and users to create and access nuclear data. The Cross Section Evaluation Working Group (CSEWG) has begun the design of a new Generalized Nuclear Data (or 'GND') structure, meant to replace older formats with a hierarchy that mirrors the underlying physics, and is aligned with modern coding and database practices. In support of this new structure, Lawrence Livermore National Laboratory (LLNL) has updated its nuclear data/reactions management package Fudge to handle GND structured nuclear data. Fudge provides tools for converting both the latest ENDF format (ENDF-6) and the LLNL Evaluated Nuclear Data Library (ENDL) format to and from GND, as well as for visualizing, modifying and processing (i.e., converting evaluated nuclear data into a form more suitable to transport codes) GND structured nuclear data. GND defines the structure needed for storing nuclear data evaluations and the type of data that needs to be stored. But unlike ENDF and ENDL, GND does not define how the data are to be stored in a file. Currently, Fudge writes the structured GND data to a file using the eXtensible Markup Language (XML), as it is ASCII based and can be viewed with any text editor. XML is a meta-language, meaning that it has a primitive set of definitions for representing hierarchical data/text in a file. Other meta-languages, like HDF5 which stores the data in binary form, can also be used to store GND in a file. In this paper, we will present an overview of the new GND data structures along with associated tools in Fudge.

SPOC Benchmark Case: SNRE Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vishal Patel; Michael Eades; Claude Russel Joyner II

The Small Nuclear Rocket Engine (SNRE) was modeled in the Center for Space Nuclear Research’s (CSNR) Space Propulsion Optimization Code (SPOC). SPOC aims to create nuclear thermal propulsion (NTP) geometries quickly to perform parametric studies on design spaces of historic and new NTP designs. The SNRE geometry was modeled in SPOC and a critical core with a reasonable amount of criticality margin was found. The fuel, tie-tubes, reflector, and control drum masses were predicted rather well. These are all very important for neutronics calculations so the active reactor geometries created with SPOC can continue to be trusted. Thermal calculations ofmore » the average and hot fuel channels agreed very well. The specific impulse calculations used historically and in SPOC disagree so mass flow rates and impulses differed. Modeling peripheral and power balance components that do not affect nuclear characteristics of the core is not a feature of SPOC and as such, these components should continue to be designed using other tools. A full paper detailing the available SNRE data and comparisons with SPOC outputs will be submitted as a follow-up to this abstract.« less

Leadership Class Configuration Interaction Code - Status and Opportunities

NASA Astrophysics Data System (ADS)

Vary, James

2011-10-01

With support from SciDAC-UNEDF (www.unedf.org) nuclear theorists have developed and are continuously improving a Leadership Class Configuration Interaction Code (LCCI) for forefront nuclear structure calculations. The aim of this project is to make state-of-the-art nuclear structure tools available to the entire community of researchers including graduate students. The project includes codes such as NuShellX, MFDn and BIGSTICK that run a range of computers from laptops to leadership class supercomputers. Codes, scripts, test cases and documentation have been assembled, are under continuous development and are scheduled for release to the entire research community in November 2011. A covering script that accesses the appropriate code and supporting files is under development. In addition, a Data Base Management System (DBMS) that records key information from large production runs and archived results of those runs has been developed (http://nuclear.physics.iastate.edu/info/) and will be released. Following an outline of the project, the code structure, capabilities, the DBMS and current efforts, I will suggest a path forward that would benefit greatly from a significant partnership between researchers who use the codes, code developers and the National Nuclear Data efforts. This research is supported in part by DOE under grant DE-FG02-87ER40371 and grant DE-FC02-09ER41582 (SciDAC-UNEDF).

US nuclear engineering education: Status and prospects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1990-01-01

This study, conducted under the auspices of the Energy Engineering Board of the National Research Council, examines the status of and outlook for nuclear engineering education in the United States. The study resulted from a widely felt concern about the downward trends in student enrollments in nuclear engineering, in both graduate and undergraduate programs. Concerns have also been expressed about the declining number of US university nuclear engineering departments and programs, the aging of their faculties, the appropriateness of their curricula and research funding for industry and government needs, the availability of scholarships and research funding, and the increasing ratiomore » of foreign to US graduate students. A fundamental issue is whether the supply of nuclear engineering graduates will be adequate for the future. Although such issues are more general, pertaining to all areas of US science and engineering education, they are especially acute for nuclear engineering education. 30 refs., 12 figs., 20 tabs.« less

Influence of flowfield and vehicle parameters on engineering aerothermal methods

NASA Technical Reports Server (NTRS)

Wurster, Kathryn E.; Zoby, E. Vincent; Thompson, Richard A.

1989-01-01

The reliability and flexibility of three engineering codes used in the aerosphace industry (AEROHEAT, INCHES, and MINIVER) were investigated by comparing the results of these codes with Reentry F flight data and ground-test heat-transfer data for a range of cone angles, and with the predictions obtained using the detailed VSL3D code; the engineering solutions were also compared. In particular, the impact of several vehicle and flow-field parameters on the heat transfer and the capability of the engineering codes to predict these results were determined. It was found that entropy, pressure gradient, nose bluntness, gas chemistry, and angle of attack all affect heating levels. A comparison of the results of the three engineering codes with Reentry F flight data and with the predictions obtained of the VSL3D code showed a very good agreement in the regions of the applicability of the codes. It is emphasized that the parameters used in this study can significantly influence the actual heating levels and the prediction capability of a code.

The Design of PSB-VVER Experiments Relevant to Accident Management

NASA Astrophysics Data System (ADS)

Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

Monte Carlo calculations of positron emitter yields in proton radiotherapy.

PubMed

Seravalli, E; Robert, C; Bauer, J; Stichelbaut, F; Kurz, C; Smeets, J; Van Ngoc Ty, C; Schaart, D R; Buvat, I; Parodi, K; Verhaegen, F

2012-03-21

Positron emission tomography (PET) is a promising tool for monitoring the three-dimensional dose distribution in charged particle radiotherapy. PET imaging during or shortly after proton treatment is based on the detection of annihilation photons following the ß(+)-decay of radionuclides resulting from nuclear reactions in the irradiated tissue. Therapy monitoring is achieved by comparing the measured spatial distribution of irradiation-induced ß(+)-activity with the predicted distribution based on the treatment plan. The accuracy of the calculated distribution depends on the correctness of the computational models, implemented in the employed Monte Carlo (MC) codes that describe the interactions of the charged particle beam with matter and the production of radionuclides and secondary particles. However, no well-established theoretical models exist for predicting the nuclear interactions and so phenomenological models are typically used based on parameters derived from experimental data. Unfortunately, the experimental data presently available are insufficient to validate such phenomenological hadronic interaction models. Hence, a comparison among the models used by the different MC packages is desirable. In this work, starting from a common geometry, we compare the performances of MCNPX, GATE and PHITS MC codes in predicting the amount and spatial distribution of proton-induced activity, at therapeutic energies, to the already experimentally validated PET modelling based on the FLUKA MC code. In particular, we show how the amount of ß(+)-emitters produced in tissue-like media depends on the physics model and cross-sectional data used to describe the proton nuclear interactions, thus calling for future experimental campaigns aiming at supporting improvements of MC modelling for clinical application of PET monitoring. © 2012 Institute of Physics and Engineering in Medicine

Educational Innovation in the Design of an Online Nuclear Engineering Curriculum

ERIC Educational Resources Information Center

Hall, Simin; Jones, Brett D.; Amelink, Catherine; Hu, Deyu

2013-01-01

The purpose of this paper is to describe the development and implementation phases of online graduate nuclear engineering courses that are part of the Graduate Nuclear Engineering Certificate program at Virginia Tech. Virginia Tech restarted its nuclear engineering program in the Fall of 2007 with 60 students, and by 2009, the enrollment had grown…

Calculated concentrations of any radionuclide deposited on the ground by release from underground nuclear detonations, tests of nuclear rockets, and tests of nuclear ramjet engines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hicks, H.G.

1981-11-01

This report presents calculated gamma radiation exposure rates and ground deposition of related radionuclides resulting from three types of event that deposited detectable radioactivity outside the Nevada Test Site complex, namely, underground nuclear detonations, tests of nuclear rocket engines and tests of nuclear ramjet engines.

Comparison of GLIMPS and HFAST Stirling engine code predictions with experimental data

NASA Technical Reports Server (NTRS)

Geng, Steven M.; Tew, Roy C.

1992-01-01

Predictions from GLIMPS and HFAST design codes are compared with experimental data for the RE-1000 and SPRE free piston Stirling engines. Engine performance and available power loss predictions are compared. Differences exist between GLIMPS and HFAST loss predictions. Both codes require engine specific calibration to bring predictions and experimental data into agreement.

The RNA Exosome Adaptor ZFC3H1 Functionally Competes with Nuclear Export Activity to Retain Target Transcripts.

PubMed

Silla, Toomas; Karadoulama, Evdoxia; Mąkosa, Dawid; Lubas, Michal; Jensen, Torben Heick

2018-05-15

Mammalian genomes are promiscuously transcribed, yielding protein-coding and non-coding products. Many transcripts are short lived due to their nuclear degradation by the ribonucleolytic RNA exosome. Here, we show that abolished nuclear exosome function causes the formation of distinct nuclear foci, containing polyadenylated (pA + ) RNA secluded from nucleocytoplasmic export. We asked whether exosome co-factors could serve such nuclear retention. Co-localization studies revealed the enrichment of pA + RNA foci with "pA-tail exosome targeting (PAXT) connection" components MTR4, ZFC3H1, and PABPN1 but no overlap with known nuclear structures such as Cajal bodies, speckles, paraspeckles, or nucleoli. Interestingly, ZFC3H1 is required for foci formation, and in its absence, selected pA + RNAs, including coding and non-coding transcripts, are exported to the cytoplasm in a process dependent on the mRNA export factor AlyREF. Our results establish ZFC3H1 as a central nuclear pA + RNA retention factor, counteracting nuclear export activity. Copyright © 2018 The Author(s). Published by Elsevier Inc. All rights reserved.

BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) - Generation Methodology and Preliminary Testing of two ENEA-Bologna Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

NASA Astrophysics Data System (ADS)

Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela

2016-02-01

Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.

HZETRN: Description of a free-space ion and nucleon transport and shielding computer program

NASA Technical Reports Server (NTRS)

Wilson, John W.; Badavi, Francis F.; Cucinotta, Francis A.; Shinn, Judy L.; Badhwar, Gautam D.; Silberberg, R.; Tsao, C. H.; Townsend, Lawrence W.; Tripathi, Ram K.

1995-01-01

The high-charge-and energy (HZE) transport computer program HZETRN is developed to address the problems of free-space radiation transport and shielding. The HZETRN program is intended specifically for the design engineer who is interested in obtaining fast and accurate dosimetric information for the design and construction of space modules and devices. The program is based on a one-dimensional space-marching formulation of the Boltzmann transport equation with a straight-ahead approximation. The effect of the long-range Coulomb force and electron interaction is treated as a continuous slowing-down process. Atomic (electronic) stopping power coefficients with energies above a few A MeV are calculated by using Bethe's theory including Bragg's rule, Ziegler's shell corrections, and effective charge. Nuclear absorption cross sections are obtained from fits to quantum calculations and total cross sections are obtained with a Ramsauer formalism. Nuclear fragmentation cross sections are calculated with a semiempirical abrasion-ablation fragmentation model. The relation of the final computer code to the Boltzmann equation is discussed in the context of simplifying assumptions. A detailed description of the flow of the computer code, input requirements, sample output, and compatibility requirements for non-VAX platforms are provided.

Current status of nuclear engineering education

DOE Office of Scientific and Technical Information (OSTI.GOV)

Palladino, N.J.

1975-09-01

The 65 colleges and universities offering undergraduate degrees in nuclear engineering and the 15 schools offering strong nuclear engineering options are, in general, doing a good job to meet the current spectrum of job opportunities. But, nuclear engineering programs are not producing enough graduates to meet growing demands. They currently receive little aid and support from their customers --industry and government--in the form of scholarships, grants, faculty research support, student thesis and project support, or student summer jobs. There is not enough interaction between industry and universities. Most nuclear engineering programs are geared too closely to the technology of themore » present family of reactors and too little to the future breeder reactors and controlled thermonuclear reactors. In addition, nuclear engineering programs attract too few women and members of minority ethnic groups. Further study of the reasons for this fact is needed so that effective corrective action can be taken. Faculty in nuclear engineering programs should assume greater initiative to provide attractive and objective nuclear energy electives for technical and nontechnical students in other disciplines to improve their technical understanding of the safety and environmental issues involved. More aggressive and persistent efforts must be made by nuclear engineering schools to obtain industry support and involvement in their programs. (auth)« less

Job Prospects for Nuclear Engineers.

ERIC Educational Resources Information Center

Basta, Nicholas

1987-01-01

Discusses trends in job opportunities for nuclear engineers. Lists some of the factors influencing increases and decreases in the demand for nuclear engineers. Describes the effects on career opportunities from recent nuclear accidents, military research and development, and projected increases of demand for electricity. (TW)

Review of current nuclear fallout codes.

PubMed

Auxier, Jerrad P; Auxier, John D; Hall, Howard L

2017-05-01

The importance of developing a robust nuclear forensics program to combat the illicit use of nuclear material that may be used as an improvised nuclear device is widely accepted. In order to decrease the threat to public safety and improve governmental response, government agencies have developed fallout-analysis codes to predict the fallout particle size, dose, and dispersion and dispersion following a detonation. This paper will review the different codes that have been developed for predicting fallout from both chemical and nuclear weapons. This will decrease the response time required for the government to respond to the event. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Neil R.; Serrano De Caro, Magdalena; Rodriguez, Edward A.

2012-08-28

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to itsmore » superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.« less

Comparison of deterministic and stochastic approaches for isotopic concentration and decay heat uncertainty quantification on elementary fission pulse

NASA Astrophysics Data System (ADS)

Lahaye, S.; Huynh, T. D.; Tsilanizara, A.

2016-03-01

Uncertainty quantification of interest outputs in nuclear fuel cycle is an important issue for nuclear safety, from nuclear facilities to long term deposits. Most of those outputs are functions of the isotopic vector density which is estimated by fuel cycle codes, such as DARWIN/PEPIN2, MENDEL, ORIGEN or FISPACT. CEA code systems DARWIN/PEPIN2 and MENDEL propagate by two different methods the uncertainty from nuclear data inputs to isotopic concentrations and decay heat. This paper shows comparisons between those two codes on a Uranium-235 thermal fission pulse. Effects of nuclear data evaluation's choice (ENDF/B-VII.1, JEFF-3.1.1 and JENDL-2011) is inspected in this paper. All results show good agreement between both codes and methods, ensuring the reliability of both approaches for a given evaluation.

CESAR: A Code for Nuclear Fuel and Waste Characterisation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vidal, J.M.; Grouiller, J.P.; Launay, A.

2006-07-01

CESAR (Simplified Evolution Code Applied to Reprocessing) is a depletion code developed through a joint program between CEA and COGEMA. In the late 1980's, the first use of this code dealt with nuclear measurement at the Laboratories of the La Hague reprocessing plant. The use of CESAR was then extended to characterizations of all entrance materials and for characterisation, via tracer, of all produced waste. The code can distinguish more than 100 heavy nuclides, 200 fission products and 100 activation products, and it can characterise both the fuel and the structural material of the fuel. CESAR can also make depletionmore » calculations from 3 months to 1 million years of cooling time. Between 2003-2005, the 5. version of the code was developed. The modifications were related to the harmonisation of the code's nuclear data with the JEF2.2 nuclear data file. This paper describes the code and explains the extensive use of this code at the La Hague reprocessing plant and also for prospective studies. The second part focuses on the modifications of the latest version, and describes the application field and the qualification of the code. Many companies and the IAEA use CESAR today. CESAR offers a Graphical User Interface, which is very user-friendly. (authors)« less

US Nuclear Engineering Education: Status and prospects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1990-01-01

This study, conducted under the auspices of the Energy Engineering Board of the National Research Council, examines the status of and outlook for nuclear engineering education in the United States. The study, as described in this report resulted from a widely felt concern about the downward trends in student enrollments in nuclear engineering, in both graduate and undergraduate programs. Concerns have also been expressed about the declining number of US university nuclear engineering departments and programs, the ageing of their faculties, the appropriateness of their curricula and research funding for industry and government needs, the availability of scholarships and researchmore » funding, and the increasing ratio of foreign to US graduate students. A fundamental issue is whether the supply of nuclear engineering graduates will be adequate for the future. Although such issues are more general, pertaining to all areas of US science and engineering education, they are especially acute for nuclear engineering education. 30 refs., 24 figs., 49 tabs.« less

Calibration and comparison of the NASA Lewis free-piston Stirling engine model predictions with RE-1000 test data

NASA Technical Reports Server (NTRS)

Geng, Steven M.

1987-01-01

A free-piston Stirling engine performance code is being upgraded and validated at the NASA Lewis Research Center under an interagency agreement between the Department of Energy's Oak Ridge National Laboratory and NASA Lewis. Many modifications were made to the free-piston code in an attempt to decrease the calibration effort. A procedure was developed that made the code calibration process more systematic. Engine-specific calibration parameters are often used to bring predictions and experimental data into better agreement. The code was calibrated to a matrix of six experimental data points. Predictions of the calibrated free-piston code are compared with RE-1000 free-piston Stirling engine sensitivity test data taken at NASA Lewis. Reasonable agreement was obtained between the code prediction and the experimental data over a wide range of engine operating conditions.

Calibration and comparison of the NASA Lewis free-piston Stirling engine model predictions with RE-1000 test data

NASA Technical Reports Server (NTRS)

Geng, Steven M.

1987-01-01

A free-piston Stirling engine performance code is being upgraded and validated at the NASA Lewis Research Center under an interagency agreement between the Department of Energy's Oak Ridge National Laboratory and NASA Lewis. Many modifications were made to the free-piston code in an attempt to decrease the calibration effort. A procedure was developed that made the code calibration process more systematic. Engine-specific calibration parameters are often used to bring predictions and experimental data into better agreement. The code was calibrated to a matrix of six experimental data points. Predictions of the calibrated free-piston code are compared with RE-1000 free-piston Stirling engine sensitivity test data taken at NASA Lewis. Resonable agreement was obtained between the code predictions and the experimental data over a wide range of engine operating conditions.

Manpower Assessment Brief #44: NUCLEAR ENGINEERING Enrollments Decreased at All Levels in 1998. Undergraduate and Doctoral Degrees Decreased, While Master's Degrees Increased Slightly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shirley, Duveen

1999-05-04

The survey of "Nuclear Engineering Enrollments and Degrees, 1998" was sent to 45 institutions offering a major in nuclear engineering or an option program in another discipline or department (for example, electrical or mechanical engineering) equivalent to a major that qualifies the graduates to perform as nuclear engineers. This document provides statistical data on undergraduate and graduate enrollments and degrees, employment and post-graduation plans, and foreign national participation.

ARC-2001-ACD01-0018

NASA Image and Video Library

2001-02-16

New Center Network Deployment ribbon Cutting: from left to right: Maryland Edwards, Code JT upgrade project deputy task manager; Ed Murphy, foundry networks systems engineer; Bohdan Cmaylo, Code JT upgrade project task manager, Scott Santiago, Division Chief, Code JT; Greg Miller, Raytheon Network engineer and Frank Daras, Raytheon network engineering manager.

Abstracts for student symposium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldman, B.

Lawrence Livermore National Laboratory Science and Engineering Research Semester (SERS) students are participants in a national program sponsored by the DOE Office of Energy Research. Presented topics from Fall 1993 include: Laser glass, wiring codes, lead in food and food containers, chromium removal from ground water, fiber optic sensors for ph measurement, CFC replacement, predator/prey simulation, detection of micronuclei in germ cells, DNA conformation, stimulated brillouin scattering, DNA sequencing, evaluation of education programs, neural network analysis of nuclear glass, lithium ion batteries, Indonesian snails, optical switching systems, and photoreceiver design. Individual papers are indexed separately on the Energy Data Base.

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.

2011-03-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repositorymore » designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.« less

Progress of teaching and learning of nuclear engineering courses at College of Engineering, Universiti Tenaga Nasional (UNITEN)

NASA Astrophysics Data System (ADS)

Hamid, Nasri A.; Mohamed, Abdul Aziz; Yusoff, Mohd. Zamri

2015-04-01

Developing human capital in nuclear with required nuclear background and professional qualifications is necessary to support the implementation of nuclear power projects in the near future. Sufficient educational and training skills are required to ensure that the human resources needed by the nuclear power industry meets its high standard. The Government of Malaysia has made the decision to include nuclear as one of the electricity generation option for the country, post 2020 in order to cater for the increasing energy demands of the country as well as to reduce CO2 emission. The commitment by the government has been made clearer with the inclusion of the development of first NPP by 2021 in the Economic Transformation Program (ETP) which was launched by the government in October 2010. The In tandem with the government initiative to promote nuclear energy, Center for Nuclear Energy, College of Engineering, Universiti Tenaga Nasional (UNITEN) is taking the responsibility in developing human capital in the area of nuclear power and technology. In the beginning, the College of Engineering has offered the Introduction to Nuclear Technology course as a technical elective course for all undergraduate engineering students. Gradually, other nuclear technical elective courses are offered such as Nuclear Policy, Security and Safeguards, Introduction to Nuclear Engineering, Radiation Detection and Nuclear Instrumentation, Introduction to Reactor Physics, Radiation Safety and Waste Management, and Nuclear Thermal-hydraulics. In addition, another course Advancement in Nuclear Energy is offered as one of the postgraduate elective courses. To enhance the capability of teaching staffs in nuclear areas at UNITEN, several junior lecturers are sent to pursue their postgraduate studies in the Republic of Korea, United States and the United Kingdom, while the others are participating in short courses and workshops in nuclear that are conducted locally and abroad. This paper describes the progress of teaching and learning in nuclear engineering and technology at UNITEN that include curriculum development, students' enrolment and performance, and teaching staff's human resource development.

A case study for retaining nuclear power experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beckjord, E.S.

1996-12-31

Nuclear engineering departments at U.S. universities are rethinking curricula to focus on essentials. Prospective engineers must know nuclear engineering disciplines, but knowing how their engineering forebears solved important problems will empower them even more by learning some history along with engineering. I suggest a way to retain experience, giving an example: the emergency core cooling system (ECCS) controversy and resolution.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Fabrication development for high-level nuclear waste containers for the tuff repository; Phase 1 final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Domian, H.A.; Holbrook, R.L.; LaCount, D.F.

1990-09-01

This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of variousmore » processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs.« less

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

Internationalizing professional codes in engineering.

PubMed

Harris, Charles E

2004-07-01

Professional engineering societies which are based in the United States, such as the American Society of Mechanical Engineers (ASME, now ASME International) are recognizing that their codes of ethics must apply to engineers working throughout the world. An examination of the ethical code of the ASME International shows that its provisions pose many problems of application, especially in societies outside the United States. In applying the codes effectively in the international environment, two principal issues must be addressed. First, some Culture Transcending Guidelines must be identified and justified. Nine such guidelines are identified Second, some methods for applying the codes to particular situations must be identified Three such methods are specification, balancing, and finding a creative middle way.

Decoding the function of nuclear long non-coding RNAs.

PubMed

Chen, Ling-Ling; Carmichael, Gordon G

2010-06-01

Long non-coding RNAs (lncRNAs) are mRNA-like, non-protein-coding RNAs that are pervasively transcribed throughout eukaryotic genomes. Rather than silently accumulating in the nucleus, many of these are now known or suspected to play important roles in nuclear architecture or in the regulation of gene expression. In this review, we highlight some recent progress in how lncRNAs regulate these important nuclear processes at the molecular level. Copyright 2010 Elsevier Ltd. All rights reserved.

DataRocket: Interactive Visualisation of Data Structures

NASA Astrophysics Data System (ADS)

Parkes, Steve; Ramsay, Craig

2010-08-01

CodeRocket is a software engineering tool that provides cognitive support to the software engineer for reasoning about a method or procedure and for documenting the resulting code [1]. DataRocket is a software engineering tool designed to support visualisation and reasoning about program data structures. DataRocket is part of the CodeRocket family of software tools developed by Rapid Quality Systems [2] a spin-out company from the Space Technology Centre at the University of Dundee. CodeRocket and DataRocket integrate seamlessly with existing architectural design and coding tools and provide extensive documentation with little or no effort on behalf of the software engineer. Comprehensive, abstract, detailed design documentation is available early on in a project so that it can be used for design reviews with project managers and non expert stakeholders. Code and documentation remain fully synchronised even when changes are implemented in the code without reference to the existing documentation. At the end of a project the press of a button suffices to produce the detailed design document. Existing legacy code can be easily imported into CodeRocket and DataRocket to reverse engineer detailed design documentation making legacy code more manageable and adding substantially to its value. This paper introduces CodeRocket. It then explains the rationale for DataRocket and describes the key features of this new tool. Finally the major benefits of DataRocket for different stakeholders are considered.

Finite element analyses for seismic shear wall international standard problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, Y.J.; Hofmayer, C.H.

Two identical reinforced concrete (RC) shear walls, which consist of web, flanges and massive top and bottom slabs, were tested up to ultimate failure under earthquake motions at the Nuclear Power Engineering Corporation`s (NUPEC) Tadotsu Engineering Laboratory, Japan. NUPEC provided the dynamic test results to the OECD (Organization for Economic Cooperation and Development), Nuclear Energy Agency (NEA) for use as an International Standard Problem (ISP). The shear walls were intended to be part of a typical reactor building. One of the major objectives of the Seismic Shear Wall ISP (SSWISP) was to evaluate various seismic analysis methods for concrete structuresmore » used for design and seismic margin assessment. It also offered a unique opportunity to assess the state-of-the-art in nonlinear dynamic analysis of reinforced concrete shear wall structures under severe earthquake loadings. As a participant of the SSWISP workshops, Brookhaven National Laboratory (BNL) performed finite element analyses under the sponsorship of the U.S. Nuclear Regulatory Commission (USNRC). Three types of analysis were performed, i.e., monotonic static (push-over), cyclic static and dynamic analyses. Additional monotonic static analyses were performed by two consultants, F. Vecchio of the University of Toronto (UT) and F. Filippou of the University of California at Berkeley (UCB). The analysis results by BNL and the consultants were presented during the second workshop in Yokohama, Japan in 1996. A total of 55 analyses were presented during the workshop by 30 participants from 11 different countries. The major findings on the presented analysis methods, as well as engineering insights regarding the applicability and reliability of the FEM codes are described in detail in this report. 16 refs., 60 figs., 16 tabs.« less

Engineering Codes of Ethics and the Duty to Set a Moral Precedent.

PubMed

Schlossberger, Eugene

2016-10-01

Each of the major engineering societies has its own code of ethics. Seven "common core" clauses and several code-specific clauses can be identified. The paper articulates objections to and rationales for two clauses that raise controversy: do engineers have a duty (a) to provide pro bono services and/or speak out on major issues, and (b) to associate only with reputable individuals and organizations? This latter "association clause" can be justified by the "proclamative principle," an alternative to Kant's universalizability requirement. At the heart of engineering codes of ethics, and implicit in what it is to be a moral agent, the "proclamative principle" asserts that one's life should proclaim one's moral stances (one's values, principles, perceptions, etc.). More specifically, it directs engineers to strive to insure that their actions, thoughts, and relationships be fit to offer to their communities as part of the body of moral precedents for how to be an engineer. Understanding codes of ethics as reflections of this principle casts light both on how to apply the codes and on the distinction between private and professional morality.

Interactive Finite Elements for General Engine Dynamics Analysis

NASA Technical Reports Server (NTRS)

Adams, M. L.; Padovan, J.; Fertis, D. G.

1984-01-01

General nonlinear finite element codes were adapted for the purpose of analyzing the dynamics of gas turbine engines. In particular, this adaptation required the development of a squeeze-film damper element software package and its implantation into a representative current generation code. The ADINA code was selected because of prior use of it and familiarity with its internal structure and logic. This objective was met and the results indicate that such use of general purpose codes is viable alternative to specialized codes for general dynamics analysis of engines.

Visual Computing Environment Workshop

NASA Technical Reports Server (NTRS)

Lawrence, Charles (Compiler)

1998-01-01

The Visual Computing Environment (VCE) is a framework for intercomponent and multidisciplinary computational simulations. Many current engineering analysis codes simulate various aspects of aircraft engine operation. For example, existing computational fluid dynamics (CFD) codes can model the airflow through individual engine components such as the inlet, compressor, combustor, turbine, or nozzle. Currently, these codes are run in isolation, making intercomponent and complete system simulations very difficult to perform. In addition, management and utilization of these engineering codes for coupled component simulations is a complex, laborious task, requiring substantial experience and effort. To facilitate multicomponent aircraft engine analysis, the CFD Research Corporation (CFDRC) is developing the VCE system. This system, which is part of NASA's Numerical Propulsion Simulation System (NPSS) program, can couple various engineering disciplines, such as CFD, structural analysis, and thermal analysis.

Utilization of recently developed codes for high power Brayton and Rankine cycle power systems

NASA Technical Reports Server (NTRS)

Doherty, Michael P.

1993-01-01

Two recently developed FORTRAN computer codes for high power Brayton and Rankine thermodynamic cycle analysis for space power applications are presented. The codes were written in support of an effort to develop a series of subsystem models for multimegawatt Nuclear Electric Propulsion, but their use is not limited just to nuclear heat sources or to electric propulsion. Code development background, a description of the codes, some sample input/output from one of the codes, and state future plans/implications for the use of these codes by NASA's Lewis Research Center are provided.

The Effect of Cold Work on Properties of Alloy 617

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Richard

2014-08-01

Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was submitted in 1992 to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed.1 Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. The mechanical and physical properties of Alloy 617 were extensively characterized for the VHTR programs in the 1980’s andmore » incorporated into the 1992 draft Code Case. Recently, the properties of modern heats of the alloy that incorporate an additional processing step, electro-slag re-melting, have been characterized both to confirm that the properties of contemporary material are consistent with those in the historical record and to increase the available database. A number of potential issues that were identified as requiring further consideration prior to the withdrawal of the 1992 Code Case are also being re-examined in the current R&D program. Code Cases are again being developed to allow use of Alloy 617 for nuclear design within the rules of the ASME Boiler and Pressure Vessel Code. In general the Code defines two temperature ranges for nuclear design with austenitic and nickel based alloys. Below 427°C (800°F) time dependent behavior is not considered, while above this temperature creep and creep-fatigue are considered to be the dominant life-limiting deformation modes. There is a corresponding differentiation in the treatment of the potential for effects associated with cold work. Below 427°C the principal issue is the relationship between the level of cold work and the propensity for stress corrosion cracking and above that temperature the primary concern is the impact of cold work on creep-rupture behavior.« less

Implanting a Discipline: The Academic Trajectory of Nuclear Engineering in the USA and UK

ERIC Educational Resources Information Center

Johnston, Sean F.

2009-01-01

The nuclear engineer emerged as a new form of recognised technical professional between 1940 and the early 1960s as nuclear fission, the chain reaction and their applications were explored. The institutionalization of nuclear engineering--channelled into new national laboratories and corporate design offices during the decade after the war, and…

Nuclear shell model code CRUNCHER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Resler, D.A.; Grimes, S.M.

1988-05-01

A new nuclear shell model code CRUNCHER, patterned after the code VLADIMIR, has been developed. While CRUNCHER and VLADIMIR employ the techniques of an uncoupled basis and the Lanczos process, improvements in the new code allow it to handle much larger problems than the previous code and to perform them more efficiently. Tests involving a moderately sized calculation indicate that CRUNCHER running on a SUN 3/260 workstation requires approximately one-half the central processing unit (CPU) time required by VLADIMIR running on a CRAY-1 supercomputer.

Development of undergraduate nuclear security curriculum at College of Engineering, Universiti Tenaga Nasional

NASA Astrophysics Data System (ADS)

Hamid, Nasri A.; Mujaini, Madihah; Mohamed, Abdul Aziz

2017-01-01

The Center for Nuclear Energy (CNE), College of Engineering, Universiti Tenaga Nasional (UNITEN) has a great responsibility to undertake educational activities that promote developing human capital in the area of nuclear engineering and technology. Developing human capital in nuclear through education programs is necessary to support the implementation of nuclear power projects in Malaysia in the near future. In addition, the educational program must also meet the nuclear power industry needs and requirements. In developing a certain curriculum, the contents must comply with the university's Outcomes Based Education (OBE) philosophy. One of the important courses in the nuclear curriculum is in the area of nuclear security. Basically the nuclear security course covers the current issues of law, politics, military strategy, and technology with regard to weapons of mass destruction and related topics in international security, and review legal regulations and political relationship that determine the state of nuclear security at the moment. In addition, the course looks into all aspects of the nuclear safeguards, builds basic knowledge and understanding of nuclear non-proliferation, nuclear forensics and nuclear safeguards in general. The course also discusses tools used to combat nuclear proliferation such as treaties, institutions, multilateral arrangements and technology controls. In this paper, we elaborate the development of undergraduate nuclear security course at the College of Engineering, Universiti Tenaga Nasional. Since the course is categorized as mechanical engineering subject, it must be developed in tandem with the program educational objectives (PEO) of the Bachelor of Mechanical Engineering program. The course outcomes (CO) and transferrable skills are also identified. Furthermore, in aligning the CO with program outcomes (PO), the PO elements need to be emphasized through the CO-PO mapping. As such, all assessments and distribution of Bloom Taxonomy levels are assigned in accordance with the CO-PO mapping. Finally, the course has to fulfill the International Engineering Alliance (IEA) Graduate Attributes of the Washington Accord.

10 CFR Appendix S to Part 50 - Earthquake Engineering Criteria for Nuclear Power Plants

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Earthquake Engineering Criteria for Nuclear Power Plants S... FACILITIES Pt. 50, App. S Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants... applicant or holder whose construction permit was issued before January 10, 1997, the earthquake engineering...

10 CFR Appendix S to Part 50 - Earthquake Engineering Criteria for Nuclear Power Plants

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Earthquake Engineering Criteria for Nuclear Power Plants S... FACILITIES Pt. 50, App. S Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants... applicant or holder whose construction permit was issued before January 10, 1997, the earthquake engineering...

10 CFR Appendix S to Part 50 - Earthquake Engineering Criteria for Nuclear Power Plants

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Earthquake Engineering Criteria for Nuclear Power Plants S... FACILITIES Pt. 50, App. S Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants... applicant or holder whose construction permit was issued before January 10, 1997, the earthquake engineering...

10 CFR Appendix S to Part 50 - Earthquake Engineering Criteria for Nuclear Power Plants

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Earthquake Engineering Criteria for Nuclear Power Plants S... FACILITIES Pt. 50, App. S Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants... applicant or holder whose construction permit was issued before January 10, 1997, the earthquake engineering...

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.

2016-09-20

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such asmore » MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

Simulation Enabled Safeguards Assessment Methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robert Bean; Trond Bjornard; Thomas Larson

2007-09-01

It is expected that nuclear energy will be a significant component of future supplies. New facilities, operating under a strengthened international nonproliferation regime will be needed. There is good reason to believe virtual engineering applied to the facility design, as well as to the safeguards system design will reduce total project cost and improve efficiency in the design cycle. Simulation Enabled Safeguards Assessment MEthodology (SESAME) has been developed as a software package to provide this capability for nuclear reprocessing facilities. The software architecture is specifically designed for distributed computing, collaborative design efforts, and modular construction to allow step improvements inmore » functionality. Drag and drop wireframe construction allows the user to select the desired components from a component warehouse, render the system for 3D visualization, and, linked to a set of physics libraries and/or computational codes, conduct process evaluations of the system they have designed.« less

Generation of a variety of stable Influenza A reporter viruses by genetic engineering of the NS gene segment

PubMed Central

Reuther, Peter; Göpfert, Kristina; Dudek, Alexandra H.; Heiner, Monika; Herold, Susanne; Schwemmle, Martin

2015-01-01

Influenza A viruses (IAV) pose a constant threat to the human population and therefore a better understanding of their fundamental biology and identification of novel therapeutics is of upmost importance. Various reporter-encoding IAV were generated to achieve these goals, however, one recurring difficulty was the genetic instability especially of larger reporter genes. We employed the viral NS segment coding for the non-structural protein 1 (NS1) and nuclear export protein (NEP) for stable expression of diverse reporter proteins. This was achieved by converting the NS segment into a single open reading frame (ORF) coding for NS1, the respective reporter and NEP. To allow expression of individual proteins, the reporter genes were flanked by two porcine Teschovirus-1 2A peptide (PTV-1 2A)-coding sequences. The resulting viruses encoding luciferases, fluorescent proteins or a Cre recombinase are characterized by a high genetic stability in vitro and in mice and can be readily employed for antiviral compound screenings, visualization of infected cells or cells that survived acute infection. PMID:26068081

A Working Model for the System Alumina-Magnesia.

DTIC Science & Technology

1983-05-01

Several regions in the resulting diagram appear rather uncertain: the liquidus ’National bureau of StandaTds. JANAF Thermochemical Tables, by D. R. Stull ...Code 131) 1 Naval Ordnance Station, Indian Head (Technical Library) 29 Naval Postgraduate School. Monterey Code 012, Dean of Research (1) Code 06... Dean of Science and Engineering (1) Code 1424. Library - Technical Reports (2) Code 33. Weapons Engineering Program Office (1) Code 61. Chairman

Application of CFD codes to the design and development of propulsion systems

NASA Technical Reports Server (NTRS)

Lord, W. K.; Pickett, G. F.; Sturgess, G. J.; Weingold, H. D.

1987-01-01

The internal flows of aerospace propulsion engines have certain common features that are amenable to analysis through Computational Fluid Dynamics (CFD) computer codes. Although the application of CFD to engineering problems in engines was delayed by the complexities associated with internal flows, many codes with different capabilities are now being used as routine design tools. This is illustrated by examples taken from the aircraft gas turbine engine of flows calculated with potential flow, Euler flow, parabolized Navier-Stokes, and Navier-Stokes codes. Likely future directions of CFD applied to engine flows are described, and current barriers to continued progress are highlighted. The potential importance of the Numerical Aerodynamic Simulator (NAS) to resolution of these difficulties is suggested.

Progress of teaching and learning of nuclear engineering courses at College of Engineering, Universiti Tenaga Nasional (UNITEN)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hamid, Nasri A., E-mail: Nasri@uniten.edu.my; Mohamed, Abdul Aziz; Yusoff, Mohd. Zamri

Developing human capital in nuclear with required nuclear background and professional qualifications is necessary to support the implementation of nuclear power projects in the near future. Sufficient educational and training skills are required to ensure that the human resources needed by the nuclear power industry meets its high standard. The Government of Malaysia has made the decision to include nuclear as one of the electricity generation option for the country, post 2020 in order to cater for the increasing energy demands of the country as well as to reduce CO{sub 2} emission. The commitment by the government has been mademore » clearer with the inclusion of the development of first NPP by 2021 in the Economic Transformation Program (ETP) which was launched by the government in October 2010. The In tandem with the government initiative to promote nuclear energy, Center for Nuclear Energy, College of Engineering, Universiti Tenaga Nasional (UNITEN) is taking the responsibility in developing human capital in the area of nuclear power and technology. In the beginning, the College of Engineering has offered the Introduction to Nuclear Technology course as a technical elective course for all undergraduate engineering students. Gradually, other nuclear technical elective courses are offered such as Nuclear Policy, Security and Safeguards, Introduction to Nuclear Engineering, Radiation Detection and Nuclear Instrumentation, Introduction to Reactor Physics, Radiation Safety and Waste Management, and Nuclear Thermal-hydraulics. In addition, another course Advancement in Nuclear Energy is offered as one of the postgraduate elective courses. To enhance the capability of teaching staffs in nuclear areas at UNITEN, several junior lecturers are sent to pursue their postgraduate studies in the Republic of Korea, United States and the United Kingdom, while the others are participating in short courses and workshops in nuclear that are conducted locally and abroad. This paper describes the progress of teaching and learning in nuclear engineering and technology at UNITEN that include curriculum development, students’ enrolment and performance, and teaching staff’s human resource development.« less

Nuclear Engineering Technologists in the Nuclear Power Era

ERIC Educational Resources Information Center

Wang, C. H.; And Others

1974-01-01

Describes manpower needs in nuclear engineering in the areas of research and development, architectural engineering and construction supervision, power reactor operations, and regulatory tasks. Outlines a suitable curriculum to prepare students for the tasks related to construction and operation of power reactors. (GS)

Alignment-based and alignment-free methods converge with experimental data on amino acids coded by stop codons at split between nuclear and mitochondrial genetic codes.

PubMed

Seligmann, Hervé

2018-05-01

Genetic codes mainly evolve by reassigning punctuation codons, starts and stops. Previous analyses assuming that undefined amino acids translate stops showed greater divergence between nuclear and mitochondrial genetic codes. Here, three independent methods converge on which amino acids translated stops at split between nuclear and mitochondrial genetic codes: (a) alignment-free genetic code comparisons inserting different amino acids at stops; (b) alignment-based blast analyses of hypothetical peptides translated from non-coding mitochondrial sequences, inserting different amino acids at stops; (c) biases in amino acid insertions at stops in proteomic data. Hence short-term protein evolution models reconstruct long-term genetic code evolution. Mitochondria reassign stops to amino acids otherwise inserted at stops by codon-anticodon mismatches (near-cognate tRNAs). Hence dual function (translation termination and translation by codon-anticodon mismatch) precedes mitochondrial reassignments of stops to amino acids. Stop ambiguity increases coded information, compensates endocellular mitogenome reduction. Mitochondrial codon reassignments might prevent viral infections. Copyright © 2018 Elsevier B.V. All rights reserved.

Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

NASA Astrophysics Data System (ADS)

Bang, Youngsuk

Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.

Evaluation of the efficiency and fault density of software generated by code generators

NASA Technical Reports Server (NTRS)

Schreur, Barbara

1993-01-01

Flight computers and flight software are used for GN&C (guidance, navigation, and control), engine controllers, and avionics during missions. The software development requires the generation of a considerable amount of code. The engineers who generate the code make mistakes and the generation of a large body of code with high reliability requires considerable time. Computer-aided software engineering (CASE) tools are available which generates code automatically with inputs through graphical interfaces. These tools are referred to as code generators. In theory, code generators could write highly reliable code quickly and inexpensively. The various code generators offer different levels of reliability checking. Some check only the finished product while some allow checking of individual modules and combined sets of modules as well. Considering NASA's requirement for reliability, an in house manually generated code is needed. Furthermore, automatically generated code is reputed to be as efficient as the best manually generated code when executed. In house verification is warranted.

Physics and nuclear power

NASA Astrophysics Data System (ADS)

Buttery, N. E.

2008-03-01

Nuclear power owes its origin to physicists. Fission was demonstrated by physicists and chemists and the first nuclear reactor project was led by physicists. However as nuclear power was harnessed to produce electricity the role of the engineer became stronger. Modern nuclear power reactors bring together the skills of physicists, chemists, chemical engineers, electrical engineers, mechanical engineers and civil engineers. The paper illustrates this by considering the Sizewell B project and the role played by physicists in this. This covers not only the roles in design and analysis but in problem solving during the commissioning of first of a kind plant. Looking forward to the challenges to provide sustainable and environmentally acceptable energy sources for the future illustrates the need for a continuing synergy between physics and engineering. This will be discussed in the context of the challenges posed by Generation IV reactors.

Integration of Dakota into the NEAMS Workbench

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swiler, Laura Painton; Lefebvre, Robert A.; Langley, Brandon R.

2017-07-01

This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on integrating Dakota into the NEAMS Workbench. The NEAMS Workbench, developed at Oak Ridge National Laboratory, is a new software framework that provides a graphical user interface, input file creation, parsing, validation, job execution, workflow management, and output processing for a variety of nuclear codes. Dakota is a tool developed at Sandia National Laboratories that provides a suite of uncertainty quantification and optimization algorithms. Providing Dakota within the NEAMS Workbench allows users of nuclear simulation codes to perform uncertainty and optimization studies on their nuclear codes frommore » within a common, integrated environment. Details of the integration and parsing are provided, along with an example of Dakota running a sampling study on the fuels performance code, BISON, from within the NEAMS Workbench.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryu, Jun-hyung

University education aims to supply qualified human resources for industries. In complex large scale engineering systems such as nuclear power plants, the importance of qualified human resources cannot be underestimated. The corresponding education program should involve many topics systematically. Recently a nuclear engineering program has been initiated in Dongguk University, South Korea. The current education program focuses on undergraduate level nuclear engineering students. Our main objective is to provide industries fresh engineers with the understanding on the interconnection of local parts and the entire systems of nuclear power plants and the associated systems. From the experience there is a hugemore » opportunity for chemical engineering disciple in the context of giving macroscopic overview on nuclear power plant and waste treatment management by strengthening the analyzing capability of fundamental situations. (authors)« less

Cascade Optimization for Aircraft Engines With Regression and Neural Network Analysis - Approximators

NASA Technical Reports Server (NTRS)

Patnaik, Surya N.; Guptill, James D.; Hopkins, Dale A.; Lavelle, Thomas M.

2000-01-01

The NASA Engine Performance Program (NEPP) can configure and analyze almost any type of gas turbine engine that can be generated through the interconnection of a set of standard physical components. In addition, the code can optimize engine performance by changing adjustable variables under a set of constraints. However, for engine cycle problems at certain operating points, the NEPP code can encounter difficulties: nonconvergence in the currently implemented Powell's optimization algorithm and deficiencies in the Newton-Raphson solver during engine balancing. A project was undertaken to correct these deficiencies. Nonconvergence was avoided through a cascade optimization strategy, and deficiencies associated with engine balancing were eliminated through neural network and linear regression methods. An approximation-interspersed cascade strategy was used to optimize the engine's operation over its flight envelope. Replacement of Powell's algorithm by the cascade strategy improved the optimization segment of the NEPP code. The performance of the linear regression and neural network methods as alternative engine analyzers was found to be satisfactory. This report considers two examples-a supersonic mixed-flow turbofan engine and a subsonic waverotor-topped engine-to illustrate the results, and it discusses insights gained from the improved version of the NEPP code.

Running SW4 On New Commodity Technology Systems (CTS-1) Platform

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodgers, Arthur J.; Petersson, N. Anders; Pitarka, Arben

We have recently been running earthquake ground motion simulations with SW4 on the new capacity computing systems, called the Commodity Technology Systems - 1 (CTS-1) at Lawrence Livermore National Laboratory (LLNL). SW4 is a fourth order time domain finite difference code developed by LLNL and distributed by the Computational Infrastructure for Geodynamics (CIG). SW4 simulates seismic wave propagation in complex three-dimensional Earth models including anelasticity and surface topography. We are modeling near-fault earthquake strong ground motions for the purposes of evaluating the response of engineered structures, such as nuclear power plants and other critical infrastructure. Engineering analysis of structures requiresmore » the inclusion of high frequencies which can cause damage, but are often difficult to include in simulations because of the need for large memory to model fine grid spacing on large domains.« less

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Themore » result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.« less

International Maritime Organizational Code For Safe Carriage Of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks On Board Ships -- IMO Resolution A.748(18)

DOT National Transportation Integrated Search

1994-05-26

This Circular calls the attention of Coast Guard field units, marine surveyors, shippers and carriers of nuclear materials to the International Maritime Organization (IMO) Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Leve...

Methodology, status, and plans for development and assessment of the RELAP5 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, G.W.; Riemke, R.A.

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

Optimizing Chemical-Vapor-Deposition Diamond for Nitrogen-Vacancy Center Ensemble Magnetrometry

DTIC Science & Technology

2017-06-01

Ju Li Battelle Energy Alliance Professor of Nuclear Science and Engineering Professor of Materials Science and Engineering...Sciences, U. S. Air Force Academy (2015) Submitted to the Department of Nuclear Science and Engineering in partial fulfillment of the requirements for the...degree of Master of Science in Nuclear Science and Engineering at the MASSACHUSETTS INSTITUTE OF TECHNOLOGY June 2017 c○ Massachusetts Institute of

NERVA dynamic analysis methodology, SPRVIB

NASA Technical Reports Server (NTRS)

Vronay, D. F.

1972-01-01

The general dynamic computer code called SPRVIB (Spring Vib) developed in support of the NERVA (nuclear engine for rocket vehicle application) program is described. Using normal mode techniques, the program computes kinematical responses of a structure caused by various combinations of harmonic and elliptic forcing functions or base excitations. Provision is made for a graphical type of force or base excitation input to the structure. A description of the required input format and a listing of the program are presented, along with several examples illustrating the use of the program. SPRVIB is written in FORTRAN 4 computer language for use on the CDC 6600 or the IBM 360/75 computers.

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

The Navy/NASA Engine Program (NNEP89): A user's manual

NASA Technical Reports Server (NTRS)

Plencner, Robert M.; Snyder, Christopher A.

1991-01-01

An engine simulation computer code called NNEP89 was written to perform 1-D steady state thermodynamic analysis of turbine engine cycles. By using a very flexible method of input, a set of standard components are connected at execution time to simulate almost any turbine engine configuration that the user could imagine. The code was used to simulate a wide range of engine cycles from turboshafts and turboprops to air turborockets and supersonic cruise variable cycle engines. Off design performance is calculated through the use of component performance maps. A chemical equilibrium model is incorporated to adequately predict chemical dissociation as well as model virtually any fuel. NNEP89 is written in standard FORTRAN77 with clear structured programming and extensive internal documentation. The standard FORTRAN77 programming allows it to be installed onto most mainframe computers and workstations without modification. The NNEP89 code was derived from the Navy/NASA Engine program (NNEP). NNEP89 provides many improvements and enhancements to the original NNEP code and incorporates features which make it easier to use for the novice user. This is a comprehensive user's guide for the NNEP89 code.

Teaching Problem-Solving Skills to Nuclear Engineering Students

ERIC Educational Resources Information Center

Waller, E.; Kaye, M. H.

2012-01-01

Problem solving is an essential skill for nuclear engineering graduates entering the workforce. Training in qualitative and quantitative aspects of problem solving allows students to conceptualise and execute solutions to complex problems. Solutions to problems in high consequence fields of study such as nuclear engineering require rapid and…

Nuclear Engineering Enrollments and Degrees, 1982.

ERIC Educational Resources Information Center

Sweeney, Deborah H.; And Others

This report presents data on the number of students enrolled and the number of bachelor's, master's, and doctoral degrees awarded in academic year 1981-82 from 72 United States institutions offering degree programs in nuclear engineering or nuclear options within other engineering fields. Presented as well are historical data for the last decade…

Brief 76 Nuclear Engineering Enrollments and Degrees Survey, 2015 Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

The 2015 Nuclear Engineering Enrollments and Degrees Survey reports degrees granted between September 1, 2014 and August 31, 2015. Enrollment information refers to the fall term 2015. The enrollments and degrees data comprises students majoring in nuclear engineering or in an option program equivalent to a major. Thirty-five academic programs reported having nuclear engineering programs during 2015, and data was received from all thirty-five programs. The report includes enrollment information on undergraduate students and graduate students and information by degree level for post-graduation plans.

Brief 74 Nuclear Engineering Enrollments and Degrees Survey, 2014 Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

2015-03-15

The 2014 survey includes degrees granted between September 1, 2013 and August 31, 2014, and enrollments for fall 2014. There are three academic programs new to this year's survey. Thirty-five academic programs reported having nuclear engineering programs during 2014, and data were provided by all thirty-five. The enrollments and degrees data include students majoring in nuclear engineering or in an option program equivalent to a major. Two nuclear engineering programs have indicated that health physics option enrollments and degrees are also reported in the health physics enrollments and degrees survey.

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

NASA Astrophysics Data System (ADS)

Gomin, E. A.; Davidenko, V. D.; Zinchenko, A. S.; Kharchenko, I. K.

2017-12-01

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

Nuclear Physics Made Very, Very Easy

NASA Technical Reports Server (NTRS)

Hanlen, D. F.; Morse, W. J.

1968-01-01

The fundamental approach to nuclear physics was prepared to introduce basic reactor principles to various groups of non-nuclear technical personnel associated with NERVA Test Operations. NERVA Test Operations functions as the field test group for the Nuclear Rocket Engine Program. Nuclear Engine for Rocket Vehicle Application (NERVA) program is the combined efforts of Aerojet-General Corporation as prime contractor, and Westinghouse Astronuclear Laboratory as the major subcontractor, for the assembly and testing of nuclear rocket engines. Development of the NERVA Program is under the direction of the Space Nuclear Propulsion Office, a joint agency of the U.S. Atomic Energy Commission and the National Aeronautics and Space Administration.

Computer Simulation of the VASIMR Engine

NASA Technical Reports Server (NTRS)

Garrison, David

2005-01-01

The goal of this project is to develop a magneto-hydrodynamic (MHD) computer code for simulation of the VASIMR engine. This code is designed be easy to modify and use. We achieve this using the Cactus framework, a system originally developed for research in numerical relativity. Since its release, Cactus has become an extremely powerful and flexible open source framework. The development of the code will be done in stages, starting with a basic fluid dynamic simulation and working towards a more complex MHD code. Once developed, this code can be used by students and researchers in order to further test and improve the VASIMR engine.

Computational Infrastructure for Engine Structural Performance Simulation

NASA Technical Reports Server (NTRS)

Chamis, Christos C.

1997-01-01

Select computer codes developed over the years to simulate specific aspects of engine structures are described. These codes include blade impact integrated multidisciplinary analysis and optimization, progressive structural fracture, quantification of uncertainties for structural reliability and risk, benefits estimation of new technology insertion and hierarchical simulation of engine structures made from metal matrix and ceramic matrix composites. Collectively these codes constitute a unique infrastructure readiness to credibly evaluate new and future engine structural concepts throughout the development cycle from initial concept, to design and fabrication, to service performance and maintenance and repairs, and to retirement for cause and even to possible recycling. Stated differently, they provide 'virtual' concurrent engineering for engine structures total-life-cycle-cost.

Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

NASA Astrophysics Data System (ADS)

Frybort, Jan

2017-09-01

Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

Towards defining the role of glycans as hardware in information storage and transfer: basic principles, experimental approaches and recent progress.

PubMed

Solís, D; Jiménez-Barbero, J; Kaltner, H; Romero, A; Siebert, H C; von der Lieth, C W; Gabius, H J

2001-01-01

The term 'code' in biological information transfer appears to be tightly and hitherto exclusively connected with the genetic code based on nucleotides and translated into functional activities via proteins. However, the recent appreciation of the enormous coding capacity of oligosaccharide chains of natural glycoconjugates has spurred to give heed to a new concept: versatile glycan assembly by the genetically encoded glycosyltransferases endows cells with a probably not yet fully catalogued array of meaningful messages. Enciphered by sugar receptors such as endogenous lectins the information of code words established by a series of covalently linked monosaccharides as letters for example guides correct intra- and intercellular routing of glycoproteins, modulates cell proliferation or migration and mediates cell adhesion. Evidently, the elucidation of the structural frameworks and the recognition strategies within the operation of the sugar code poses a fascinating conundrum. The far-reaching impact of this recognition mode on the level of cells, tissues and organs has fueled vigorous investigations to probe the subtleties of protein-carbohydrate interactions. This review presents information on the necessarily concerted approach using X-ray crystallography, molecular modeling, nuclear magnetic resonance spectroscopy, thermodynamic analysis and engineered ligands and receptors. This part of the treatise is flanked by exemplarily chosen insights made possible by these techniques. Copyright 2001 S. Karger AG, Basel

A Programmatic and Engineering Approach to the Development of a Nuclear Thermal Rocket for Space Exploration

NASA Technical Reports Server (NTRS)

Bordelon, Wayne J., Jr.; Ballard, Rick O.; Gerrish, Harold P., Jr.

2006-01-01

With the announcement of the Vision for Space Exploration on January 14, 2004, there has been a renewed interest in nuclear thermal propulsion. Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions; however, the cost to develop a nuclear thermal rocket engine system is uncertain. Key to determining the engine development cost will be the engine requirements, the technology used in the development and the development approach. The engine requirements and technology selection have not been defined and are awaiting definition of the Mars architecture and vehicle definitions. The paper discusses an engine development approach in light of top-level strategic questions and considerations for nuclear thermal propulsion and provides a suggested approach based on work conducted at the NASA Marshall Space Flight Center to support planning and requirements for the Prometheus Power and Propulsion Office. This work is intended to help support the development of a comprehensive strategy for nuclear thermal propulsion, to help reduce the uncertainty in the development cost estimate, and to help assess the potential value of and need for nuclear thermal propulsion for a human Mars mission.

Subgroup A : nuclear model codes report to the Sixteenth Meeting of the WPEC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talou, P.; Chadwick, M. B.; Dietrich, F. S.

2004-01-01

The Subgroup A activities focus on the development of nuclear reaction models and codes, used in evaluation work for nuclear reactions from the unresolved energy region up to the pion threshold production limit, and for target nuclides from the low teens and heavier. Much of the efforts are devoted by each participant to the continuing development of their own Institution codes. Progresses in this arena are reported in detail for each code in the present document. EMPIRE-II is of public access. The release of the TALYS code has been announced for the ND2004 Conference in Santa Fe, NM, October 2004.more » McGNASH is still under development and is not expected to be released in the very near future. In addition, Subgroup A members have demonstrated a growing interest in working on common modeling and codes capabilities, which would significantly reduce the amount of duplicate work, help manage efficiently the growing lines of existing codes, and render codes inter-comparison much easier. A recent and important activity of the Subgroup A has therefore been to develop the framework and the first bricks of the ModLib library, which is constituted of mostly independent pieces of codes written in Fortran 90 (and above) to be used in existing and future nuclear reaction codes. Significant progresses in the development of ModLib have been made during the past year. Several physics modules have been added to the library, and a few more have been planned in detail for the coming year.« less

Object-oriented approach for gas turbine engine simulation

NASA Technical Reports Server (NTRS)

Curlett, Brian P.; Felder, James L.

1995-01-01

An object-oriented gas turbine engine simulation program was developed. This program is a prototype for a more complete, commercial grade engine performance program now being proposed as part of the Numerical Propulsion System Simulator (NPSS). This report discusses architectural issues of this complex software system and the lessons learned from developing the prototype code. The prototype code is a fully functional, general purpose engine simulation program, however, only the component models necessary to model a transient compressor test rig have been written. The production system will be capable of steady state and transient modeling of almost any turbine engine configuration. Chief among the architectural considerations for this code was the framework in which the various software modules will interact. These modules include the equation solver, simulation code, data model, event handler, and user interface. Also documented in this report is the component based design of the simulation module and the inter-component communication paradigm. Object class hierarchies for some of the code modules are given.

Radiological Protection and Nuclear Engineering Studies in Multi-MW Target Systems

NASA Astrophysics Data System (ADS)

Luis, Raul Fernandes

Several innovative projects involving nuclear technology have emerged around the world in recent years, for applications such as spallation neutron sources, accelerator-driven systems for the transmutation of nuclear waste and radioactive ion beam (RIB) production. While the available neutron Wuxes from nuclear reactors did not increase substantially in intensity over the past three decades, the intensities of neutron sources produced in spallation targets have increased steadily, and should continue to do so during the 21st century. Innovative projects like ESS, MYRRHA and EURISOL lie at the forefront of the ongoing pursuit for increasingly bright neutron sources; driven by proton beams with energies up to 2 GeV and intensities up to several mA, the construction of their proposed facilities involves complex Nuclear Technology and Radiological Protection design studies executed by multidisciplinary teams of scientists and engineers from diUerent branches of Science. The intense neutron Wuxes foreseen for those facilities can be used in several scientiVc research Velds, such as Nuclear Physics and Astrophysics, Medicine and Materials Science. In this work, the target systems of two facilitites for the production of RIBs using the Isotope Separation On-Line (ISOL) method were studied in detail: ISOLDE, operating at CERN since 1967, and EURISOL, the next-generation ISOL facility to be built in Europe. For the EURISOL multi-MW target station, a detailed study of Radiological Protection was carried out using the Monte Carlo code FLUKA. Simulations were done to assess neutron Wuences, Vssion rates, ambient dose equivalent rates during operation and after shutdown and the production of radioactive nuclei in the targets and surrounding materials. DiUerent materials were discussed for diUerent components of the target system, aiming at improving its neutronics performance while keeping the residual activities resulting from material activation as low as possible. The second goal of this work was to perform an optimisation study for the ISOLDE neutron converter and Vssion target system. The target system was simulated using FLUKA and the cross section codes TALYS and ABRABLA, with the objective of maximising the performance of the system for the production of pure beams of neutron-rich isotopes, suppressing the contaminations by undesired neutron-deficient isobars. Two alternative target systems were proposed in the optimisation studies; the simplest of the two, with some modiVcations, was built as a prototype and tested at ISOLDE. The experimental results clearly show that it is possible, with simple changes in the layouts of the target systems, to produce purer beams of neutron-rich isotopes around the doubly magic nuclei 78Ni and 132Sn. A study of Radiological Protection was also performed, comparing the performances of the prototype target system and the standard ISOLDE target system. None

Software Engineering Education Directory

DTIC Science & Technology

1990-04-01

and Engineering (CMSC 735) Codes: GPEV2 * Textiooks: IEEE Tutoria on Models and Metrics for Software Management and Engameeing by Basi, Victor R...Software Engineering (Comp 227) Codes: GPRY5 Textbooks: IEEE Tutoria on Software Design Techniques by Freeman, Peter and Wasserman, Anthony 1. Software

Multi-Zone Liquid Thrust Chamber Performance Code with Domain Decomposition for Parallel Processing

NASA Technical Reports Server (NTRS)

Navaz, Homayun K.

2002-01-01

Computational Fluid Dynamics (CFD) has considerably evolved in the last decade. There are many computer programs that can perform computations on viscous internal or external flows with chemical reactions. CFD has become a commonly used tool in the design and analysis of gas turbines, ramjet combustors, turbo-machinery, inlet ducts, rocket engines, jet interaction, missile, and ramjet nozzles. One of the problems of interest to NASA has always been the performance prediction for rocket and air-breathing engines. Due to the complexity of flow in these engines it is necessary to resolve the flowfield into a fine mesh to capture quantities like turbulence and heat transfer. However, calculation on a high-resolution grid is associated with a prohibitively increasing computational time that can downgrade the value of the CFD for practical engineering calculations. The Liquid Thrust Chamber Performance (LTCP) code was developed for NASA/MSFC (Marshall Space Flight Center) to perform liquid rocket engine performance calculations. This code is a 2D/axisymmetric full Navier-Stokes (NS) solver with fully coupled finite rate chemistry and Eulerian treatment of liquid fuel and/or oxidizer droplets. One of the advantages of this code has been the resemblance of its input file to the JANNAF (Joint Army Navy NASA Air Force Interagency Propulsion Committee) standard TDK code, and its automatic grid generation for JANNAF defined combustion chamber wall geometry. These options minimize the learning effort for TDK users, and make the code a good candidate for performing engineering calculations. Although the LTCP code was developed for liquid rocket engines, it is a general-purpose code and has been used for solving many engineering problems. However, the single zone formulation of the LTCP has limited the code to be applicable to problems with complex geometry. Furthermore, the computational time becomes prohibitively large for high-resolution problems with chemistry, two-equation turbulence model, and two-phase flow. To overcome these limitations, the LTCP code is rewritten to include the multi-zone capability with domain decomposition that makes it suitable for parallel processing, i.e., enabling the code to run every zone or sub-domain on a separate processor. This can reduce the run time by a factor of 6 to 8, depending on the problem.

Preface to the Special Issue on TOUGH Symposium 2015

NASA Astrophysics Data System (ADS)

Blanco-Martín, Laura

2017-11-01

The TOUGH Symposium 2015 was held in Berkeley, California, September 28-30, 2015. The TOUGH family of codes, developed at the Energy Geosciences Division of Lawrence Berkeley National Laboratory (LBNL), is a suite of computer programs for the simulation of multiphase and multicomponent fluid and heat flows in porous and fractured media with applications in many geosciences fields, such as geothermal reservoir engineering, nuclear waste disposal, geological carbon sequestration, oil and gas reservoirs, gas hydrate research, vadose zone hydrology and environmental remediation. Since the first release in the 1980s, many modifications and enhancements have been continuously made to TOUGH and its various descendants (iTOUGH2, TOUGH+, TOUGH-MP, TOUGHREACT, TOUGH+HYDRATE, TMVOC...), at LBNL and elsewhere. Today, these codes are used worldwide in academia, government organizations and private companies in problems involving coupled hydrological, thermal, biogeochemical and geomechanical processes. The Symposia, organized every 2-3 years, bring together developers and users for an open exchange on recent code enhancements and applications. In 2015, the Symposium was attended by one hundred participants, representing thirty-four nationalities. This Special Issue in Computers & Geosciences gathers extended versions of selected Symposium proceedings related to (i) recent enhancements to the TOUGH family of codes and (ii) coupled flow and geomechanics processes modeling.

New Tool Released for Engine-Airframe Blade-Out Structural Simulations

NASA Technical Reports Server (NTRS)

Lawrence, Charles

2004-01-01

Researchers at the NASA Glenn Research Center have enhanced a general-purpose finite element code, NASTRAN, for engine-airframe structural simulations during steady-state and transient operating conditions. For steady-state simulations, the code can predict critical operating speeds, natural modes of vibration, and forced response (e.g., cabin noise and component fatigue). The code can be used to perform static analysis to predict engine-airframe response and component stresses due to maneuver loads. For transient response, the simulation code can be used to predict response due to bladeoff events and subsequent engine shutdown and windmilling conditions. In addition, the code can be used as a pretest analysis tool to predict the results of the bladeout test required for FAA certification of new and derivative aircraft engines. Before the present analysis code was developed, all the major aircraft engine and airframe manufacturers in the United States and overseas were performing similar types of analyses to ensure the structural integrity of engine-airframe systems. Although there were many similarities among the analysis procedures, each manufacturer was developing and maintaining its own structural analysis capabilities independently. This situation led to high software development and maintenance costs, complications with manufacturers exchanging models and results, and limitations in predicting the structural response to the desired degree of accuracy. An industry-NASA team was formed to overcome these problems by developing a common analysis tool that would satisfy all the structural analysis needs of the industry and that would be available and supported by a commercial software vendor so that the team members would be relieved of maintenance and development responsibilities. Input from all the team members was used to ensure that everyone's requirements were satisfied and that the best technology was incorporated into the code. Furthermore, because the code would be distributed by a commercial software vendor, it would be more readily available to engine and airframe manufacturers, as well as to nonaircraft companies that did not previously have access to this capability.

Program For Optimization Of Nuclear Rocket Engines

NASA Technical Reports Server (NTRS)

Plebuch, R. K.; Mcdougall, J. K.; Ridolphi, F.; Walton, James T.

1994-01-01

NOP is versatile digital-computer program devoloped for parametric analysis of beryllium-reflected, graphite-moderated nuclear rocket engines. Facilitates analysis of performance of engine with respect to such considerations as specific impulse, engine power, type of engine cycle, and engine-design constraints arising from complications of fuel loading and internal gradients of temperature. Predicts minimum weight for specified performance.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rizwan-uddin

Recently, various branches of engineering and science have seen a rapid increase in the number of dynamical analyses undertaken. This modern phenomenon often obscures the fact that such analyses were sometimes carried out even before the current trend began. Moreover, these earlier analyses, which even now seem very ingenuous, were carried out at a time when the available information about dynamical systems was not as well disseminated as it is today. One such analysis, carried out in the early 1960s, showed the existence of stable limit cycles in a simple model for space-independent xenon dynamics in nuclear reactors. The authors,more » apparently unaware of the now well-known bifurcation theorem by Hopf, could not numerically discover unstable limit cycles, though they did find regions in parameter space where the fixed points are stable for small perturbations but unstable for very large perturbations. The analysis was carried out both analytically and numerically. As a tribute to these early nonlinear dynamicists in the field of nuclear engineering, in this paper, the Hopf theorem and its conclusions are briefly described, and then the solution of the space-independent xenon oscillation problem is presented, which was obtained using the bifurcation analysis BIFDD code. These solutions are presented along with a discussion of the earlier results.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

NASA Astrophysics Data System (ADS)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Auto-Coding UML Statecharts for Flight Software

NASA Technical Reports Server (NTRS)

Benowitz, Edward G; Clark, Ken; Watney, Garth J.

2006-01-01

Statecharts have been used as a means to communicate behaviors in a precise manner between system engineers and software engineers. Hand-translating a statechart to code, as done on some previous space missions, introduces the possibility of errors in the transformation from chart to code. To improve auto-coding, we have developed a process that generates flight code from UML statecharts. Our process is being used for the flight software on the Space Interferometer Mission (SIM).

Requirements for construction of nuclear system components at elevated temperatures (supplement to ASME Code Cases 1592, 1593, 1594, 1595, and 1596)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This standard provides rules for the construction of Class 1 nuclear components, parts, and appurtenances for use at elevated temperatures. This standard is a complete set of requirements only when used in conjunction with Section III of the ASME Boiler and Pressure Vessel Code (ASME Code) and addenda, ASME Code Cases 1592, 1593, 1594, 1595, and 1596, and RDT E 15-2NB. Unmodified paragraphs of the referenced Code Cases are not repeated in this standard but are a part of the requirements of this standard.

Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

PubMed

Li, Fusheng; Han, Xiaogang

2012-07-01

The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. Copyright © 2011 Elsevier Ltd. All rights reserved.

EMPIRE: A Reaction Model Code for Nuclear Astrophysics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Palumbo, A., E-mail: apalumbo@bnl.gov; Herman, M.; Capote, R.

The correct modeling of abundances requires knowledge of nuclear cross sections for a variety of neutron, charged particle and γ induced reactions. These involve targets far from stability and are therefore difficult (or currently impossible) to measure. Nuclear reaction theory provides the only way to estimate values of such cross sections. In this paper we present application of the EMPIRE reaction code to nuclear astrophysics. Recent measurements are compared to the calculated cross sections showing consistent agreement for n-, p- and α-induced reactions of strophysical relevance.

Study of components and statistical reaction mechanism in simulation of nuclear process for optimized production of {sup 64}Cu and {sup 67}Ga medical radioisotopes using TALYS, EMPIRE and LISE++ nuclear reaction and evaporation codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nasrabadi, M. N., E-mail: mnnasrabadi@ast.ui.ac.ir; Sepiani, M.

2015-03-30

Production of medical radioisotopes is one of the most important tasks in the field of nuclear technology. These radioactive isotopes are mainly produced through variety nuclear process. In this research, excitation functions and nuclear reaction mechanisms are studied for simulation of production of these radioisotopes in the TALYS, EMPIRE and LISE++ reaction codes, then parameters and different models of nuclear level density as one of the most important components in statistical reaction models are adjusted for optimum production of desired radioactive yields.

Study of components and statistical reaction mechanism in simulation of nuclear process for optimized production of 64Cu and 67Ga medical radioisotopes using TALYS, EMPIRE and LISE++ nuclear reaction and evaporation codes

NASA Astrophysics Data System (ADS)

Nasrabadi, M. N.; Sepiani, M.

2015-03-01

Production of medical radioisotopes is one of the most important tasks in the field of nuclear technology. These radioactive isotopes are mainly produced through variety nuclear process. In this research, excitation functions and nuclear reaction mechanisms are studied for simulation of production of these radioisotopes in the TALYS, EMPIRE & LISE++ reaction codes, then parameters and different models of nuclear level density as one of the most important components in statistical reaction models are adjusted for optimum production of desired radioactive yields.

75 FR 43208 - Withdrawal of Regulatory Guide 5.17

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-23

... Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555.... Introduction The U.S. Nuclear Regulatory Commission (NRC) is withdrawing Regulatory Guide 5.17, ``Truck... Development Branch, Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc. 2010-18077 Filed...

Researcher Poses with a Nuclear Rocket Model

NASA Image and Video Library

1961-11-21

A researcher at the NASA Lewis Research Center with slide ruler poses with models of the earth and a nuclear-propelled rocket. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The nuclear rocket model in this photograph includes a reactor at the far right with a hydrogen propellant tank and large radiator below. The payload or crew would be at the far left, distanced from the reactor.

Harbors.

DTIC Science & Technology

1981-07-01

CONTRACT OR GRANT NUMBER(e) Naval Facilities Engineering Command 200 Stovall Street r Alexandria, VA 22332 (Code 0453) s. PERFORMING ORGANIZATION NAME...AND ADDRESS 10. PROGRAM ELEMENT. PROJECT. TASK • Naval Facilities Engineering Command AREA & WORK UNIT NUMBERS < 200 Stovall Street Engineering and...Design Alexandria, VA 22332 It. CONTROLLING OFFICE NAME AND ADDRESS 12. REPORT DATE ~ Naval Facilities Engineering Command (Code10432) July 1981 200

Optimizing a liquid propellant rocket engine with an automated combustor design code (AUTOCOM)

NASA Technical Reports Server (NTRS)

Hague, D. S.; Reichel, R. H.; Jones, R. T.; Glatt, C. R.

1972-01-01

A procedure for automatically designing a liquid propellant rocket engine combustion chamber in an optimal fashion is outlined. The procedure is contained in a digital computer code, AUTOCOM. The code is applied to an existing engine, and design modifications are generated which provide a substantial potential payload improvement over the existing design. Computer time requirements for this payload improvement were small, approximately four minutes in the CDC 6600 computer.

The NJOY Nuclear Data Processing System, Version 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Macfarlane, Robert; Muir, Douglas W.; Boicourt, R. M.

The NJOY Nuclear Data Processing System, version 2016, is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from evaluated nuclear data in the ENDF-4 through ENDF-6 legacy card-image formats. NJOY works with evaluated files for incident neutrons, photons, and charged particles, producing libraries for a wide variety of particle transport and reactor analysis codes.

OWL: A code for the two-center shell model with spherical Woods-Saxon potentials

NASA Astrophysics Data System (ADS)

Diaz-Torres, Alexis

2018-03-01

A Fortran-90 code for solving the two-center nuclear shell model problem is presented. The model is based on two spherical Woods-Saxon potentials and the potential separable expansion method. It describes the single-particle motion in low-energy nuclear collisions, and is useful for characterizing a broad range of phenomena from fusion to nuclear molecular structures.

Multiphysics Code Demonstrated for Propulsion Applications

NASA Technical Reports Server (NTRS)

Lawrence, Charles; Melis, Matthew E.

1998-01-01

The utility of multidisciplinary analysis tools for aeropropulsion applications is being investigated at the NASA Lewis Research Center. The goal of this project is to apply Spectrum, a multiphysics code developed by Centric Engineering Systems, Inc., to simulate multidisciplinary effects in turbomachinery components. Many engineering problems today involve detailed computer analyses to predict the thermal, aerodynamic, and structural response of a mechanical system as it undergoes service loading. Analysis of aerospace structures generally requires attention in all three disciplinary areas to adequately predict component service behavior, and in many cases, the results from one discipline substantially affect the outcome of the other two. There are numerous computer codes currently available in the engineering community to perform such analyses in each of these disciplines. Many of these codes are developed and used in-house by a given organization, and many are commercially available. However, few, if any, of these codes are designed specifically for multidisciplinary analyses. The Spectrum code has been developed for performing fully coupled fluid, thermal, and structural analyses on a mechanical system with a single simulation that accounts for all simultaneous interactions, thus eliminating the requirement for running a large number of sequential, separate, disciplinary analyses. The Spectrum code has a true multiphysics analysis capability, which improves analysis efficiency as well as accuracy. Centric Engineering, Inc., working with a team of Lewis and AlliedSignal Engines engineers, has been evaluating Spectrum for a variety of propulsion applications including disk quenching, drum cavity flow, aeromechanical simulations, and a centrifugal compressor flow simulation.

Performance assessment methodology and preliminary results for low-level radioactive waste disposal in Taiwan.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arnold, Bill Walter; Chang, Fu-lin; Mattie, Patrick D.

2006-02-01

Sandia National Laboratories (SNL) and Taiwan's Institute for Nuclear Energy Research (INER) have teamed together to evaluate several candidate sites for Low-Level Radioactive Waste (LLW) disposal in Taiwan. Taiwan currently has three nuclear power plants, with another under construction. Taiwan also has a research reactor, as well as medical and industrial wastes to contend with. Eventually the reactors will be decomissioned. Operational and decommissioning wastes will need to be disposed in a licensed disposal facility starting in 2014. Taiwan has adopted regulations similar to the US Nuclear Regulatory Commission's (NRC's) low-level radioactive waste rules (10 CFR 61) to govern themore » disposal of LLW. Taiwan has proposed several potential sites for the final disposal of LLW that is now in temporary storage on Lanyu Island and on-site at operating nuclear power plants, and for waste generated in the future through 2045. The planned final disposal facility will have a capacity of approximately 966,000 55-gallon drums. Taiwan is in the process of evaluating the best candidate site to pursue for licensing. Among these proposed sites there are basically two disposal concepts: shallow land burial and cavern disposal. A representative potential site for shallow land burial is located on a small island in the Taiwan Strait with basalt bedrock and interbedded sedimentary rocks. An engineered cover system would be constructed to limit infiltration for shallow land burial. A representative potential site for cavern disposal is located along the southeastern coast of Taiwan in a tunnel system that would be about 500 to 800 m below the surface. Bedrock at this site consists of argillite and meta-sedimentary rocks. Performance assessment analyses will be performed to evaluate future performance of the facility and the potential dose/risk to exposed populations. Preliminary performance assessment analyses will be used in the site-selection process and to aid in design of the disposal system. Final performance assessment analyses will be used in the regulatory process of licensing a site. The SNL/INER team has developed a performance assessment methodology that is used to simulate processes associated with the potential release of radionuclides to evaluate these sites. The following software codes are utilized in the performance assessment methodology: GoldSim (to implement a probabilistic analysis that will explicitly address uncertainties); the NRC's Breach, Leach, and Transport - Multiple Species (BLT-MS) code (to simulate waste-container degradation, waste-form leaching, and transport through the host rock); the Finite Element Heat and Mass Transfer code (FEHM) (to simulate groundwater flow and estimate flow velocities); the Hydrologic Evaluation of Landfill performance Model (HELP) code (to evaluate infiltration through the disposal cover); the AMBER code (to evaluate human health exposures); and the NRC's Disposal Unit Source Term -- Multiple Species (DUST-MS) code (to screen applicable radionuclides). Preliminary results of the evaluations of the two disposal concept sites are presented.« less

CELCAP: A Computer Model for Cogeneration System Analysis

NASA Technical Reports Server (NTRS)

1985-01-01

A description of the CELCAP cogeneration analysis program is presented. A detailed description of the methodology used by the Naval Civil Engineering Laboratory in developing the CELCAP code and the procedures for analyzing cogeneration systems for a given user are given. The four engines modeled in CELCAP are: gas turbine with exhaust heat boiler, diesel engine with waste heat boiler, single automatic-extraction steam turbine, and back-pressure steam turbine. Both the design point and part-load performances are taken into account in the engine models. The load model describes how the hourly electric and steam demand of the user is represented by 24 hourly profiles. The economic model describes how the annual and life-cycle operating costs that include the costs of fuel, purchased electricity, and operation and maintenance of engines and boilers are calculated. The CELCAP code structure and principal functions of the code are described to how the various components of the code are related to each other. Three examples of the application of the CELCAP code are given to illustrate the versatility of the code. The examples shown represent cases of system selection, system modification, and system optimization.

Progress on China nuclear data processing code system

NASA Astrophysics Data System (ADS)

Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

2017-09-01

China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

NASA Astrophysics Data System (ADS)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

PARALLEL PERTURBATION MODEL FOR CYCLE TO CYCLE VARIABILITY PPM4CCV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ameen, Muhsin Mohammed; Som, Sibendu

This code consists of a Fortran 90 implementation of the parallel perturbation model to compute cyclic variability in spark ignition (SI) engines. Cycle-to-cycle variability (CCV) is known to be detrimental to SI engine operation resulting in partial burn and knock, and result in an overall reduction in the reliability of the engine. Numerical prediction of cycle-to-cycle variability (CCV) in SI engines is extremely challenging for two key reasons: (i) high-fidelity methods such as large eddy simulation (LES) are required to accurately capture the in-cylinder turbulent flow field, and (ii) CCV is experienced over long timescales and hence the simulations needmore » to be performed for hundreds of consecutive cycles. In the new technique, the strategy is to perform multiple parallel simulations, each of which encompasses 2-3 cycles, by effectively perturbing the simulation parameters such as the initial and boundary conditions. The PPM4CCV code is a pre-processing code and can be coupled with any engine CFD code. PPM4CCV was coupled with Converge CFD code and a 10-time speedup was demonstrated over the conventional multi-cycle LES in predicting the CCV for a motored engine. Recently, the model is also being applied to fired engines including port fuel injected (PFI) and direct injection spark ignition engines and the preliminary results are very encouraging.« less

75 FR 79049 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-17

..., Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission... INFORMATION: I. Introduction The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to an existing... Development Branch, Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc. 2010-31731 Filed...

Performance Evaluation of Axial Flow AG-1 FC and Prototype FM (High Strength) HEPA Filters - 13123

DOE Office of Scientific and Technical Information (OSTI.GOV)

Giffin, Paxton K.; Parsons, Michael S.; Wilson, John A.

High efficiency particulate air (HEPA) filters are routinely used in DOE nuclear containment activities. The Nuclear Air Cleaning Handbook (NACH) stipulates that air cleaning devices and equipment used in DOE nuclear applications must meet the American Society of Mechanical Engineers (ASME) Code on Nuclear Air and Gas Treatment (AG-1) standard. This testing activity evaluates two different axial flow HEPA filters, those from AG-1 Sections FC and FM. Section FM is under development and has not yet been added to AG-1 due to a lack of qualification data available for these filters. Section FC filters are axial flow units that utilizemore » a fibrous glass filtering medium. The section FM filters utilize a similar fibrous glass medium, but also have scrim backing. The scrim-backed filters have demonstrated the ability to endure pressure impulses capable of completely destroying FC filters. The testing activities presented herein will examine the total lifetime loading for both FC and FM filters under ambient conditions and at elevated conditions of temperature and relative humidity. Results will include loading curves, penetration curves, and testing condition parameters. These testing activities have been developed through collaborations with representatives from the National Nuclear Security Administration (NNSA), DOE Office of Environmental Management (DOE-EM), New Mexico State University, and Mississippi State University. (authors)« less

A Semantic Analysis Method for Scientific and Engineering Code

NASA Technical Reports Server (NTRS)

Stewart, Mark E. M.

1998-01-01

This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

A Computer Code for Gas Turbine Engine Weight And Disk Life Estimation

NASA Technical Reports Server (NTRS)

Tong, Michael T.; Ghosn, Louis J.; Halliwell, Ian; Wickenheiser, Tim (Technical Monitor)

2002-01-01

Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. In this paper, the major enhancements to NASA's engine-weight estimate computer code (WATE) are described. These enhancements include the incorporation of improved weight-calculation routines for the compressor and turbine disks using the finite-difference technique. Furthermore, the stress distribution for various disk geometries was also incorporated, for a life-prediction module to calculate disk life. A material database, consisting of the material data of most of the commonly-used aerospace materials, has also been incorporated into WATE. Collectively, these enhancements provide a more realistic and systematic way to calculate the engine weight. They also provide additional insight into the design trade-off between engine life and engine weight. To demonstrate the new capabilities, the enhanced WATE code is used to perform an engine weight/life trade-off assessment on a production aircraft engine.

10 CFR 851.27 - Reference sources.

Code of Federal Regulations, 2013 CFR

2013-01-01

...) American Society of Mechanical Engineers (ASME), P.O. Box 2300 Fairfield, NJ 07007. Telephone: 800-843-2763... Electrical Code,” (2005). (5) NFPA 70E, “Standard for Electrical Safety in the Workplace,” (2004). (6... Engineers (ASME) Boilers and Pressure Vessel Code, sections I through XII including applicable Code Cases...

10 CFR 851.27 - Reference sources.

Code of Federal Regulations, 2014 CFR

2014-01-01

...) American Society of Mechanical Engineers (ASME), P.O. Box 2300 Fairfield, NJ 07007. Telephone: 800-843-2763... Electrical Code,” (2005). (5) NFPA 70E, “Standard for Electrical Safety in the Workplace,” (2004). (6... Engineers (ASME) Boilers and Pressure Vessel Code, sections I through XII including applicable Code Cases...

Developments in REDES: The rocket engine design expert system

NASA Technical Reports Server (NTRS)

Davidian, Kenneth O.

1990-01-01

The Rocket Engine Design Expert System (REDES) is being developed at the NASA-Lewis to collect, automate, and perpetuate the existing expertise of performing a comprehensive rocket engine analysis and design. Currently, REDES uses the rigorous JANNAF methodology to analyze the performance of the thrust chamber and perform computational studies of liquid rocket engine problems. The following computer codes were included in REDES: a gas properties program named GASP, a nozzle design program named RAO, a regenerative cooling channel performance evaluation code named RTE, and the JANNAF standard liquid rocket engine performance prediction code TDK (including performance evaluation modules ODE, ODK, TDE, TDK, and BLM). Computational analyses are being conducted by REDES to provide solutions to liquid rocket engine thrust chamber problems. REDES is built in the Knowledge Engineering Environment (KEE) expert system shell and runs on a Sun 4/110 computer.

Developments in REDES: The Rocket Engine Design Expert System

NASA Technical Reports Server (NTRS)

Davidian, Kenneth O.

1990-01-01

The Rocket Engine Design Expert System (REDES) was developed at NASA-Lewis to collect, automate, and perpetuate the existing expertise of performing a comprehensive rocket engine analysis and design. Currently, REDES uses the rigorous JANNAF methodology to analyze the performance of the thrust chamber and perform computational studies of liquid rocket engine problems. The following computer codes were included in REDES: a gas properties program named GASP; a nozzle design program named RAO; a regenerative cooling channel performance evaluation code named RTE; and the JANNAF standard liquid rocket engine performance prediction code TDK (including performance evaluation modules ODE, ODK, TDE, TDK, and BLM). Computational analyses are being conducted by REDES to provide solutions to liquid rocket engine thrust chamber problems. REDES was built in the Knowledge Engineering Environment (KEE) expert system shell and runs on a Sun 4/110 computer.

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

National Nuclear Data Center

Science.gov Websites

reaction data Sigma Retrieval & Plotting Nuclear structure & decay Data Nuclear Science References Experimental Unevaluated Nuclear Data List Evaluated Nuclear Structure Data File NNDC databases Ground and isomeric states properties Nuclear structure & decay data journal Nuclear reaction model code Tools and

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Evaluation of Spacecraft Shielding Effectiveness for Radiation Protection

NASA Technical Reports Server (NTRS)

Cucinotta, Francis A.; Wilson, John W.

1999-01-01

The potential for serious health risks from solar particle events (SPE) and galactic cosmic rays (GCR) is a critical issue in the NASA strategic plan for the Human Exploration and Development of Space (HEDS). The excess cost to protect against the GCR and SPE due to current uncertainties in radiation transmission properties and cancer biology could be exceedingly large based on the excess launch costs to shield against uncertainties. The development of advanced shielding concepts is an important risk mitigation area with the potential to significantly reduce risk below conventional mission designs. A key issue in spacecraft material selection is the understanding of nuclear reactions on the transmission properties of materials. High-energy nuclear particles undergo nuclear reactions in passing through materials and tissue altering their composition and producing new radiation types. Spacecraft and planetary habitat designers can utilize radiation transport codes to identify optimal materials for lowering exposures and to optimize spacecraft design to reduce astronaut exposures. To reach these objectives will require providing design engineers with accurate data bases and computationally efficient software for describing the transmission properties of space radiation in materials. Our program will reduce the uncertainty in the transmission properties of space radiation by improving the theoretical description of nuclear reactions and radiation transport, and provide accurate physical descriptions of the track structure of microscopic energy deposition.

STEM Leader from the Roeper School: An Interview with Nuclear Engineer Clair J. Sullivan

ERIC Educational Resources Information Center

Ambrose, Don

2016-01-01

Clair J. Sullivan is an assistant professor in the Department of Nuclear, Plasma and Radiological Engineering at the University of Illinois at Urbana-Champaign (UIUC). Her research interests include radiation detection and measurements; gamma-ray spectroscopy; automated isotope identification algorithms; nuclear forensics; nuclear security;…

Hybrid discrete ordinates and characteristics method for solving the linear Boltzmann equation

NASA Astrophysics Data System (ADS)

Yi, Ce

With the ability of computer hardware and software increasing rapidly, deterministic methods to solve the linear Boltzmann equation (LBE) have attracted some attention for computational applications in both the nuclear engineering and medical physics fields. Among various deterministic methods, the discrete ordinates method (SN) and the method of characteristics (MOC) are two of the most widely used methods. The SN method is the traditional approach to solve the LBE for its stability and efficiency. While the MOC has some advantages in treating complicated geometries. However, in 3-D problems requiring a dense discretization grid in phase space (i.e., a large number of spatial meshes, directions, or energy groups), both methods could suffer from the need for large amounts of memory and computation time. In our study, we developed a new hybrid algorithm by combing the two methods into one code, TITAN. The hybrid approach is specifically designed for application to problems containing low scattering regions. A new serial 3-D time-independent transport code has been developed. Under the hybrid approach, the preferred method can be applied in different regions (blocks) within the same problem model. Since the characteristics method is numerically more efficient in low scattering media, the hybrid approach uses a block-oriented characteristics solver in low scattering regions, and a block-oriented SN solver in the remainder of the physical model. In the TITAN code, a physical problem model is divided into a number of coarse meshes (blocks) in Cartesian geometry. Either the characteristics solver or the SN solver can be chosen to solve the LBE within a coarse mesh. A coarse mesh can be filled with fine meshes or characteristic rays depending on the solver assigned to the coarse mesh. Furthermore, with its object-oriented programming paradigm and layered code structure, TITAN allows different individual spatial meshing schemes and angular quadrature sets for each coarse mesh. Two quadrature types (level-symmetric and Legendre-Chebyshev quadrature) along with the ordinate splitting techniques (rectangular splitting and PN-TN splitting) are implemented. In the S N solver, we apply a memory-efficient 'front-line' style paradigm to handle the fine mesh interface fluxes. In the characteristics solver, we have developed a novel 'backward' ray-tracing approach, in which a bi-linear interpolation procedure is used on the incoming boundaries of a coarse mesh. A CPU-efficient scattering kernel is shared in both solvers within the source iteration scheme. Angular and spatial projection techniques are developed to transfer the angular fluxes on the interfaces of coarse meshes with different discretization grids. The performance of the hybrid algorithm is tested in a number of benchmark problems in both nuclear engineering and medical physics fields. Among them are the Kobayashi benchmark problems and a computational tomography (CT) device model. We also developed an extra sweep procedure with the fictitious quadrature technique to calculate angular fluxes along directions of interest. The technique is applied in a single photon emission computed tomography (SPECT) phantom model to simulate the SPECT projection images. The accuracy and efficiency of the TITAN code are demonstrated in these benchmarks along with its scalability. A modified version of the characteristics solver is integrated in the PENTRAN code and tested within the parallel engine of PENTRAN. The limitations on the hybrid algorithm are also studied.

Scientific and Technical Manpower Requirements of Selected Segments of the Atomic Energy Field. Final Report.

ERIC Educational Resources Information Center

Voight, Keith L.

The primary purpose of the study was to develop a supply/demand ratio for nuclear degree scientists and engineers from July 1969 through 1973. The need by private industry and electric utilities for scientists and engineers with degrees in disciplines other than nuclear science or engineering, as well as for technicians, nuclear reactor operators,…

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

Incorporating Manual and Autonomous Code Generation

NASA Technical Reports Server (NTRS)

McComas, David

1998-01-01

Code can be generated manually or using code-generated software tools, but how do you interpret the two? This article looks at a design methodology that combines object-oriented design with autonomic code generation for attitude control flight software. Recent improvements in space flight computers are allowing software engineers to spend more time engineering the applications software. The application developed was the attitude control flight software for an astronomical satellite called the Microwave Anisotropy Probe (MAP). The MAP flight system is being designed, developed, and integrated at NASA's Goddard Space Flight Center. The MAP controls engineers are using Integrated Systems Inc.'s MATRIXx for their controls analysis. In addition to providing a graphical analysis for an environment, MATRIXx includes an autonomic code generation facility called AutoCode. This article examines the forces that shaped the final design and describes three highlights of the design process: (1) Defining the manual to autonomic code interface; (2) Applying object-oriented design to the manual flight code; (3) Implementing the object-oriented design in C.

National Electrical Code in Power Engineering Course for Electrical Engineering Curriculum

ERIC Educational Resources Information Center

Azizur, Rahman M. M.

2011-01-01

In order to ensure the safety of their inhabitants and properties, the residential, industrial and business installations require complying with NEC (national electrical code) for electrical systems. Electrical design engineers and technicians rely heavily on these very important design guidelines. However, these design guidelines are not formally…

Fracture Analysis of Welded Type 304 Stainless Steel Pipe

DTIC Science & Technology

1986-11-01

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code . In order to accomplish these objectives, a series of seven full...Mechanical Engineers Boiler and Pressure Vessel Code , Section XI IWB-3640 (Winter Addenda 1983). 5. Ranganath, S., and U.S. Mehta, "Engineering Methods for

78 FR 79363 - Hazardous Materials: Adoption of ASME Code Section XII and the National Board Inspection Code

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

... Rulemaking Division, (202) 366-8553, or Stanley Staniszewski, Engineering and Research [[Page 79364... increased capacity to transport product. A review of previous research by PHMSA's Engineering and Research..., knowledge-sharing, and skill development across all engineering disciplines. ASME is recognized globally for...

Error-correction coding for digital communications

NASA Astrophysics Data System (ADS)

Clark, G. C., Jr.; Cain, J. B.

This book is written for the design engineer who must build the coding and decoding equipment and for the communication system engineer who must incorporate this equipment into a system. It is also suitable as a senior-level or first-year graduate text for an introductory one-semester course in coding theory. Fundamental concepts of coding are discussed along with group codes, taking into account basic principles, practical constraints, performance computations, coding bounds, generalized parity check codes, polynomial codes, and important classes of group codes. Other topics explored are related to simple nonalgebraic decoding techniques for group codes, soft decision decoding of block codes, algebraic techniques for multiple error correction, the convolutional code structure and Viterbi decoding, syndrome decoding techniques, and sequential decoding techniques. System applications are also considered, giving attention to concatenated codes, coding for the white Gaussian noise channel, interleaver structures for coded systems, and coding for burst noise channels.

Conceptual design studies and experiments related to cavity exhaust systems for nuclear light bulb configurations

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Stoeffler, R. C.

1972-01-01

Investigations of various phases of gaseous nuclear rocket technology have been conducted. The principal research efforts have recently been directed toward the closed-cycle, vortex-stabilized nuclear light bulb engine and toward a small-scale fissioning uranium plasma experiment that could be conducted in the Los Alamos Scientific Laboratory's Nuclear Furnace. The engine concept is based on the transfer of energy by thermal radiation from gaseous fissioning uranium, through a transparent wall, to hydrogen propellant. The reference engine configuration is comprised of seven unit cavities, each having its own fuel transparent wall and propellant duct. The basic design of the engine is described. Subsequent studies performed to supplement and investigate the basic design are reported. Summaries of other nuclear light bulb research programs are included.

Modeling the Effects of Ice Accretion on the Low Pressure Compressor and the Overall Turbofan Engine System Performance

NASA Technical Reports Server (NTRS)

Veres, Joseph P.; Jorgenson, Philip C. E.; Wright, William B.

2011-01-01

The focus of this study is on utilizing a mean line compressor flow analysis code coupled to an engine system thermodynamic code, to estimate the effects of ice accretion on the low pressure compressor, and quantifying its effects on the engine system throughout a notional flight trajectory. In this paper a temperature range in which engine icing would occur was assumed. This provided a mechanism to locate potential component icing sites and allow the computational tools to add blockages due to ice accretion in a parametric fashion. Ultimately the location and level of blockage due to icing would be provided by an ice accretion code. To proceed, an engine system modeling code and a mean line compressor flow analysis code were utilized to calculate the flow conditions in the fan-core and low pressure compressor and to identify potential locations within the compressor where ice may accrete. In this study, an "additional blockage" due to the accretion of ice on the metal surfaces, has been added to the baseline aerodynamic blockage due to boundary layer, as well as the blade metal blockage. Once the potential locations of ice accretion are identified, the levels of additional blockage due to accretion were parametrically varied to estimate the effects on the low pressure compressor blade row performance operating within the engine system environment. This study includes detailed analysis of compressor and engine performance during cruise and descent operating conditions at several altitudes within the notional flight trajectory. The purpose of this effort is to develop the computer codes to provide a predictive capability to forecast the onset of engine icing events, such that they could ultimately help in the avoidance of these events.

NASA Lewis Stirling engine computer code evaluation

NASA Technical Reports Server (NTRS)

Sullivan, Timothy J.

1989-01-01

In support of the U.S. Department of Energy's Stirling Engine Highway Vehicle Systems program, the NASA Lewis Stirling engine performance code was evaluated by comparing code predictions without engine-specific calibration factors to GPU-3, P-40, and RE-1000 Stirling engine test data. The error in predicting power output was -11 percent for the P-40 and 12 percent for the Re-1000 at design conditions and 16 percent for the GPU-3 at near-design conditions (2000 rpm engine speed versus 3000 rpm at design). The efficiency and heat input predictions showed better agreement with engine test data than did the power predictions. Concerning all data points, the error in predicting the GPU-3 brake power was significantly larger than for the other engines and was mainly a result of inaccuracy in predicting the pressure phase angle. Analysis into this pressure phase angle prediction error suggested that improvements to the cylinder hysteresis loss model could have a significant effect on overall Stirling engine performance predictions.

Nuclear Energy Knowledge and Validation Center (NEKVaC) Needs Workshop Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans

2015-02-01

The Department of Energy (DOE) has made significant progress developing simulation tools to predict the behavior of nuclear systems with greater accuracy and of increasing our capability to predict the behavior of these systems outside of the standard range of applications. These analytical tools require a more complex array of validation tests to accurately simulate the physics and multiple length and time scales. Results from modern simulations will allow experiment designers to narrow the range of conditions needed to bound system behavior and to optimize the deployment of instrumentation to limit the breadth and cost of the campaign. Modern validation,more » verification and uncertainty quantification (VVUQ) techniques enable analysts to extract information from experiments in a systematic manner and provide the users with a quantified uncertainty estimate. Unfortunately, the capability to perform experiments that would enable taking full advantage of the formalisms of these modern codes has progressed relatively little (with some notable exceptions in fuels and thermal-hydraulics); the majority of the experimental data available today is the "historic" data accumulated over the last decades of nuclear systems R&D. A validated code-model is a tool for users. An unvalidated code-model is useful for code developers to gain understanding, publish research results, attract funding, etc. As nuclear analysis codes have become more sophisticated, so have the measurement and validation methods and the challenges that confront them. A successful yet cost-effective validation effort requires expertise possessed only by a few, resources possessed only by the well-capitalized (or a willing collective), and a clear, well-defined objective (validating a code that is developed to satisfy the need(s) of an actual user). To that end, the Idaho National Laboratory established the Nuclear Energy Knowledge and Validation Center to address the challenges of modern code validation and to manage the knowledge from past, current, and future experimental campaigns. By pulling together the best minds involved in code development, experiment design, and validation to establish and disseminate best practices and new techniques, the Nuclear Energy Knowledge and Validation Center (NEKVaC or the ‘Center’) will be a resource for industry, DOE Programs, and academia validation efforts.« less

Joint Conference on Marine Safety and Environment/Ship Production (1st), held 1-5 June 1992

DTIC Science & Technology

1992-06-05

have to face fresh challenges in the future. But as long as we regard these challenges as opportunities, we will remain masters of our own destiny ...has recently published for discussion an “ Embryo Code of Practice” (1991) entitled “Engineers and Risk Issues”. Thii draft Code is intended to...Department of Energy5 Her Majesty’s Stationery Office. Cm.1310. The Engineering Counci1 (1991) Engineers and Risk Issues : An Embryo Code of Practice. London

An Historical Perspective of the NERVA Nuclear Rocket Engine Technology Program

NASA Technical Reports Server (NTRS)

Robbins, W. H.; Finger, H. B.

1991-01-01

Nuclear rocket research and development was initiated in the United States in 1955 and is still being pursued to a limited extent. The major technology emphasis occurred in the decade of the 1960s and was primarily associated with the Rover/NERVA Program where the technology for a nuclear rocket engine system for space application was developed and demonstrated. The NERVA (Nuclear Engine for Rocket Vehicle Application) technology developed twenty years ago provides a comprehensive and viable propulsion technology base that can be applied and will prove to be valuable for application to the NASA Space Exploration Initiative (SEI). This paper, which is historical in scope, provides an overview of the conduct of the NERVA Engine Program, its organization and management, development philosophy, the engine configuration, and significant accomplishments.

Transmutation Fuel Performance Code Thermal Model Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

Simulation of prompt gamma-ray emission during proton radiotherapy.

PubMed

Verburg, Joost M; Shih, Helen A; Seco, Joao

2012-09-07

The measurement of prompt gamma rays emitted from proton-induced nuclear reactions has been proposed as a method to verify in vivo the range of a clinical proton radiotherapy beam. A good understanding of the prompt gamma-ray emission during proton therapy is key to develop a clinically feasible technique, as it can facilitate accurate simulations and uncertainty analysis of gamma detector designs. Also, the gamma production cross-sections may be incorporated as prior knowledge in the reconstruction of the proton range from the measurements. In this work, we performed simulations of proton-induced nuclear reactions with the main elements of human tissue, carbon-12, oxygen-16 and nitrogen-14, using the nuclear reaction models of the GEANT4 and MCNP6 Monte Carlo codes and the dedicated nuclear reaction codes TALYS and EMPIRE. For each code, we made an effort to optimize the input parameters and model selection. The results of the models were compared to available experimental data of discrete gamma line cross-sections. Overall, the dedicated nuclear reaction codes reproduced the experimental data more consistently, while the Monte Carlo codes showed larger discrepancies for a number of gamma lines. The model differences lead to a variation of the total gamma production near the end of the proton range by a factor of about 2. These results indicate a need for additional theoretical and experimental study of proton-induced gamma emission in human tissue.

Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Da Cruz, D. F.; Rochman, D.; Koning, A. J.

2012-07-01

This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

Final report to DOE: Matching Grant Program for the Penn State University Nuclear Engineering Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jack S. Brenizer, Jr.

2003-01-17

The DOE/Industry Matching Grant Program is designed to encourage collaborative support for nuclear engineering education as well as research between the nation's nuclear industry and the U.S. Department of Energy (DOE). Despite a serious decline in student enrollments in the 1980s and 1990s, the discipline of nuclear engineering remained important to the advancement of the mission goals of DOE. The program is designed to ensure that academic programs in nuclear engineering are maintained and enhanced in universities throughout the U.S. At Penn State, the Matching Grant Program played a critical role in the survival of the Nuclear Engineering degree programs.more » Funds were used in a variety of ways to support both undergraduate and graduate students directly. Some of these included providing seed funding for new graduate research initiatives, funding the development of new course materials, supporting new teaching facilities, maintenance and purchase of teaching laboratory equipment, and providing undergraduate scholarships, graduate fellowships, and wage payroll positions for students.« less

Early Program Development

NASA Image and Video Library

2004-04-15

This artist's concept illustrates the NERVA (Nuclear Engine for Rocket Vehicle Application) engine's hot bleed cycle in which a small amount of hydrogen gas is diverted from the thrust nozzle, thus eliminating the need for a separate system to drive the turbine. The NERVA engine, based on KIWI nuclear reactor technology, would power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which the Marshall Space Flight Center had development responsibility.

RELAP5-3D Resolution of Known Restart/Backup Issues

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mesina, George L.; Anderson, Nolan A.

2014-12-01

The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequentialmore » verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.« less

Planning U.S. General Purpose Forces: The Theater Nuclear Forces

DTIC Science & Technology

1977-01-01

usefulness in combat. All U.S. nuclear weapons deployed in Europe are fitted with Permissive Action Links (PAL), coded devices designed to impede...may be proposed. The Standard Missile 2, the Harpoon missile, the Mk48 tor- pedo , and the SUBROC anti-submarine rocket are all being considered for...Permissive Action Link . A coded device attached to nuclear weapons deployed abroad that impedes the unauthorized arming or firing of the weapon. Pershing

Radiation Transport Calculation of the UGXR Collimators for the Jules Horowitz Reactor (JHR)

NASA Astrophysics Data System (ADS)

Chento, Yelko; Hueso, César; Zamora, Imanol; Fabbri, Marco; Fuente, Cristina De La; Larringan, Asier

2017-09-01

Jules Horowitz Reactor (JHR), a major infrastructure of European interest in the fission domain, will be built and operated in the framework of an international cooperation, including the development and qualification of materials and nuclear fuel used in nuclear industry. For this purpose UGXR Collimators, two multi slit gamma and X-ray collimation mechatronic systems, will be installed at the JHR pool and at the Irradiated Components Storage pool. Expected amounts of radiation produced by the spent fuel and X-ray accelerator implies diverse aspects need to be verified to ensure adequate radiological zoning and personnel radiation protection. A computational methodology was devised to validate the Collimators design by means of coupling different engineering codes. In summary, several assessments were performed by means of MCNP5v1.60 to fulfil all the radiological requirements in Nominal scenario (TEDE < 25µSv/h) and in Maintenance scenario (TEDE < 2mSv/h) among others, detailing the methodology, hypotheses and assumptions employed.

15 CFR Appendix B to Part 30 - AES Filing Codes

Code of Federal Regulations, 2014 CFR

2014-01-01

... charity FS—Foreign Military Sales ZD—North American Free Trade Agreements (NAFTA) duty deferral shipments...—Validated End User Authorization C58CCD—Consumer Communication Devices C59STA—Strategic Trade Authorization Department of Energy/National Nuclear Security Administration (DOE/NNSA) Codes E01—DOE/NNSA Nuclear...

HINCOF-1: a Code for Hail Ingestion in Engine Inlets

NASA Technical Reports Server (NTRS)

Gopalaswamy, N.; Murthy, S. N. B.

1995-01-01

One of the major concerns during hail ingestion into an engine is the resulting amount and space- and time-wise distribution of hail at the engine face for a given geometry of inlet and set of atmospheric and flight conditions. The appearance of hail in the capture streamtube is invariably random in space and time, with respect to size and momentum. During the motion of a hailstone through an inlet, a hailstone undergoes several processes, namely impact with other hailstones and material surfaces of the inlet and spinner, rolling and rebound following impact; heat and mass transfer; phase change; and shattering, the latter three due to friction and impact. Taking all of these factors into account, a numerical code, designated HINCOF-I, has been developed for determining the motion hailstones from the atmosphere, through an inlet, and up to the engine face. The numerical procedure is based on the Monte-Carlo method. The report presents a description of the code, along with several illustrative cases. The code can be utilized to relate the spinner geometry - conical or, more effective, elliptical - to the possible diversion of hail at the engine face into the bypass stream. The code is also useful for assessing the influence of various hail characteristics on the ingestion and distribution of hailstones over the engine face.

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N., E-mail: zizin@adis.vver.kiae.ru

2010-12-15

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit ofmore » the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

2010-12-01

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

Nuclear Security Education Program at the Pennsylvania State University

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uenlue, Kenan; The Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park, PA 16802-2304; Jovanovic, Igor

The availability of trained and qualified nuclear and radiation security experts worldwide has decreased as those with hands-on experience have retired while the demand for these experts and skills have increased. The U.S. Department of Energy's National Nuclear Security Administration's (NNSA) Global Threat Reduction Initiative (GTRI) has responded to the continued loss of technical and policy expertise amongst personnel and students in the security field by initiating the establishment of a Nuclear Security Education Initiative, in partnership with Pennsylvania State University (PSU), Texas A and M (TAMU), and Massachusetts Institute of Technology (MIT). This collaborative, multi-year initiative forms the basismore » of specific education programs designed to educate the next generation of personnel who plan on careers in the nonproliferation and security fields with both domestic and international focus. The three universities worked collaboratively to develop five core courses consistent with the GTRI mission, policies, and practices. These courses are the following: Global Nuclear Security Policies, Detectors and Source Technologies, Applications of Detectors/Sensors/Sources for Radiation Detection and Measurements Nuclear Security Laboratory, Threat Analysis and Assessment, and Design and Analysis of Security Systems for Nuclear and Radiological Facilities. The Pennsylvania State University (PSU) Nuclear Engineering Program is a leader in undergraduate and graduate-level nuclear engineering education in the USA. The PSU offers undergraduate and graduate programs in nuclear engineering. The PSU undergraduate program in nuclear engineering is the largest nuclear engineering programs in the USA. The PSU Radiation Science and Engineering Center (RSEC) facilities are being used for most of the nuclear security education program activities. Laboratory space and equipment was made available for this purpose. The RSEC facilities include the Penn State Breazeale Reactor (PSBR), gamma irradiation facilities (in-pool irradiator, dry irradiator, and hot cells), neutron beam laboratory, radiochemistry laboratories, and various radiation detection and measurement laboratories. A new nuclear security education laboratory was created with DOE NNSA- GTRI funds at RSEC. The nuclear security graduate level curriculum enables the PSU to educate and train future nuclear security experts, both within the United States as well as worldwide. The nuclear security education program at Penn State will grant a Master's degree in nuclear security starting fall 2015. The PSU developed two courses: Nuclear Security- Detector And Source Technologies and Nuclear Security- Applications of Detectors/Sensors/Sources for Radiation Detection and Measurements (Laboratory). Course descriptions and course topics of these courses are described briefly: - Nuclear Security - Detector and Source Technologies; - Nuclear Security - Applications of Detectors/Sensors/Sources for Radiation Detection and Measurements Laboratory.« less

Computer codes for checking, plotting and processing of neutron cross-section covariance data and their application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sartori, E.; Roussin, R.W.

This paper presents a brief review of computer codes concerned with checking, plotting, processing and using of covariances of neutron cross-section data. It concentrates on those available from the computer code information centers of the United States and the OECD/Nuclear Energy Agency. Emphasis will be placed also on codes using covariances for specific applications such as uncertainty analysis, data adjustment and data consistency analysis. Recent evaluations contain neutron cross section covariance information for all isotopes of major importance for technological applications of nuclear energy. It is therefore important that the available software tools needed for taking advantage of this informationmore » are widely known as hey permit the determination of better safety margins and allow the optimization of more economic, I designs of nuclear energy systems.« less

The Manhattan Project

NASA Astrophysics Data System (ADS)

Reed, B. Cameron

2014-10-01

The Manhattan Project was the United States Army’s program to develop and deploy nuclear weapons during World War II. In these devices, which are known popularly as ‘atomic bombs’, energy is released not by a chemical explosion but by the much more violent process of fission of nuclei of heavy elements via a neutron-mediated chain-reaction. Three years after taking on this project in mid-1942, the Army’s Manhattan Engineer District produced three nuclear bombs of two different designs. Two of these devices were fueled with the 239 isotope of the synthetic element plutonium, while the third employed the rare 235 isotope of uranium. One of the plutonium devices, code-named Trinity, was detonated in a test in southern New Mexico on 16 July 1945; this was the world’s first nuclear explosion. Three weeks later, on 6 August, the uranium bomb, Little Boy, was dropped on the Japanese city of Hiroshima. On 9 August the second plutonium device, Fat Man, was dropped on Nagasaki. Together, the two bombings killed over 100 000 people and were at least partially responsible for the Japanese government’s 14 August decision to surrender. This article surveys, at an undergraduate level, the science and history of the Manhattan Project.

Sandia Engineering Analysis Code Access System v. 2.0.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sjaardema, Gregory D.

The Sandia Engineering Analysis Code Access System (SEACAS) is a suite of preprocessing, post processing, translation, visualization, and utility applications supporting finite element analysis software using the Exodus database file format.

Aerothermo-Structural Analysis of Low Cost Composite Nozzle/Inlet Components

NASA Technical Reports Server (NTRS)

Shivakumar, Kuwigai; Challa, Preeli; Sree, Dave; Reddy, D.

1999-01-01

This research is a cooperative effort among the Turbomachinery and Propulsion Division of NASA Glenn, CCMR of NC A&T State University, and the Tuskegee University. The NC A&T is the lead center and Tuskegee University is the participating institution. Objectives of the research were to develop an integrated aerodynamic, thermal and structural analysis code for design of aircraft engine components, such as, nozzles and inlets made of textile composites; conduct design studies on typical inlets for hypersonic transportation vehicles and setup standards test examples and finally manufacture a scaled down composite inlet. These objectives are accomplished through the following seven tasks: (1) identify the relevant public domain codes for all three types of analysis; (2) evaluate the codes for the accuracy of results and computational efficiency; (3) develop aero-thermal and thermal structural mapping algorithms; (4) integrate all the codes into one single code; (5) write a graphical user interface to improve the user friendliness of the code; (6) conduct test studies for rocket based combined-cycle engine inlet; and finally (7) fabricate a demonstration inlet model using textile preform composites. Tasks one, two and six are being pursued. Selected and evaluated NPARC for flow field analysis, CSTEM for in-depth thermal analysis of inlets and nozzles and FRAC3D for stress analysis. These codes have been independently verified for accuracy and performance. In addition, graphical user interface based on micromechanics analysis for laminated as well as textile composites was developed. Demonstration of this code will be made at the conference. A rocket based combined cycle engine was selected for test studies. Flow field analysis of various inlet geometries were studied. Integration of codes is being continued. The codes developed are being applied to a candidate example of trailblazer engine proposed for space transportation. A successful development of the code will provide a simpler, faster and user-friendly tool for conducting design studies of aircraft and spacecraft engines, applicable in high speed civil transport and space missions.

Mitochondrial genome evolution in the Saccharomyces sensu stricto complex.

PubMed

Ruan, Jiangxing; Cheng, Jian; Zhang, Tongcun; Jiang, Huifeng

2017-01-01

Exploring the evolutionary patterns of mitochondrial genomes is important for our understanding of the Saccharomyces sensu stricto (SSS) group, which is a model system for genomic evolution and ecological analysis. In this study, we first obtained the complete mitochondrial sequences of two important species, Saccharomyces mikatae and Saccharomyces kudriavzevii. We then compared the mitochondrial genomes in the SSS group with those of close relatives, and found that the non-coding regions evolved rapidly, including dramatic expansion of intergenic regions, fast evolution of introns and almost 20-fold higher rearrangement rates than those of the nuclear genomes. However, the coding regions, and especially the protein-coding genes, are more conserved than those in the nuclear genomes of the SSS group. The different evolutionary patterns of coding and non-coding regions in the mitochondrial and nuclear genomes may be related to the origin of the aerobic fermentation lifestyle in this group. Our analysis thus provides novel insights into the evolution of mitochondrial genomes.

Performance assessment of KORAT-3D on the ANL IBM-SP computer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alexeyev, A.V.; Zvenigorodskaya, O.A.; Shagaliev, R.M.

1999-09-01

The TENAR code is currently being developed at the Russian Federal Nuclear Center (VNIIEF) as a coupled dynamics code for the simulation of transients in VVER and RBMK systems and other nuclear systems. The neutronic module in this code system is KORAT-3D. This module is also one of the most computationally intensive components of the code system. A parallel version of KORAT-3D has been implemented to achieve the goal of obtaining transient solutions in reasonable computational time, particularly for RBMK calculations that involve the application of >100,000 nodes. An evaluation of the KORAT-3D code performance was recently undertaken on themore » Argonne National Laboratory (ANL) IBM ScalablePower (SP) parallel computer located in the Mathematics and Computer Science Division of ANL. At the time of the study, the ANL IBM-SP computer had 80 processors. This study was conducted under the auspices of a technical staff exchange program sponsored by the International Nuclear Safety Center (INSC).« less

Dynamic Simulation of a Wave Rotor Topped Turboshaft Engine

NASA Technical Reports Server (NTRS)

Greendyke, R. B.; Paxson, D. E.; Schobeiri, M. T.

1997-01-01

The dynamic behavior of a wave rotor topped turboshaft engine is examined using a numerical simulation. The simulation utilizes an explicit, one-dimensional, multi-passage, CFD based wave rotor code in combination with an implicit, one-dimensional, component level dynamic engine simulation code. Transient responses to rapid fuel flow rate changes and compressor inlet pressure changes are simulated and compared with those of a similarly sized, untopped, turboshaft engine. Results indicate that the wave rotor topped engine responds in a stable, and rapid manner. Furthermore, during certain transient operations, the wave rotor actually tends to enhance engine stability. In particular, there is no tendency toward surge in the compressor of the wave rotor topped engine during rapid acceleration. In fact, the compressor actually moves slightly away from the surge line during this transient. This behavior is precisely the opposite to that of an untopped engine. The simulation is described. Issues associated with integrating CFD and component level codes are discussed. Results from several transient simulations are presented and discussed.

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

NASA Astrophysics Data System (ADS)

Hosseini, Seyed Abolfazl; Afrakoti, Iman Esmaili Paeen

2017-04-01

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The 241Am-9Be and 252Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions.

40 CFR 62.15265 - How do I monitor the load of my municipal waste combustion unit?

Code of Federal Regulations, 2013 CFR

2013-07-01

... Mechanical Engineers (ASME PTC 4.1—1964): Test Code for Steam Generating Units, Power Test Code 4.1-1964... of Mechanical Engineers, Service Center, 22 Law Drive, Post Office Box 2900, Fairfield, NJ 07007. You....archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html. (4) Design, construct...

40 CFR 62.15265 - How do I monitor the load of my municipal waste combustion unit?

Code of Federal Regulations, 2011 CFR

2011-07-01

... Mechanical Engineers (ASME PTC 4.1—1964): Test Code for Steam Generating Units, Power Test Code 4.1-1964... of Mechanical Engineers, Service Center, 22 Law Drive, Post Office Box 2900, Fairfield, NJ 07007. You....archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html. (4) Design, construct...

40 CFR 62.15265 - How do I monitor the load of my municipal waste combustion unit?

Code of Federal Regulations, 2010 CFR

2010-07-01

... Mechanical Engineers (ASME PTC 4.1—1964): Test Code for Steam Generating Units, Power Test Code 4.1-1964... of Mechanical Engineers, Service Center, 22 Law Drive, Post Office Box 2900, Fairfield, NJ 07007. You....archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html. (4) Design, construct...

40 CFR 62.15265 - How do I monitor the load of my municipal waste combustion unit?

Code of Federal Regulations, 2012 CFR

2012-07-01

... Mechanical Engineers (ASME PTC 4.1—1964): Test Code for Steam Generating Units, Power Test Code 4.1-1964... of Mechanical Engineers, Service Center, 22 Law Drive, Post Office Box 2900, Fairfield, NJ 07007. You....archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html. (4) Design, construct...

40 CFR 62.15265 - How do I monitor the load of my municipal waste combustion unit?

Code of Federal Regulations, 2014 CFR

2014-07-01

... Mechanical Engineers (ASME PTC 4.1—1964): Test Code for Steam Generating Units, Power Test Code 4.1-1964... of Mechanical Engineers, Service Center, 22 Law Drive, Post Office Box 2900, Fairfield, NJ 07007. You....archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html. (4) Design, construct...

Survey of Codes Employing Nuclear Damage Assessment

DTIC Science & Technology

1977-10-01

surveyed codes were com- DO 73Mu 1473 ETN OF 1NOVSSSOLETE UNCLASSIFIED 1 SECURITY CLASSIFICATION OF THIS f AGE (Wh*11 Date Efntered)S<>-~C. I UNCLASSIFIED...level and above) TALLEY/TOTEM not nuclear TARTARUS too highly aggregated (battalion level and above) UNICORN highly aggregated force allocation code...vulnerability data can bq input by the user as he receives them, and there is the abil ’ity to replay any situation using hindsight. The age of target

Nuclear Materials Science

NASA Astrophysics Data System (ADS)

Whittle, Karl

2016-06-01

Concerns around global warming have led to a nuclear renaissance in many countries, meanwhile the nuclear industry is warning already of a need to train more nuclear engineers and scientists, who are needed in a range of areas from healthcare and radiation detection to space exploration and advanced materials as well as for the nuclear power industry. Here Karl Whittle provides a solid overview of the intersection of nuclear engineering and materials science at a level approachable by advanced students from materials, engineering and physics. The text explains the unique aspects needed in the design and implementation of materials for use in demanding nuclear settings. In addition to material properties and their interaction with radiation the book covers a range of topics including reactor design, fuels, fusion, future technologies and lessons learned from past incidents. Accompanied by problems, videos and teaching aids the book is suitable for a course text in nuclear materials and a reference for those already working in the field.

Navigation and vessel inspection circular No. 3-94. International maritime organization code for the safe carriage of irradiated nuclear fuel, plutonium and high-level radioactive wastes in flasks on board ships (IMO resolution a.748(18)). Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1994-05-26

The Circular calls the attention of Coast Guard field units, marine surveyors, shippers and carriers of nuclear materials to the International Maritime Organization (IMO) Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board Ships (IMO Resolution A.748(18)).

UMCP-BG and E collaboration in nuclear power engineering in the framework of DOE-Utility Nuclear Power Engineering Education Matching Grant Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wolfe, Lothar PhD

2000-03-01

The DOE-Utility Nuclear Power Engineering Education Matching Grant Program has been established to support the education of students in Nuclear Engineering Programs to maintain a knowledgeable workforce in the United States in order to keep nuclear power as a viable component in a mix of energy sources for the country. The involvement of the utility industry ensures that this grant program satisfies the needs and requirements of local nuclear energy producers and at the same time establishes a strong linkage between education and day-to-day nuclear power generation. As of 1997, seventeen pairs of university-utility partners existed. UMCP was never amore » member of that group of universities, but applied for the first time with a proposal to Baltimore Gas and Electric Company in January 1999 [1]. This proposal was generously granted by BG&E [2,3] in the form of a gift in the amount of $25,000 from BG&E's Corporate Contribution Program. Upon the arrival of a newly appointed Director of Administration in the Department of Materials and Nuclear Engineering, the BG&E check was deposited into the University's Maryland Foundation Fund. The receipt of the letter and the check enabled UMCP to apply for DOE's matching funds in the same amount by a proposal.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov; Tomé, Carlos, E-mail: tome@lanl.gov; Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC)more » polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Visual Computing Environment

NASA Technical Reports Server (NTRS)

Lawrence, Charles; Putt, Charles W.

1997-01-01

The Visual Computing Environment (VCE) is a NASA Lewis Research Center project to develop a framework for intercomponent and multidisciplinary computational simulations. Many current engineering analysis codes simulate various aspects of aircraft engine operation. For example, existing computational fluid dynamics (CFD) codes can model the airflow through individual engine components such as the inlet, compressor, combustor, turbine, or nozzle. Currently, these codes are run in isolation, making intercomponent and complete system simulations very difficult to perform. In addition, management and utilization of these engineering codes for coupled component simulations is a complex, laborious task, requiring substantial experience and effort. To facilitate multicomponent aircraft engine analysis, the CFD Research Corporation (CFDRC) is developing the VCE system. This system, which is part of NASA's Numerical Propulsion Simulation System (NPSS) program, can couple various engineering disciplines, such as CFD, structural analysis, and thermal analysis. The objectives of VCE are to (1) develop a visual computing environment for controlling the execution of individual simulation codes that are running in parallel and are distributed on heterogeneous host machines in a networked environment, (2) develop numerical coupling algorithms for interchanging boundary conditions between codes with arbitrary grid matching and different levels of dimensionality, (3) provide a graphical interface for simulation setup and control, and (4) provide tools for online visualization and plotting. VCE was designed to provide a distributed, object-oriented environment. Mechanisms are provided for creating and manipulating objects, such as grids, boundary conditions, and solution data. This environment includes parallel virtual machine (PVM) for distributed processing. Users can interactively select and couple any set of codes that have been modified to run in a parallel distributed fashion on a cluster of heterogeneous workstations. A scripting facility allows users to dictate the sequence of events that make up the particular simulation.

40 CFR 1048.110 - How must my engines diagnose malfunctions?

Code of Federal Regulations, 2010 CFR

2010-07-01

..., the MIL may stay off during later engine operation. (d) Store trouble codes in computer memory. Record and store in computer memory any diagnostic trouble codes showing a malfunction that should illuminate...

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

Scientific and Technical Publishing at Goddard Space Flight Center in Fiscal Year 1994

NASA Technical Reports Server (NTRS)

1994-01-01

This publication is a compilation of scientific and technical material that was researched, written, prepared, and disseminated by the Center's scientists and engineers during FY94. It is presented in numerical order of the GSFC author's sponsoring technical directorate; i.e., Code 300 is the Office of Flight Assurance, Code 400 is the Flight Projects Directorate, Code 500 is the Mission Operations and Data Systems Directorate, Code 600 is the Space Sciences Directorate, Code 700 is the Engineering Directorate, Code 800 is the Suborbital Projects and Operations Directorate, and Code 900 is the Earth Sciences Directorate. The publication database contains publication or presentation title, author(s), document type, sponsor, and organizational code. This is the second annual compilation for the Center.

Overview of the Graphical User Interface for the GERM Code (GCR Event-Based Risk Model

NASA Technical Reports Server (NTRS)

Kim, Myung-Hee; Cucinotta, Francis A.

2010-01-01

The descriptions of biophysical events from heavy ions are of interest in radiobiology, cancer therapy, and space exploration. The biophysical description of the passage of heavy ions in tissue and shielding materials is best described by a stochastic approach that includes both ion track structure and nuclear interactions. A new computer model called the GCR Event-based Risk Model (GERM) code was developed for the description of biophysical events from heavy ion beams at the NASA Space Radiation Laboratory (NSRL). The GERM code calculates basic physical and biophysical quantities of high-energy protons and heavy ions that have been studied at NSRL for the purpose of simulating space radiobiological effects. For mono-energetic beams, the code evaluates the linear-energy transfer (LET), range (R), and absorption in tissue equivalent material for a given Charge (Z), Mass Number (A) and kinetic energy (E) of an ion. In addition, a set of biophysical properties are evaluated such as the Poisson distribution of ion or delta-ray hits for a specified cellular area, cell survival curves, and mutation and tumor probabilities. The GERM code also calculates the radiation transport of the beam line for either a fixed number of user-specified depths or at multiple positions along the Bragg curve of the particle. The contributions from primary ion and nuclear secondaries are evaluated. The GERM code accounts for the major nuclear interaction processes of importance for describing heavy ion beams, including nuclear fragmentation, elastic scattering, and knockout-cascade processes by using the quantum multiple scattering fragmentation (QMSFRG) model. The QMSFRG model has been shown to be in excellent agreement with available experimental data for nuclear fragmentation cross sections, and has been used by the GERM code for application to thick target experiments. The GERM code provides scientists participating in NSRL experiments with the data needed for the interpretation of their experiments, including the ability to model the beam line, the shielding of samples and sample holders, and the estimates of basic physical and biological outputs of the designed experiments. We present an overview of the GERM code GUI, as well as providing training applications.

HUFF, a One-Dimensional Hydrodynamics Code for Strong Shocks

DTIC Science & Technology

1978-12-01

results for two sample problems. The first problem discussed is a one-kiloton nuclear burst in infinite sea level air. The second problem is the one...of HUFF as an effective first order hydro- dynamic computer code. 1 KT Explosion The one-kiloton nuclear explosion in infinite sea level air was

Computational Fluid Dynamics Analysis Method Developed for Rocket-Based Combined Cycle Engine Inlet

NASA Technical Reports Server (NTRS)

1997-01-01

Renewed interest in hypersonic propulsion systems has led to research programs investigating combined cycle engines that are designed to operate efficiently across the flight regime. The Rocket-Based Combined Cycle Engine is a propulsion system under development at the NASA Lewis Research Center. This engine integrates a high specific impulse, low thrust-to-weight, airbreathing engine with a low-impulse, high thrust-to-weight rocket. From takeoff to Mach 2.5, the engine operates as an air-augmented rocket. At Mach 2.5, the engine becomes a dual-mode ramjet; and beyond Mach 8, the rocket is turned back on. One Rocket-Based Combined Cycle Engine variation known as the "Strut-Jet" concept is being investigated jointly by NASA Lewis, the U.S. Air Force, Gencorp Aerojet, General Applied Science Labs (GASL), and Lockheed Martin Corporation. Work thus far has included wind tunnel experiments and computational fluid dynamics (CFD) investigations with the NPARC code. The CFD method was initiated by modeling the geometry of the Strut-Jet with the GRIDGEN structured grid generator. Grids representing a subscale inlet model and the full-scale demonstrator geometry were constructed. These grids modeled one-half of the symmetric inlet flow path, including the precompression plate, diverter, center duct, side duct, and combustor. After the grid generation, full Navier-Stokes flow simulations were conducted with the NPARC Navier-Stokes code. The Chien low-Reynolds-number k-e turbulence model was employed to simulate the high-speed turbulent flow. Finally, the CFD solutions were postprocessed with a Fortran code. This code provided wall static pressure distributions, pitot pressure distributions, mass flow rates, and internal drag. These results were compared with experimental data from a subscale inlet test for code validation; then they were used to help evaluate the demonstrator engine net thrust.

CSEWG

Science.gov Websites

Nuclear Data Evaluation Co-operation (WPEC) Nuclear Reaction Data Centers, NRDC (IAEA Vienna) EMPIRE , Nuclear Reaction Model Code Atlas of Neutron Resonances The Cross Section Evaluation Working Group (CSEWG

FRENDY: A new nuclear data processing system being developed at JAEA

NASA Astrophysics Data System (ADS)

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

2017-09-01

JAEA has provided an evaluated nuclear data library JENDL and nuclear application codes such as MARBLE, SRAC, MVP and PHITS. These domestic codes have been widely used in many universities and industrial companies in Japan. However, we sometimes find problems in imported processing systems and need to revise them when the new JENDL is released. To overcome such problems and immediately process the nuclear data when it is released, JAEA started developing a new nuclear data processing system, FRENDY in 2013. This paper describes the outline of the development of FRENDY and both its capabilities and performances by the analyses of criticality experiments. The verification results indicate that FRENDY properly generates ACE files.

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2007-01-01

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

Fluid dynamic modeling of junctions in internal combustion engine inlet and exhaust systems

NASA Astrophysics Data System (ADS)

Chalet, David; Chesse, Pascal

2010-10-01

The modeling of inlet and exhaust systems of internal combustion engine is very important in order to evaluate the engine performance. This paper presents new pressure losses models which can be included in a one dimensional engine simulation code. In a first part, a CFD analysis is made in order to show the importance of the density in the modeling approach. Then, the CFD code is used, as a numerical test bench, for the pressure losses models development. These coefficients depend on the geometrical characteristics of the junction and an experimental validation is made with the use of a shock tube test bench. All the models are then included in the engine simulation code of the laboratory. The numerical calculation of unsteady compressible flow, in each pipe of the inlet and exhaust systems, is made and the calculated engine torque is compared with experimental measurements.

The nuclear thermal electric rocket: a proposed innovative propulsion concept for manned interplanetary missions

NASA Astrophysics Data System (ADS)

Dujarric, C.; Santovincenzo, A.; Summerer, L.

2013-03-01

Conventional propulsion technology (chemical and electric) currently limits the possibilities for human space exploration to the neighborhood of the Earth. If farther destinations (such as Mars) are to be reached with humans on board, a more capable interplanetary transfer engine featuring high thrust, high specific impulse is required. The source of energy which could in principle best meet these engine requirements is nuclear thermal. However, the nuclear thermal rocket technology is not yet ready for flight application. The development of new materials which is necessary for the nuclear core will require further testing on ground of full-scale nuclear rocket engines. Such testing is a powerful inhibitor to the nuclear rocket development, as the risks of nuclear contamination of the environment cannot be entirely avoided with current concepts. Alongside already further matured activities in the field of space nuclear power sources for generating on-board power, a low level investigation on nuclear propulsion has been running since long within ESA, and innovative concepts have already been proposed at an IAF conference in 1999 [1, 2]. Following a slow maturation process, a new concept was defined which was submitted to a concurrent design exercise in ESTEC in 2007. Great care was taken in the selection of the design parameters to ensure that this quite innovative concept would in all respects likely be feasible with margins. However, a thorough feasibility demonstration will require a more detailed design including the selection of appropriate materials and the verification that these can withstand the expected mechanical, thermal, and chemical environment. So far, the predefinition work made clear that, based on conservative technology assumptions, a specific impulse of 920 s could be obtained with a thrust of 110 kN. Despite the heavy engine dry mass, a preliminary mission analysis using conservative assumptions showed that the concept was reducing the required Initial Mass in Low Earth Orbit compared to conventional nuclear thermal rockets for a human mission to Mars. Of course, the realization of this concept still requires proper engineering and the dimensioning of quite unconventional machinery. A patent was filed on the concept. Because of the operating parameters of the nuclear core, which are very specific to this type of concept, it seems possible to test on ground this kind of engine at full scale in close loop using a reasonable size test facility with safe and clean conditions. Such tests can be conducted within fully confined enclosure, which would substantially increase the associated inherent nuclear safety levels. This breakthrough removes a showstopper for nuclear rocket engines development. The present paper will disclose the NTER (Nuclear Thermal Electric Rocket) engine concept, will present some of the results of the ESTEC concurrent engineering exercise, and will explain the concept for the NTER on-ground testing facility. Regulations and safety issues related to the development and implementation of the NTER concept will be addressed as well.

WE-AB-204-11: Development of a Nuclear Medicine Dosimetry Module for the GPU-Based Monte Carlo Code ARCHER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, T; Lin, H; Xu, X

Purpose: To develop a nuclear medicine dosimetry module for the GPU-based Monte Carlo code ARCHER. Methods: We have developed a nuclear medicine dosimetry module for the fast Monte Carlo code ARCHER. The coupled electron-photon Monte Carlo transport kernel included in ARCHER is built upon the Dose Planning Method code (DPM). The developed module manages the radioactive decay simulation by consecutively tracking several types of radiation on a per disintegration basis using the statistical sampling method. Optimization techniques such as persistent threads and prefetching are studied and implemented. The developed module is verified against the VIDA code, which is based onmore » Geant4 toolkit and has previously been verified against OLINDA/EXM. A voxelized geometry is used in the preliminary test: a sphere made of ICRP soft tissue is surrounded by a box filled with water. Uniform activity distribution of I-131 is assumed in the sphere. Results: The self-absorption dose factors (mGy/MBqs) of the sphere with varying diameters are calculated by ARCHER and VIDA respectively. ARCHER’s result is in agreement with VIDA’s that are obtained from a previous publication. VIDA takes hours of CPU time to finish the computation, while it takes ARCHER 4.31 seconds for the 12.4-cm uniform activity sphere case. For a fairer CPU-GPU comparison, more effort will be made to eliminate the algorithmic differences. Conclusion: The coupled electron-photon Monte Carlo code ARCHER has been extended to radioactive decay simulation for nuclear medicine dosimetry. The developed code exhibits good performance in our preliminary test. The GPU-based Monte Carlo code is developed with grant support from the National Institute of Biomedical Imaging and Bioengineering through an R01 grant (R01EB015478)« less

Critical evaluation of reverse engineering tool Imagix 4D!

PubMed

Yadav, Rashmi; Patel, Ravindra; Kothari, Abhay

2016-01-01

The comprehension of legacy codes is difficult to understand. Various commercial reengineering tools are available that have unique working styles, and are equipped with their inherent capabilities and shortcomings. The focus of the available tools is in visualizing static behavior not the dynamic one. Therefore, it is difficult for people who work in software product maintenance, code understanding reengineering/reverse engineering. Consequently, the need for a comprehensive reengineering/reverse engineering tool arises. We found the usage of Imagix 4D to be good as it generates the maximum pictorial representations in the form of flow charts, flow graphs, class diagrams, metrics and, to a partial extent, dynamic visualizations. We evaluated Imagix 4D with the help of a case study involving a few samples of source code. The behavior of the tool was analyzed on multiple small codes and a large code gcc C parser. Large code evaluation was performed to uncover dead code, unstructured code, and the effect of not including required files at preprocessing level. The utility of Imagix 4D to prepare decision density and complexity metrics for a large code was found to be useful in getting to know how much reengineering is required. At the outset, Imagix 4D offered limitations in dynamic visualizations, flow chart separation (large code) and parsing loops. The outcome of evaluation will eventually help in upgrading Imagix 4D and posed a need of full featured tools in the area of software reengineering/reverse engineering. It will also help the research community, especially those who are interested in the realm of software reengineering tool building.

Programming (Tips) for Physicists & Engineers

ScienceCinema

Ozcan, Erkcan

2018-02-19

Programming for today's physicists and engineers. Work environment: today's astroparticle, accelerator experiments and information industry rely on large collaborations. Need more than ever: code sharing/resuse, code building--framework integration, documentation and good visualization, working remotely, not reinventing the wheel.

MedlinePlus Connect: Web Application

MedlinePlus

... will result in a query to the MedlinePlus search engine. If you specify a code and the name/ ... system or problem code, will use the MedlinePlus search engine (English only): https://connect.medlineplus.gov/application?mainSearchCriteria. ...

Programming (Tips) for Physicists & Engineers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ozcan, Erkcan

2010-07-13

Programming for today's physicists and engineers. Work environment: today's astroparticle, accelerator experiments and information industry rely on large collaborations. Need more than ever: code sharing/resuse, code building--framework integration, documentation and good visualization, working remotely, not reinventing the wheel.

Design considerations in clustering nuclear rocket engines

NASA Technical Reports Server (NTRS)

Sager, Paul H.

1992-01-01

An initial investigation of the design considerations in clustering nuclear rocket engines for space transfer vehicles has been made. The clustering of both propulsion modules (which include start tanks) and nuclear rocket engines installed directly to a vehicle core tank appears to be feasible. Special provisions to shield opposite run tanks and the opposite side of a core tank - in the case of the boost pump concept - are required; the installation of a circumferential reactor side shield sector appears to provide an effective solution to this problem. While the time response to an engine-out event does not appear to be critical, the gimbal displacement required appears to be important. Since an installation of three engines offers a substantial reduction in gimbal requirements for engine-out and it may be possible to further enhance mission reliability with the greater number of engines, it is recommended that a cluster of four engines be considered.

Design considerations in clustering nuclear rocket engines

NASA Astrophysics Data System (ADS)

Sager, Paul H.

1992-07-01

An initial investigation of the design considerations in clustering nuclear rocket engines for space transfer vehicles has been made. The clustering of both propulsion modules (which include start tanks) and nuclear rocket engines installed directly to a vehicle core tank appears to be feasible. Special provisions to shield opposite run tanks and the opposite side of a core tank - in the case of the boost pump concept - are required; the installation of a circumferential reactor side shield sector appears to provide an effective solution to this problem. While the time response to an engine-out event does not appear to be critical, the gimbal displacement required appears to be important. Since an installation of three engines offers a substantial reduction in gimbal requirements for engine-out and it may be possible to further enhance mission reliability with the greater number of engines, it is recommended that a cluster of four engines be considered.

Performance potential of gas-core and fusion rockets - A mission applications survey.

NASA Technical Reports Server (NTRS)

Fishbach, L. H.; Willis, E. A., Jr.

1971-01-01

This paper reports an evaluation of the performance potential of five nuclear rocket engines for four mission classes. These engines are: the regeneratively cooled gas-core nuclear rocket; the light bulb gas-core nuclear rocket; the space-radiator cooled gas-core nuclear rocket; the fusion rocket; and an advanced solid-core nuclear rocket which is included for comparison. The missions considered are: earth-to-orbit launch; near-earth space missions; close interplanetary missions; and distant interplanetary missions. For each of these missions, the capabilities of each rocket engine type are compared in terms of payload ratio for the earth launch mission or by the initial vehicle mass in earth orbit for space missions (a measure of initial cost). Other factors which might determine the engine choice are discussed. It is shown that a 60 day manned round trip to Mars is conceivable.-

Nuclear electric propulsion technologies - Overview of the NASA/DoE/DoD Nuclear Electric Propulsion Workshop

NASA Technical Reports Server (NTRS)

Barnett, John W.

1991-01-01

Nuclear propulsion technology offers substantial benefits to the ambitious piloted and robotic solar system exploration missions of the Space Exploration Initiative (SEI). This paper summarizes a workshop jointly sponsored by NASA, DoE, and DoD to assess candidate nuclear electric propulsion technologies. Twenty-one power and propulsion concepts are reviewed. Nuclear power concepts include solid and gaseous fuel concepts, with static and dynamic power conversion. Propulsion concepts include steady state and pulsed electromagnetic engines, a pulsed electrothermal engine, and a steady state electrostatic engine. The technologies vary widely in maturity. The workshop review panels concluded that compelling benefits would accrue from the development of nuclear electric propulsion systems, and that a focused, well-funded program is required to prepare the technologies for SEI missions.

Successful Recovery of Nuclear Protein-Coding Genes from Small Insects in Museums Using Illumina Sequencing.

PubMed

Kanda, Kojun; Pflug, James M; Sproul, John S; Dasenko, Mark A; Maddison, David R

2015-01-01

In this paper we explore high-throughput Illumina sequencing of nuclear protein-coding, ribosomal, and mitochondrial genes in small, dried insects stored in natural history collections. We sequenced one tenebrionid beetle and 12 carabid beetles ranging in size from 3.7 to 9.7 mm in length that have been stored in various museums for 4 to 84 years. Although we chose a number of old, small specimens for which we expected low sequence recovery, we successfully recovered at least some low-copy nuclear protein-coding genes from all specimens. For example, in one 56-year-old beetle, 4.4 mm in length, our de novo assembly recovered about 63% of approximately 41,900 nucleotides in a target suite of 67 nuclear protein-coding gene fragments, and 70% using a reference-based assembly. Even in the least successfully sequenced carabid specimen, reference-based assembly yielded fragments that were at least 50% of the target length for 34 of 67 nuclear protein-coding gene fragments. Exploration of alternative references for reference-based assembly revealed few signs of bias created by the reference. For all specimens we recovered almost complete copies of ribosomal and mitochondrial genes. We verified the general accuracy of the sequences through comparisons with sequences obtained from PCR and Sanger sequencing, including of conspecific, fresh specimens, and through phylogenetic analysis that tested the placement of sequences in predicted regions. A few possible inaccuracies in the sequences were detected, but these rarely affected the phylogenetic placement of the samples. Although our sample sizes are low, an exploratory regression study suggests that the dominant factor in predicting success at recovering nuclear protein-coding genes is a high number of Illumina reads, with success at PCR of COI and killing by immersion in ethanol being secondary factors; in analyses of only high-read samples, the primary significant explanatory variable was body length, with small beetles being more successfully sequenced.

Successful Recovery of Nuclear Protein-Coding Genes from Small Insects in Museums Using Illumina Sequencing

PubMed Central

Dasenko, Mark A.

2015-01-01

In this paper we explore high-throughput Illumina sequencing of nuclear protein-coding, ribosomal, and mitochondrial genes in small, dried insects stored in natural history collections. We sequenced one tenebrionid beetle and 12 carabid beetles ranging in size from 3.7 to 9.7 mm in length that have been stored in various museums for 4 to 84 years. Although we chose a number of old, small specimens for which we expected low sequence recovery, we successfully recovered at least some low-copy nuclear protein-coding genes from all specimens. For example, in one 56-year-old beetle, 4.4 mm in length, our de novo assembly recovered about 63% of approximately 41,900 nucleotides in a target suite of 67 nuclear protein-coding gene fragments, and 70% using a reference-based assembly. Even in the least successfully sequenced carabid specimen, reference-based assembly yielded fragments that were at least 50% of the target length for 34 of 67 nuclear protein-coding gene fragments. Exploration of alternative references for reference-based assembly revealed few signs of bias created by the reference. For all specimens we recovered almost complete copies of ribosomal and mitochondrial genes. We verified the general accuracy of the sequences through comparisons with sequences obtained from PCR and Sanger sequencing, including of conspecific, fresh specimens, and through phylogenetic analysis that tested the placement of sequences in predicted regions. A few possible inaccuracies in the sequences were detected, but these rarely affected the phylogenetic placement of the samples. Although our sample sizes are low, an exploratory regression study suggests that the dominant factor in predicting success at recovering nuclear protein-coding genes is a high number of Illumina reads, with success at PCR of COI and killing by immersion in ethanol being secondary factors; in analyses of only high-read samples, the primary significant explanatory variable was body length, with small beetles being more successfully sequenced. PMID:26716693

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Evaluation of Production Cross Sections of Li, Be, B in CR

NASA Technical Reports Server (NTRS)

Moskalenko, I. V.; Mashnik, S. G.

2003-01-01

Accurate evaluation of the production cross section of light elements is important for models of cosmic ray (CR) propagation, galactic chemical evolution, and cosmological studies. However, the experimental spallation cross section data are scarce and often unavailable to CR community while semi-empirical systematics are frequently wrong by a significant factor. Running sophisticated nuclear codes is not an option of choice for everyone either. We use the Los Alamos versions of the Quark-Gluon String Model code LAQGSM and the improved Cascade-Exciton Model code CEM2k together with all available data from Los Alamos Nuclear Laboratory (LANL) nuclear database to produce evaluated production cross sections of isotopes of Li, Be, and B suitable for astrophysical applications. The LAQGSM and CEM2k models have been shown to reproduce well nuclear reactions and hadronic data in the range 0.01-800 GeV/nucleon.

Assessment of the Effects of Entrainment and Wind Shear on Nuclear Cloud Rise Modeling

NASA Astrophysics Data System (ADS)

Zalewski, Daniel; Jodoin, Vincent

2001-04-01

Accurate modeling of nuclear cloud rise is critical in hazard prediction following a nuclear detonation. This thesis recommends improvements to the model currently used by DOD. It considers a single-term versus a three-term entrainment equation, the value of the entrainment and eddy viscous drag parameters, as well as the effect of wind shear in the cloud rise following a nuclear detonation. It examines departures from the 1979 version of the Department of Defense Land Fallout Interpretive Code (DELFIC) with the current code used in the Hazard Prediction and Assessment Capability (HPAC) code version 3.2. The recommendation for a single-term entrainment equation, with constant value parameters, without wind shear corrections, and without cloud oscillations is based on both a statistical analysis using 67 U.S. nuclear atmospheric test shots and the physical representation of the modeling. The statistical analysis optimized the parameter values of interest for four cases: the three-term entrainment equation with wind shear and without wind shear as well as the single-term entrainment equation with and without wind shear. The thesis then examines the effect of cloud oscillations as a significant departure in the code. Modifications to user input atmospheric tables are identified as a potential problem in the calculation of stabilized cloud dimensions in HPAC.

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Fission Activities of the Nuclear Reactions Group in Uppsala

NASA Astrophysics Data System (ADS)

Al-Adili, A.; Alhassan, E.; Gustavsson, C.; Helgesson, P.; Jansson, K.; Koning, A.; Lantz, M.; Mattera, A.; Prokofiev, A. V.; Rakopoulos, V.; Sjöstrand, H.; Solders, A.; Tarrío, D.; Österlund, M.; Pomp, S.

This paper highlights some of the main activities related to fission of the nuclear reactions group at Uppsala University. The group is involved for instance in fission yield experiments at the IGISOL facility, cross-section measurements at the NFS facility, as well as fission dynamics studies at the IRMM JRC-EC. Moreover, work is ongoing on the Total Monte Carlo (TMC) methodology and on including the GEF fission code into the TALYS nuclear reaction code. Selected results from these projects are discussed.

Structural mechanics simulations

NASA Technical Reports Server (NTRS)

Biffle, Johnny H.

1992-01-01

Sandia National Laboratory has a very broad structural capability. Work has been performed in support of reentry vehicles, nuclear reactor safety, weapons systems and components, nuclear waste transport, strategic petroleum reserve, nuclear waste storage, wind and solar energy, drilling technology, and submarine programs. The analysis environment contains both commercial and internally developed software. Included are mesh generation capabilities, structural simulation codes, and visual codes for examining simulation results. To effectively simulate a wide variety of physical phenomena, a large number of constitutive models have been developed.

Creep of A508/533 Pressure Vessel Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard Wright

2014-08-01

ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with themore » very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.« less

75 FR 45171 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-02

... Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555...-3040. This guide describes some engineering practices and methods generally considered by the NRC to be... they reflect the latest general engineering approaches that are acceptable to the NRC staff. If future...

76 FR 50275 - Guidance for the Assessment of Beyond-Design-Basis Aircraft Impacts

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-12

...: Mekonen M. Bayssie, Regulatory Guide Development Branch, Division of Engineering, Office of Nuclear... e-mail to [email protected] . SUPPLEMENTARY INFORMATION: I. Introduction The Nuclear..., Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc. 2011-20513 Filed 8-11-11; 8:45 am...

Subscale Validation of the Subsurface Active Filtration of Exhaust (SAFE) Approach to NTP Ground Testing

NASA Technical Reports Server (NTRS)

Marshall, William M.; Borowski, Stanley K.; Bulman, Mel; Joyner, Russell; Martin, Charles R.

2015-01-01

Brief History of NTP: Project Rover Began in 1950s by Los Alamos Scientific Labs (now Los Alamos National Labs) and ran until 1970s Tested a series of nuclear reactor engines of varying size at Nevada Test Site (now Nevada National Security Site) Ranged in scale from 111 kN (25 klbf) to 1.1 MN (250 klbf) Included Nuclear Furnace-1 tests Demonstrated the viability and capability of a nuclear rocket engine test program One of Kennedys 4 goals during famous moon speech to Congress Nuclear Engines for Rocket Vehicle Applications (NERVA) Atomic Energy Commission and NASA joint venture started in 1964 Parallel effort to Project Rover was focused on technology demonstration Tested XE engine, a 245-kN (55-klbf) engine to demonstrate startup shutdown sequencing. Hot-hydrogen stream is passed directly through fuel elements potential for radioactive material to be eroded into gaseous fuel flow as identified in previous programs NERVA and Project Rover (1950s-70s) were able to test in open atmosphere similar to conventional rocket engine test stands today Nuclear Furance-1 tests employed a full scrubber system Increased government and environmental regulations prohibit the modern testing in open atmosphere. Since the 1960s, there has been an increasing cessation on open air testing of nuclear material Political and national security concerns further compound the regulatory environment

Just sustainability? Sustainability and social justice in professional codes of ethics for engineers.

PubMed

Brauer, Cletus S

2013-09-01

Should environmental, social, and economic sustainability be of primary concern to engineers? Should social justice be among these concerns? Although the deterioration of our natural environment and the increase in social injustices are among today's most pressing and important issues, engineering codes of ethics and their paramountcy clause, which contains those values most important to engineering and to what it means to be an engineer, do not yet put either concept on a par with the safety, health, and welfare of the public. This paper addresses a recent proposal by Michelfelder and Jones (2011) to include sustainability in the paramountcy clause as a way of rectifying the current disregard for social justice issues in the engineering codes. That proposal builds on a certain notion of sustainability that includes social justice as one of its dimensions and claims that social justice is a necessary condition for sustainability, not vice versa. The relationship between these concepts is discussed, and the original proposal is rejected. Drawing on insights developed throughout the paper, some suggestions are made as to how one should address the different requirements that theory and practice demand of the value taxonomy of professional codes of ethics.

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Pullout of a Rigid Insert Adhesively Bonded to an Elastic Half Plane.

DTIC Science & Technology

1983-12-01

COMMAND UNITED STATES AIR FORCE C-= °84 02 13 071. C,, W % d 6 This document was prepared by the Department of Engineering Mechanics, USAF Academy Faculty...THOMAS E. KULLGREN, Lt Col, USAF Project Engineer /Scientist Professor and Acting Head, Department of Engineering Mechanics KENNETH E. SIEGETH Lt Col...Department of Engineering (Ifapphicable) Mechanics USAFA/DFEM 6c. ADDRESS (City. State and ZIP Code) 7b. ADDRESS (City, Slate and ZIP Code) USAF Academy

Experimental validation of depletion calculations with VESTA 2.1.5 using JEFF-3.2

NASA Astrophysics Data System (ADS)

Haeck, Wim; Ichou, Raphaëlle

2017-09-01

The removal of decay heat is a significant safety concern in nuclear engineering for the operation of a nuclear reactor both in normal and accidental conditions and for intermediate and long term waste storage facilities. The correct evaluation of the decay heat produced by an irradiated material requires first of all the calculation of the composition of the irradiated material by depletion codes such as VESTA 2.1, currently under development at IRSN in France. A set of PWR assembly decay heat measurements performed by the Swedish Central Interim Storage Facility (CLAB) located in Oskarshamm (Sweden) have been calculated using different nuclear data libraries: ENDF/B-VII.0, JEFF-3.1, JEFF-3.2 and JEFF-3.3T1. Using these nuclear data libraries, VESTA 2.1 calculates the assembly decay heat for almost all cases within 4% of the measured decay heat. On average, the ENDF/B-VII.0 calculated decay heat values appear to give a systematic underestimation of only 0.5%. When using the JEFF-3.1 library, this results a systematic underestimation of about 2%. By switching to the JEFF-3.2 library, this systematic underestimation is improved slighty (up to 1.5%). The changes made in the JEFF-3.3T1 beta library appear to be overcorrecting, as the systematic underestimation is transformed into a systematic overestimation of about 1.5%.

Optimized scalar promotion with load and splat SIMD instructions

DOEpatents

Eichenberger, Alexander E; Gschwind, Michael K; Gunnels, John A

2013-10-29

Mechanisms for optimizing scalar code executed on a single instruction multiple data (SIMD) engine are provided. Placement of vector operation-splat operations may be determined based on an identification of scalar and SIMD operations in an original code representation. The original code representation may be modified to insert the vector operation-splat operations based on the determined placement of vector operation-splat operations to generate a first modified code representation. Placement of separate splat operations may be determined based on identification of scalar and SIMD operations in the first modified code representation. The first modified code representation may be modified to insert or delete separate splat operations based on the determined placement of the separate splat operations to generate a second modified code representation. SIMD code may be output based on the second modified code representation for execution by the SIMD engine.

Optimized scalar promotion with load and splat SIMD instructions

DOEpatents

Eichenberger, Alexandre E [Chappaqua, NY; Gschwind, Michael K [Chappaqua, NY; Gunnels, John A [Yorktown Heights, NY

2012-08-28

Mechanisms for optimizing scalar code executed on a single instruction multiple data (SIMD) engine are provided. Placement of vector operation-splat operations may be determined based on an identification of scalar and SIMD operations in an original code representation. The original code representation may be modified to insert the vector operation-splat operations based on the determined placement of vector operation-splat operations to generate a first modified code representation. Placement of separate splat operations may be determined based on identification of scalar and SIMD operations in the first modified code representation. The first modified code representation may be modified to insert or delete separate splat operations based on the determined placement of the separate splat operations to generate a second modified code representation. SIMD code may be output based on the second modified code representation for execution by the SIMD engine.

Second International Workshop on Software Engineering and Code Design in Parallel Meteorological and Oceanographic Applications

NASA Technical Reports Server (NTRS)

OKeefe, Matthew (Editor); Kerr, Christopher L. (Editor)

1998-01-01

This report contains the abstracts and technical papers from the Second International Workshop on Software Engineering and Code Design in Parallel Meteorological and Oceanographic Applications, held June 15-18, 1998, in Scottsdale, Arizona. The purpose of the workshop is to bring together software developers in meteorology and oceanography to discuss software engineering and code design issues for parallel architectures, including Massively Parallel Processors (MPP's), Parallel Vector Processors (PVP's), Symmetric Multi-Processors (SMP's), Distributed Shared Memory (DSM) multi-processors, and clusters. Issues to be discussed include: (1) code architectures for current parallel models, including basic data structures, storage allocation, variable naming conventions, coding rules and styles, i/o and pre/post-processing of data; (2) designing modular code; (3) load balancing and domain decomposition; (4) techniques that exploit parallelism efficiently yet hide the machine-related details from the programmer; (5) tools for making the programmer more productive; and (6) the proliferation of programming models (F--, OpenMP, MPI, and HPF).

A Program for Cultivating Nuclear Talent at Engineering Educational Institute in a Remote Area from Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Takahashi, Tsuyoshi

Recently, in Japan, the number of students who hope for finding employment at the nuclear power company has decreased as students‧ concern for the nuclear power industry decreases. To improve the situation, Ministry of Education, Culture, Sports, Science and Technology launched the program of cultivating talent for nuclear power which supports research and education of nuclear power in the academic year of 2007. Supported by the program, Kushiro College of Technology conducted several activities concerning nuclear power for about a year. The students came to be interested in nuclear engineering through these activities and its results.

Research Prototype: Automated Analysis of Scientific and Engineering Semantics

NASA Technical Reports Server (NTRS)

Stewart, Mark E. M.; Follen, Greg (Technical Monitor)

2001-01-01

Physical and mathematical formulae and concepts are fundamental elements of scientific and engineering software. These classical equations and methods are time tested, universally accepted, and relatively unambiguous. The existence of this classical ontology suggests an ideal problem for automated comprehension. This problem is further motivated by the pervasive use of scientific code and high code development costs. To investigate code comprehension in this classical knowledge domain, a research prototype has been developed. The prototype incorporates scientific domain knowledge to recognize code properties (including units, physical, and mathematical quantity). Also, the procedure implements programming language semantics to propagate these properties through the code. This prototype's ability to elucidate code and detect errors will be demonstrated with state of the art scientific codes.

An Experiment in Scientific Code Semantic Analysis

NASA Technical Reports Server (NTRS)

Stewart, Mark E. M.

1998-01-01

This paper concerns a procedure that analyzes aspects of the meaning or semantics of scientific and engineering code. This procedure involves taking a user's existing code, adding semantic declarations for some primitive variables, and parsing this annotated code using multiple, distributed expert parsers. These semantic parser are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. The parsers will automatically recognize and document some static, semantic concepts and locate some program semantic errors. Results are shown for a subroutine test case and a collection of combustion code routines. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

ODECS -- A computer code for the optimal design of S.I. engine control strategies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arsie, I.; Pianese, C.; Rizzo, G.

1996-09-01

The computer code ODECS (Optimal Design of Engine Control Strategies) for the design of Spark Ignition engine control strategies is presented. This code has been developed starting from the author`s activity in this field, availing of some original contributions about engine stochastic optimization and dynamical models. This code has a modular structure and is composed of a user interface for the definition, the execution and the analysis of different computations performed with 4 independent modules. These modules allow the following calculations: (1) definition of the engine mathematical model from steady-state experimental data; (2) engine cycle test trajectory corresponding to amore » vehicle transient simulation test such as ECE15 or FTP drive test schedule; (3) evaluation of the optimal engine control maps with a steady-state approach; (4) engine dynamic cycle simulation and optimization of static control maps and/or dynamic compensation strategies, taking into account dynamical effects due to the unsteady fluxes of air and fuel and the influences of combustion chamber wall thermal inertia on fuel consumption and emissions. Moreover, in the last two modules it is possible to account for errors generated by a non-deterministic behavior of sensors and actuators and the related influences on global engine performances, and compute robust strategies, less sensitive to stochastic effects. In the paper the four models are described together with significant results corresponding to the simulation and the calculation of optimal control strategies for dynamic transient tests.« less

Comparative analysis of cooling systems for energy equipment of combined heat and power plants and nuclear power plants

NASA Astrophysics Data System (ADS)

Reutov, B. F.; Lazarev, M. V.; Ermakova, S. V.; Zisman, S. L.; Kaplanovich, L. S.; Svetushkov, V. V.

2016-07-01

In the 20th century, the thermal power engineering in this country was oriented toward oncethrough cooling systems. More than 50% of the CHPP and NPP capacities with once-through cooling systems put into operation before the 1990s were large-scale water consumers but with minimum irretrievable water consumption. In 1995, the Water Code of the Russian Federation was adopted in which restrictions on application of once-through cooling systems for newly designed combined heat and power plants (CHPPs) were introduced for the first time. A ban on application of once-through systems was imposed by the current Water Code of the Russian Federation (Federal law no. 74-FZ, Art. 60 Cl. 4) not only for new CHPPs but also for those to be modified. Clause 4 of Article 60 of the Water Code of the Russian Federation contravenes law no. 7-FZ "On Protection of the Environment" that has priority significance, since the water environment is only part of the natural environment and those articles of the Water Code of the Russian Federation that are related directly to electric power engineering, viz., Articles 46 and 62. In recent decades, the search for means to increase revenue charges and the economic pressure on the thermal power industry caused introduction by law of charges for use of water by cooling systems irrespective of the latter's impact on the water quality of the source, the environment, the economic efficiency of the power production, and the living conditions of the people. The long-range annual increase in the water use charges forces the power generating companies to switch transfer once-through service water supply installations to recirculating water supply systems and once-through-recirculating systems with multiple reuse of warm water, which drastically reduces the technical, economic, and ecological characteristic of the power plant operation and also results in increasing power rates for the population. This work comprehensively substantiates the demands of power engineering specialists that the ban on development and construction of once-through service water supply systems should be lifted and the proposals for new parameters, e.g., temperature and back pressure, for designing low-potential equipment of steam-gas and steam-power plants.

Overview of High-Fidelity Modeling Activities in the Numerical Propulsion System Simulations (NPSS) Project

NASA Technical Reports Server (NTRS)

Veres, Joseph P.

2002-01-01

A high-fidelity simulation of a commercial turbofan engine has been created as part of the Numerical Propulsion System Simulation Project. The high-fidelity computer simulation utilizes computer models that were developed at NASA Glenn Research Center in cooperation with turbofan engine manufacturers. The average-passage (APNASA) Navier-Stokes based viscous flow computer code is used to simulate the 3D flow in the compressors and turbines of the advanced commercial turbofan engine. The 3D National Combustion Code (NCC) is used to simulate the flow and chemistry in the advanced aircraft combustor. The APNASA turbomachinery code and the NCC combustor code exchange boundary conditions at the interface planes at the combustor inlet and exit. This computer simulation technique can evaluate engine performance at steady operating conditions. The 3D flow models provide detailed knowledge of the airflow within the fan and compressor, the high and low pressure turbines, and the flow and chemistry within the combustor. The models simulate the performance of the engine at operating conditions that include sea level takeoff and the altitude cruise condition.

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Karyopherin-mediated nuclear import of the homing endonuclease VMA1-derived endonuclease is required for self-propagation of the coding region.

PubMed

Nagai, Yuri; Nogami, Satoru; Kumagai-Sano, Fumi; Ohya, Yoshikazu

2003-03-01

VMA1-derived endonuclease (VDE), a site-specific endonuclease in Saccharomyces cerevisiae, enters the nucleus to generate a double-strand break in the VDE-negative allelic locus, mediating the self-propagating gene conversion called homing. Although VDE is excluded from the nucleus in mitotic cells, it relocalizes at premeiosis, becoming localized in both the nucleus and the cytoplasm in meiosis. The nuclear localization of VDE is induced by inactivation of TOR kinases, which constitute central regulators of cell differentiation in S. cerevisiae, and by nutrient depletion. A functional genomic approach revealed that at least two karyopherins, Srp1p and Kap142p, are required for the nuclear localization pattern. Genetic and physical interactions between Srp1p and VDE imply direct involvement of karyopherin-mediated nuclear transport in this process. Inactivation of TOR signaling or acquisition of an extra nuclear localization signal in the VDE coding region leads to artificial nuclear localization of VDE and thereby induces homing even during mitosis. These results serve as evidence that VDE utilizes the host systems of nutrient signal transduction and nucleocytoplasmic transport to ensure the propagation of its coding region.

Karyopherin-Mediated Nuclear Import of the Homing Endonuclease VMA1-Derived Endonuclease Is Required for Self-Propagation of the Coding Region

PubMed Central

Nagai, Yuri; Nogami, Satoru; Kumagai-Sano, Fumi; Ohya, Yoshikazu

2003-01-01

VMA1-derived endonuclease (VDE), a site-specific endonuclease in Saccharomyces cerevisiae, enters the nucleus to generate a double-strand break in the VDE-negative allelic locus, mediating the self-propagating gene conversion called homing. Although VDE is excluded from the nucleus in mitotic cells, it relocalizes at premeiosis, becoming localized in both the nucleus and the cytoplasm in meiosis. The nuclear localization of VDE is induced by inactivation of TOR kinases, which constitute central regulators of cell differentiation in S. cerevisiae, and by nutrient depletion. A functional genomic approach revealed that at least two karyopherins, Srp1p and Kap142p, are required for the nuclear localization pattern. Genetic and physical interactions between Srp1p and VDE imply direct involvement of karyopherin-mediated nuclear transport in this process. Inactivation of TOR signaling or acquisition of an extra nuclear localization signal in the VDE coding region leads to artificial nuclear localization of VDE and thereby induces homing even during mitosis. These results serve as evidence that VDE utilizes the host systems of nutrient signal transduction and nucleocytoplasmic transport to ensure the propagation of its coding region. PMID:12588991

Comparison of Engine Cycle Codes for Rocket-Based Combined Cycle Engines

NASA Technical Reports Server (NTRS)

Waltrup, Paul J.; Auslender, Aaron H.; Bradford, John E.; Carreiro, Louis R.; Gettinger, Christopher; Komar, D. R.; McDonald, J.; Snyder, Christopher A.

2002-01-01

This paper summarizes the results from a one day workshop on Rocket-Based Combined Cycle (RBCC) Engine Cycle Codes held in Monterey CA in November of 2000 at the 2000 JANNAF JPM with the authors as primary participants. The objectives of the workshop were to discuss and compare the merits of existing Rocket-Based Combined Cycle (RBCC) engine cycle codes being used by government and industry to predict RBCC engine performance and interpret experimental results. These merits included physical and chemical modeling, accuracy and user friendliness. The ultimate purpose of the workshop was to identify the best codes for analyzing RBCC engines and to document any potential shortcomings, not to demonstrate the merits or deficiencies of any particular engine design. Five cases representative of the operating regimes of typical RBCC engines were used as the basis of these comparisons. These included Mach 0 sea level static and Mach 1.0 and Mach 2.5 Air-Augmented-Rocket (AAR), Mach 4 subsonic combustion ramjet or dual-mode scramjet, and Mach 8 scramjet operating modes. Specification of a generic RBCC engine geometry and concomitant component operating efficiencies, bypass ratios, fuel/oxidizer/air equivalence ratios and flight dynamic pressures were provided. The engine included an air inlet, isolator duct, axial rocket motor/injector, axial wall fuel injectors, diverging combustor, and exit nozzle. Gaseous hydrogen was used as the fuel with the rocket portion of the system using a gaseous H2/O2 propellant system to avoid cryogenic issues. The results of the workshop, even after post-workshop adjudication of differences, were surprising. They showed that the codes predicted essentially the same performance at the Mach 0 and I conditions, but progressively diverged from a common value (for example, for fuel specific impulse, Isp) as the flight Mach number increased, with the largest differences at Mach 8. The example cases and results are compared and discussed in this paper.

Performance Criteria of Nuclear Space Propulsion Systems

NASA Astrophysics Data System (ADS)

Shepherd, L. R.

Future exploration of the solar system on a major scale will require propulsion systems capable of performance far greater than is achievable with the present generation of rocket engines using chemical propellants. Viable missions going deeper into interstellar space will be even more demanding. Propulsion systems based on nuclear energy sources, fission or (eventually) fusion offer the best prospect for meeting the requirements. The most obvious gain coming from the application of nuclear reactions is the possibility, at least in principle, of obtaining specific impulses a thousandfold greater than can be achieved in chemically energised rockets. However, practical considerations preclude the possibility of exploiting the full potential of nuclear energy sources in any engines conceivable in terms of presently known technology. Achievable propulsive power is a particularly limiting factor, since this determines the acceleration that may be obtained. Conventional chemical rocket engines have specific propulsive powers (power per unit engine mass) in the order of gigawatts per tonne. One cannot envisage the possibility of approaching such a level of performance by orders of magnitude in presently conceivable nuclear propulsive systems. The time taken, under power, to reach a given terminal velocity is proportional to the square of the engine's exhaust velocity and the inverse of its specific power. An assessment of various nuclear propulsion concepts suggests that, even with the most optimistic assumptions, it could take many hundreds of years to attain the velocities necessary to reach the nearest stars. Exploration within a range of the order of a thousand AU, however, would appear to offer viable prospects, even with the low levels of specific power of presently conceivable nuclear engines.

Nuclear engineering enrollments and degrees, 1981

DOE Office of Scientific and Technical Information (OSTI.GOV)

Little, J R; Shirley, D L

1982-05-01

This report presents data on the number of students enrolled and the degrees awarded in academic year 1980-81 from 73 US institutions offering degree programs in nuclear engineering or nuclear options within other engineering fields. Presented here are historical data for the last decade, which provide information such as trends by degree level, foreign national student participation, female and minority student participation, and placement of graduates. Also included is a listing of the universities by type of program and number of students.

Copper Doping of Zinc Oxide by Nuclear Transmutation

DTIC Science & Technology

2014-03-27

Copper Doping of Zinc Oxide by Nuclear Transmutation THESIS Matthew C. Recker, Captain, USAF AFIT-ENP-14-M-30 DEPARTMENT OF THE AIR FORCE AIR...NUCLEAR TRANSMUTATION THESIS Presented to the Faculty Department of Engineering Physics Graduate School of Engineering and Management Air Force...COPPER DOPING OF ZINC OXIDE BY NUCLEAR TRANSMUTATION Matthew C. Recker, BS Captain, USAF Approved: //signed// 27 February 2014 John W. McClory, PhD

Nuclear Thermal Propulsion (NTP) Development Activities at the NASA Marshall Space Flight Center - 2006 Accomplishments

NASA Technical Reports Server (NTRS)

Ballard, Richard O.

2007-01-01

In 2005-06, the Prometheus program funded a number of tasks at the NASA-Marshall Space Flight Center (MSFC) to support development of a Nuclear Thermal Propulsion (NTP) system for future manned exploration missions. These tasks include the following: 1. NTP Design Develop Test & Evaluate (DDT&E) Planning 2. NTP Mission & Systems Analysis / Stage Concepts & Engine Requirements 3. NTP Engine System Trade Space Analysis and Studies 4. NTP Engine Ground Test Facility Assessment 5. Non-Nuclear Environmental Simulator (NTREES) 6. Non-Nuclear Materials Fabrication & Evaluation 7. Multi-Physics TCA Modeling. This presentation is a overview of these tasks and their accomplishments

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

World Energy Data System (WENDS). Volume X. Nuclear facility profiles, PO--ZA. [Brief tabulated information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then bymore » facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile.« less

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

DOE Office of Scientific and Technical Information (OSTI.GOV)

Susan Stacy; Hollie K. Gilbert

2005-02-01

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to housemore » the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.« less

Nonlinear dynamic simulation of single- and multi-spool core engines

NASA Technical Reports Server (NTRS)

Schobeiri, T.; Lippke, C.; Abouelkheir, M.

1993-01-01

In this paper a new computational method for accurate simulation of the nonlinear dynamic behavior of single- and multi-spool core engines, turbofan engines, and power generation gas turbine engines is presented. In order to perform the simulation, a modularly structured computer code has been developed which includes individual mathematical modules representing various engine components. The generic structure of the code enables the dynamic simulation of arbitrary engine configurations ranging from single-spool thrust generation to multi-spool thrust/power generation engines under adverse dynamic operating conditions. For precise simulation of turbine and compressor components, row-by-row calculation procedures were implemented that account for the specific turbine and compressor cascade and blade geometry and characteristics. The dynamic behavior of the subject engine is calculated by solving a number of systems of partial differential equations, which describe the unsteady behavior of the individual components. In order to ensure the capability, accuracy, robustness, and reliability of the code, comprehensive critical performance assessment and validation tests were performed. As representatives, three different transient cases with single- and multi-spool thrust and power generation engines were simulated. The transient cases range from operating with a prescribed fuel schedule, to extreme load changes, to generator and turbine shut down.

Parametric Model of an Aerospike Rocket Engine

NASA Technical Reports Server (NTRS)

Korte, J. J.

2000-01-01

A suite of computer codes was assembled to simulate the performance of an aerospike engine and to generate the engine input for the Program to Optimize Simulated Trajectories. First an engine simulator module was developed that predicts the aerospike engine performance for a given mixture ratio, power level, thrust vectoring level, and altitude. This module was then used to rapidly generate the aerospike engine performance tables for axial thrust, normal thrust, pitching moment, and specific thrust. Parametric engine geometry was defined for use with the engine simulator module. The parametric model was also integrated into the iSIGHTI multidisciplinary framework so that alternate designs could be determined. The computer codes were used to support in-house conceptual studies of reusable launch vehicle designs.

Parametric Model of an Aerospike Rocket Engine

NASA Technical Reports Server (NTRS)

Korte, J. J.

2000-01-01

A suite of computer codes was assembled to simulate the performance of an aerospike engine and to generate the engine input for the Program to Optimize Simulated Trajectories. First an engine simulator module was developed that predicts the aerospike engine performance for a given mixture ratio, power level, thrust vectoring level, and altitude. This module was then used to rapidly generate the aerospike engine performance tables for axial thrust, normal thrust, pitching moment, and specific thrust. Parametric engine geometry was defined for use with the engine simulator module. The parametric model was also integrated into the iSIGHT multidisciplinary framework so that alternate designs could be determined. The computer codes were used to support in-house conceptual studies of reusable launch vehicle designs.

Program user's manual for optimizing the design of a liquid or gaseous propellant rocket engine with the automated combustor design code AUTOCOM

NASA Technical Reports Server (NTRS)

Reichel, R. H.; Hague, D. S.; Jones, R. T.; Glatt, C. R.

1973-01-01

This computer program manual describes in two parts the automated combustor design optimization code AUTOCOM. The program code is written in the FORTRAN 4 language. The input data setup and the program outputs are described, and a sample engine case is discussed. The program structure and programming techniques are also described, along with AUTOCOM program analysis.

Three-dimensional pin-to-pin analyses of VVER-440 cores by the MOBY-DICK code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lehmann, M.; Mikolas, P.

1994-12-31

Nuclear design for the Dukovany (EDU) VVER-440s nuclear power plant is routinely performed by the MOBY-DICK system. After its implementation on Hewlett Packard series 700 workstations, it is able to perform routinely three-dimensional pin-to-pin core analyses. For purposes of code validation, the benchmark prepared from EDU operational data was solved.

EASY-II Renaissance: n, p, d, α, γ-induced Inventory Code System

NASA Astrophysics Data System (ADS)

Sublet, J.-Ch.; Eastwood, J. W.; Morgan, J. G.

2014-04-01

The European Activation SYstem has been re-engineered and re-written in modern programming languages so as to answer today's and tomorrow's needs in terms of activation, transmutation, depletion, decay and processing of radioactive materials. The new FISPACT-II inventory code development project has allowed us to embed many more features in terms of energy range: up to GeV; incident particles: alpha, gamma, proton, deuteron and neutron; and neutron physics: self-shielding effects, temperature dependence and covariance, so as to cover all anticipated application needs: nuclear fission and fusion, accelerator physics, isotope production, stockpile and fuel cycle stewardship, materials characterization and life, and storage cycle management. In parallel, the maturity of modern, truly general purpose libraries encompassing thousands of target isotopes such as TENDL-2012, the evolution of the ENDF-6 format and the capabilities of the latest generation of processing codes PREPRO, NJOY and CALENDF have allowed the activation code to be fed with more robust, complete and appropriate data: cross sections with covariance, probability tables in the resonance ranges, kerma, dpa, gas and radionuclide production and 24 decay types. All such data for the five most important incident particles (n, p, d, α, γ), are placed in evaluated data files up to an incident energy of 200 MeV. The resulting code system, EASY-II is designed as a functional replacement for the previous European Activation System, EASY-2010. It includes many new features and enhancements, but also benefits already from the feedback from extensive validation and verification activities performed with its predecessor.

Engine dynamic analysis with general nonlinear finite element codes. Part 2: Bearing element implementation overall numerical characteristics and benchmaking

NASA Technical Reports Server (NTRS)

Padovan, J.; Adams, M.; Fertis, J.; Zeid, I.; Lam, P.

1982-01-01

Finite element codes are used in modelling rotor-bearing-stator structure common to the turbine industry. Engine dynamic simulation is used by developing strategies which enable the use of available finite element codes. benchmarking the elements developed are benchmarked by incorporation into a general purpose code (ADINA); the numerical characteristics of finite element type rotor-bearing-stator simulations are evaluated through the use of various types of explicit/implicit numerical integration operators. Improving the overall numerical efficiency of the procedure is improved.

Automatic mathematical modeling for space application

NASA Technical Reports Server (NTRS)

Wang, Caroline K.

1987-01-01

A methodology for automatic mathematical modeling is described. The major objective is to create a very friendly environment for engineers to design, maintain and verify their model and also automatically convert the mathematical model into FORTRAN code for conventional computation. A demonstration program was designed for modeling the Space Shuttle Main Engine simulation mathematical model called Propulsion System Automatic Modeling (PSAM). PSAM provides a very friendly and well organized environment for engineers to build a knowledge base for base equations and general information. PSAM contains an initial set of component process elements for the Space Shuttle Main Engine simulation and a questionnaire that allows the engineer to answer a set of questions to specify a particular model. PSAM is then able to automatically generate the model and the FORTRAN code. A future goal is to download the FORTRAN code to the VAX/VMS system for conventional computation.

Parameter study of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239c

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, W.W.; Layton, J.P.

1976-09-13

The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The NUROC3A systems analysis code was designed to provide the user with performance characteristics of the dual-mode system. Volume 3 describes utilization of the NUROC3A code to produce a detailed parameter study of the system.

Seals Code Development Workshop

NASA Technical Reports Server (NTRS)

Hendricks, Robert C. (Compiler); Liang, Anita D. (Compiler)

1996-01-01

Seals Workshop of 1995 industrial code (INDSEAL) release include ICYL, GCYLT, IFACE, GFACE, SPIRALG, SPIRALI, DYSEAL, and KTK. The scientific code (SCISEAL) release includes conjugate heat transfer and multidomain with rotordynamic capability. Several seals and bearings codes (e.g., HYDROFLEX, HYDROTRAN, HYDROB3D, FLOWCON1, FLOWCON2) are presented and results compared. Current computational and experimental emphasis includes multiple connected cavity flows with goals of reducing parasitic losses and gas ingestion. Labyrinth seals continue to play a significant role in sealing with face, honeycomb, and new sealing concepts under investigation for advanced engine concepts in view of strict environmental constraints. The clean sheet approach to engine design is advocated with program directions and anticipated percentage SFC reductions cited. Future activities center on engine applications with coupled seal/power/secondary flow streams.

Particle trajectory computation on a 3-dimensional engine inlet. Final Report Ph.D. Thesis

NASA Technical Reports Server (NTRS)

Kim, J. J.

1986-01-01

A 3-dimensional particle trajectory computer code was developed to compute the distribution of water droplet impingement efficiency on a 3-dimensional engine inlet. The computed results provide the essential droplet impingement data required for the engine inlet anti-icing system design and analysis. The droplet trajectories are obtained by solving the trajectory equation using the fourth order Runge-Kutta and Adams predictor-corrector schemes. A compressible 3-D full potential flow code is employed to obtain a cylindrical grid definition of the flowfield on and about the engine inlet. The inlet surface is defined mathematically through a system of bi-cubic parametric patches in order to compute the droplet impingement points accurately. Analysis results of the 3-D trajectory code obtained for an axisymmetric droplet impingement problem are in good agreement with NACA experimental data. Experimental data are not yet available for the engine inlet impingement problem analyzed. Applicability of the method to solid particle impingement problems, such as engine sand ingestion, is also demonstrated.

Users manual for program NYQUIST: Liquid rocket nyquist plots developed for use on a PC computer

NASA Astrophysics Data System (ADS)

Armstrong, Wilbur C.

1992-06-01

The piping in a liquid rocket can assume complex configurations due to multiple tanks, multiple engines, and structures that must be piped around. The capability to handle some of these complex configurations have been incorporated into the NYQUIST code. The capability to modify the input on line has been implemented. The configurations allowed include multiple tanks, multiple engines, and the splitting of a pipe into unequal segments going to different (or the same) engines. This program will handle the following type elements: straight pipes, bends, inline accumulators, tuned stub accumulators, Helmholtz resonators, parallel resonators, pumps, split pipes, multiple tanks, and multiple engines. The code is too large to compile as one program using Microsoft FORTRAN 5; therefore, the code was broken into two segments: NYQUIST1.FOR and NYQUIST2.FOR. These are compiled separately and then linked together. The final run code is not too large (approximately equals 344,000 bytes).

Computerized systems analysis and optimization of aircraft engine performance, weight, and life cycle costs

NASA Technical Reports Server (NTRS)

Fishbach, L. H.

1979-01-01

The computational techniques utilized to determine the optimum propulsion systems for future aircraft applications and to identify system tradeoffs and technology requirements are described. The characteristics and use of the following computer codes are discussed: (1) NNEP - a very general cycle analysis code that can assemble an arbitrary matrix fans, turbines, ducts, shafts, etc., into a complete gas turbine engine and compute on- and off-design thermodynamic performance; (2) WATE - a preliminary design procedure for calculating engine weight using the component characteristics determined by NNEP; (3) POD DRG - a table look-up program to calculate wave and friction drag of nacelles; (4) LIFCYC - a computer code developed to calculate life cycle costs of engines based on the output from WATE; and (5) INSTAL - a computer code developed to calculate installation effects, inlet performance and inlet weight. Examples are given to illustrate how these computer techniques can be applied to analyze and optimize propulsion system fuel consumption, weight, and cost for representative types of aircraft and missions.

Users manual for program NYQUIST: Liquid rocket nyquist plots developed for use on a PC computer

NASA Technical Reports Server (NTRS)

Armstrong, Wilbur C.

1992-01-01

The piping in a liquid rocket can assume complex configurations due to multiple tanks, multiple engines, and structures that must be piped around. The capability to handle some of these complex configurations have been incorporated into the NYQUIST code. The capability to modify the input on line has been implemented. The configurations allowed include multiple tanks, multiple engines, and the splitting of a pipe into unequal segments going to different (or the same) engines. This program will handle the following type elements: straight pipes, bends, inline accumulators, tuned stub accumulators, Helmholtz resonators, parallel resonators, pumps, split pipes, multiple tanks, and multiple engines. The code is too large to compile as one program using Microsoft FORTRAN 5; therefore, the code was broken into two segments: NYQUIST1.FOR and NYQUIST2.FOR. These are compiled separately and then linked together. The final run code is not too large (approximately equals 344,000 bytes).

Innovations in Nuclear Infrastructure and Education

DOE Office of Scientific and Technical Information (OSTI.GOV)

John Bernard

The decision to implement the Innovation in Nuclear Infrastructure and Engineering Program (INIE) was an important first step towards ensuring that the United States preserves its worldwide leadership role in the field of nuclear science and engineering. Prior to INIE, university nuclear science and engineering programs were waning, undergraduate student enrollment was down, university research reactors were being shut down, while others faced the real possibility of closure. For too long, cutting edge research in the areas of nuclear medicine, neutron scattering, radiochemistry, and advanced materials was undervalued and therefore underfunded. The INIE program corrected this lapse in focus andmore » direction and started the process of drawing a new blueprint with positive goals and objectives that supports existing as well the next generation of educators, students and researchers.« less

Ducted-Fan Engine Acoustic Predictions using a Navier-Stokes Code

NASA Technical Reports Server (NTRS)

Rumsey, C. L.; Biedron, R. T.; Farassat, F.; Spence, P. L.

1998-01-01

A Navier-Stokes computer code is used to predict one of the ducted-fan engine acoustic modes that results from rotor-wake/stator-blade interaction. A patched sliding-zone interface is employed to pass information between the moving rotor row and the stationary stator row. The code produces averaged aerodynamic results downstream of the rotor that agree well with a widely used average-passage code. The acoustic mode of interest is generated successfully by the code and is propagated well upstream of the rotor; temporal and spatial numerical resolution are fine enough such that attenuation of the signal is small. Two acoustic codes are used to find the far-field noise. Near-field propagation is computed by using Eversman's wave envelope code, which is based on a finite-element model. Propagation to the far field is accomplished by using the Kirchhoff formula for moving surfaces with the results of the wave envelope code as input data. Comparison of measured and computed far-field noise levels show fair agreement in the range of directivity angles where the peak radiation lobes from the inlet are observed. Although only a single acoustic mode is targeted in this study, the main conclusion is a proof-of-concept: Navier-Stokes codes can be used both to generate and propagate rotor/stator acoustic modes forward through an engine, where the results can be coupled to other far-field noise prediction codes.

Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kahler, Albert Comstock

We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

75 FR 16517 - FirstEnergy Nuclear Operating Company; Environmental Assessment and Finding of No Significant Impact

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

... Perry. FOR FURTHER INFORMATION CONTACT: Michael Mahoney, Office of Nuclear Reactor Regulation, U.S... Nuclear Reactor Regulation. [FR Doc. 2010-7331 Filed 3-31-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-440; NRC-2010-0124] FirstEnergy Nuclear Operating...

Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution

NASA Astrophysics Data System (ADS)

Abolfazl Hosseini, Seyed; Vosoughi, Naser; Zangian, Mehdi

2015-05-01

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304-309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137Cs source is also compared with the experimental data.

Physical Processes and Applications of the Monte Carlo Radiative Energy Deposition (MRED) Code

NASA Astrophysics Data System (ADS)

Reed, Robert A.; Weller, Robert A.; Mendenhall, Marcus H.; Fleetwood, Daniel M.; Warren, Kevin M.; Sierawski, Brian D.; King, Michael P.; Schrimpf, Ronald D.; Auden, Elizabeth C.

2015-08-01

MRED is a Python-language scriptable computer application that simulates radiation transport. It is the computational engine for the on-line tool CRÈME-MC. MRED is based on c++ code from Geant4 with additional Fortran components to simulate electron transport and nuclear reactions with high precision. We provide a detailed description of the structure of MRED and the implementation of the simulation of physical processes used to simulate radiation effects in electronic devices and circuits. Extensive discussion and references are provided that illustrate the validation of models used to implement specific simulations of relevant physical processes. Several applications of MRED are summarized that demonstrate its ability to predict and describe basic physical phenomena associated with irradiation of electronic circuits and devices. These include effects from single particle radiation (including both direct ionization and indirect ionization effects), dose enhancement effects, and displacement damage effects. MRED simulations have also helped to identify new single event upset mechanisms not previously observed by experiment, but since confirmed, including upsets due to muons and energetic electrons.

Interdisciplinary Team-Teaching Experience for a Computer and Nuclear Energy Course for Electrical and Computer Engineering Students

ERIC Educational Resources Information Center

Kim, Charles; Jackson, Deborah; Keiller, Peter

2016-01-01

A new, interdisciplinary, team-taught course has been designed to educate students in Electrical and Computer Engineering (ECE) so that they can respond to global and urgent issues concerning computer control systems in nuclear power plants. This paper discusses our experience and assessment of the interdisciplinary computer and nuclear energy…

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

NASA Astrophysics Data System (ADS)

Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

2016-02-01

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level.

NUclear EVacuation Analysis Code (NUEVAC) : a tool for evaluation of sheltering and evacuation responses following urban nuclear detonations.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoshimura, Ann S.; Brandt, Larry D.

2009-11-01

The NUclear EVacuation Analysis Code (NUEVAC) has been developed by Sandia National Laboratories to support the analysis of shelter-evacuate (S-E) strategies following an urban nuclear detonation. This tool can model a range of behaviors, including complex evacuation timing and path selection, as well as various sheltering or mixed evacuation and sheltering strategies. The calculations are based on externally generated, high resolution fallout deposition and plume data. Scenario setup and calculation outputs make extensive use of graphics and interactive features. This software is designed primarily to produce quantitative evaluations of nuclear detonation response options. However, the outputs have also proven usefulmore » in the communication of technical insights concerning shelter-evacuate tradeoffs to urban planning or response personnel.« less

Does CTCF mediate between nuclear organization and gene expression?

PubMed

Ohlsson, Rolf; Lobanenkov, Victor; Klenova, Elena

2010-01-01

The multifunctional zinc-finger protein CCCTC-binding factor (CTCF) is a very strong candidate for the role of coordinating the expression level of coding sequences with their three-dimensional position in the nucleus, apparently responding to a "code" in the DNA itself. Dynamic interactions between chromatin fibers in the context of nuclear architecture have been implicated in various aspects of genome functions. However, the molecular basis of these interactions still remains elusive and is a subject of intense debate. Here we discuss the nature of CTCF-DNA interactions, the CTCF-binding specificity to its binding sites and the relationship between CTCF and chromatin, and we examine data linking CTCF with gene regulation in the three-dimensional nuclear space. We discuss why these features render CTCF a very strong candidate for the role and propose a unifying model, the "CTCF code," explaining the mechanistic basis of how the information encrypted in DNA may be interpreted by CTCF into diverse nuclear functions.

Convection and thermal radiation analytical models applicable to a nuclear waste repository room

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davis, B.W.

1979-01-17

Time-dependent temperature distributions in a deep geologic nuclear waste repository have a direct impact on the physical integrity of the emplaced canisters and on the design of retrievability options. This report (1) identifies the thermodynamic properties and physical parameters of three convection regimes - forced, natural, and mixed; (2) defines the convection correlations applicable to calculating heat flow in a ventilated (forced-air) and in a nonventilated nuclear waste repository room; and (3) delineates a computer code that (a) computes and compares the floor-to-ceiling heat flow by convection and radiation, and (b) determines the nonlinear equivalent conductivity table for a repositorymore » room. (The tables permit the use of the ADINAT code to model surface-to-surface radiation and the TRUMP code to employ two different emissivity properties when modeling radiation exchange between the surface of two different materials.) The analysis shows that thermal radiation dominates heat flow modes in a nuclear waste repository room.« less

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

Nuclear Resonance Fluorescence for Materials Assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quiter, Brian; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX's photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

Comparison of Models for Spacer Grid Pressure Loss in Nuclear Fuel Bundles for One and Two-Phase Flows

NASA Astrophysics Data System (ADS)

Maskal, Alan B.

Spacer grids maintain the structural integrity of the fuel rods within fuel bundles of nuclear power plants. They can also improve flow characteristics within the nuclear reactor core. However, spacer grids add reactor coolant pressure losses, which require estimation and engineering into the design. Several mathematical models and computer codes were developed over decades to predict spacer grid pressure loss. Most models use generalized characteristics, measured by older, less precise equipment. The study of OECD/US-NRC BWR Full-Size Fine Mesh Bundle Tests (BFBT) provides updated and detailed experimental single and two-phase results, using technically advanced flow measurements for a wide range of boundary conditions. This thesis compares the predictions from the mathematical models to the BFBT experimental data by utilizing statistical formulae for accuracy and precision. This thesis also analyzes the effects of BFBT flow characteristics on spacer grids. No single model has been identified as valid for all flow conditions. However, some models' predictions perform better than others within a range of flow conditions, based on the accuracy and precision of the models' predictions. This study also demonstrates that pressure and flow quality have a significant effect on two-phase flow spacer grid models' biases.

RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumptionmore » of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. Themore » assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

An astrophysical engine that stores gravitational work as nuclear Coulomb energy

NASA Astrophysics Data System (ADS)

Clayton, Donald

2014-03-01

I describe supernovae gravity machines that store large internal nuclear Coulomb energy, 0.80Z2A- 1 / 3MeV per nucleus. Excess of it is returned later by electron capture and positron emission. Decay energy manifests as (1) observable gamma-ray lines (2) light curves of supernovae (3) chemical energy of free carbon dissociated from CO molecules (4) huge abundances of radiogenic daughters. I illustrate by rapid silicon burning, a natural epoch in SN II. Gravitational work produces the high temperatures that photoeject nucleons and alpha particles from heavy nuclei. These are retained by other nuclei to balance photoejection rates (quasiequilibrium). The abundance distribution adjusts slowly as remaining abundance of Z = N 28Si decomposes, so p, n, α recaptures hug the Z = N line. This occurs in milliseconds, too rapidly for weak decay to alter bulk Z/N ratio. The figure displays those quasiequilibrium abundances color-coded to their decays. Z = N = 2k nuclei having k < 11 are stable, whereas k > 10 are radioactive owing to excess Coulomb energy. Weak decays radiate that excess energy weeks later to fuel the four macroscopic energetic phenomena cited. How startling to think of the Coulomb nuclear force as storing cosmic energy and its weak decay releasing macroscopic activation to SNII.

Structural Life and Reliability Metrics: Benchmarking and Verification of Probabilistic Life Prediction Codes

NASA Technical Reports Server (NTRS)

Litt, Jonathan S.; Soditus, Sherry; Hendricks, Robert C.; Zaretsky, Erwin V.

2002-01-01

Over the past two decades there has been considerable effort by NASA Glenn and others to develop probabilistic codes to predict with reasonable engineering certainty the life and reliability of critical components in rotating machinery and, more specifically, in the rotating sections of airbreathing and rocket engines. These codes have, to a very limited extent, been verified with relatively small bench rig type specimens under uniaxial loading. Because of the small and very narrow database the acceptance of these codes within the aerospace community has been limited. An alternate approach to generating statistically significant data under complex loading and environments simulating aircraft and rocket engine conditions is to obtain, catalog and statistically analyze actual field data. End users of the engines, such as commercial airlines and the military, record and store operational and maintenance information. This presentation describes a cooperative program between the NASA GRC, United Airlines, USAF Wright Laboratory, U.S. Army Research Laboratory and Australian Aeronautical & Maritime Research Laboratory to obtain and analyze these airline data for selected components such as blades, disks and combustors. These airline data will be used to benchmark and compare existing life prediction codes.

75 FR 69137 - Southern Nuclear Operating Company Inc. Edwin I. Hatch Nuclear Plant, Unit No. 2 Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-10

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-366; NRC-2010-0345] Southern Nuclear Operating Company Inc. Edwin I. Hatch Nuclear Plant, Unit No. 2 Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering the issuance of an exemption from Title 10 of the Code of Federal Regulations, ...

75 FR 73135 - Southern Nuclear Operating Company, Inc. Joseph M. Farley Nuclear Plant, Environmental Assessment...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-29

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-348 and 50-364; NRC-2009-0375] Southern Nuclear Operating Company, Inc. Joseph M. Farley Nuclear Plant, Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an Exemption, pursuant to Title 10 of the Code of Federal...

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

Improvements to the nuclear model code GNASH for cross section calculations at higher energies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Young, P.G.; Chadwick, M.B.

1994-05-01

The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV.

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Xyce parallel electronic simulator design.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thornquist, Heidi K.; Rankin, Eric Lamont; Mei, Ting

2010-09-01

This document is the Xyce Circuit Simulator developer guide. Xyce has been designed from the 'ground up' to be a SPICE-compatible, distributed memory parallel circuit simulator. While it is in many respects a research code, Xyce is intended to be a production simulator. As such, having software quality engineering (SQE) procedures in place to insure a high level of code quality and robustness are essential. Version control, issue tracking customer support, C++ style guildlines and the Xyce release process are all described. The Xyce Parallel Electronic Simulator has been under development at Sandia since 1999. Historically, Xyce has mostly beenmore » funded by ASC, the original focus of Xyce development has primarily been related to circuits for nuclear weapons. However, this has not been the only focus and it is expected that the project will diversify. Like many ASC projects, Xyce is a group development effort, which involves a number of researchers, engineers, scientists, mathmaticians and computer scientists. In addition to diversity of background, it is to be expected on long term projects for there to be a certain amount of staff turnover, as people move on to different projects. As a result, it is very important that the project maintain high software quality standards. The point of this document is to formally document a number of the software quality practices followed by the Xyce team in one place. Also, it is hoped that this document will be a good source of information for new developers.« less

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

Analysis of internal flows relative to the space shuttle main engine

NASA Technical Reports Server (NTRS)

1987-01-01

Cooperative efforts between the Lockheed-Huntsville Computational Mechanics Group and the NASA-MSFC Computational Fluid Dynamics staff has resulted in improved capabilities for numerically simulating incompressible flows generic to the Space Shuttle Main Engine (SSME). A well established and documented CFD code was obtained, modified, and applied to laminar and turbulent flows of the type occurring in the SSME Hot Gas Manifold. The INS3D code was installed on the NASA-MSFC CRAY-XMP computer system and is currently being used by NASA engineers. Studies to perform a transient analysis of the FPB were conducted. The COBRA/TRAC code is recommended for simulating the transient flow of oxygen into the LOX manifold. Property data for modifying the code to represent LOX/GOX flow was collected. The ALFA code was developed and recommended for representing the transient combustion in the preburner. These two codes will couple through the transient boundary conditions to simulate the startup and/or shutdown of the fuel preburner. A study, NAS8-37461, is currently being conducted to implement this modeling effort.

The roles of engineering notebooks in shaping elementary engineering student discourse and practice

NASA Astrophysics Data System (ADS)

Hertel, Jonathan D.; Cunningham, Christine M.; Kelly, Gregory J.

2017-06-01

Engineering design challenges offer important opportunities for students to learn science and engineering knowledge and practices. This study examines how students' engineering notebooks across four units of the curriculum Engineering is Elementary (EiE) support student work during design challenges. Through educational ethnography and discourse analysis, transcripts of student talk and action were created and coded around the uses of notebooks in the accomplishment of engineering tasks. Our coding process identified two broad categories of roles of the notebooks: they scaffold student activity and support epistemic practices of engineering. The study showed the importance of prompts to engage students in effective uses of writing, the roles the notebook assumes in the students' small groups, and the ways design challenges motivate children to write and communicate.

Activation cross section and isomeric cross section ratio for the 76Ge(n,2n)75m,gGe process

NASA Astrophysics Data System (ADS)

Luo, Junhua; Jiang, Li; Wang, Xinxing

2018-04-01

We measured neutron-induced reaction cross sections for the 76Ge(n,2n)75m,gGe reactions and their isomeric cross section ratios σm/σg at three neutron energies between 13 and 15MeV by an activation and off-line γ-ray spectrometric technique using the K-400 Neutron Generator at the Chinese Academy of Engineering Physics (CAEP). Ge samples and Nb monitor foils were activated together to determine the reaction cross section and the incident neutron flux. The monoenergetic neutron beams were formed via the 3H( d, n)4He reaction. The pure cross section of the ground state was derived from the absolute cross section of the metastable state and the residual nuclear decay analysis. The cross sections were also calculated using the nuclear model code TALYS-1.8 with different level density options at neutron energies varying from the reaction threshold to 20MeV. Results are discussed and compared with the corresponding literature data.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sattison, M.B.

The Idaho National Engineering Laboratory (INEL) over the three years has created 75 plant-specific Accident Sequence Precursor (ASP) models using the SAPHIRE suite of PRA codes. Along with the new models, the INEL has also developed a new module for SAPHIRE which is tailored specifically to the unique needs of ASP evaluations. These models and software will be the next generation of risk tools for the evaluation of accident precursors by both the U.S. Nuclear Regulatory Commission`s (NRC`s) Office of Nuclear Reactor Regulation (NRR) and the Office for Analysis and Evaluation of Operational Data (AEOD). This paper presents an overviewmore » of the models and software. Key characteristics include: (1) classification of the plant models according to plant response with a unique set of event trees for each plant class, (2) plant-specific fault trees using supercomponents, (3) generation and retention of all system and sequence cutsets, (4) full flexibility in modifying logic, regenerating cutsets, and requantifying results, and (5) user interface for streamlined evaluation of ASP events. Future plans for the ASP models is also presented.« less

Large-scale large eddy simulation of nuclear reactor flows: Issues and perspectives

DOE Office of Scientific and Technical Information (OSTI.GOV)

Merzari, Elia; Obabko, Aleks; Fischer, Paul

Numerical simulation has been an intrinsic part of nuclear engineering research since its inception. In recent years a transition is occurring toward predictive, first-principle-based tools such as computational fluid dynamics. Even with the advent of petascale computing, however, such tools still have significant limitations. In the present work some of these issues, and in particular the presence of massive multiscale separation, are discussed, as well as some of the research conducted to mitigate them. Petascale simulations at high fidelity (large eddy simulation/direct numerical simulation) were conducted with the massively parallel spectral element code Nek5000 on a series of representative problems.more » These simulations shed light on the requirements of several types of simulation: (1) axial flow around fuel rods, with particular attention to wall effects; (2) natural convection in the primary vessel; and (3) flow in a rod bundle in the presence of spacing devices. Finally, the focus of the work presented here is on the lessons learned and the requirements to perform these simulations at exascale. Additional physical insight gained from these simulations is also emphasized.« less

Large-scale large eddy simulation of nuclear reactor flows: Issues and perspectives

DOE PAGES

Merzari, Elia; Obabko, Aleks; Fischer, Paul; ...

2016-11-03

Numerical simulation has been an intrinsic part of nuclear engineering research since its inception. In recent years a transition is occurring toward predictive, first-principle-based tools such as computational fluid dynamics. Even with the advent of petascale computing, however, such tools still have significant limitations. In the present work some of these issues, and in particular the presence of massive multiscale separation, are discussed, as well as some of the research conducted to mitigate them. Petascale simulations at high fidelity (large eddy simulation/direct numerical simulation) were conducted with the massively parallel spectral element code Nek5000 on a series of representative problems.more » These simulations shed light on the requirements of several types of simulation: (1) axial flow around fuel rods, with particular attention to wall effects; (2) natural convection in the primary vessel; and (3) flow in a rod bundle in the presence of spacing devices. Finally, the focus of the work presented here is on the lessons learned and the requirements to perform these simulations at exascale. Additional physical insight gained from these simulations is also emphasized.« less

Development of code evaluation criteria for assessing predictive capability and performance

NASA Technical Reports Server (NTRS)

Lin, Shyi-Jang; Barson, S. L.; Sindir, M. M.; Prueger, G. H.

1993-01-01

Computational Fluid Dynamics (CFD), because of its unique ability to predict complex three-dimensional flows, is being applied with increasing frequency in the aerospace industry. Currently, no consistent code validation procedure is applied within the industry. Such a procedure is needed to increase confidence in CFD and reduce risk in the use of these codes as a design and analysis tool. This final contract report defines classifications for three levels of code validation, directly relating the use of CFD codes to the engineering design cycle. Evaluation criteria by which codes are measured and classified are recommended and discussed. Criteria for selecting experimental data against which CFD results can be compared are outlined. A four phase CFD code validation procedure is described in detail. Finally, the code validation procedure is demonstrated through application of the REACT CFD code to a series of cases culminating in a code to data comparison on the Space Shuttle Main Engine High Pressure Fuel Turbopump Impeller.

Reformation of Regulatory Technical Standards for Nuclear Power Generation Equipments in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mikio Kurihara; Masahiro Aoki; Yu Maruyama

2006-07-01

Comprehensive reformation of the regulatory system has been introduced in Japan in order to apply recent technical progress in a timely manner. 'The Technical Standards for Nuclear Power Generation Equipments', known as the Ordinance No.622) of the Ministry of International Trade and Industry, which is used for detailed design, construction and operating stage of Nuclear Power Plants, was being modified to performance specifications with the consensus codes and standards being used as prescriptive specifications, in order to facilitate prompt review of the Ordinance with response to technological innovation. The activities on modification were performed by the Nuclear and Industrial Safetymore » Agency (NISA), the regulatory body in Japan, with support of the Japan Nuclear Energy Safety Organization (JNES), a technical support organization. The revised Ordinance No.62 was issued on July 1, 2005 and is enforced from January 1 2006. During the period from the issuance to the enforcement, JNES carried out to prepare enforceable regulatory guide which complies with each provisions of the Ordinance No.62, and also made technical assessment to endorse the applicability of consensus codes and standards, in response to NISA's request. Some consensus codes and standards were re-assessed since they were already used in regulatory review of the construction plan submitted by licensee. Other consensus codes and standards were newly assessed for endorsement. In case that proper consensus code or standards were not prepared, details of regulatory requirements were described in the regulatory guide as immediate measures. At the same time, appropriate standards developing bodies were requested to prepare those consensus code or standards. Supplementary note which provides background information on the modification, applicable examples etc. was prepared for convenience to the users of the Ordinance No. 62. This paper shows the activities on modification and the results, following the NISA's presentation at ICONE-13 that introduced the framework of the performance specifications and the modification process of the Ordinance NO. 62. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sublet, J.-Ch., E-mail: jean-christophe.sublet@ukaea.uk; Eastwood, J.W.; Morgan, J.G.

Fispact-II is a code system and library database for modelling activation-transmutation processes, depletion-burn-up, time dependent inventory and radiation damage source terms caused by nuclear reactions and decays. The Fispact-II code, written in object-style Fortran, follows the evolution of material irradiated by neutrons, alphas, gammas, protons, or deuterons, and provides a wide range of derived radiological output quantities to satisfy most needs for nuclear applications. It can be used with any ENDF-compliant group library data for nuclear reactions, particle-induced and spontaneous fission yields, and radioactive decay (including but not limited to TENDL-2015, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0u, CENDL-3.1 processed into fine-group-structure files, GEFY-5.2more » and UKDD-16), as well as resolved and unresolved resonance range probability tables for self-shielding corrections and updated radiological hazard indices. The code has many novel features including: extension of the energy range up to 1 GeV; additional neutron physics including self-shielding effects, temperature dependence, thin and thick target yields; pathway analysis; and sensitivity and uncertainty quantification and propagation using full covariance data. The latest ENDF libraries such as TENDL encompass thousands of target isotopes. Nuclear data libraries for Fispact-II are prepared from these using processing codes PREPRO, NJOY and CALENDF. These data include resonance parameters, cross sections with covariances, probability tables in the resonance ranges, PKA spectra, kerma, dpa, gas and radionuclide production and energy-dependent fission yields, supplemented with all 27 decay types. All such data for the five most important incident particles are provided in evaluated data tables. The Fispact-II simulation software is described in detail in this paper, together with the nuclear data libraries. The Fispact-II system also includes several utility programs for code-use optimisation, visualisation and production of secondary radiological quantities. Included in the paper are summaries of results from the suite of verification and validation reports available with the code.« less

FISPACT-II: An Advanced Simulation System for Activation, Transmutation and Material Modelling

NASA Astrophysics Data System (ADS)

Sublet, J.-Ch.; Eastwood, J. W.; Morgan, J. G.; Gilbert, M. R.; Fleming, M.; Arter, W.

2017-01-01

Fispact-II is a code system and library database for modelling activation-transmutation processes, depletion-burn-up, time dependent inventory and radiation damage source terms caused by nuclear reactions and decays. The Fispact-II code, written in object-style Fortran, follows the evolution of material irradiated by neutrons, alphas, gammas, protons, or deuterons, and provides a wide range of derived radiological output quantities to satisfy most needs for nuclear applications. It can be used with any ENDF-compliant group library data for nuclear reactions, particle-induced and spontaneous fission yields, and radioactive decay (including but not limited to TENDL-2015, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0u, CENDL-3.1 processed into fine-group-structure files, GEFY-5.2 and UKDD-16), as well as resolved and unresolved resonance range probability tables for self-shielding corrections and updated radiological hazard indices. The code has many novel features including: extension of the energy range up to 1 GeV; additional neutron physics including self-shielding effects, temperature dependence, thin and thick target yields; pathway analysis; and sensitivity and uncertainty quantification and propagation using full covariance data. The latest ENDF libraries such as TENDL encompass thousands of target isotopes. Nuclear data libraries for Fispact-II are prepared from these using processing codes PREPRO, NJOY and CALENDF. These data include resonance parameters, cross sections with covariances, probability tables in the resonance ranges, PKA spectra, kerma, dpa, gas and radionuclide production and energy-dependent fission yields, supplemented with all 27 decay types. All such data for the five most important incident particles are provided in evaluated data tables. The Fispact-II simulation software is described in detail in this paper, together with the nuclear data libraries. The Fispact-II system also includes several utility programs for code-use optimisation, visualisation and production of secondary radiological quantities. Included in the paper are summaries of results from the suite of verification and validation reports available with the code.

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco; ...

2017-07-19

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

7 CFR 1755.503 - General.

Code of Federal Regulations, 2010 CFR

2010-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2012 CFR

2012-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2011 CFR

2011-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2013 CFR

2013-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2014 CFR

2014-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

Fundamental Studies in the Molecular Basis of Laser Induced Retinal Damage

DTIC Science & Technology

1988-01-01

Cornell University School of Applied & Engineering Physics Ithaca, NY 14853 DOD DISTRIBUTION STATEMENT Approved for public release; distribution unlimited...Code) 7b. ADDRESS (City, State, and ZIP Code) School of Applied & Engineering Physics Ithaca, NY 14853 8a. NAME OF FUNDING/SPONSORING Bb. OFFICE SYMBOL

Fundamental Studies in the Molecular Basis of Laser Induced Retinal Damage

DTIC Science & Technology

1988-01-01

Cornell University .LECT l School of Applied & Engineering PhysicsIthaca, NY 14853 0 JAN 198D DOD DISTRIBUTION STATEMENT Approved for public release...State, and ZIP Code) 7b. ADDRESS (City, State, and ZIP Code) School of Applied & Engineering Physics Ithaca, NY 14853 Ba. NAME OF FUNDING/ SPONSORING

Studies of Heat Transfer in Complex Internal Flows.

DTIC Science & Technology

1982-01-01

D.C. 20362 (Tel 202-692-6874) Mr. Richard S. Carlton Director, Engines Division, Code 523 NC #4 Naval Sea Systems Command Washington, D.C. 20362...Walter Ritz Code 033C Naval Ships Systems Engineering Station Philadelphia, Pennsylvania 19112 (Tel. 215-755-3841) Dr. Simion Kuo United Tech. Res

Your Career and Nuclear Weapons: A Guide for Young Scientists and Engineers.

ERIC Educational Resources Information Center

Albrecht, Andreas; And Others

This four-part booklet examines various issues related to nuclear weapons and how they will affect an individual working as a scientist or engineer. It provides information about the history of nuclear weapons, about the weapons industry which produces them, and about new weapons programs. Issues are raised so that new or future graduates may make…

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. D. Staiger

2007-06-01

This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

78 FR 17450 - Entergy Nuclear Operations, Inc.; Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-21

... INFORMATION CONTACT: Nageswaran Kalyanam, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2013-06510 Filed 3-20-13; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0012] [Docket Nos. 50-458, 50-155, 72-043, 50-003, 50-247...

QRAP: A numerical code for projected (Q)uasiparticle (RA)ndom (P)hase approximation

NASA Astrophysics Data System (ADS)

Samana, A. R.; Krmpotić, F.; Bertulani, C. A.

2010-06-01

A computer code for quasiparticle random phase approximation - QRPA and projected quasiparticle random phase approximation - PQRPA models of nuclear structure is explained in details. The residual interaction is approximated by a simple δ-force. An important application of the code consists in evaluating nuclear matrix elements involved in neutrino-nucleus reactions. As an example, cross sections for 56Fe and 12C are calculated and the code output is explained. The application to other nuclei and the description of other nuclear and weak decay processes are also discussed. Program summaryTitle of program: QRAP ( Quasiparticle RAndom Phase approximation) Computers: The code has been created on a PC, but also runs on UNIX or LINUX machines Operating systems: WINDOWS or UNIX Program language used: Fortran-77 Memory required to execute with typical data: 16 Mbytes of RAM memory and 2 MB of hard disk space No. of lines in distributed program, including test data, etc.: ˜ 8000 No. of bytes in distributed program, including test data, etc.: ˜ 256 kB Distribution format: tar.gz Nature of physical problem: The program calculates neutrino- and antineutrino-nucleus cross sections as a function of the incident neutrino energy, and muon capture rates, using the QRPA or PQRPA as nuclear structure models. Method of solution: The QRPA, or PQRPA, equations are solved in a self-consistent way for even-even nuclei. The nuclear matrix elements for the neutrino-nucleus interaction are treated as the beta inverse reaction of odd-odd nuclei as function of the transfer momentum. Typical running time: ≈ 5 min on a 3 GHz processor for Data set 1.

Minimizing human error in radiopharmaceutical preparation and administration via a bar code-enhanced nuclear pharmacy management system.

PubMed

Hakala, John L; Hung, Joseph C; Mosman, Elton A

2012-09-01

The objective of this project was to ensure correct radiopharmaceutical administration through the use of a bar code system that links patient and drug profiles with on-site information management systems. This new combined system would minimize the amount of manual human manipulation, which has proven to be a primary source of error. The most common reason for dosing errors is improper patient identification when a dose is obtained from the nuclear pharmacy or when a dose is administered. A standardized electronic transfer of information from radiopharmaceutical preparation to injection will further reduce the risk of misadministration. Value stream maps showing the flow of the patient dose information, as well as potential points of human error, were developed. Next, a future-state map was created that included proposed corrections for the most common critical sites of error. Transitioning the current process to the future state will require solutions that address these sites. To optimize the future-state process, a bar code system that links the on-site radiology management system with the nuclear pharmacy management system was proposed. A bar-coded wristband connects the patient directly to the electronic information systems. The bar code-enhanced process linking the patient dose with the electronic information reduces the number of crucial points for human error and provides a framework to ensure that the prepared dose reaches the correct patient. Although the proposed flowchart is designed for a site with an in-house central nuclear pharmacy, much of the framework could be applied by nuclear medicine facilities using unit doses. An electronic connection between information management systems to allow the tracking of a radiopharmaceutical from preparation to administration can be a useful tool in preventing the mistakes that are an unfortunate reality for any facility.

Analysis of Defenses Against Code Reuse Attacks on Modern and New Architectures

DTIC Science & Technology

2015-09-01

soundness or completeness. An incomplete analysis will produce extra edges in the CFG that might allow an attacker to slip through. An unsound analysis...Analysis of Defenses Against Code Reuse Attacks on Modern and New Architectures by Isaac Noah Evans Submitted to the Department of Electrical...Engineering and Computer Science in partial fulfillment of the requirements for the degree of Master of Engineering in Electrical Engineering and Computer

Centrifugal and Axial Pump Design and Off-Design Performance Prediction

NASA Technical Reports Server (NTRS)

Veres, Joseph P.

1995-01-01

A meanline pump-flow modeling method has been developed to provide a fast capability for modeling pumps of cryogenic rocket engines. Based on this method, a meanline pump-flow code PUMPA was written that can predict the performance of pumps at off-design operating conditions, given the loss of the diffusion system at the design point. The design-point rotor efficiency and slip factors are obtained from empirical correlations to rotor-specific speed and geometry. The pump code can model axial, inducer, mixed-flow, and centrifugal pumps and can model multistage pumps in series. The rapid input setup and computer run time for this meanline pump flow code make it an effective analysis and conceptual design tool. The map-generation capabilities of the code provide the information needed for interfacing with a rocket engine system modeling code. The off-design and multistage modeling capabilities of PUMPA permit the user to do parametric design space exploration of candidate pump configurations and to provide head-flow maps for engine system evaluation.

RTE: A computer code for Rocket Thermal Evaluation

NASA Technical Reports Server (NTRS)

Naraghi, Mohammad H. N.

1995-01-01

The numerical model for a rocket thermal analysis code (RTE) is discussed. RTE is a comprehensive thermal analysis code for thermal analysis of regeneratively cooled rocket engines. The input to the code consists of the composition of fuel/oxidant mixture and flow rates, chamber pressure, coolant temperature and pressure. dimensions of the engine, materials and the number of nodes in different parts of the engine. The code allows for temperature variation in axial, radial and circumferential directions. By implementing an iterative scheme, it provides nodal temperature distribution, rates of heat transfer, hot gas and coolant thermal and transport properties. The fuel/oxidant mixture ratio can be varied along the thrust chamber. This feature allows the user to incorporate a non-equilibrium model or an energy release model for the hot-gas-side. The user has the option of bypassing the hot-gas-side calculations and directly inputting the gas-side fluxes. This feature is used to link RTE to a boundary layer module for the hot-gas-side heat flux calculations.

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Verification of BWR Turbine Skyshine Dose with the MCNP5 Code Based on an Experiment Made at SHIMANE Nuclear Power Station

NASA Astrophysics Data System (ADS)

Tayama, Ryuichi; Wakasugi, Kenichi; Kawanaka, Ikunori; Kadota, Yoshinobu; Murakami, Yasuhiro

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benjamin Spencer; Yongfeng Zhang; Pritam Chakraborty

2013-09-01

The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. Themore » reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i.e. late blooming phases) to become active later in life beyond the current operational experience. To develop a tool that can eventually serve a role in decision-making, it is clear that research and development must be perfomed at multiple scales. At the engineering scale, a multiphysics analysis code that can capture the thermomechanical response of the RPV under accident conditions, including detailed fracture mechanics evaluations of flaws with arbitrary geometry and orientation, is needed to assess whether the fracture toughness, as defined by the master curve, including the effects of embrittlement, is exceeded. At the atomistic scale, the fundamental mechanisms of degradation need to be understood, including the effects of that degradation on the relevant material properties. In addition, there is a need to better understand the mechanisms leading to the transition from ductile to brittle fracture through improved continuum mechanics modeling at the fracture coupon scale. Work is currently being conducted at all of these levels with the goal of creating a usable engineering tool informed by lower length-scale modeling. This report summarizes progress made in these efforts during FY 2013.« less

75 FR 3761 - Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant, Units 1 and 2...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-22

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-321 and 50-366; NRC-2010-0024] Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant, Units 1 and 2; Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an Exemption, pursuant to Title 10 of the Code of...

Computational Nuclear Physics and Post Hartree-Fock Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lietz, Justin; Sam, Novario; Hjorth-Jensen, M.

We present a computational approach to infinite nuclear matter employing Hartree-Fock theory, many-body perturbation theory and coupled cluster theory. These lectures are closely linked with those of chapters 9, 10 and 11 and serve as input for the correlation functions employed in Monte Carlo calculations in chapter 9, the in-medium similarity renormalization group theory of dense fermionic systems of chapter 10 and the Green's function approach in chapter 11. We provide extensive code examples and benchmark calculations, allowing thereby an eventual reader to start writing her/his own codes. We start with an object-oriented serial code and end with discussions onmore » strategies for porting the code to present and planned high-performance computing facilities.« less

Reactor engineering support of operations at the Davis-Besse nuclear power station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

A Report of the Nuclear Engineering Division Sessions at the 1971 ASEE Annual Conference

ERIC Educational Resources Information Center

Eckley, Wayne; Nelson, George W.

1972-01-01

Summarizes the discussions at the conference under the topics, Objective Criteria for the Future" and Teaching Concepts Basic to Nuclear Engineering." Includes comments from personnel representing universities, industries, and government laboratories. (TS)

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

A novel nuclear genetic code alteration in yeasts and the evolution of codon reassignment in eukaryotes

PubMed Central

Mühlhausen, Stefanie; Findeisen, Peggy; Plessmann, Uwe; Urlaub, Henning; Kollmar, Martin

2016-01-01

The genetic code is the cellular translation table for the conversion of nucleotide sequences into amino acid sequences. Changes to the meaning of sense codons would introduce errors into almost every translated message and are expected to be highly detrimental. However, reassignment of single or multiple codons in mitochondria and nuclear genomes, although extremely rare, demonstrates that the code can evolve. Several models for the mechanism of alteration of nuclear genetic codes have been proposed (including “codon capture,” “genome streamlining,” and “ambiguous intermediate” theories), but with little resolution. Here, we report a novel sense codon reassignment in Pachysolen tannophilus, a yeast related to the Pichiaceae. By generating proteomics data and using tRNA sequence comparisons, we show that Pachysolen translates CUG codons as alanine and not as the more usual leucine. The Pachysolen tRNACAG is an anticodon-mutated tRNAAla containing all major alanine tRNA recognition sites. The polyphyly of the CUG-decoding tRNAs in yeasts is best explained by a tRNA loss driven codon reassignment mechanism. Loss of the CUG-tRNA in the ancient yeast is followed by gradual decrease of respective codons and subsequent codon capture by tRNAs whose anticodon is not part of the aminoacyl-tRNA synthetase recognition region. Our hypothesis applies to all nuclear genetic code alterations and provides several testable predictions. We anticipate more codon reassignments to be uncovered in existing and upcoming genome projects. PMID:27197221

Large liquid rocket engine transient performance simulation system

NASA Technical Reports Server (NTRS)

Mason, J. R.; Southwick, R. D.

1989-01-01

Phase 1 of the Rocket Engine Transient Simulation (ROCETS) program consists of seven technical tasks: architecture; system requirements; component and submodel requirements; submodel implementation; component implementation; submodel testing and verification; and subsystem testing and verification. These tasks were completed. Phase 2 of ROCETS consists of two technical tasks: Technology Test Bed Engine (TTBE) model data generation; and system testing verification. During this period specific coding of the system processors was begun and the engineering representations of Phase 1 were expanded to produce a simple model of the TTBE. As the code was completed, some minor modifications to the system architecture centering on the global variable common, GLOBVAR, were necessary to increase processor efficiency. The engineering modules completed during Phase 2 are listed: INJTOO - main injector; MCHBOO - main chamber; NOZLOO - nozzle thrust calculations; PBRNOO - preburner; PIPE02 - compressible flow without inertia; PUMPOO - polytropic pump; ROTROO - rotor torque balance/speed derivative; and TURBOO - turbine. Detailed documentation of these modules is in the Appendix. In addition to the engineering modules, several submodules were also completed. These submodules include combustion properties, component performance characteristics (maps), and specific utilities. Specific coding was begun on the system configuration processor. All functions necessary for multiple module operation were completed but the SOLVER implementation is still under development. This system, the Verification Checkout Facility (VCF) allows interactive comparison of module results to store data as well as provides an intermediate checkout of the processor code. After validation using the VCF, the engineering modules and submodules were used to build a simple TTBE.

Thermal-hydraulic interfacing code modules for CANDU reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Shock-driven fluid-structure interaction for civil design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, Stephen L; Deiterding, Ralf

The multiphysics fluid-structure interaction simulation of shock-loaded structures requires the dynamic coupling of a shock-capturing flow solver to a solid mechanics solver for large deformations. The Virtual Test Facility combines a Cartesian embedded boundary approach with dynamic mesh adaptation in a generic software framework of flow solvers using hydrodynamic finite volume upwind schemes that are coupled to various explicit finite element solid dynamics solvers (Deiterding et al., 2006). This paper gives a brief overview of the computational approach and presents first simulations that utilize the general purpose solid dynamics code DYNA3D for complex 3D structures of interest in civil engineering.more » Results from simulations of a reinforced column, highway bridge, multistory building, and nuclear reactor building are presented.« less

Midwest Nuclear Science and Engineering Consortium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. Wynn Volkert; Dr. Arvind Kumar; Dr. Bryan Becker

2010-12-08

The objective of the Midwest Nuclear Science and Engineering Consortium (MNSEC) is to enhance the scope, quality and integration of educational and research capabilities of nuclear sciences and engineering (NS/E) programs at partner schools in support of the U.S. nuclear industry (including DOE laboratories). With INIE support, MNSEC had a productive seven years and made impressive progress in achieving these goals. Since the past three years have been no-cost-extension periods, limited -- but notable -- progress has been made in FY10. Existing programs continue to be strengthened and broadened at Consortium partner institutions. The enthusiasm generated by the academic, state,more » federal, and industrial communities for the MNSEC activities is reflected in the significant leveraging that has occurred for our programs.« less

Nuclear Targeting Terms for Engineers and Scientists

DOE Office of Scientific and Technical Information (OSTI.GOV)

St Ledger, John W.

The Department of Defense has a methodology for targeting nuclear weapons, and a jargon that is used to communicate between the analysts, planners, aircrews, and missile crews. The typical engineer or scientist in the Department of Energy may not have been exposed to the nuclear weapons targeting terms and methods. This report provides an introduction to the terms and methodologies used for nuclear targeting. Its purpose is to prepare engineers and scientists to participate in wargames, exercises, and discussions with the Department of Defense. Terms such as Circular Error Probable, probability of hit and damage, damage expectancy, and the physicalmore » vulnerability system are discussed. Methods for compounding damage from multiple weapons applied to one target are presented.« less

A model describing intra-granular fission gas behaviour in oxide fuel for advanced engineering tools

NASA Astrophysics Data System (ADS)

Pizzocri, D.; Pastore, G.; Barani, T.; Magni, A.; Luzzi, L.; Van Uffelen, P.; Pitts, S. A.; Alfonsi, A.; Hales, J. D.

2018-04-01

The description of intra-granular fission gas behaviour is a fundamental part of any model for the prediction of fission gas release and swelling in nuclear fuel. In this work we present a model describing the evolution of intra-granular fission gas bubbles in terms of bubble number density and average size, coupled to gas release to grain boundaries. The model considers the fundamental processes of single gas atom diffusion, gas bubble nucleation, re-solution and gas atom trapping at bubbles. The model is derived from a detailed cluster dynamics formulation, yet it consists of only three differential equations in its final form; hence, it can be efficiently applied in engineering fuel performance codes while retaining a physical basis. We discuss improvements relative to previous single-size models for intra-granular bubble evolution. We validate the model against experimental data, both in terms of bubble number density and average bubble radius. Lastly, we perform an uncertainty and sensitivity analysis by propagating the uncertainties in the parameters to model results.

Advanced transportation system studies. Alternate propulsion subsystem concepts: Propulsion database

NASA Technical Reports Server (NTRS)

Levack, Daniel

1993-01-01

The Advanced Transportation System Studies alternate propulsion subsystem concepts propulsion database interim report is presented. The objective of the database development task is to produce a propulsion database which is easy to use and modify while also being comprehensive in the level of detail available. The database is to be available on the Macintosh computer system. The task is to extend across all three years of the contract. Consequently, a significant fraction of the effort in this first year of the task was devoted to the development of the database structure to ensure a robust base for the following years' efforts. Nonetheless, significant point design propulsion system descriptions and parametric models were also produced. Each of the two propulsion databases, parametric propulsion database and propulsion system database, are described. The descriptions include a user's guide to each code, write-ups for models used, and sample output. The parametric database has models for LOX/H2 and LOX/RP liquid engines, solid rocket boosters using three different propellants, a hybrid rocket booster, and a NERVA derived nuclear thermal rocket engine.

40 CFR 86.085-37 - Production vehicles and engines.

Code of Federal Regulations, 2014 CFR

2014-07-01

... transmission class. (2) Base level means a unique combination of basic engine, inertia weight, and transmission class. (3) Vehicle configuration means a unique combination of basic engine, engine code, inertia weight...

Software for Collaborative Engineering of Launch Rockets

NASA Technical Reports Server (NTRS)

Stanley, Thomas Troy

2003-01-01

The Rocket Evaluation and Cost Integration for Propulsion and Engineering software enables collaborative computing with automated exchange of information in the design and analysis of launch rockets and other complex systems. RECIPE can interact with and incorporate a variety of programs, including legacy codes, that model aspects of a system from the perspectives of different technological disciplines (e.g., aerodynamics, structures, propulsion, trajectory, aeroheating, controls, and operations) and that are used by different engineers on different computers running different operating systems. RECIPE consists mainly of (1) ISCRM a file-transfer subprogram that makes it possible for legacy codes executed in their original operating systems on their original computers to exchange data and (2) CONES an easy-to-use filewrapper subprogram that enables the integration of legacy codes. RECIPE provides a tightly integrated conceptual framework that emphasizes connectivity among the programs used by the collaborators, linking these programs in a manner that provides some configuration control while facilitating collaborative engineering tradeoff studies, including design to cost studies. In comparison with prior collaborative-engineering schemes, one based on the use of RECIPE enables fewer engineers to do more in less time.

Power control of SAFE reactor using fuzzy logic

NASA Astrophysics Data System (ADS)

Irvine, Claude

2002-01-01

Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc. .

Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

NASA Astrophysics Data System (ADS)

Ródenas, José

2017-11-01

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

AIRCRAFT REACTOR CONTROL SYSTEM APPLICABLE TO TURBOJET AND TURBOPROP POWER PLANTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorker, G.E.

1955-07-19

Control systems proposed for direct cycle nuclear powered aircraft commonly involve control of engine speed, nuclear energy input, and chcmical energy input. A system in which these parameters are controlled by controlling the total energy input, the ratio of nuclear and chemical energy input, and the engine speed is proposed. The system is equally applicable to turbojet or turboprop applications. (auth)

Nuclear Resonance Fluorescence for Materials Assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quiter, Brian J.; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX?s photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

Multi-University Southeast INIE Consortium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ayman Hawari; Nolan Hertel; Mohamed Al-Sheikhly

2 Project Summary: The Multi-University Southeast INIE Consortium (MUSIC) was established in response to the US Department of Energy’s (DOE) Innovations in Nuclear Infrastructure and Education (INIE) program. MUSIC was established as a consortium composed of academic members and national laboratory partners. The members of MUSIC are the nuclear engineering programs and research reactors of Georgia Institute of Technology (GIT), North Carolina State University (NCSU), University of Maryland (UMD), University of South Carolina (USC), and University of Tennessee (UTK). The University of Florida (UF), and South Carolina State University (SCSU) were added to the MUSIC membership in the second year.more » In addition, to ensure proper coordination between the academic community and the nation’s premier research and development centers in the fields of nuclear science and engineering, MUSIC created strategic partnerships with Oak Ridge National Laboratory (ORNL) including the Spallation Neutron Source (SNS) project and the Joint Institute for Neutron Scattering (JINS), and the National Institute of Standards and Technology (NIST). A partnership was also created with the Armed Forces Radiobiology Research Institute (AFRRI) with the aim of utilizing their reactor in research if funding becomes available. Consequently, there are three university research reactors (URRs) within MUSIC, which are located at NCSU (1-MW PULSTAR), UMD (0.25-MW TRIGA) and UF (0.10-MW Argonaut), and the AFRRI reactor (1-MW TRIGA MARK F). The overall objectives of MUSIC are: a) Demonstrate that University Research Reactors (URR) can be used as modern and innovative instruments of research in the basic and applied sciences, which include applications in fundamental physics, materials science and engineering, nondestructive examination, elemental analysis, and contributions to research in the health and medical sciences, b) Establish a strong technical collaboration between the nuclear engineering faculty and the MUSIC URRs. This will be achieved by involving the faculty in the development of state-of-the-art research facilities at the URRs and subsequently, in the utilization of these facilities, c) Facilitate the use of the URRs by the science and engineering faculty within the individual institutions and by the general community of science and engineering, d) Develop a far-reaching educational component that is capable of addressing the needs of the nuclear science and engineering community. Specifically, the aim of this component will be to perform public outreach activities, contribute to the active recruitment of the next generation of nuclear professionals, strengthen the education of nuclear engineering students, and promote nuclear engineering education for minority students.« less

Report of the Nuclear Propulsion Mission Analysis, Figures of Merit Subpanel: Quantifiable figures of merit for nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Haynes, Davy A.

1991-01-01

The results of an inquiry by the Nuclear Propulsion Mission Analysis, Figures of Merit subpanel are given. The subpanel was tasked to consider the question of what are the appropriate and quantifiable parameters to be used in the definition of an overall figure of merit (FoM) for Mars transportation system (MTS) nuclear thermal rocket engines (NTR). Such a characterization is needed to resolve the NTR engine design trades by a logical and orderly means, and to provide a meaningful method for comparison of the various NTR engine concepts. The subpanel was specifically tasked to identify the quantifiable engine parameters which would be the most significant engine factors affecting an overall FoM for a MTS and was not tasked with determining 'acceptable' or 'recommended' values for the identified parameters. In addition, the subpanel was asked not to define an overall FoM for a MTS. Thus, the selection of a specific approach, applicable weighting factors, to any interrelationships, for establishing an overall numerical FoM were considered beyond the scope of the subpanel inquiry.

Parametric analysis of diffuser requirements for high expansion ratio space engine

NASA Technical Reports Server (NTRS)

Wojciechowski, C. J.; Anderson, P. G.

1981-01-01

A supersonic diffuser ejector design computer program was developed. Using empirically modified one dimensional flow methods the diffuser ejector geometry is specified by the code. The design code results for calculations up to the end of the diffuser second throat were verified. Diffuser requirements for sea level testing of high expansion ratio space engines were defined. The feasibility of an ejector system using two commonly available turbojet engines feeding two variable area ratio ejectors was demonstrated.

Performance Benefits for Wave Rotor-Topped Gas Turbine Engines

NASA Technical Reports Server (NTRS)

Jones, Scott M.; Welch, Gerard E.

1996-01-01

The benefits of wave rotor-topping in turboshaft engines, subsonic high-bypass turbofan engines, auxiliary power units, and ground power units are evaluated. The thermodynamic cycle performance is modeled using a one-dimensional steady-state code; wave rotor performance is modeled using one-dimensional design/analysis codes. Design and off-design engine performance is calculated for baseline engines and wave rotor-topped engines, where the wave rotor acts as a high pressure spool. The wave rotor-enhanced engines are shown to have benefits in specific power and specific fuel flow over the baseline engines without increasing turbine inlet temperature. The off-design steady-state behavior of a wave rotor-topped engine is shown to be similar to a conventional engine. Mission studies are performed to quantify aircraft performance benefits for various wave rotor cycle and weight parameters. Gas turbine engine cycles most likely to benefit from wave rotor-topping are identified. Issues of practical integration and the corresponding technical challenges with various engine types are discussed.

Understanding Engineers' Responsibilities: A Prerequisite to Designing Engineering Education : Commentary on "Educating Engineers for the Public Good Through International Internships: Evidence from a Case Study at Universitat Politècnica de València".

PubMed

Murphy, Colleen; Gardoni, Paolo

2017-07-18

The development of the curriculum for engineering education (course requirements as well as extra-curricular activities like study abroad and internships) should be based on a comprehensive understanding of engineers' responsibilities. The responsibilities that are constitutive of being an engineer include striving to fulfill the standards of excellence set by technical codes; to improve the idealized models that engineers use to predict, for example, the behavior of alternative designs; and to achieve the internal goods such as safety and sustainability as they are reflected in the design codes. Globalization has implications for these responsibilities and, in turn, for engineering education, by, for example, modifying the collection of possible solutions recognized for existing problems. In addition, international internships can play an important role in fostering the requisite moral imagination of engineering students.

Publicizing Your Web Resources for Maximum Exposure.

ERIC Educational Resources Information Center

Smith, Kerry J.

2001-01-01

Offers advice to librarians for marketing their Web sites on Internet search engines. Advises against relying solely on spiders and recommends adding metadata to the source code and delivering that information directly to the search engines. Gives an overview of metadata and typical coding for meta tags. Includes Web addresses for a number of…

Ethics in Engineering: Student Perceptions and Their Professional Identity Development

ERIC Educational Resources Information Center

Stappenbelt, Brad

2013-01-01

Professional ethics instruction in engineering is commonly conducted by examining case studies in light of the code of conduct of a suitable professional body. Although graphical presentations of spectacular failures, sobering stories of the repercussions and the solid framework provided by the tenets of a code of ethics may leave a lasting…

A Model for On-Offshore Sediment Transport in the Surfzone.

DTIC Science & Technology

1982-12-01

34 Journal of Waterway Port, Coastal and Ocean Engineering, American Society of Civil Engineers, vol 108, no. WW2 , pp 163-179. Short, A. D. (1978) "Wave...PWO, Mayport FL; Utilities Engr Off. Rota Spain NAVTECHTRACEN SCE. Pensacola FL NAVWPNCEN Code 2636 China Lake; Code 3803 China Lake, CA NAVWPNSTA

A Coding Scheme for Analysing Problem-Solving Processes of First-Year Engineering Students

ERIC Educational Resources Information Center

Grigg, Sarah J.; Benson, Lisa C.

2014-01-01

This study describes the development and structure of a coding scheme for analysing solutions to well-structured problems in terms of cognitive processes and problem-solving deficiencies for first-year engineering students. A task analysis approach was used to assess students' problem solutions using the hierarchical structure from a…

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

Future space transport

NASA Technical Reports Server (NTRS)

Grishin, S. D.; Chekalin, S. V.

1984-01-01

Prospects for the mastery of space and the basic problems which must be solved in developing systems for both manned and cargo spacecraft are examined. The achievements and flaws of rocket boosters are discussed as well as the use of reusable spacecraft. The need for orbiting satellite solar power plants and related astrionics for active control of large space structures for space stations and colonies in an age of space industrialization is demonstrated. Various forms of spacecraft propulsion are described including liquid propellant rocket engines, nuclear reactors, thermonuclear rocket engines, electrorocket engines, electromagnetic engines, magnetic gas dynamic generators, electromagnetic mass accelerators (rail guns), laser rocket engines, pulse nuclear rocket engines, ramjet thermonuclear rocket engines, and photon rockets. The possibilities of interstellar flight are assessed.

Final Progress Report for Award DE-FG07-05ID14637.pdf

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cathy Dixon

2012-03-09

2004-2011 Final Report for AFCI University Fellowship Program. The goal of this effort was to be supportive of university students and university programs - particularly those students and programs that will help to strengthen the development of nuclear-related fields. The program also supported the stability of the nuclear infrastructure and developed research partnerships that are helping to enlarge the national nuclear science technology base. In this fellowship program, the U.S. Department of Energy sought master's degree students in nuclear, mechanical, or chemical engineering, engineering/applied physics, physics, chemistry, radiochemistry, or fields of science and engineering applicable to the AFCI/Gen IV/GNEP missionsmore » in order to meet future U.S. nuclear program needs. The fellowship program identified candidates and selected full time students of high-caliber who were taking nuclear courses as part of their degree programs. The DOE Academic Program Managers encouraged fellows to pursue summer internships at national laboratories and supported the students with appropriate information so that both the fellows and the nation's nuclear energy objectives were successful.« less

75 FR 11575 - James A. Fitzpatrick Nuclear Power Plant Environmental Assessment and Finding of No Significant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-11

... Power Plant Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory... Code of Federal Regulations (10 CFR), Appendix R, ``Fire Protection Program for Nuclear Power...), for the operation of the James A. FitzPatrick Nuclear Power Plant (JAFNPP) located in Oswego County...

75 FR 42469 - Firstenergy Nuclear Operating Company; Request for Licensing Action

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

... nuclear plant in Ohio, preventing the reactor from restarting until such time that the NRC determines... Commission's regulations. The request has been referred to the Director of the Office of Nuclear Reactor... of Nuclear Reactor Regulation. [FR Doc. 2010-17834 Filed 7-20-10; 8:45 am] BILLING CODE 7590-01-P ...

CFD Modeling of Free-Piston Stirling Engines

NASA Technical Reports Server (NTRS)

Ibrahim, Mounir B.; Zhang, Zhi-Guo; Tew, Roy C., Jr.; Gedeon, David; Simon, Terrence W.

2001-01-01

NASA Glenn Research Center (GRC) is funding Cleveland State University (CSU) to develop a reliable Computational Fluid Dynamics (CFD) code that can predict engine performance with the goal of significant improvements in accuracy when compared to one-dimensional (1-D) design code predictions. The funding also includes conducting code validation experiments at both the University of Minnesota (UMN) and CSU. In this paper a brief description of the work-in-progress is provided in the two areas (CFD and Experiments). Also, previous test results are compared with computational data obtained using (1) a 2-D CFD code obtained from Dr. Georg Scheuerer and further developed at CSU and (2) a multidimensional commercial code CFD-ACE+. The test data and computational results are for (1) a gas spring and (2) a single piston/cylinder with attached annular heat exchanger. The comparisons among the codes are discussed. The paper also discusses plans for conducting code validation experiments at CSU and UMN.

Integrative Curriculum Development in Nuclear Education and Research Vertical Enhancement Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Egarievwe, Stephen U.; Jow, Julius O.; Edwards, Matthew E.

Using a vertical education enhancement model, a Nuclear Education and Research Vertical Enhancement (NERVE) program was developed. The NERVE program is aimed at developing nuclear engineering education and research to 1) enhance skilled workforce development in disciplines relevant to nuclear power, national security and medical physics, and 2) increase the number of students and faculty from underrepresented groups (women and minorities) in fields related to the nuclear industry. The program uses multi-track training activities that vertically cut across the several education domains: undergraduate degree programs, graduate schools, and post-doctoral training. In this paper, we present the results of an integrativemore » curriculum development in the NERVE program. The curriculum development began with nuclear content infusion into existing science, engineering and technology courses. The second step involved the development of nuclear engineering courses: 1) Introduction to Nuclear Engineering, 2) Nuclear Engineering I, and 2) Nuclear Engineering II. The third step is the establishment of nuclear engineering concentrations in two engineering degree programs: 1) electrical engineering, and 2) mechanical engineering. A major outcome of the NERVE program is a collaborative infrastructure that uses laboratory work, internships at nuclear facilities, on-campus research, and mentoring in collaboration with industry and government partners to provide hands-on training for students. The major activities of the research and education collaborations include: - One-week spring training workshop at Brookhaven National Laboratory: The one-week training and workshop is used to enhance research collaborations and train faculty and students on user facilities/equipment at Brookhaven National Laboratory, and for summer research internships. Participants included students, faculty members at Alabama A and M University and research collaborators at BNL. The activities include 1) tour and introduction to user facilities/equipment at BNL that are used for research in room-temperature semiconductor nuclear detectors, 2) presentations on advances on this project and on wide band-gap semiconductor nuclear detectors in general, and 3) graduate students' research presentations. - Invited speakers and lectures: This brings collaborating research scientist from BNL to give talks and lectures on topics directly related to the project. Attendance includes faculty members, researchers and students throughout the university. - Faculty-students team summer research at BNL: This DOE and National Science Foundation (NSF) program help train students and faculty members in research. Faculty members go on to establish research collaborations with scientists at BNL, develop and submit research proposals to funding agencies, transform research experience at BNL to establish and enhance reach capabilities at home institution, and integrate their research into teaching through class projects and hands-on training for students. The students go on to participate in research work at BNL and at home institution, co-author research papers for conferences and technical journals, and transform their experiences into developing senior and capstone projects. - Grant proposal development: Faculty members in the NERVE program collaborate with BNL scientists to develop proposals, which often help to get external funding needed to expand and sustain the continuity of research activities and supports for student's wages and scholarships (stipends, tuition and fees). - Faculty development and mentoring: The above collaboration activities help faculty professional development. The experiences, grants, joint publications in technical journals, and supervision of student's research, including thesis and dissertation research projects, contribute greatly to faculty development. Senior scientists at BNL and senior faculty members on campus jointly mentor junior faculty members to enhance their professional growth. - Graduate thesis and dissertation research: Brookhaven National Laboratory provides unique opportunities and outstanding research resources for the NERVE program graduate research. Scientists from BNL serve in master's degree thesis and PhD dissertation committees, where they play active roles in the supervision of the research. (authors)« less

Computational electronics and electromagnetics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shang, C. C.

The Computational Electronics and Electromagnetics thrust area at Lawrence Livermore National Laboratory serves as the focal point for engineering R&D activities for developing computer-based design, analysis, and tools for theory. Key representative applications include design of particle accelerator cells and beamline components; engineering analysis and design of high-power components, photonics, and optoelectronics circuit design; EMI susceptibility analysis; and antenna synthesis. The FY-96 technology-base effort focused code development on (1) accelerator design codes; (2) 3-D massively parallel, object-oriented time-domain EM codes; (3) material models; (4) coupling and application of engineering tools for analysis and design of high-power components; (5) 3-D spectral-domainmore » CEM tools; and (6) enhancement of laser drilling codes. Joint efforts with the Power Conversion Technologies thrust area include development of antenna systems for compact, high-performance radar, in addition to novel, compact Marx generators. 18 refs., 25 figs., 1 tab.« less

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

Waste and Recycling

ScienceCinema

McCarthy, Kathy

2018-01-01

Nuclear engineer Dr. Kathy McCarthy talks about nuclear energy, the challenge of nuclear waste and the research aimed at solutions. For more information about nuclear energy research, visit http://www.facebook.com/idahonationallaboratory.

Standardized Radiation Shield Design Methods: 2005 HZETRN

NASA Technical Reports Server (NTRS)

Wilson, John W.; Tripathi, Ram K.; Badavi, Francis F.; Cucinotta, Francis A.

2006-01-01

Research committed by the Langley Research Center through 1995 resulting in the HZETRN code provides the current basis for shield design methods according to NASA STD-3000 (2005). With this new prominence, the database, basic numerical procedures, and algorithms are being re-examined with new methods of verification and validation being implemented to capture a well defined algorithm for engineering design processes to be used in this early development phase of the Bush initiative. This process provides the methodology to transform the 1995 HZETRN research code into the 2005 HZETRN engineering code to be available for these early design processes. In this paper, we will review the basic derivations including new corrections to the codes to insure improved numerical stability and provide benchmarks for code verification.

Toward Intelligent Software Defect Detection

NASA Technical Reports Server (NTRS)

Benson, Markland J.

2011-01-01

Source code level software defect detection has gone from state of the art to a software engineering best practice. Automated code analysis tools streamline many of the aspects of formal code inspections but have the drawback of being difficult to construct and either prone to false positives or severely limited in the set of defects that can be detected. Machine learning technology provides the promise of learning software defects by example, easing construction of detectors and broadening the range of defects that can be found. Pinpointing software defects with the same level of granularity as prominent source code analysis tools distinguishes this research from past efforts, which focused on analyzing software engineering metrics data with granularity limited to that of a particular function rather than a line of code.

Space Nuclear Reactor Engineering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Quick reproduction of blast-wave flow-field properties of nuclear, TNT, and ANFO explosions

NASA Astrophysics Data System (ADS)

Groth, C. P. T.

1986-04-01

In many instances, extensive blast-wave flow-field properties are required in gasdynamics research studies of blast-wave loading and structure response, and in evaluating the effects of explosions on their environment. This report provides a very useful computer code, which can be used in conjunction with the DNA Nuclear Blast Standard subroutines and code, to quickly reconstruct complete and fairly accurate blast-wave data for almost any free-air (spherical) and surface-burst (hemispherical) nuclear, trinitrotoluene (TNT), or ammonium nitrate-fuel oil (ANFO) explosion. This code is capable of computing all of the main flow properties as functions of radius and time, as well as providing additional information regarding air viscosity, reflected shock-wave properties, and the initial decay of the flow properties just behind the shock front. Both spatial and temporal distributions of the major blast-wave flow properties are also made readily available. Finally, provisions are also included in the code to provide additional information regarding the peak or shock-front flow properties over a range of radii, for a specific explosion of interest.

Long non-coding RNA nuclear paraspeckle assembly transcript 1 inhibits the apoptosis of retina Müller cells after diabetic retinopathy through regulating miR-497/brain-derived neurotrophic factor axis.

PubMed

Li, Xiu-Juan

2018-05-01

The role of long non-coding RNA in diabetic retinopathy, a serious complication of diabetes mellitus, has attracted increasing attention in recent years. The purpose of this study was to explore whether long non-coding RNA nuclear paraspeckle assembly transcript 1 was involved in the context of diabetic retinopathy and its underlying mechanisms. Our results revealed that nuclear paraspeckle assembly transcript 1 was significantly downregulated in the retina of diabetes mellitus rats. Meanwhile, miR-497 was significantly increased in diabetes mellitus rats' retina and high glucose-treated Müller cells, but brain-derived neurotrophic factor was increased. We also found that high glucose-induced apoptosis of Müller cells was accompanied by the significant downregulation of nuclear paraspeckle assembly transcript 1 in vitro. Further study demonstrated that high glucose-promoted Müller cells apoptosis through downregulating nuclear paraspeckle assembly transcript 1 and downregulated nuclear paraspeckle assembly transcript 1 mediated this effect via negative regulating miR-497. Moreover, brain-derived neurotrophic factor was negatively regulated by miR-497 and associated with the apoptosis of Müller cells under high glucose. Our results suggested that under diabetic conditions, downregulated nuclear paraspeckle assembly transcript 1 decreased the expression of brain-derived neurotrophic factor through elevating miR-497, thereby promoting Müller cells apoptosis and aggravating diabetic retinopathy.

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meng Lin; Dong Hou; Zhihong Xu

2006-07-01

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, justmore » can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I and C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results. (authors)« less

The ENSDF Java Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sonzogni, A.A.

2005-05-24

A package of computer codes has been developed to process and display nuclear structure and decay data stored in the ENSDF (Evaluated Nuclear Structure Data File) library. The codes were written in an object-oriented fashion using the java language. This allows for an easy implementation across multiple platforms as well as deployment on web pages. The structure of the different java classes that make up the package is discussed as well as several different implementations.

Hexagonal Uniformly Redundant Arrays (HURAs) for scintillator based coded aperture neutron imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamage, K.A.A.; Zhou, Q.

2015-07-01

A series of Monte Carlo simulations have been conducted, making use of the EJ-426 neutron scintillator detector, to investigate the potential of using hexagonal uniformly redundant arrays (HURAs) for scintillator based coded aperture neutron imaging. This type of scintillator material has a low sensitivity to gamma rays, therefore, is of particular use in a system with a source that emits both neutrons and gamma rays. The simulations used an AmBe source, neutron images have been produced using different coded-aperture materials (boron- 10, cadmium-113 and gadolinium-157) and location error has also been estimated. In each case the neutron image clearly showsmore » the location of the source with a relatively small location error. Neutron images with high resolution can be easily used to identify and locate nuclear materials precisely in nuclear security and nuclear decommissioning applications. (authors)« less

Axial deformed solution of the Skyrme-Hartree-Fock-Bogolyubov equations using the transformed harmonic oscillator Basis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perez, R. Navarro; Schunck, N.; Lasseri, R.

2017-03-09

HFBTHO is a physics computer code that is used to model the structure of the nucleus. It is an implementation of the nuclear energy Density Functional Theory (DFT), where the energy of the nucleus is obtained by integration over space of some phenomenological energy density, which is itself a functional of the neutron and proton densities. In HFBTHO, the energy density derives either from the zero-range Dkyrme or the finite-range Gogny effective two-body interaction between nucleons. Nuclear superfluidity is treated at the Hartree-Fock-Bogoliubov (HFB) approximation, and axial-symmetry of the nuclear shape is assumed. This version is the 3rd release ofmore » the program; the two previous versions were published in Computer Physics Communications [1,2]. The previous version was released at LLNL under GPL 3 Open Source License and was given release code LLNL-CODE-573953.« less

A systems engineering initiative for NASA's space communications

NASA Technical Reports Server (NTRS)

Hornstein, Rhoda S.; Hei, Donald J., Jr.; Kelly, Angelita C.; Lightfoot, Patricia C.; Bell, Holland T.; Cureton-Snead, Izeller E.; Hurd, William J.; Scales, Charles H.

1993-01-01

In addition to but separate from the Red and Blue Teams commissioned by the NASA Administrator, NASA's Associate Administrator for Space Communications commissioned a Blue Team to review the Office of Space Communications (Code O) Core Program and determine how the program could be conducted faster, better, and cheaper, without compromising safety. Since there was no corresponding Red Team for the Code O Blue Team, the Blue Team assumed a Red Team independent attitude and challenged the status quo. The Blue Team process and results are summarized. The Associate Administrator for Space Communications subsequently convened a special management session to discuss the significance and implications of the Blue Team's report and to lay the groundwork and teamwork for the next steps, including the transition from engineering systems to systems engineering. The methodology and progress toward realizing the Code O Family vision and accomplishing the systems engineering initiative for NASA's space communications are presented.

Radiation shielding estimates for manned Mars space flight.

PubMed

Dudkin, V E; Kovalev, E E; Kolomensky, A V; Sakovich, V A; Semenov, V F; Demin, V P; Benton, E V

1992-01-01

In the analysis of the required radiation shielding protection of spacecraft during a Mars flight, specific effects of solar activity (SA) on the intensity of galactic and solar cosmic rays were taken into consideration. Three spaceflight periods were considered: (1) maximum SA; (2) minimum SA; and (3) intermediate SA, when intensities of both galactic and solar cosmic rays are moderately high. Scenarios of spaceflights utilizing liquid-propellant rocket engines, low- and intermediate-thrust nuclear electrojet engines, and nuclear rocket engines, all of which have been designed in the Soviet Union, are reviewed. Calculations were performed on the basis of a set of standards for radiation protection approved by the U.S.S.R. State Committee for Standards. It was found that the lowest estimated mass of a Mars spacecraft, including the radiation shielding mass, obtained using a combination of a liquid propellant engine with low and intermediate thrust nuclear electrojet engines, would be 500-550 metric tons.

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Borowski, Stanley K.; Schnitzler, Bruce

2015-01-01

In an effort to further refine potential point of departure nuclear thermal rocket engine designs, four proposed engine designs representing two thrust classes and utilizing two different fuel matrix types are designed and analyzed from both a neutronics and thermodynamic cycle perspective. Two of these nuclear rocket engine designs employ a tungsten and uranium dioxide cermet (ceramic-metal) fuel with a prismatic geometry based on the ANL-200 and the GE-710, while the other two designs utilize uranium-zirconium-carbide in a graphite composite fuel and a prismatic fuel element geometry developed during the Rover/NERVA Programs. Two engines are analyzed for each fuel type, a small criticality limited design and a 111 kN (25 klbf) thrust class engine design, which has been the focus of numerous manned mission studies, including NASA's Design Reference Architecture 5.0. slightly higher T/W ratios, but they required substantially more 235U.

Design implications for task-specific search utilities for retrieval and re-engineering of code

NASA Astrophysics Data System (ADS)

Iqbal, Rahat; Grzywaczewski, Adam; Halloran, John; Doctor, Faiyaz; Iqbal, Kashif

2017-05-01

The importance of information retrieval systems is unquestionable in the modern society and both individuals as well as enterprises recognise the benefits of being able to find information effectively. Current code-focused information retrieval systems such as Google Code Search, Codeplex or Koders produce results based on specific keywords. However, these systems do not take into account developers' context such as development language, technology framework, goal of the project, project complexity and developer's domain expertise. They also impose additional cognitive burden on users in switching between different interfaces and clicking through to find the relevant code. Hence, they are not used by software developers. In this paper, we discuss how software engineers interact with information and general-purpose information retrieval systems (e.g. Google, Yahoo!) and investigate to what extent domain-specific search and recommendation utilities can be developed in order to support their work-related activities. In order to investigate this, we conducted a user study and found that software engineers followed many identifiable and repeatable work tasks and behaviours. These behaviours can be used to develop implicit relevance feedback-based systems based on the observed retention actions. Moreover, we discuss the implications for the development of task-specific search and collaborative recommendation utilities embedded with the Google standard search engine and Microsoft IntelliSense for retrieval and re-engineering of code. Based on implicit relevance feedback, we have implemented a prototype of the proposed collaborative recommendation system, which was evaluated in a controlled environment simulating the real-world situation of professional software engineers. The evaluation has achieved promising initial results on the precision and recall performance of the system.

New French Regulation for NPPs and Code Consequences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Faidy, Claude

2006-07-01

On December 2005, the French regulator issued a new regulation for French nuclear power plants, in particular for pressure equipment (PE). This regulation need first to agree with non-nuclear PE regulation and add to that some specific requirements, in particular radiation protection requirements. Different advantages are in these proposal, it's more qualitative risk oriented and it's an important link with non-nuclear industry. Only few components are nuclear specific. But, the general philosophy of the existing Codes (RCC-M [15], KTA [16] or ASME [17]) have to be improved. For foreign Codes, it's plan to define the differences in the user specifications.more » In parallel to that, a new safety classification has been developed by French utility. The consequences is the need to cross all these specifications to define a minimum quality level for each components or systems. In the same time a new concept has been developed to replace the well known 'Leak Before Break methodology': the 'Break Exclusion' methodology. This paper will summarize the key aspects of these different topics. (authors)« less

Prompt gamma-ray imaging for small animals

NASA Astrophysics Data System (ADS)

Xu, Libai

Small animal imaging is recognized as a powerful discovery tool for small animal modeling of human diseases, which is providing an important clue to complete understanding of disease mechanisms and is helping researchers develop and test new treatments. The current small animal imaging techniques include positron emission tomography (PET), single photon emission tomography (SPECT), computed tomography (CT), magnetic resonance imaging (MRI), and ultrasound (US). A new imaging modality called prompt gamma-ray imaging (PGI) has been identified and investigated primarily by Monte Carlo simulation. Currently it is suggested for use on small animals. This new technique could greatly enhance and extend the present capabilities of PET and SPECT imaging from ingested radioisotopes to the imaging of selected non-radioactive elements, such as Gd, Cd, Hg, and B, and has the great potential to be used in Neutron Cancer Therapy to monitor neutron distribution and neutron-capture agent distribution. This approach consists of irradiating small animals in the thermal neutron beam of a nuclear reactor to produce prompt gamma rays from the elements in the sample by the radiative capture (n, gamma) reaction. These prompt gamma rays are emitted in energies that are characteristic of each element and they are also produced in characteristic coincident chains. After measuring these prompt gamma rays by surrounding spectrometry array, the distribution of each element of interest in the sample is reconstructed from the mapping of each detected signature gamma ray by either electronic collimations or mechanical collimations. In addition, the transmitted neutrons from the beam can be simultaneously used for very sensitive anatomical imaging, which provides the registration for the elemental distributions obtained from PGI. The primary approach is to use Monte Carlo simulation methods either with the specific purpose code CEARCPG, developed at NC State University or with the general purpose codes GEANT4 or MCNP5, to predict results and investigate the feasibility of this new imaging idea. Benchmark experiments have been conducted to test the capability of the code to simulate prompt gamma rays, which are produced by following the nuclear structures of each irradiated isotope, and coincidence counting techniques, which are considered the most important improvement in neutron-related gamma-ray detection applications to reduce gamma background and improve system signal-to-noise ratios. With coincidence prompt gamma rays available, two major imaging techniques, electronic collimations and mechanic collimations, are implemented in the simulation to illustrate the feasibility of imaging elemental distribution by this new technique. The expectation maximization algorithm is employed in electronic collimation to reconstruct images. The common SPECT imaging algorithms are used in mechanical collimation to get an image. Several critical topics concerning practical applications have already been discussed, such as the radiation dose to the mouse and the detection efficiency of high-energy gamma rays. The funding of this work is provided by the Center for Engineering Application of Radioisotopes (CEAR) at North Carolina State University (NCSU) and Nuclear Engineering Education Research.

Nuclear Thermal Propulsion Mars Mission Systems Analysis and Requirements Definition

NASA Technical Reports Server (NTRS)

Mulqueen, Jack; Chiroux, Robert C.; Thomas, Dan; Crane, Tracie

2007-01-01

This paper describes the Mars transportation vehicle design concepts developed by the Marshall Space Flight Center (MSFC) Advanced Concepts Office. These vehicle design concepts provide an indication of the most demanding and least demanding potential requirements for nuclear thermal propulsion systems for human Mars exploration missions from years 2025 to 2035. Vehicle concept options vary from large "all-up" vehicle configurations that would transport all of the elements for a Mars mission on one vehicle. to "split" mission vehicle configurations that would consist of separate smaller vehicles that would transport cargo elements and human crew elements to Mars separately. Parametric trades and sensitivity studies show NTP stage and engine design options that provide the best balanced set of metrics based on safety, reliability, performance, cost and mission objectives. Trade studies include the sensitivity of vehicle performance to nuclear engine characteristics such as thrust, specific impulse and nuclear reactor type. Tbe associated system requirements are aligned with the NASA Exploration Systems Mission Directorate (ESMD) Reference Mars mission as described in the Explorations Systems Architecture Study (ESAS) report. The focused trade studies include a detailed analysis of nuclear engine radiation shield requirements for human missions and analysis of nuclear thermal engine design options for the ESAS reference mission.

Automated real-time software development

NASA Technical Reports Server (NTRS)

Jones, Denise R.; Walker, Carrie K.; Turkovich, John J.

1993-01-01

A Computer-Aided Software Engineering (CASE) system has been developed at the Charles Stark Draper Laboratory (CSDL) under the direction of the NASA Langley Research Center. The CSDL CASE tool provides an automated method of generating source code and hard copy documentation from functional application engineering specifications. The goal is to significantly reduce the cost of developing and maintaining real-time scientific and engineering software while increasing system reliability. This paper describes CSDL CASE and discusses demonstrations that used the tool to automatically generate real-time application code.

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

NASA Astrophysics Data System (ADS)

Wemple, Charles; Zwermann, Winfried

2017-09-01

Uncertainty quantification has been extensively applied to nuclear criticality analyses for many years and has recently begun to be applied to depletion calculations. However, regulatory bodies worldwide are trending toward requiring such analyses for reactor fuel cycle calculations, which also requires uncertainty propagation for isotopics and nuclear reaction rates. XSUSA is a proven methodology for cross section uncertainty propagation based on random sampling of the nuclear data according to covariance data in multi-group representation; HELIOS2 is a lattice code widely used for commercial and research reactor fuel cycle calculations. This work describes a technique to automatically propagate the nuclear data uncertainties via the XSUSA approach through fuel lattice calculations in HELIOS2. Application of the XSUSA methodology in HELIOS2 presented some unusual challenges because of the highly-processed multi-group cross section data used in commercial lattice codes. Currently, uncertainties based on the SCALE 6.1 covariance data file are being used, but the implementation can be adapted to other covariance data in multi-group structure. Pin-cell and assembly depletion calculations, based on models described in the UAM-LWR Phase I and II benchmarks, are performed and uncertainties in multiplication factor, reaction rates, isotope concentrations, and delayed-neutron data are calculated. With this extension, it will be possible for HELIOS2 users to propagate nuclear data uncertainties directly from the microscopic cross sections to subsequent core simulations.

Sustaining engineering codes of ethics for the twenty-first century.

PubMed

Michelfelder, Diane; Jones, Sharon A

2013-03-01

How much responsibility ought a professional engineer to have with regard to supporting basic principles of sustainable development? While within the United States, professional engineering societies, as reflected in their codes of ethics, differ in their responses to this question, none of these professional societies has yet to put the engineer's responsibility toward sustainability on a par with commitments to public safety, health, and welfare. In this paper, we aim to suggest that sustainability should be included in the paramountcy clause because it is a necessary condition to ensure the safety, health, and welfare of the public. Part of our justification rests on the fact that to engineer sustainably means among many things to consider social justice, understood as the fair and equitable distribution of social goods, as a design constraint similar to technical, economic, and environmental constraints. This element of social justice is not explicit in the current paramountcy clause. Our argument rests on demonstrating that social justice in terms of both inter- and intra-generational equity is an important dimension of sustainability (and engineering). We also propose that embracing sustainability in the codes while recognizing the role that social justice plays may elevate the status of the engineer as public intellectual and agent of social good. This shift will then need to be incorporated in how we teach undergraduate engineering students about engineering ethics.

48 CFR 2001.104-1 - Publication and code arrangement.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 6 2011-10-01 2011-10-01 false Publication and code arrangement. 2001.104-1 Section 2001.104-1 Federal Acquisition Regulations System NUCLEAR REGULATORY... 2001.104-1 Publication and code arrangement. (a) The NRCAR and its subsequent changes are: (1...

Design engineer perceptions and attitudes regarding human factors application to nuclear power plant design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ma, R.; Jones, J. M.

2006-07-01

With the renewed interest in nuclear power and the possibility of constructing new reactors within the next decade in the U.S., there are several challenges for the regulators, designers, and vendors. One challenge is to ensure that Human Factors Engineering (HFE) is involved, and correctly applied in the life-cycle design of the Nuclear Power Plant (NPP). As an important part of the effort, people would ask: 'is the system-design engineer effectively incorporating HFE in the NPPs design?' The present study examines the sagacity of Instrumentation and Control design engineers on issues relating to awareness, attitude, and application of HFE inmore » NPP design. A questionnaire was developed and distributed, focusing on the perceptions and attitudes of the design engineers. The responses revealed that, while the participants had a relatively high positive attitude about HFE, their awareness and application of HFE were moderate. The results also showed that senior engineers applied HFE more frequently in their design work than young engineers. This study provides some preliminary results and implications for improved HFE education and application in NPP design. (authors)« less

Flow Analysis of a Gas Turbine Low- Pressure Subsystem

NASA Technical Reports Server (NTRS)

Veres, Joseph P.

1997-01-01

The NASA Lewis Research Center is coordinating a project to numerically simulate aerodynamic flow in the complete low-pressure subsystem (LPS) of a gas turbine engine. The numerical model solves the three-dimensional Navier-Stokes flow equations through all components within the low-pressure subsystem as well as the external flow around the engine nacelle. The Advanced Ducted Propfan Analysis Code (ADPAC), which is being developed jointly by Allison Engine Company and NASA, is the Navier-Stokes flow code being used for LPS simulation. The majority of the LPS project is being done under a NASA Lewis contract with Allison. Other contributors to the project are NYMA and the University of Toledo. For this project, the Energy Efficient Engine designed by GE Aircraft Engines is being modeled. This engine includes a low-pressure system and a high-pressure system. An inlet, a fan, a booster stage, a bypass duct, a lobed mixer, a low-pressure turbine, and a jet nozzle comprise the low-pressure subsystem within this engine. The tightly coupled flow analysis evaluates aerodynamic interactions between all components of the LPS. The high-pressure core engine of this engine is simulated with a one-dimensional thermodynamic cycle code in order to provide boundary conditions to the detailed LPS model. This core engine consists of a high-pressure compressor, a combustor, and a high-pressure turbine. The three-dimensional LPS flow model is coupled to the one-dimensional core engine model to provide a "hybrid" flow model of the complete gas turbine Energy Efficient Engine. The resulting hybrid engine model evaluates the detailed interaction between the LPS components at design and off-design engine operating conditions while considering the lumped-parameter performance of the core engine.

Development of Northeast Asia Nuclear Power Plant Accident Simulator.

PubMed

Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

2017-06-15

A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Proceedings of the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M and C 2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2013-07-01

The Mathematics and Computation Division of the American Nuclear (ANS) and the Idaho Section of the ANS hosted the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M and C 2013). This proceedings contains over 250 full papers with topics ranging from reactor physics; radiation transport; materials science; nuclear fuels; core performance and optimization; reactor systems and safety; fluid dynamics; medical applications; analytical and numerical methods; algorithms for advanced architectures; and validation verification, and uncertainty quantification.

Reactor physics teaching and research in the Swiss nuclear engineering master

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chawla, R.; Paul Scherrer Inst., CH-5232 Villigen PSI

Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

Performance Benefits for a Turboshaft Engine Using Nonlinear Engine Control Technology Investigated

NASA Technical Reports Server (NTRS)

Jones, Scott M.

2004-01-01

The potential benefits of nonlinear engine control technology applied to a General Electric T700 helicopter engine were investigated. This technology is being developed by the U.S. Navy SPAWAR Systems Center for a variety of applications. When used as a means of active stability control, nonlinear engine control technology uses sensors and small amounts of injected air to allow compressors to operate with reduced stall margin, which can improve engine pressure ratio. The focus of this study was to determine the best achievable reduction in fuel consumption for the T700 turboshaft engine. A customer deck (computer code) was provided by General Electric to calculate the T700 engine performance, and the NASA Glenn Research Center used this code to perform the analysis. The results showed a 2- to 5-percent reduction in brake specific fuel consumption (BSFC) at the three Sikorsky H-60 helicopter operating points of cruise, loiter, and hover.

75 FR 37842 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-30

... INFORMATION CONTACT: R. A. Jervey, Regulatory Guide Development Branch, Division of Engineering, Office of...) 251-7404 or e-mail [email protected] . SUPPLEMENTARY INFORMATION: I. Introduction The U.S. Nuclear..., Regulatory Guide Development Branch, Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc...

Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

2013-01-01

Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

Engineering High Assurance Distributed Cyber Physical Systems

DTIC Science & Technology

2015-01-15

decisions: number of interacting agents and co-dependent decisions made in real-time without causing interference . To engineer a high assurance DART...environment specification, architecture definition, domain-specific languages, design patterns, code - generation, analysis, test-generation, and simulation...include synchronization between the models and source code , debugging at the model level, expression of the design intent, and quality of service

78 FR 54635 - Notice of Intent To Prepare an Environmental Impact Statement for EA-18G Growler Airfield...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-05

... the potential environmental effects associated with the introduction of two additional EA-18G Growler... CONTACT: EA-18G EIS Project Manager (Code EV21/ SS); Naval Facilities Engineering Command (NAVFAC... request should be submitted to: EA-18G EIS Project Manager (Code EV21/SS); Naval Facilities Engineering...

Thermal Propulsion Capture System Heat Exchanger Design

NASA Technical Reports Server (NTRS)

Richard, Evan M.

2016-01-01

One of the biggest challenges of manned spaceflight beyond low earth orbit and the moon is harmful radiation that astronauts would be exposed to on their long journey to Mars and further destinations. Using nuclear energy has the potential to be a more effective means of propulsion compared to traditional chemical engines (higher specific impulse). An upper stage nuclear engine would allow astronauts to reach their destination faster and more fuel efficiently. Testing these engines poses engineering challenges due to the need to totally capture the engine exhaust. The Thermal Propulsion Capture System is a concept for cost effectively and safely testing Nuclear Thermal Engines. Nominally, hydrogen exhausted from the engine is not radioactive, but is treated as such in case of fuel element failure. The Thermal Propulsion Capture System involves injecting liquid oxygen to convert the hydrogen exhaust into steam. The steam is then cooled and condensed into liquid water to allow for storage. The Thermal Propulsion Capture System concept for ground testing of a nuclear powered engine involves capturing the engine exhaust to be cooled and condensed before being stored. The hydrogen exhaust is injected with liquid oxygen and burned to form steam. That steam must be cooled to saturation temperatures before being condensed into liquid water. A crossflow heat exchanger using water as a working fluid will be designed to accomplish this goal. Design a cross flow heat exchanger for the Thermal Propulsion Capture System testing which: Eliminates the need for water injection cooling, Cools steam from 5800 F to saturation temperature, and Is efficient and minimizes water requirement.

Comparison of FDNS liquid rocket engine plume computations with SPF/2

NASA Technical Reports Server (NTRS)

Kumar, G. N.; Griffith, D. O., II; Warsi, S. A.; Seaford, C. M.

1993-01-01

Prediction of a plume's shape and structure is essential to the evaluation of base region environments. The JANNAF standard plume flowfield analysis code SPF/2 predicts plumes well, but cannot analyze base regions. Full Navier-Stokes CFD codes can calculate both zones; however, before they can be used, they must be validated. The CFD code FDNS3D (Finite Difference Navier-Stokes Solver) was used to analyze the single plume of a Space Transportation Main Engine (STME) and comparisons were made with SPF/2 computations. Both frozen and finite rate chemistry models were employed as well as two turbulence models in SPF/2. The results indicate that FDNS3D plume computations agree well with SPF/2 predictions for liquid rocket engine plumes.

Report of the Secretary of Defense Task Force on DoD Nuclear Weapons Management. Phase 1. The Air Force’s Nuclear Mission

DTIC Science & Technology

2008-09-01

under- resourced. • Missile transfer vans /warhead transfer vans require upgrades. • ICBM weapon system test sets under-funded; the coding system...Air Force’s Nuclear Mission D-1 Appendix D. Current B-52 Basing Status Barksdale AFB, LA 64 B-52Hs Minot AFB, ND 27 B-52Hs Edwards AFB, CA 3...Barksdale – 64 B-52s 2 BW (ACC) 15 TF; 24 CC; 7 BAI 53 WG (ACC) 2 Test Coded 917 WG (AFRC) 8 CC; 1 BAI 7 Unfunded AR Edwards - 3 B-52s 412 TW 2 Test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Detilleux, Michel; Centner, Baudouin

The paper describes different methodologies and tools developed in-house by Tractebel Engineering to facilitate the engineering works to be carried out especially in the frame of decommissioning projects. Three examples of tools with their corresponding results are presented: - The LLWAA-DECOM code, a software developed for the radiological characterization of contaminated systems and equipment. The code constitutes a specific module of more general software that was originally developed to characterize radioactive waste streams in order to be able to declare the radiological inventory of critical nuclides, in particular difficult-to-measure radionuclides, to the Authorities. In the case of LLWAA-DECOM, deposited activitiesmore » inside contaminated equipment (piping, tanks, heat exchangers...) and scaling factors between nuclides, at any given time of the decommissioning time schedule, are calculated on the basis of physical characteristics of the systems and of operational parameters of the nuclear power plant. This methodology was applied to assess decommissioning costs of Belgian NPPs, to characterize the primary system of Trino NPP in Italy, to characterize the equipment of miscellaneous circuits of Ignalina NPP and of Kozloduy unit 1 and, to calculate remaining dose rates around equipment in the frame of the preparation of decommissioning activities; - The VISIMODELLER tool, a user friendly CAD interface developed to ease the introduction of lay-out areas in a software named VISIPLAN. VISIPLAN is a 3D dose rate assessment tool for ALARA work planning, developed by the Belgian Nuclear Research Centre SCK.CEN. Both softwares were used for projects such as the steam generators replacements in Belgian NPPs or the preparation of the decommissioning of units 1 and 2 of Kozloduy NPP; - The DBS software, a software developed to manage the different kinds of activities that are part of the general time schedule of a decommissioning project. For each activity, when relevant, algorithms allow to estimate, on the basis of local inputs, radiological exposures of the operators (collective and individual doses), production of primary, secondary and tertiary waste and their characterization, production of conditioned waste, release of effluents,... and enable the calculation and the presentation (histograms) of the global results for all activities together. An example of application in the frame of the Ignalina decommissioning project is given. (authors)« less