Neutron Capture Gamma-Ray Libraries for Nuclear Applications

NASA Astrophysics Data System (ADS)

Sleaford, B. W.; Firestone, R. B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H. D.

2011-06-01

The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

Kiwi: An Evaluated Library of Uncertainties in Nuclear Data and Package for Nuclear Sensitivity Studies

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

Pruet, J

2007-06-23

This report describes Kiwi, a program developed at Livermore to enable mature studies of the relation between imperfectly known nuclear physics and uncertainties in simulations of complicated systems. Kiwi includes a library of evaluated nuclear data uncertainties, tools for modifying data according to these uncertainties, and a simple interface for generating processed data used by transport codes. As well, Kiwi provides access to calculations of k eigenvalues for critical assemblies. This allows the user to check implications of data modifications against integral experiments for multiplying systems. Kiwi is written in python. The uncertainty library has the same format and directorymore » structure as the native ENDL used at Livermore. Calculations for critical assemblies rely on deterministic and Monte Carlo codes developed by B division.« less

Tumor hypoxia induces nuclear paraspeckle formation through HIF-2α dependent transcriptional activation of NEAT1 leading to cancer cell survival

PubMed Central

Choudhry, H; Albukhari, A; Morotti, M; Haider, S; Moralli, D; Smythies, J; Schödel, J; Green, C M; Camps, C; Buffa, F; Ratcliffe, P; Ragoussis, J; Harris, A L; Mole, D R

2015-01-01

Activation of cellular transcriptional responses, mediated by hypoxia-inducible factor (HIF), is common in many types of cancer, and generally confers a poor prognosis. Known to induce many hundreds of protein-coding genes, HIF has also recently been shown to be a key regulator of the non-coding transcriptional response. Here, we show that NEAT1 long non-coding RNA (lncRNA) is a direct transcriptional target of HIF in many breast cancer cell lines and in solid tumors. Unlike previously described lncRNAs, NEAT1 is regulated principally by HIF-2 rather than by HIF-1. NEAT1 is a nuclear lncRNA that is an essential structural component of paraspeckles and the hypoxic induction of NEAT1 induces paraspeckle formation in a manner that is dependent upon both NEAT1 and on HIF-2. Paraspeckles are multifunction nuclear structures that sequester transcriptionally active proteins as well as RNA transcripts that have been subjected to adenosine-to-inosine (A-to-I) editing. We show that the nuclear retention of one such transcript, F11R (also known as junctional adhesion molecule 1, JAM1), in hypoxia is dependent upon the hypoxic increase in NEAT1, thereby conferring a novel mechanism of HIF-dependent gene regulation. Induction of NEAT1 in hypoxia also leads to accelerated cellular proliferation, improved clonogenic survival and reduced apoptosis, all of which are hallmarks of increased tumorigenesis. Furthermore, in patients with breast cancer, high tumor NEAT1 expression correlates with poor survival. Taken together, these results indicate a new role for HIF transcriptional pathways in the regulation of nuclear structure and that this contributes to the pro-tumorigenic hypoxia-phenotype in breast cancer. PMID:25417700

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

NASTRAN Analysis Comparison to Shock Tube Tests Used to Simulate Nuclear Overpressures

NASA Technical Reports Server (NTRS)

Wheless, T. K.

1985-01-01

This report presents a study of the effectiveness of the NASTRAN computer code for predicting structural response to nuclear blast overpressures. NASTRAN's effectiveness is determined by comparing results against shock tube tests used to simulate nuclear overpressures. Seven panels of various configurations are compared in this study. Panel deflections are the criteria used to measure NASTRAN's effectiveness. This study is a result of needed improvements in the survivability/vulnerability analyses subjected to nuclear blast.

GRIZZLY Model of Multi-Reactive Species Diffusion, Moisture/Heat Transfer and Alkali-Silica Reaction for Simulating Concrete Aging and Degradation

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

Huang, Hai; Spencer, Benjamin W.; Cai, Guowei

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear power plants for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have accurate and reliable predictive tools to address concerns related to various aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to document themore » progress of the development and implementation of a fully coupled thermo-hydro-mechanical-chemical model in GRIZZLY code with the ultimate goal to reliably simulate and predict long-term performance and response of aged NPP concrete structures subjected to a number of aging mechanisms including external chemical attacks and volume-changing chemical reactions within concrete structures induced by alkali-silica reactions and long-term exposure to irradiation. Based on a number of survey reports of concrete aging mechanisms relevant to nuclear power plants and recommendations from researchers in concrete community, we’ve implemented three modules during FY15 in GRIZZLY code, (1) multi-species reactive diffusion model within cement materials; (2) coupled moisture and heat transfer model in concrete; and (3) anisotropic, stress-dependent, alkali-silica reaction induced swelling model. The multi-species reactive diffusion model was implemented with the objective to model aging of concrete structures subjected to aggressive external chemical attacks (e.g., chloride attack, sulfate attack, etc.). It considers multiple processes relevant to external chemical attacks such as diffusion of ions in aqueous phase within pore spaces, equilibrium chemical speciation reactions and kinetic mineral dissolution/precipitation. The moisture/heat transfer module was implemented to simulate long-term spatial and temporal evolutions of the moisture and temperature fields within concrete structures at both room and elevated temperatures. The ASR swelling model implemented in GRIZZLY code can simulate anisotropic expansions of ASR gel under either uniaxial, biaxial and triaxial stress states, and can be run simultaneously with the moisture/heat transfer model and coupled with various elastic/inelastic solid mechanics models that were implemented in GRIZZLY code previously. This report provides detailed descriptions of the governing equations, constitutive equations and numerical algorithms of the three modules implemented in GRIZZLY during FY15, simulation results of example problems and model validation results by comparing simulations with available experimental data reported in the literature. The close match between the experiments and simulations clearly demonstrate the potential of GRIZZLY code for reliable evaluation and prediction of long-term performance and response of aged concrete structures in nuclear power plants.« less

The use of the SRIM code for calculation of radiation damage induced by neutrons

NASA Astrophysics Data System (ADS)

Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

2017-12-01

Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

Current and anticipated uses of the thermal hydraulics codes at the NRC

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

Caruso, R.

1997-07-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support thesemore » needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.« less

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.

Nuclear Computational Low Energy Initiative (NUCLEI)

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

Reddy, Sanjay K.

This is the final report for University of Washington for the NUCLEI SciDAC-3. The NUCLEI -project, as defined by the scope of work, will develop, implement and run codes for large-scale computations of many topics in low-energy nuclear physics. Physics to be studied include the properties of nuclei and nuclear decays, nuclear structure and reactions, and the properties of nuclear matter. The computational techniques to be used include Quantum Monte Carlo, Configuration Interaction, Coupled Cluster, and Density Functional methods. The research program will emphasize areas of high interest to current and possible future DOE nuclear physics facilities, including ATLAS andmore » FRIB (nuclear structure and reactions, and nuclear astrophysics), TJNAF (neutron distributions in nuclei, few body systems, and electroweak processes), NIF (thermonuclear reactions), MAJORANA and FNPB (neutrino-less double-beta decay and physics beyond the Standard Model), and LANSCE (fission studies).« less

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

Development of an object-oriented ORIGEN for advanced nuclear fuel modeling applications

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

Skutnik, S.; Havloej, F.; Lago, D.

2013-07-01

The ORIGEN package serves as the core depletion and decay calculation module within the SCALE code system. A recent major re-factor to the ORIGEN code architecture as part of an overall modernization of the SCALE code system has both greatly enhanced its maintainability as well as afforded several new capabilities useful for incorporating depletion analysis into other code frameworks. This paper will present an overview of the improved ORIGEN code architecture (including the methods and data structures introduced) as well as current and potential future applications utilizing the new ORIGEN framework. (authors)

Multi-scale modeling of irradiation effects in spallation neutron source materials

NASA Astrophysics Data System (ADS)

Yoshiie, T.; Ito, T.; Iwase, H.; Kaneko, Y.; Kawai, M.; Kishida, I.; Kunieda, S.; Sato, K.; Shimakawa, S.; Shimizu, F.; Hashimoto, S.; Hashimoto, N.; Fukahori, T.; Watanabe, Y.; Xu, Q.; Ishino, S.

2011-07-01

Changes in mechanical property of Ni under irradiation by 3 GeV protons were estimated by multi-scale modeling. The code consisted of four parts. The first part was based on the Particle and Heavy-Ion Transport code System (PHITS) code for nuclear reactions, and modeled the interactions between high energy protons and nuclei in the target. The second part covered atomic collisions by particles without nuclear reactions. Because the energy of the particles was high, subcascade analysis was employed. The direct formation of clusters and the number of mobile defects were estimated using molecular dynamics (MD) and kinetic Monte-Carlo (kMC) methods in each subcascade. The third part considered damage structural evolutions estimated by reaction kinetic analysis. The fourth part involved the estimation of mechanical property change using three-dimensional discrete dislocation dynamics (DDD). Using the above four part code, stress-strain curves for high energy proton irradiated Ni were obtained.

PHASE I MATERIALS PROPERTY DATABASE DEVELOPMENT FOR ASME CODES AND STANDARDS

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

Ren, Weiju; Lin, Lianshan

2013-01-01

To support the ASME Boiler and Pressure Vessel Codes and Standard (BPVC) in modern information era, development of a web-based materials property database is initiated under the supervision of ASME Committee on Materials. To achieve efficiency, the project heavily draws upon experience from development of the Gen IV Materials Handbook and the Nuclear System Materials Handbook. The effort is divided into two phases. Phase I is planned to deliver a materials data file warehouse that offers a depository for various files containing raw data and background information, and Phase II will provide a relational digital database that provides advanced featuresmore » facilitating digital data processing and management. Population of the database will start with materials property data for nuclear applications and expand to data covering the entire ASME Code and Standards including the piping codes as the database structure is continuously optimized. The ultimate goal of the effort is to establish a sound cyber infrastructure that support ASME Codes and Standards development and maintenance.« less

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.

Rocketdyne/Westinghouse nuclear thermal rocket engine modeling

NASA Technical Reports Server (NTRS)

Glass, James F.

1993-01-01

The topics are presented in viewgraph form and include the following: systems approach needed for nuclear thermal rocket (NTR) design optimization; generic NTR engine power balance codes; rocketdyne nuclear thermal system code; software capabilities; steady state model; NTR engine optimizer code-logic; reactor power calculation logic; sample multi-component configuration; NTR design code output; generic NTR code at Rocketdyne; Rocketdyne NTR model; and nuclear thermal rocket modeling directions.

Response surface method in geotechnical/structural analysis, phase 1

NASA Astrophysics Data System (ADS)

Wong, F. S.

1981-02-01

In the response surface approach, an approximating function is fit to a long running computer code based on a limited number of code calculations. The approximating function, called the response surface, is then used to replace the code in subsequent repetitive computations required in a statistical analysis. The procedure of the response surface development and feasibility of the method are shown using a sample problem in slop stability which is based on data from centrifuge experiments of model soil slopes and involves five random soil parameters. It is shown that a response surface can be constructed based on as few as four code calculations and that the response surface is computationally extremely efficient compared to the code calculation. Potential applications of this research include probabilistic analysis of dynamic, complex, nonlinear soil/structure systems such as slope stability, liquefaction, and nuclear reactor safety.

The gene coding for small ribosomal subunit RNA in the basidiomycete Ustilago maydis contains a group I intron.

PubMed Central

De Wachter, R; Neefs, J M; Goris, A; Van de Peer, Y

1992-01-01

The nucleotide sequence of the gene coding for small ribosomal subunit RNA in the basidiomycete Ustilago maydis was determined. It revealed the presence of a group I intron with a length of 411 nucleotides. This is the third occurrence of such an intron discovered in a small subunit rRNA gene encoded by a eukaryotic nuclear genome. The other two occurrences are in Pneumocystis carinii, a fungus of uncertain taxonomic status, and Ankistrodesmus stipitatus, a green alga. The nucleotides of the conserved core structure of 101 group I intron sequences present in different genes and genome types were aligned and their evolutionary relatedness was examined. This revealed a cluster including all group I introns hitherto found in eukaryotic nuclear genes coding for small and large subunit rRNAs. A secondary structure model was designed for the area of the Ustilago maydis small ribosomal subunit RNA precursor where the intron is situated. It shows that the internal guide sequence pairing with the intron boundaries fits between two helices of the small subunit rRNA, and that minimal rearrangement of base pairs suffices to achieve the definitive secondary structure of the 18S rRNA upon splicing. PMID:1561081

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.

Light Water Reactor Sustainability Program: Survey of Models for Concrete Degradation

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

Spencer, Benjamin W.; Huang, Hai

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have predictive tools to address concerns related to aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to review and document the main aging mechanismsmore » of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.« less

The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose

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

Grebe, A.; Leveling, A.; Lu, T.

The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay gamma-quanta by the residuals in the activated structures and scoring the prompt doses of these gamma-quanta at arbitrary distances frommore » those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and showed a good agreement. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.« less

The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose

NASA Astrophysics Data System (ADS)

Grebe, A.; Leveling, A.; Lu, T.; Mokhov, N.; Pronskikh, V.

2018-01-01

The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay γ-quanta by the residuals in the activated structures and scoring the prompt doses of these γ-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and against experimental data from the CERF facility at CERN, and FermiCORD showed reasonable agreement with these. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.

Stalled RNAP-II molecules bound to non-coding rDNA spacers are required for normal nucleolus architecture.

PubMed

Freire-Picos, M A; Landeira-Ameijeiras, V; Mayán, María D

2013-07-01

The correct distribution of nuclear domains is critical for the maintenance of normal cellular processes such as transcription and replication, which are regulated depending on their location and surroundings. The most well-characterized nuclear domain, the nucleolus, is essential for cell survival and metabolism. Alterations in nucleolar structure affect nuclear dynamics; however, how the nucleolus and the rest of the nuclear domains are interconnected is largely unknown. In this report, we demonstrate that RNAP-II is vital for the maintenance of the typical crescent-shaped structure of the nucleolar rDNA repeats and rRNA transcription. When stalled RNAP-II molecules are not bound to the chromatin, the nucleolus loses its typical crescent-shaped structure. However, the RNAP-II interaction with Seh1p, or cryptic transcription by RNAP-II, is not critical for morphological changes. Copyright © 2013 John Wiley & Sons, Ltd.

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

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.

Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

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

Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library ofmore » uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.« less

Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

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

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.

1998-04-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library ofmore » uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.« less

Assessment and Requirements of Nuclear Reaction Databases for GCR Transport in the Atmosphere and Structures

NASA Technical Reports Server (NTRS)

Cucinotta, F. A.; Wilson, J. W.; Shinn, J. L.; Tripathi, R. K.

1998-01-01

The transport properties of galactic cosmic rays (GCR) in the atmosphere, material structures, and human body (self-shielding) am of interest in risk assessment for supersonic and subsonic aircraft and for space travel in low-Earth orbit and on interplanetary missions. Nuclear reactions, such as knockout and fragmentation, present large modifications of particle type and energies of the galactic cosmic rays in penetrating materials. We make an assessment of the current nuclear reaction models and improvements in these model for developing required transport code data bases. A new fragmentation data base (QMSFRG) based on microscopic models is compared to the NUCFRG2 model and implications for shield assessment made using the HZETRN radiation transport code. For deep penetration problems, the build-up of light particles, such as nucleons, light clusters and mesons from nuclear reactions in conjunction with the absorption of the heavy ions, leads to the dominance of the charge Z = 0, 1, and 2 hadrons in the exposures at large penetration depths. Light particles are produced through nuclear or cluster knockout and in evaporation events with characteristically distinct spectra which play unique roles in the build-up of secondary radiation's in shielding. We describe models of light particle production in nucleon and heavy ion induced reactions and make an assessment of the importance of light particle multiplicity and spectral parameters in these exposures.

Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

DOE PAGES

Maljovec, D.; Liu, S.; Wang, B.; ...

2015-07-14

Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

Progress Towards a Rad-Hydro Code for Modern Computing Architectures LA-UR-10-02825

NASA Astrophysics Data System (ADS)

Wohlbier, J. G.; Lowrie, R. B.; Bergen, B.; Calef, M.

2010-11-01

We are entering an era of high performance computing where data movement is the overwhelming bottleneck to scalable performance, as opposed to the speed of floating-point operations per processor. All multi-core hardware paradigms, whether heterogeneous or homogeneous, be it the Cell processor, GPGPU, or multi-core x86, share this common trait. In multi-physics applications such as inertial confinement fusion or astrophysics, one may be solving multi-material hydrodynamics with tabular equation of state data lookups, radiation transport, nuclear reactions, and charged particle transport in a single time cycle. The algorithms are intensely data dependent, e.g., EOS, opacity, nuclear data, and multi-core hardware memory restrictions are forcing code developers to rethink code and algorithm design. For the past two years LANL has been funding a small effort referred to as Multi-Physics on Multi-Core to explore ideas for code design as pertaining to inertial confinement fusion and astrophysics applications. The near term goals of this project are to have a multi-material radiation hydrodynamics capability, with tabular equation of state lookups, on cartesian and curvilinear block structured meshes. In the longer term we plan to add fully implicit multi-group radiation diffusion and material heat conduction, and block structured AMR. We will report on our progress to date.

Development of Safety Assessment Code for Decommissioning of Nuclear Facilities

NASA Astrophysics Data System (ADS)

Shimada, Taro; Ohshima, Soichiro; Sukegawa, Takenori

A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (currently JAEA). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning, and also evaluates the public dose at accidental situations including fire and explosion. As for the public, both the internal and the external doses are calculated by considering inhalation, ingestion, direct radiation from radioactive aerosols and radioactive depositions, and skyshine radiation from waste containers. For external dose for workers, the dose rate from contaminated components and structures to be dismantled is calculated. Internal dose for workers is calculated by considering dismantling conditions, e.g. cutting speed, cutting length of the components and exhaust velocity. Estimation models for dose rate and staying time were verified by comparison with the actual external dose of workers which were acquired during JPDR decommissioning project. DecDose code is expected to contribute the safety assessment for decommissioning of nuclear facilities.

Transport of Light Ions in Matter

NASA Technical Reports Server (NTRS)

Wilson, J. W.; Cucinotta, F. A.; Tai, H.; Shinn, J. L.; Chun, S. Y.; Tripathi, R. K.; Sihver, L.

1998-01-01

A recent set of light ion experiments are analyzed using the Green's function method of solving the Boltzmann equation for ions of high charge and energy (the GRNTRN transport code) and the NUCFRG2 fragmentation database generator code. Although the NUCFRG2 code reasonably represents the fragmentation of heavy ions, the effects of light ion fragmentation requires a more detailed nuclear model including shell structure and short range correlations appearing as tightly bound clusters in the light ion nucleus. The most recent NTJCFRG2 code is augmented with a quasielastic alpha knockout model and semiempirical adjustments (up to 30 percent in charge removal) in the fragmentation process allowing reasonable agreement with the experiments to be obtained. A final resolution of the appropriate cross sections must await the full development of a coupled channel reaction model in which shell structure and clustering can be accurately evaluated.

Structure-Based Low-Rank Model With Graph Nuclear Norm Regularization for Noise Removal.

PubMed

Ge, Qi; Jing, Xiao-Yuan; Wu, Fei; Wei, Zhi-Hui; Xiao, Liang; Shao, Wen-Ze; Yue, Dong; Li, Hai-Bo

2017-07-01

Nonlocal image representation methods, including group-based sparse coding and block-matching 3-D filtering, have shown their great performance in application to low-level tasks. The nonlocal prior is extracted from each group consisting of patches with similar intensities. Grouping patches based on intensity similarity, however, gives rise to disturbance and inaccuracy in estimation of the true images. To address this problem, we propose a structure-based low-rank model with graph nuclear norm regularization. We exploit the local manifold structure inside a patch and group the patches by the distance metric of manifold structure. With the manifold structure information, a graph nuclear norm regularization is established and incorporated into a low-rank approximation model. We then prove that the graph-based regularization is equivalent to a weighted nuclear norm and the proposed model can be solved by a weighted singular-value thresholding algorithm. Extensive experiments on additive white Gaussian noise removal and mixed noise removal demonstrate that the proposed method achieves a better performance than several state-of-the-art algorithms.

Iowa State University – Final Report for SciDAC3/NUCLEI

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

Vary, James P

The Iowa State University (ISU) contributions to the NUCLEI project are focused on developing, implementing and running an efficient and scalable configuration interaction code (Many-Fermion Dynamics – nuclear or MFDn) for leadership class supercomputers addressing forefront research problems in low-energy nuclear physics. We investigate nuclear structure and reactions with realistic nucleon-nucleon (NN) and three-nucleon (3N) interactions. We select a few highlights from our work that has produced a total of more than 82 refereed publications and more than 109 invited talks under SciDAC3/NUCLEI.

Summary Report of a Specialized Workshop on Nuclear Structure and Decay Data (NSDD) Evaluations

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

Nichols, Alan L.; Dimitrious, P.; Kondev, F. G.

2015-04-27

A three-day specialised workshop on Nuclear Structure and Decay Data Evaluations was organised and held at the headquarters of the International Atomic Energy Agency in Vienna, Austria, from 27 to 29 April 2015. This workshop covered a wide range of important topics and issues addressed when evaluating and maintaining the Evaluated Nuclear Structure Data File (ENSDF). The primary aim was to improve evaluators’ abilities to identify and understand the most appropriate evaluation processes to adopt in the formulation of individual ENSDF data sets. Participants assessed and reviewed existing policies, procedures and codes, and round-table discussions included the debate and resolutionmore » of specific difficulties experienced by ENSDF evaluators (i.e., all workshop participants). The contents of this report constitute a record of this workshop, based on the presentations and subsequent discussions.« less

KAOS/LIB-V: A library of nuclear response functions generated by KAOS-V code from ENDF/B-V and other data files

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

Farawila, Y.; Gohar, Y.; Maynard, C.

1989-04-01

KAOS/LIB-V: A library of processed nuclear responses for neutronics analyses of nuclear systems has been generated. The library was prepared using the KAOS-V code and nuclear data from ENDF/B-V. The library includes kerma (kinetic energy released in materials) factors and other nuclear response functions for all materials presently of interest in fusion and fission applications for 43 nonfissionable and 15 fissionable isotopes and elements. The nuclear response functions include gas production and tritium-breeding functions, and all important reaction cross sections. KAOS/LIB-V employs the VITAMIN-E weighting function and energy group structure of 174 neutron groups. Auxiliary nuclear data bases, e.g., themore » Japanese evaluated nuclear data library JENDL-2 were used as a source of isotopic cross sections when these data are not provided in ENDF/B-V files for a natural element. These are needed mainly to estimate average quantities such as effective Q-values for the natural element. This analysis of local energy deposition was instrumental in detecting and understanding energy balance deficiencies and other problems in the ENDF/B-V data. Pertinent information about the library and a graphical display of the main nuclear response functions for all materials in the library are given. 35 refs.« less

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.

Code qualification of structural materials for AFCI advanced recycling reactors.

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

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

Structural map of Kaposi sarcoma-associated herpesvirus RNA provides clues to molecular interactions | Center for Cancer Research

Cancer.gov

Scientists from CCR have generated a comprehensive structural map of Kaposi sarcoma-associated herpesvirus polyadenylated nuclear (PAN) RNA, a long non-coding RNA that helps the virus evade detection by its host’s immune system. The findings open new oppportunites to study the life cycle of this cancer-causing virus.  Learn more...

Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

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

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library ofmore » uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.« 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

WESCOM. A Fortran Code for Evaluation of Nuclear Weapon Effects on Satellite Communications. Volume 2. Code Structure

DTIC Science & Technology

1981-01-31

quantities for h i ;.;h-:t It i 1 ndc hurst s 1BMI.I Determines t ime-independent fireball quantities for low-altitude bursts 10 Table 1...of reference Oval of Cassini (km) LAFBP - vortex longitudinal radius (km) LAFBP - vortex transverse radius (km) Power law exponent Inner scale...Maximum slant range of ionization from transmitter (km) Power law exponent Frequency (Hz) Striation velocity flag Propagation path index Radius

RNA Nuclear Export: From Neurological Disorders to Cancer.

PubMed

Hautbergue, Guillaume M

2017-01-01

The presence of a nuclear envelope, also known as nuclear membrane, defines the structural framework of all eukaryotic cells by separating the nucleus, which contains the genetic material, from the cytoplasm where the synthesis of proteins takes place. Translation of proteins in Eukaryotes is thus dependent on the active transport of DNA-encoded RNA molecules through pores embedded within the nuclear membrane. Several mechanisms are involved in this process generally referred to as RNA nuclear export or nucleocytoplasmic transport of RNA. The regulated expression of genes requires the nuclear export of protein-coding messenger RNA molecules (mRNAs) as well as non-coding RNAs (ncRNAs) together with proteins and pre-assembled ribosomal subunits. The nuclear export of mRNAs is intrinsically linked to the co-transcriptional processing of nascent transcripts synthesized by the RNA polymerase II. This functional coupling is essential for the survival of cells allowing for timely nuclear export of fully processed transcripts, which could otherwise cause the translation of abnormal proteins such as the polymeric repeat proteins produced in some neurodegenerative diseases. Alterations of the mRNA nuclear export pathways can also lead to genome instability and to various forms of cancer. This chapter will describe the molecular mechanisms driving the nuclear export of RNAs with a particular emphasis on mRNAs. It will also review their known alterations in neurological disorders and cancer, and the recent opportunities they offer for the potential development of novel therapeutic strategies.

Moving Towards a State of the Art Charge-Exchange Reaction Code

NASA Astrophysics Data System (ADS)

Poxon-Pearson, Terri; Nunes, Filomena; Potel, Gregory

2017-09-01

Charge-exchange reactions have a wide range of applications, including late stellar evolution, constraining the matrix elements for neutrinoless double β-decay, and exploring symmetry energy and other aspects of exotic nuclear matter. Still, much of the reaction theory needed to describe these transitions is underdeveloped and relies on assumptions and simplifications that are often extended outside of their region of validity. In this work, we have begun to move towards a state of the art charge-exchange reaction code. As a first step, we focus on Fermi transitions using a Lane potential in a few body, Distorted Wave Born Approximation (DWBA) framework. We have focused on maintaining a modular structure for the code so we can later incorporate complications such as nonlocality, breakup, and microscopic inputs. Results using this new charge-exchange code will be shown compared to the analysis in for the case of 48Ca(p,n)48Sc. This work was supported in part by the National Nuclear Security Administration under the Stewardship Science Academic Alliances program through the U.S. DOE Cooperative Agreement No. DE- FG52-08NA2855.

Cracking the Sugar Code by Mass Spectrometry - An Invited Perspective in Honor of Dr. Catherine E. Costello, Recipient of the 2017 ASMS Distinguished Contribution Award

NASA Astrophysics Data System (ADS)

Mirgorodskaya, Ekaterina; Karlsson, Niclas G.; Sihlbom, Carina; Larson, Göran; Nilsson, Carol L.

2018-04-01

The structural study of glycans and glycoconjugates is essential to assign their roles in homeostasis, health, and disease. Once dominated by nuclear magnetic resonance spectroscopy, mass spectrometric methods have become the preferred toolbox for the determination of glycan structures at high sensitivity. The patterns of such structures in different cellular states now allow us to interpret the sugar codes in health and disease, based on structure-function relationships. Dr. Catherine E. Costello was the 2017 recipient of the American Society for Mass Spectrometry's Distinguished Contribution Award. In this Perspective article, we describe her seminal work in a historical and geographical context and review the impact of her research accomplishments in the field. 8[Figure not available: see fulltext.

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.

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.

Geophysical Plasmas and Atmospheric Modeling.

DTIC Science & Technology

1982-01-01

analysis of structuring observed in the STARFISH event and early time structure formation in the LWIR predicted for a standard event, (2) IR structure...for the LWIR problem, the SCORPIO code was coupled to a numerical module which describes the behavior of the Rayleigh-Taylor insta- bility, as discussed...the LWIR C1-14p) induced by sunlight and earthshine. Of the constituents in nuclear debris the uranium atom is of particular interest because it

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)

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

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.

A comparison of total reaction cross section models used in particle and heavy ion transport codes

NASA Astrophysics Data System (ADS)

Sihver, Lembit; Lantz, M.; Takechi, M.; Kohama, A.; Ferrari, A.; Cerutti, F.; Sato, T.

To be able to calculate the nucleon-nucleus and nucleus-nucleus total reaction cross sections with precision is very important for studies of basic nuclear properties, e.g. nuclear structure. This is also of importance for particle and heavy ion transport calculations because, in all particle and heavy ion transport codes, the probability function that a projectile particle will collide within a certain distance x in the matter depends on the total reaction cross sections. Furthermore, the total reaction cross sections will also scale the calculated partial fragmentation cross sections. It is therefore crucial that accurate total reaction cross section models are used in the transport calculations. In this paper, different models for calculating nucleon-nucleus and nucleus-nucleus total reaction cross sections are compared and discussed.

The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

NASA Astrophysics Data System (ADS)

Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

2018-05-01

Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

NASA Astrophysics Data System (ADS)

Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

2018-05-01

CESAR stands in French for "simplified depletion applied to reprocessing". The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ˜400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with "industrial nuclear" constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR's). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA - Cadarache.

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

Contributions to the NUCLEI SciDAC-3 Project

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

Bogner, Scott; Nazarewicz, Witek

This is the Final Report for Michigan State University for the NUCLEI SciDAC-3 project. The NUCLEI project, as defined by the scope of work, has developed, implemented and run codes for large-scale computations of many topics in low-energy nuclear physics. Physics studied included the properties of nuclei and nuclear decays, nuclear structure and reactions, and the properties of nuclear matter. The computational techniques used included Configuration Interaction, Coupled Cluster, and Density Functional methods. The research program emphasized areas of high interest to current and possible future DOE nuclear physics facilities, including ATLAS at ANL and FRIB at MSU (nuclear structuremore » and reactions, and nuclear astrophysics), TJNAF (neutron distributions in nuclei, few body systems, and electroweak processes), NIF (thermonuclear reactions), MAJORANA and FNPB (neutrinoless double-beta decay and physics beyond the Standard Model), and LANSCE (fission studies).« less

Determination of the Order of Passes of AN Austenitic Weld by Optimization of AN Inversion Process of Ultrasound Data

NASA Astrophysics Data System (ADS)

Gueudré, C.; Marrec, L. Le; Chekroun, M.; Moysan, J.; Chassignole, B.; Corneloup, G.

2011-06-01

Multipass welds made in austenitic stainless steel, in the primary circuit of nuclear power plants with pressurized water reactors, are characterized by an anisotropic and heterogeneous structure that disturbs the ultrasonic propagation and challenge the ultrasonic non-destructive testing. The simulation in this type of structure is now possible thanks to the MINA code which allows the grain orientation modeling taking into account the welding process, and the ATHENA code to exactly simulate the ultrasonic propagation. We propose studying the case where the order of the passes is unknown to estimate the possibility of reconstructing this important parameter by ultrasound measures. The first results are presented.

Calculations vs. measurements of remnant dose rates for SNS spent structures

NASA Astrophysics Data System (ADS)

Popova, I. I.; Gallmeier, F. X.; Trotter, S.; Dayton, M.

2018-06-01

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction. Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.

Calculations vs. measurements of remnant dose rates for SNS spent structures

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

Popova, Irina I.; Gallmeier, Franz X.; Trotter, Steven M.

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction.more » Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.« less

Tornado and extreme wind design criteria for nuclear power plants

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

None

1973-12-01

Nuclear power plant design criteria for tornadoes and extreme winds are presented. Data, formulas, and procedures for determining maximum wind loading on structures and parts of structures are included. Extreme wind loading is applied to structures using methods and procedures consistent with ANSI Building Code A58.1- 1972. The design wind velocities specified generally exceed 100-year recurrent interval winds. Tornado wind loading is applied to structures using procedures paralleling those for extrene winds with additional criteria resulting from the atmospheric pressure change accompanying tornadoes and tornado missile inipact effects. Tornado loading for the 48 contiguous United States is specified for twomore » major zones separated by the Continental Divide. A cross reference listing items related to Atomic Energy Commission Safety Analysis Report format is provided. Development supporting tornado criteria is included. (auth)« less

Atomic structure of the Y complex of the nuclear pore

DOE PAGES

Kelley, Kotaro; Knockenhauer, Kevin E.; Kabachinski, Greg; ...

2015-03-30

The nuclear pore complex (NPC) is the principal gateway for transport into and out of the nucleus. Selectivity is achieved through the hydrogel-like core of the NPC. The structural integrity of the NPC depends on ~15 architectural proteins, which are organized in distinct subcomplexes to form the >40-MDa ring-like structure. In this paper, we present the 4.1-Å crystal structure of a heterotetrameric core element ('hub') of the Y complex, the essential NPC building block, from Myceliophthora thermophila. Using the hub structure together with known Y-complex fragments, we built the entire ~0.5-MDa Y complex. Our data reveal that the conserved coremore » of the Y complex has six rather than seven members. Finally, evolutionarily distant Y-complex assemblies share a conserved core that is very similar in shape and dimension, thus suggesting that there are closely related architectural codes for constructing the NPC in all eukaryotes.« less

Standardized verification of fuel cycle modeling

DOE PAGES

Feng, B.; Dixon, B.; Sunny, E.; ...

2016-04-05

A nuclear fuel cycle systems modeling and code-to-code comparison effort was coordinated across multiple national laboratories to verify the tools needed to perform fuel cycle analyses of the transition from a once-through nuclear fuel cycle to a sustainable potential future fuel cycle. For this verification study, a simplified example transition scenario was developed to serve as a test case for the four systems codes involved (DYMOND, VISION, ORION, and MARKAL), each used by a different laboratory participant. In addition, all participants produced spreadsheet solutions for the test case to check all the mass flows and reactor/facility profiles on a year-by-yearmore » basis throughout the simulation period. The test case specifications describe a transition from the current US fleet of light water reactors to a future fleet of sodium-cooled fast reactors that continuously recycle transuranic elements as fuel. After several initial coordinated modeling and calculation attempts, it was revealed that most of the differences in code results were not due to different code algorithms or calculation approaches, but due to different interpretations of the input specifications among the analysts. Therefore, the specifications for the test case itself were iteratively updated to remove ambiguity and to help calibrate interpretations. In addition, a few corrections and modifications were made to the codes as well, which led to excellent agreement between all codes and spreadsheets for this test case. Although no fuel cycle transition analysis codes matched the spreadsheet results exactly, all remaining differences in the results were due to fundamental differences in code structure and/or were thoroughly explained. As a result, the specifications and example results are provided so that they can be used to verify additional codes in the future for such fuel cycle transition scenarios.« less

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

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

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

i-PI: A Python interface for ab initio path integral molecular dynamics simulations

NASA Astrophysics Data System (ADS)

Ceriotti, Michele; More, Joshua; Manolopoulos, David E.

2014-03-01

Recent developments in path integral methodology have significantly reduced the computational expense of including quantum mechanical effects in the nuclear motion in ab initio molecular dynamics simulations. However, the implementation of these developments requires a considerable programming effort, which has hindered their adoption. Here we describe i-PI, an interface written in Python that has been designed to minimise the effort required to bring state-of-the-art path integral techniques to an electronic structure program. While it is best suited to first principles calculations and path integral molecular dynamics, i-PI can also be used to perform classical molecular dynamics simulations, and can just as easily be interfaced with an empirical forcefield code. To give just one example of the many potential applications of the interface, we use it in conjunction with the CP2K electronic structure package to showcase the importance of nuclear quantum effects in high-pressure water. Catalogue identifier: AERN_v1_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AERN_v1_0.html Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 No. of lines in distributed program, including test data, etc.: 138626 No. of bytes in distributed program, including test data, etc.: 3128618 Distribution format: tar.gz Programming language: Python. Computer: Multiple architectures. Operating system: Linux, Mac OSX, Windows. RAM: Less than 256 Mb Classification: 7.7. External routines: NumPy Nature of problem: Bringing the latest developments in the modelling of nuclear quantum effects with path integral molecular dynamics to ab initio electronic structure programs with minimal implementational effort. Solution method: State-of-the-art path integral molecular dynamics techniques are implemented in a Python interface. Any electronic structure code can be patched to receive the atomic coordinates from the Python interface, and to return the forces and energy that are used to integrate the equations of motion. Restrictions: This code only deals with distinguishable particles. It does not include fermonic or bosonic exchanges between equivalent nuclei, which can become important at very low temperatures. Running time: Depends dramatically on the nature of the simulation being performed. A few minutes for short tests with empirical force fields, up to several weeks for production calculations with ab initio forces. The examples provided with the code run in less than an hour.

Multiscale integral analysis of a HT leakage in a fusion nuclear power plant

NASA Astrophysics Data System (ADS)

Velarde, M.; Fradera, J.; Perlado, J. M.; Zamora, I.; Martínez-Saban, E.; Colomer, C.; Briani, P.

2016-05-01

The present work presents an example of the application of an integral methodology based on a multiscale analysis that covers the whole tritium cycle within a nuclear fusion power plant, from a micro scale, analyzing key components where tritium is leaked through permeation, to a macro scale, considering its atmospheric transport. A leakage from the Nuclear Power Plants, (NPP) primary to the secondary side of a heat exchanger (HEX) is considered for the present example. Both primary and secondary loop coolants are assumed to be He. Leakage is placed inside the HEX, leaking tritium in elementary tritium (HT) form to the secondary loop where it permeates through the piping structural material to the exterior. The Heating Ventilation and Air Conditioning (HVAC) system removes the leaked tritium towards the NPP exhaust. The HEX is modelled with system codes and coupled to Computational Fluid Dynamic (CFD) to account for tritium dispersion inside the nuclear power plants buildings and in site environment. Finally, tritium dispersion is calculated with an atmospheric transport code and a dosimetry analysis is carried out. Results show how the implemented methodology is capable of assessing the impact of tritium from the microscale to the atmospheric scale including the dosimetric aspect.

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.

77 FR 3073 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-23

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2008-0554] RIN 3150-AI35 American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases; Corrections AGENCY: Nuclear Regulatory... the American Society of Mechanical Engineers, Three Park Avenue, New York, NY 10016, phone (800) 843...

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.

78 FR 48727 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-09

... Analysis Reports for Nuclear Power Plants: LWR Edition,'' Section 3.9.3 ``ASME Code Class 1, 2, and 3...'s Agencywide Documents Access and Management System (ADAMS): You may access publicly available... operational readiness of snubbers (ADAMS Accession No. ML070720041), and review interfaces have been updated...

Preliminary Numerical Simulation of IR Structure Development in a Hypothetical Uranium Release.

DTIC Science & Technology

1981-11-16

art Identify by block nAsb.’) IR Structure Power spectrum Uranium release Parallax effects Numerical simulation PHARO code Isophots LWIR 20. _PSTRACT...release at 200 km altitude. Of interest is the LWIR emission from uranium oxide ions, induced by sunlight and earthshine. Assuming a one-level fluid...defense systems of long wave infrared ( LWIR ) emissions from metallic oxides in the debris from a high altitude nuclear explosion (HANE) is an

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

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

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

Identification of Putative Nuclear Receptors and Steroidogenic Enzymes in Murray-Darling Rainbowfish (Melanotaenia fluviatilis) Using RNA-Seq and De Novo Transcriptome Assembly.

PubMed

Bain, Peter A; Papanicolaou, Alexie; Kumar, Anupama

2015-01-01

Murray-Darling rainbowfish (Melanotaenia fluviatilis [Castelnau, 1878]; Atheriniformes: Melanotaeniidae) is a small-bodied teleost currently under development in Australasia as a test species for aquatic toxicological studies. To date, efforts towards the development of molecular biomarkers of contaminant exposure have been hindered by the lack of available sequence data. To address this, we sequenced messenger RNA from brain, liver and gonads of mature male and female fish and generated a high-quality draft transcriptome using a de novo assembly approach. 149,742 clusters of putative transcripts were obtained, encompassing 43,841 non-redundant protein-coding regions. Deduced amino acid sequences were annotated by functional inference based on similarity with sequences from manually curated protein sequence databases. The draft assembly contained protein-coding regions homologous to 95.7% of the complete cohort of predicted proteins from the taxonomically related species, Oryzias latipes (Japanese medaka). The mean length of rainbowfish protein-coding sequences relative to their medaka homologues was 92.1%, indicating that despite the limited number of tissues sampled a large proportion of the total expected number of protein-coding genes was captured in the study. Because of our interest in the effects of environmental contaminants on endocrine pathways, we manually curated subsets of coding regions for putative nuclear receptors and steroidogenic enzymes in the rainbowfish transcriptome, revealing 61 candidate nuclear receptors encompassing all known subfamilies, and 41 putative steroidogenic enzymes representing all major steroidogenic enzymes occurring in teleosts. The transcriptome presented here will be a valuable resource for researchers interested in biomarker development, protein structure and function, and contaminant-response genomics in Murray-Darling rainbowfish.

HZETRN: A heavy ion/nucleon transport code for space radiations

NASA Technical Reports Server (NTRS)

Wilson, John W.; Chun, Sang Y.; Badavi, Forooz F.; Townsend, Lawrence W.; Lamkin, Stanley L.

1991-01-01

The galactic heavy ion transport code (GCRTRN) and the nucleon transport code (BRYNTRN) are integrated into a code package (HZETRN). The code package is computer efficient and capable of operating in an engineering design environment for manned deep space mission studies. The nuclear data set used by the code is discussed including current limitations. Although the heavy ion nuclear cross sections are assumed constant, the nucleon-nuclear cross sections of BRYNTRN with full energy dependence are used. The relation of the final code to the Boltzmann equation is discussed in the context of simplifying assumptions. Error generation and propagation is discussed, and comparison is made with simplified analytic solutions to test numerical accuracy of the final results. A brief discussion of biological issues and their impact on fundamental developments in shielding technology is given.

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.

Code System to Calculate Tornado-Induced Flow Material Transport.

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

ANDRAE, R. W.

1999-11-18

Version: 00 TORAC models tornado-induced flows, pressures, and material transport within structures. Its use is directed toward nuclear fuel cycle facilities and their primary release pathway, the ventilation system. However, it is applicable to other structures and can model other airflow pathways within a facility. In a nuclear facility, this network system could include process cells, canyons, laboratory offices, corridors, and offgas systems. TORAC predicts flow through a network system that also includes ventilation system components such as filters, dampers, ducts, and blowers. These ventilation system components are connected to the rooms and corridors of the facility to form amore » complete network for moving air through the structure and, perhaps, maintaining pressure levels in certain areas. The material transport capability in TORAC is very basic and includes convection, depletion, entrainment, and filtration of material.« less

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.

Modelling the attenuation in the ATHENA finite elements code for the ultrasonic testing of austenitic stainless steel welds.

PubMed

Chassignole, B; Duwig, V; Ploix, M-A; Guy, P; El Guerjouma, R

2009-12-01

Multipass welds made in austenitic stainless steel, in the primary circuit of nuclear power plants with pressurized water reactors, are characterized by an anisotropic and heterogeneous structure that disturbs the ultrasonic propagation and makes ultrasonic non-destructive testing difficult. The ATHENA 2D finite element simulation code was developed to help understand the various physical phenomena at play. In this paper, we shall describe the attenuation model implemented in this code to give an account of wave scattering phenomenon through polycrystalline materials. This model is in particular based on the optimization of two tensors that characterize this material on the basis of experimental values of ultrasonic velocities attenuation coefficients. Three experimental configurations, two of which are representative of the industrial welds assessment case, are studied in view of validating the model through comparison with the simulation results. We shall thus provide a quantitative proof that taking into account the attenuation in the ATHENA code dramatically improves the results in terms of the amplitude of the echoes. The association of the code and detailed characterization of a weld's structure constitutes a remarkable breakthrough in the interpretation of the ultrasonic testing on this type of component.

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.

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.

Error correcting coding-theory for structured light illumination systems

NASA Astrophysics Data System (ADS)

Porras-Aguilar, Rosario; Falaggis, Konstantinos; Ramos-Garcia, Ruben

2017-06-01

Intensity discrete structured light illumination systems project a series of projection patterns for the estimation of the absolute fringe order using only the temporal grey-level sequence at each pixel. This work proposes the use of error-correcting codes for pixel-wise correction of measurement errors. The use of an error correcting code is advantageous in many ways: it allows reducing the effect of random intensity noise, it corrects outliners near the border of the fringe commonly present when using intensity discrete patterns, and it provides a robustness in case of severe measurement errors (even for burst errors where whole frames are lost). The latter aspect is particular interesting in environments with varying ambient light as well as in critical safety applications as e.g. monitoring of deformations of components in nuclear power plants, where a high reliability is ensured even in case of short measurement disruptions. A special form of burst errors is the so-called salt and pepper noise, which can largely be removed with error correcting codes using only the information of a given pixel. The performance of this technique is evaluated using both simulations and experiments.

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

Simulation of multi-photon emission isotopes using time-resolved SimSET multiple photon history generator

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

Chiang, Chih-Chieh; Lin, Hsin-Hon; Lin, Chang-Shiun

Abstract-Multiple-photon emitters, such as In-111 or Se-75, have enormous potential in the field of nuclear medicine imaging. For example, Se-75 can be used to investigate the bile acid malabsorption and measure the bile acid pool loss. The simulation system for emission tomography (SimSET) is a well-known Monte Carlo simulation (MCS) code in nuclear medicine for its high computational efficiency. However, current SimSET cannot simulate these isotopes due to the lack of modeling of complex decay scheme and the time-dependent decay process. To extend the versatility of SimSET for simulation of those multi-photon emission isotopes, a time-resolved multiple photon history generatormore » based on SimSET codes is developed in present study. For developing the time-resolved SimSET (trSimSET) with radionuclide decay process, the new MCS model introduce new features, including decay time information and photon time-of-flight information, into this new code. The half-life of energy states were tabulated from the Evaluated Nuclear Structure Data File (ENSDF) database. The MCS results indicate that the overall percent difference is less than 8.5% for all simulation trials as compared to GATE. To sum up, we demonstrated that time-resolved SimSET multiple photon history generator can have comparable accuracy with GATE and keeping better computational efficiency. The new MCS code is very useful to study the multi-photon imaging of novel isotopes that needs the simulation of lifetime and the time-of-fight measurements. (authors)« less

Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

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

Williams, P. T.; Dickson, T. L.; Yin, S.

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include themore » NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.« less

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.

Systems for the Intermodal Routing of Spent Nuclear Fuel

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

Peterson, Steven K; Liu, Cheng

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable systemmore » for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.« less

Risk Informed Design and Analysis Criteria for Nuclear Structures

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

Salmon, Michael W.

2015-06-17

Target performance can be achieved by defining design basis ground motion from results of a probabilistic seismic hazards assessment, and introducing known levels of conservatism in the design above the DBE. ASCE 4, 43, DOE-STD-1020 defined the DBE at 4x10-4 and introduce only slight levels of conservatism in response. ASCE 4, 43, DOE-STD-1020 assume code capacities shoot for about 98% NEP. There is a need to have a uniform target (98% NEP) for code developers (ACI, AISC, etc.) to aim for. In considering strengthening options, one must also consider cost/risk reduction achieved.

Double differential light charged particle emission cross sections for some structural fusion materials

NASA Astrophysics Data System (ADS)

Sarpün, Ismail Hakki; n, Abdullah Aydı; Tel, Eyyup

2017-09-01

In fusion reactors, neutron induced radioactivity strongly depends on the irradiated material. So, a proper selection of structural materials will have been limited the radioactive inventory in a fusion reactor. First-wall and blanket components have high radioactivity concentration due to being the most flux-exposed structures. The main objective of fusion structural material research is the development and selection of materials for reactor components with good thermo-mechanical and physical properties, coupled with low-activation characteristics. Double differential light charged particle emission cross section, which is a fundamental data to determine nuclear heating and material damages in structural fusion material research, for some elements target nuclei have been calculated by the TALYS 1.8 nuclear reaction code at 14-15 MeV neutron incident energy and compared with available experimental data in EXFOR library. Direct, compound and pre-equilibrium reaction contribution have been theoretically calculated and dominant contribution have been determined for each emission of proton, deuteron and alpha particle.

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.

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

Summary of papers on current and anticipated uses of thermal-hydraulic codes

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

Caruso, R.

1997-07-01

The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especiallymore » faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).« less

Online Monitoring of Concrete Structures in Nuclear Power Plants: Interim Report

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

Mahadevan, Sankaran; Cai, Guowei; Agarwal, Vivek

The existing fleet of nuclear power plants in the United States have initial operating licenses of 40 years, and many of these plants have applied for and received license extensions. As plant structures, systems, and components age, their useful life—considering both structural integrity and performance—is reduced as a result of deterioration of the materials. Assessment and management of aging concrete structures in nuclear plants require a more systematic approach than simple reliance on existing code-based design margins of safety. Structural health monitoring is required to produce actionable information regarding structural integrity that supports operational and maintenance decisions. The online monitoringmore » of concrete structures project conducted under the Advanced Instrumentation, Information, and Control Technologies Pathway of the Light Water Reactor Sustainability program at Idaho National Laboratory is seeking to develop and demonstrate capabilities for concrete structures health monitoring. Through this research project, several national laboratories and Vanderbilt University propose to develop a framework of research activities for the health monitoring of nuclear power plant concrete structures that includes the integration of four elements—damage modeling, monitoring, data analytics, and uncertainty quantification. This report briefly discusses activities in this project during October-December, 2014. The most significant activity during this period was the organizing of a two-day workshop on research needs in online monitoring of concrete structures, hosted by Vanderbilt University in November 2014. Thirty invitees from academia, industry and government participated in the workshop. The presentations and discussions at the workshop surveyed current activities related to concrete structures deterioration modeling and monitoring, and identified the challenges, knowledge gaps, and opportunities for advancing the state of the art; these discussions are summarized in this report« less

A Stochastic Model of Space Radiation Transport as a Tool in the Development of Time-Dependent Risk Assessment

NASA Technical Reports Server (NTRS)

Kim, Myung-Hee Y.; Nounu, Hatem N.; Ponomarev, Artem L.; Cucinotta, Francis A.

2011-01-01

A new computer model, the GCR Event-based Risk Model code (GERMcode), was developed to describe biophysical events from high-energy protons and heavy ions that have been studied at the NASA Space Radiation Laboratory (NSRL) [1] for the purpose of simulating space radiation biological effects. In the GERMcode, the biophysical description of the passage of heavy ions in tissue and shielding materials is made with a stochastic approach that includes both ion track structure and nuclear interactions. The GERMcode 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 [2]. The QMSFRG model has been shown to be in excellent agreement with available experimental data for nuclear fragmentation cross sections

Crashworthiness: Planes, trains, and automobiles

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

Logan, R.W.; Tokarz, F.J.; Whirley, R.G.

A powerful DYNA3D computer code simulates the dynamic effects of stress traveling through structures. It is the most advanced modeling tool available to study crashworthiness problems and to analyze impacts. Now used by some 1000 companies, government research laboratories, and universities in the U.S. and abroad, DYNA3D is also a preeminent example of successful technology transfer. The initial interest in such a code was to simulate the structural response of weapons systems. The need was to model not the explosive or nuclear events themselves but rather the impacts of weapons systems with the ground, tracking the stress waves as theymore » move through the object. This type of computer simulation augmented or, in certain cases, reduced the need for expensive and time-consuming crash testing.« 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...

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.

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.

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

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

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.

Role of nuclear reactions on stellar evolution of intermediate-mass stars

NASA Astrophysics Data System (ADS)

Möller, H.; Jones, S.; Fischer, T.; Martínez-Pinedo, G.

2018-01-01

The evolution of intermediate-mass stars (8 - 12 solar masses) represents one of the most challenging subjects in nuclear astrophysics. Their final fate is highly uncertain and strongly model dependent. They can become white dwarfs, they can undergo electron-capture or core-collapse supernovae or they might even proceed towards explosive oxygen burning and a subsequent thermonuclear explosion. We believe that an accurate description of nuclear reactions is crucial for the determination of the pre-supernova structure of these stars. We argue that due to the possible development of an oxygen-deflagration, a hydrodynamic description has to be used. We implement a nuclear reaction network with ∼200 nuclear species into the implicit hydrodynamic code AGILE. The reaction network considers all relevant nuclear electron captures and beta-decays. For selected relevant nuclear species, we include a set of updated reaction rates, for which we discuss the role for the evolution of the stellar core, at the example of selected stellar models. We find that the final fate of these intermediate-mass stars depends sensitively on the density threshold for weak processes that deleptonize the core.

Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

NASA Astrophysics Data System (ADS)

Piro, Markus Hans Alexander

Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system components at each iterative step, and the objective is to minimize the residuals of the mass balance equations. Several numerical advantages are achieved through this simplification. In particular, computational expense is reduced and the rate of convergence is enhanced. Furthermore, the software has demonstrated the ability to solve systems involving as many as 118 component elements. An early version of the code has already been integrated into the Advanced Multi-Physics (AMP) code under development by the Oak Ridge National Laboratory, Los Alamos National Laboratory, Idaho National Laboratory and Argonne National Laboratory. Keywords: Engineering, Nuclear -- 0552, Engineering, Material Science -- 0794, Chemistry, Mathematics -- 0405, Computer Science -- 0984

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.

A new algorithm to handle finite nuclear mass effects in electronic calculations: the ISOTOPE program.

PubMed

Gonçalves, Cristina P; Mohallem, José R

2004-11-15

We report the development of a simple algorithm to modify quantum chemistry codes based on the LCAO procedure, to account for the isotope problem in electronic structure calculations. No extra computations are required compared to standard Born-Oppenheimer calculations. An upgrade of the Gamess package called ISOTOPE is presented, and its applicability is demonstrated in some examples.

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.

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.

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.

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.

Seismic assessment of Technical Area V (TA-V).

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

Medrano, Carlos S.

The Technical Area V (TA-V) Seismic Assessment Report was commissioned as part of Sandia National Laboratories (SNL) Self Assessment Requirement per DOE O 414.1, Quality Assurance, for seismic impact on existing facilities at Technical Area-V (TA-V). SNL TA-V facilities are located on an existing Uniform Building Code (UBC) Seismic Zone IIB Site within the physical boundary of the Kirtland Air Force Base (KAFB). The document delineates a summary of the existing facilities with their safety-significant structure, system and components, identifies DOE Guidance, conceptual framework, past assessments and the present Geological and Seismic conditions. Building upon the past information and themore » evolution of the new seismic design criteria, the document discusses the potential impact of the new standards and provides recommendations based upon the current International Building Code (IBC) per DOE O 420.1B, Facility Safety and DOE G 420.1-2, Guide for the Mitigation of Natural Phenomena Hazards for DOE Nuclear Facilities and Non-Nuclear Facilities.« less

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.

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

Methodology comparison for gamma-heating calculations in material-testing reactors

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

Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A.

2015-07-01

The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physicalmore » models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear heating is represented by the physical quantity called absorbed dose (energy deposition induced by particle-matter interactions, divided by mass). Its calculation with Monte Carlo codes is possible but computationally expensive as it requires transport simulation of charged particles, along with neutrons and photons. For that reason, the calculation of another physical quantity, called KERMA, is often preferred, as KERMA calculation with Monte Carlo codes only requires transport of neutral particles. However, KERMA is only an estimator of the absorbed dose and many conditions must be fulfilled for KERMA to be equal to absorbed dose, including so-called condition of electronic equilibrium. Also, Monte Carlo computations of absorbed dose still present some physical approximations, even though there is only a limited number of them. Some of these approximations are linked to the way how Monte Carlo codes apprehend the transport simulation of charged particles and the productive and destructive interactions between photons, electrons and positrons. There exists a huge variety of electromagnetic shower models which tackle this topic. Differences in the implementation of these models can lead to discrepancies in calculated values of absorbed dose between different Monte Carlo codes. The magnitude of order of such potential discrepancies should be quantified for JHR gamma-heating calculations. We consequently present a two-pronged plan. In a first phase, we intend to perform compared absorbed dose / KERMA Monte Carlo calculations in the JHR. This way, we will study the presence or absence of electronic equilibrium in the different JHR structures and experimental devices and we will give recommendations for the choice of KERMA or absorbed dose when calculating gamma heating in the JHR. In a second phase, we intend to perform compared TRIPOLI4 / MCNP absorbed dose calculations in a simplified JHR-representative geometry. For this comparison, we will use the same nuclear data library for both codes (the European library JEFF3.1.1 and photon library EPDL97) so as to isolate the effects from electromagnetic shower models on absorbed dose calculation. This way, we hope to get insightful feedback on these models and their implementation in Monte Carlo codes. (authors)« less

Structure of the classical scrape-off layer of a tokamak

NASA Astrophysics Data System (ADS)

Rozhansky, V.; Kaveeva, E.; Senichenkov, I.; Vekshina, E.

2018-03-01

The structure of the scrape-off layer (SOL) of a tokamak with little or no turbulent transport is analyzed. The analytical estimates of the density and electron temperature fall-off lengths of the SOL are put forward. It is demonstrated that the SOL width could be of the order of the ion poloidal gyroradius, as suggested in Goldston (2012 Nuclear Fusion 52 013009). The analytical results are supported by the results of the 2D simulations of the edge plasma with reduced transport coefficients performed by SOLPS-ITER transport code.

Energy and Technology Review

NASA Astrophysics Data System (ADS)

Poggio, Andrew J.

1988-10-01

This issue of Energy and Technology Review contains: Neutron Penumbral Imaging of Laser-Fusion Targets--using our new penumbral-imaging diagnostic, we have obtained the first images that can be used to measure directly the deuterium-tritium burn region in laser-driven fusion targets; Computed Tomography for Nondestructive Evaluation--various computed tomography systems and computational techniques are used in nondestructive evaluation; Three-Dimensional Image Analysis for Studying Nuclear Chromatin Structure--we have developed an optic-electronic system for acquiring cross-sectional views of cell nuclei, and computer codes to analyze these images and reconstruct the three-dimensional structures they represent; Imaging in the Nuclear Test Program--advanced techniques produce images of unprecedented detail and resolution from Nevada Test Site data; and Computational X-Ray Holography--visible-light experiments and numerically simulated holograms test our ideas about an X-ray microscope for biological research.

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.

Evaluation of Recent Upgrades to the NESS (Nuclear Engine System Simulation) Code

NASA Technical Reports Server (NTRS)

Fittje, James E.; Schnitzler, Bruce G.

2008-01-01

The Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for exploratory expeditions to the moon, Mars, and beyond. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the Rover/NERVA program from 1955 to 1973. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design, and a comparison of its results to the Small Nuclear Rocket Engine (SNRE) design.

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

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

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

Code dependencies of pre-supernova evolution and nucleosynthesis in massive stars: evolution to the end of core helium burning

DOE PAGES

Jones, S.; Hirschi, R.; Pignatari, M.; ...

2015-01-15

We present a comparison of 15M ⊙ , 20M ⊙ and 25M ⊙ stellar models from three different codes|GENEC, KEPLER and MESA|and their nucleosynthetic yields. The models are calculated from the main sequence up to the pre-supernova (pre-SN) stage and do not include rotation. The GENEC and KEPLER models hold physics assumptions that are characteristic of the two codes. The MESA code is generally more flexible; overshooting of the convective core during the hydrogen and helium burning phases in MESA is chosen such that the CO core masses are consistent with those in the GENEC models. Full nucleosynthesis calculations aremore » performed for all models using the NuGrid post-processing tool MPPNP and the key energy-generating nuclear reaction rates are the same for all codes. We are thus able to highlight the key diferences between the models that are caused by the contrasting physics assumptions and numerical implementations of the three codes. A reasonable agreement is found between the surface abundances predicted by the models computed using the different codes, with GENEC exhibiting the strongest enrichment of H-burning products and KEPLER exhibiting the weakest. There are large variations in both the structure and composition of the models—the 15M ⊙ and 20M ⊙ in particular—at the pre-SN stage from code to code caused primarily by convective shell merging during the advanced stages. For example the C-shell abundances of O, Ne and Mg predicted by the three codes span one order of magnitude in the 15M ⊙ models. For the alpha elements between Si and Fe the differences are even larger. The s-process abundances in the C shell are modified by the merging of convective shells; the modification is strongest in the 15M ⊙ model in which the C-shell material is exposed to O-burning temperatures and the γ -process is activated. The variation in the s-process abundances across the codes is smallest in the 25M ⊙ models, where it is comparable to the impact of nuclear reaction rate uncertainties. In general the differences in the results from the three codes are due to their contrasting physics assumptions (e.g. prescriptions for mass loss and convection). The broadly similar evolution of the 25M ⊙ models gives us reassurance that different stellar evolution codes do produce similar results. For the 15M ⊙ and 20M ⊙ models, however, the different input physics and the interplay between the various convective zones lead to important differences in both the pre-supernova structure and nucleosynthesis predicted by the three codes. For the KEPLER models the core masses are different and therefore an exact match could not be expected.« less

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...

MEMOPS: data modelling and automatic code generation.

PubMed

Fogh, Rasmus H; Boucher, Wayne; Ionides, John M C; Vranken, Wim F; Stevens, Tim J; Laue, Ernest D

2010-03-25

In recent years the amount of biological data has exploded to the point where much useful information can only be extracted by complex computational analyses. Such analyses are greatly facilitated by metadata standards, both in terms of the ability to compare data originating from different sources, and in terms of exchanging data in standard forms, e.g. when running processes on a distributed computing infrastructure. However, standards thrive on stability whereas science tends to constantly move, with new methods being developed and old ones modified. Therefore maintaining both metadata standards, and all the code that is required to make them useful, is a non-trivial problem. Memops is a framework that uses an abstract definition of the metadata (described in UML) to generate internal data structures and subroutine libraries for data access (application programming interfaces--APIs--currently in Python, C and Java) and data storage (in XML files or databases). For the individual project these libraries obviate the need for writing code for input parsing, validity checking or output. Memops also ensures that the code is always internally consistent, massively reducing the need for code reorganisation. Across a scientific domain a Memops-supported data model makes it easier to support complex standards that can capture all the data produced in a scientific area, share them among all programs in a complex software pipeline, and carry them forward to deposition in an archive. The principles behind the Memops generation code will be presented, along with example applications in Nuclear Magnetic Resonance (NMR) spectroscopy and structural biology.

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.

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

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

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

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

Merzari, E.; Yuan, Haomin; Kraus, A.

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable componentsmore » in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

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

Vidal, Jean-Marc; Eschbach, Romain; Launay, Agnes

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of coolingmore » time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the CEA reference calculation codes for neutron physics with the JEFF-3.1.1 nuclear data set. (authors)« less

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

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

Delchini, Marc-Olivier; Popov, Emilian L.; Pointer, William David

This report presents the state of advancement of a Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the computational fluid dynamics (CFD) code Nek5000 using the Dakota package for flows encountered in the nuclear engineering industry. Nek5000 is a high-order spectral element CFD code developed at Argonne National Laboratory for high-resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds-averaged Navier-Stokes (URANS) simulations.

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

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.

Correlated prompt fission data in transport simulations

DOE PAGES

Talou, P.; Vogt, R.; Randrup, J.; ...

2018-01-24

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Correlated prompt fission data in transport simulations

NASA Astrophysics Data System (ADS)

Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

2018-01-01

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. This review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Finally, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.

Correlated prompt fission data in transport simulations

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

Talou, P.; Vogt, R.; Randrup, J.

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Calculating Absolute Transition Probabilities for Deformed Nuclei in the Rare-Earth Region

NASA Astrophysics Data System (ADS)

Stratman, Anne; Casarella, Clark; Aprahamian, Ani

2017-09-01

Absolute transition probabilities are the cornerstone of understanding nuclear structure physics in comparison to nuclear models. We have developed a code to calculate absolute transition probabilities from measured lifetimes, using a Python script and a Mathematica notebook. Both of these methods take pertinent quantities such as the lifetime of a given state, the energy and intensity of the emitted gamma ray, and the multipolarities of the transitions to calculate the appropriate B(E1), B(E2), B(M1) or in general, any B(σλ) values. The program allows for the inclusion of mixing ratios of different multipolarities and the electron conversion of gamma-rays to correct for their intensities, and yields results in absolute units or results normalized to Weisskopf units. The code has been tested against available data in a wide range of nuclei from the rare earth region (28 in total), including 146-154Sm, 154-160Gd, 158-164Dy, 162-170Er, 168-176Yb, and 174-182Hf. It will be available from the Notre Dame Nuclear Science Laboratory webpage for use by the community. This work was supported by the University of Notre Dame College of Science, and by the National Science Foundation, under Contract PHY-1419765.

Upgrades to the NESS (Nuclear Engine System Simulation) Code

NASA Technical Reports Server (NTRS)

Fittje, James E.

2007-01-01

In support of the President's Vision for Space Exploration, the Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for human expeditions to the moon and Mars. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the 1960's and 1970's. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design.

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)).

The structure of human SFPQ reveals a coiled-coil mediated polymer essential for functional aggregation in gene regulation

PubMed Central

Lee, Mihwa; Sadowska, Agata; Bekere, Indra; Ho, Diwei; Gully, Benjamin S.; Lu, Yanling; Iyer, K. Swaminathan; Trewhella, Jill; Fox, Archa H.; Bond, Charles S.

2015-01-01

SFPQ, (a.k.a. PSF), is a human tumor suppressor protein that regulates many important functions in the cell nucleus including coordination of long non-coding RNA molecules into nuclear bodies. Here we describe the first crystal structures of Splicing Factor Proline and Glutamine Rich (SFPQ), revealing structural similarity to the related PSPC1/NONO heterodimer and a strikingly extended structure (over 265 Å long) formed by an unusual anti-parallel coiled-coil that results in an infinite linear polymer of SFPQ dimers within the crystals. Small-angle X-ray scattering and transmission electron microscopy experiments show that polymerization is reversible in solution and can be templated by DNA. We demonstrate that the ability to polymerize is essential for the cellular functions of SFPQ: disruptive mutation of the coiled-coil interaction motif results in SFPQ mislocalization, reduced formation of nuclear bodies, abrogated molecular interactions and deficient transcriptional regulation. The coiled-coil interaction motif thus provides a molecular explanation for the functional aggregation of SFPQ that directs its role in regulating many aspects of cellular nucleic acid metabolism. PMID:25765647

Paraspeckles: nuclear bodies built on long noncoding RNA

PubMed Central

Bond, Charles S.

2009-01-01

Paraspeckles are ribonucleoprotein bodies found in the interchromatin space of mammalian cell nuclei. These structures play a role in regulating the expression of certain genes in differentiated cells by nuclear retention of RNA. The core paraspeckle proteins (PSF/SFPQ, P54NRB/NONO, and PSPC1 [paraspeckle protein 1]) are members of the DBHS (Drosophila melanogaster behavior, human splicing) family. These proteins, together with the long nonprotein-coding RNA NEAT1 (MEN-ϵ/β), associate to form paraspeckles and maintain their integrity. Given the large numbers of long noncoding transcripts currently being discovered through whole transcriptome analysis, paraspeckles may be a paradigm for a class of subnuclear bodies formed around long noncoding RNA. PMID:19720872

The argon nuclear quadrupole moments

NASA Astrophysics Data System (ADS)

Sundholm, Dage; Pyykkö, Pekka

2018-07-01

New standard values -116(2) mb and 76(3) mb are suggested for the nuclear quadrupole moments (Q) of the 39Ar and 37Ar nuclei, respectively. The Q values were obtained by combining optical measurements of the quadrupole coupling constant (B or eqQ/h) of the 3s23p54s[3/2]2 (3Po) and 3s23p54p[5/2]3 (3De) states of argon with large scale numerical complete active space self-consistent field and restricted active space self-consistent field calculations of the electric field gradient at the nucleus (q) using the LUCAS code, which is a finite-element based multiconfiguration Hartree-Fock program for atomic structure calculations.

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

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.

Deterministic and probabilistic analysis of damping device resistance under impact loads from nuclear fuel container drop

NASA Astrophysics Data System (ADS)

Kala, J.; Bajer, M.; Barnat, J.; Smutný, J.

2010-12-01

Pedestrian-induced vibrations are a criterion for serviceability. This loading is significant for light-weight footbridge structures, but was established as a basic loading for the ceilings of various ordinary buildings. Wide variations of this action exist. To verify the different conclusions of various authors, vertical pressure measurements invoked during walking were performed. In the article the approaches of different design codes are also shown.

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

FY2017 Pilot Project Plan for the Nuclear Energy Knowledge and Validation Center Initiative

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

Ren, Weiju

To prepare for technical development of computational code validation under the Nuclear Energy Knowledge and Validation Center (NEKVAC) initiative, several meetings were held by a group of experts of the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory (ORNL) to develop requirements of, and formulate a structure for, a transient fuel database through leveraging existing resources. It was concluded in discussions of these meetings that a pilot project is needed to address the most fundamental issues that can generate immediate stimulus to near-future validation developments as well as long-lasting benefits to NEKVAC operation. The present project is proposedmore » based on the consensus of these discussions. Analysis of common scenarios in code validation indicates that the incapability of acquiring satisfactory validation data is often a showstopper that must first be tackled before any confident validation developments can be carried out. Validation data are usually found scattered in different places most likely with interrelationships among the data not well documented, incomplete with information for some parameters missing, nonexistent, or unrealistic to experimentally generate. Furthermore, with very different technical backgrounds, the modeler, the experimentalist, and the knowledgebase developer that must be involved in validation data development often cannot communicate effectively without a data package template that is representative of the data structure for the information domain of interest to the desired code validation. This pilot project is proposed to use the legendary TREAT Experiments Database to provide core elements for creating an ideal validation data package. Data gaps and missing data interrelationships will be identified from these core elements. All the identified missing elements will then be filled in with experimental data if available from other existing sources or with dummy data if nonexistent. The resulting hybrid validation data package (composed of experimental and dummy data) will provide a clear and complete instance delineating the structure of the desired validation data and enabling effective communication among the modeler, the experimentalist, and the knowledgebase developer. With a good common understanding of the desired data structure by the three parties of subject matter experts, further existing data hunting will be effectively conducted, new experimental data generation will be realistically pursued, knowledgebase schema will be practically designed; and code validation will be confidently planned.« less

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.

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.

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

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.

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.

Investigation of the mechanism of meiotic DNA cleavage by VMA1-derived endonuclease uncovers a meiotic alteration in chromatin structure around the target site.

PubMed

Fukuda, Tomoyuki; Ohta, Kunihiro; Ohya, Yoshikazu

2006-06-01

VMA1-derived endonuclease (VDE), a homing endonuclease in Saccharomyces cerevisiae, is encoded by the mobile intein-coding sequence within the nuclear VMA1 gene. VDE recognizes and cleaves DNA at the 31-bp VDE recognition sequence (VRS) in the VMA1 gene lacking the intein-coding sequence during meiosis to insert a copy of the intein-coding sequence at the cleaved site. The mechanism underlying the meiosis specificity of VMA1 intein-coding sequence homing remains unclear. We studied various factors that might influence the cleavage activity in vivo and found that VDE binding to the VRS can be detected only when DNA cleavage by VDE takes place, implying that meiosis-specific DNA cleavage is regulated by the accessibility of VDE to its target site. As a possible candidate for the determinant of this accessibility, we analyzed chromatin structure around the VRS and revealed that local chromatin structure near the VRS is altered during meiosis. Although the meiotic chromatin alteration exhibits correlations with DNA binding and cleavage by VDE at the VMA1 locus, such a chromatin alteration is not necessarily observed when the VRS is embedded in ectopic gene loci. This suggests that nucleosome positioning or occupancy around the VRS by itself is not the sole mechanism for the regulation of meiosis-specific DNA cleavage by VDE and that other mechanisms are involved in the regulation.

Investigation of the Mechanism of Meiotic DNA Cleavage by VMA1-Derived Endonuclease Uncovers a Meiotic Alteration in Chromatin Structure around the Target Site

PubMed Central

Fukuda, Tomoyuki; Ohta, Kunihiro; Ohya, Yoshikazu

2006-01-01

VMA1-derived endonuclease (VDE), a homing endonuclease in Saccharomyces cerevisiae, is encoded by the mobile intein-coding sequence within the nuclear VMA1 gene. VDE recognizes and cleaves DNA at the 31-bp VDE recognition sequence (VRS) in the VMA1 gene lacking the intein-coding sequence during meiosis to insert a copy of the intein-coding sequence at the cleaved site. The mechanism underlying the meiosis specificity of VMA1 intein-coding sequence homing remains unclear. We studied various factors that might influence the cleavage activity in vivo and found that VDE binding to the VRS can be detected only when DNA cleavage by VDE takes place, implying that meiosis-specific DNA cleavage is regulated by the accessibility of VDE to its target site. As a possible candidate for the determinant of this accessibility, we analyzed chromatin structure around the VRS and revealed that local chromatin structure near the VRS is altered during meiosis. Although the meiotic chromatin alteration exhibits correlations with DNA binding and cleavage by VDE at the VMA1 locus, such a chromatin alteration is not necessarily observed when the VRS is embedded in ectopic gene loci. This suggests that nucleosome positioning or occupancy around the VRS by itself is not the sole mechanism for the regulation of meiosis-specific DNA cleavage by VDE and that other mechanisms are involved in the regulation. PMID:16757746

Turning limited experimental information into 3D models of RNA.

PubMed

Flores, Samuel Coulbourn; Altman, Russ B

2010-09-01

Our understanding of RNA functions in the cell is evolving rapidly. As for proteins, the detailed three-dimensional (3D) structure of RNA is often key to understanding its function. Although crystallography and nuclear magnetic resonance (NMR) can determine the atomic coordinates of some RNA structures, many 3D structures present technical challenges that make these methods difficult to apply. The great flexibility of RNA, its charged backbone, dearth of specific surface features, and propensity for kinetic traps all conspire with its long folding time, to challenge in silico methods for physics-based folding. On the other hand, base-pairing interactions (either in runs to form helices or isolated tertiary contacts) and motifs are often available from relatively low-cost experiments or informatics analyses. We present RNABuilder, a novel code that uses internal coordinate mechanics to satisfy user-specified base pairing and steric forces under chemical constraints. The code recapitulates the topology and characteristic L-shape of tRNA and obtains an accurate noncrystallographic structure of the Tetrahymena ribozyme P4/P6 domain. The algorithm scales nearly linearly with molecule size, opening the door to the modeling of significantly larger structures.

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

Vins, M.

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe.more » Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)« less

Monitoring, Modeling, and Diagnosis of Alkali-Silica Reaction in Small Concrete Samples

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

Agarwal, Vivek; Cai, Guowei; Gribok, Andrei V.

Assessment and management of aging concrete structures in nuclear power plants require a more systematic approach than simple reliance on existing code margins of safety. Structural health monitoring of concrete structures aims to understand the current health condition of a structure based on heterogeneous measurements to produce high-confidence actionable information regarding structural integrity that supports operational and maintenance decisions. This report describes alkali-silica reaction (ASR) degradation mechanisms and factors influencing the ASR. A fully coupled thermo-hydro-mechanical-chemical model developed by Saouma and Perotti by taking into consideration the effects of stress on the reaction kinetics and anisotropic volumetric expansion is presentedmore » in this report. This model is implemented in the GRIZZLY code based on the Multiphysics Object Oriented Simulation Environment. The implemented model in the GRIZZLY code is randomly used to initiate ASR in a 2D and 3D lattice to study the percolation aspects of concrete. The percolation aspects help determine the transport properties of the material and therefore the durability and service life of concrete. This report summarizes the effort to develop small-size concrete samples with embedded glass to mimic ASR. The concrete samples were treated in water and sodium hydroxide solution at elevated temperature to study how ingress of sodium ions and hydroxide ions at elevated temperature impacts concrete samples embedded with glass. Thermal camera was used to monitor the changes in the concrete sample and results are summarized.« less

Numerical analysis of nuclear power plant structure subjected to aircraft crash

NASA Astrophysics Data System (ADS)

Saberi, Reza; Alinejad, Majid; Mahdavi, Mir Omid; Sepanloo, Kamran

2017-12-01

An aircraft crashing into a nuclear containment may induce a series of disasters related to containment capacity, including local penetration and perforation of the containment, intensive vibrations, and fire ignited after jet fuel leakage. In this study, structural safety of a reinforced concrete containment vessel (RCCV) has been studied against the direct hit of Airbus A320, Boeing 707-320 and Phantom F4 aircrafts. ABAQUS/explicit finite element code has been used to carry out the three-dimensional numerical simulations. The impact locations identified on the nuclear containment structure are mid height of containment, center of the cylindrical portion, junction of dome and cylinder, and over the cylindrical portion close to the foundation level. The loading of the aircraft has been assigned through the corresponding reaction-time response curve. The concrete damaged plasticity model was predicted to simulate the behavior of concrete while the behavior of steel reinforcement was incorporated using elastoplastic material model. Dynamic loading conditions were considered using dynamic increase factor. The mid height of containment and center of cylindrical portion have been found to experience most severe deformation against each aircraft crash. It has also been found that compression damage in concrete is not critical at none of the impact locations.

The X chromosome in space.

PubMed

Jégu, Teddy; Aeby, Eric; Lee, Jeannie T

2017-06-01

Extensive 3D folding is required to package a genome into the tiny nuclear space, and this packaging must be compatible with proper gene expression. Thus, in the well-hierarchized nucleus, chromosomes occupy discrete territories and adopt specific 3D organizational structures that facilitate interactions between regulatory elements for gene expression. The mammalian X chromosome exemplifies this structure-function relationship. Recent studies have shown that, upon X-chromosome inactivation, active and inactive X chromosomes localize to different subnuclear positions and adopt distinct chromosomal architectures that reflect their activity states. Here, we review the roles of long non-coding RNAs, chromosomal organizational structures and the subnuclear localization of chromosomes as they relate to X-linked gene expression.

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

Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

NASA Astrophysics Data System (ADS)

Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

2017-12-01

This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

Thermomechanical analysis of fast-burst reactors

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

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

Additive manufacturing of 316L stainless steel by electron beam melting for nuclear fusion applications

NASA Astrophysics Data System (ADS)

Zhong, Yuan; Rännar, Lars-Erik; Liu, Leifeng; Koptyug, Andrey; Wikman, Stefan; Olsen, Jon; Cui, Daqing; Shen, Zhijian

2017-04-01

A feasibility study was performed to fabricate ITER In-Vessel components by one of the metal additive manufacturing methods, Electron Beam Melting® (EBM®). Solid specimens of SS316L with 99.8% relative density were prepared from gas atomized precursor powder granules. After the EBM® process the phase remains as austenite and the composition has practically not been changed. The RCC-MR code used for nuclear pressure vessels provides guidelines for this study and tensile tests and Charpy-V tests were carried out at 22 °C (RT) and 250 °C (ET). This work provides the first set of mechanical and microstructure data of EBM® SS316L for nuclear fusion applications. The mechanical testing shows that the yield strength, ductility and toughness are well above the acceptance criteria and only the ultimate tensile strength of EBM® SS316L is below the RCC-MR code. Microstructure characterizations reveal the presence of hierarchical structures consisting of solidified melt pools, columnar grains and irregular shaped sub-grains. Lots of precipitates enriched in Cr and Mo are observed at columnar grain boundaries while no sign of element segregation is shown at the sub-grain boundaries. Such a unique microstructure forms during a non-equilibrium process, comprising rapid solidification and a gradient 'annealing' process due to anisotropic thermal flow of accumulated heat inside the powder granule matrix. Relations between process parameters, specimen geometry (total building time) and sub-grain structure are discussed. Defects are formed mainly due to the large layer thickness (100 μm) which generates insufficient bonding between a few of the adjacently formed melt pools during the process. Further studies should focus on adjusting layer thickness to improve the strength of EBM® SS316L and optimizing total building time.

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

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

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

NASA Astrophysics Data System (ADS)

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

Development of a patient-specific dosimetry estimation system in nuclear medicine examination

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

Lin, H. H.; Dong, S. L.; Yang, H. J.

2011-07-01

The purpose of this study is to develop a patient-specific dosimetry estimation system in nuclear medicine examination using a SimSET-based Monte Carlo code. We added a dose deposition routine to store the deposited energy of the photons during their flights in SimSET and developed a user-friendly interface for reading PET and CT images. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The S values for {sup 99m}Tc, {sup 18}F and {sup 131}I obtained by the system were compared to those from the MCNP4C code and OLINDA. The ratios of S values computed by thismore » system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which are comparable to that obtained from MCNP4C code (0.94 to 1.20). The average ratios of S value were 0.99{+-}0.04, 1.03{+-}0.05, and 1.00{+-}0.07 for isotopes {sup 131}I, {sup 18}F, and {sup 99m}Tc, respectively. The simulation time of SimSET was two times faster than MCNP4C's for various isotopes. A 3D dose calculation was also performed on a patient data set with PET/CT examination using this system. Results from the patient data showed that the estimated S values using this system differed slightly from those of OLINDA for ORNL phantom. In conclusion, this system can generate patient-specific dose distribution and display the isodose curves on top of the anatomic structure through a friendly graphic user interface. It may also provide a useful tool to establish an appropriate dose-reduction strategy to patients in nuclear medicine environments. (authors)« less

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.

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.

GASFLOW: A Computational Fluid Dynamics Code for Gases, Aerosols, and Combustion, Volume 3: Assessment Manual

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

Müller, C.; Hughes, E. D.; Niederauer, G. F.

1998-10-01

Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best- estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the wallsmore » and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containment and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included. Volume III contains some of the assessments performed by LANL and FzK« less

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.

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

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...

A Dual Origin of the Xist Gene from a Protein-Coding Gene and a Set of Transposable Elements

PubMed Central

Elisaphenko, Eugeny A.; Kolesnikov, Nikolay N.; Shevchenko, Alexander I.; Rogozin, Igor B.; Nesterova, Tatyana B.; Brockdorff, Neil; Zakian, Suren M.

2008-01-01

X-chromosome inactivation, which occurs in female eutherian mammals is controlled by a complex X-linked locus termed the X-inactivation center (XIC). Previously it was proposed that genes of the XIC evolved, at least in part, as a result of pseudogenization of protein-coding genes. In this study we show that the key XIC gene Xist, which displays fragmentary homology to a protein-coding gene Lnx3, emerged de novo in early eutherians by integration of mobile elements which gave rise to simple tandem repeats. The Xist gene promoter region and four out of ten exons found in eutherians retain homology to exons of the Lnx3 gene. The remaining six Xist exons including those with simple tandem repeats detectable in their structure have similarity to different transposable elements. Integration of mobile elements into Xist accompanies the overall evolution of the gene and presumably continues in contemporary eutherian species. Additionally we showed that the combination of remnants of protein-coding sequences and mobile elements is not unique to the Xist gene and is found in other XIC genes producing non-coding nuclear RNA. PMID:18575625

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

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...

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

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.

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.

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

NASA Astrophysics Data System (ADS)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

TCOF1 gene encodes a putative nucleolar phosphoprotein that exhibits mutations in Treacher Collins Syndrome throughout its coding region.

PubMed

Wise, C A; Chiang, L C; Paznekas, W A; Sharma, M; Musy, M M; Ashley, J A; Lovett, M; Jabs, E W

1997-04-01

Treacher Collins Syndrome (TCS) is the most common of the human mandibulofacial dysostosis disorders. Recently, a partial TCOF1 cDNA was identified and shown to contain mutations in TCS families. Here we present the entire exon/intron genomic structure and the complete coding sequence of TCOF1. TCOF1 encodes a low complexity protein of 1,411 amino acids, whose predicted protein structure reveals repeated motifs that mirror the organization of its exons. These motifs are shared with nucleolar trafficking proteins in other species and are predicted to be highly phosphorylated by casein kinase. Consistent with this, the full-length TCOF1 protein sequence also contains putative nuclear and nucleolar localization signals. Throughout the open reading frame, we detected an additional eight mutations in TCS families and several polymorphisms. We postulate that TCS results from defects in a nucleolar trafficking protein that is critically required during human craniofacial development.

APORT: a program for the area-based apportionment of county variables to cells of a polar grid. [Airborne pollutant transport models

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

Fields, D.E.; Little, C.A.

1978-11-01

The APORT computer code was developed to apportion variables tabulated for polygon-structured civil districts onto cells of a polar grid. The apportionment is based on fractional overlap between the polygon and the grid cells. Centering the origin of the polar system at a pollutant source site yields results that are very useful for assessing and interpreting the effects of airborne pollutant dissemination. The APOPLT graphics code, which uses the same data set as APORT, provides a convenient visual display of the polygon structure and the extent of the polar grid. The APORT/APOPLT methodology was verified by application to county summariesmore » of cattle population for counties surrounding the Oyster Creek, New Jersey, nuclear power plant. These numerical results, which were obtained using approximately 2-min computer time on an IBM System 360/91 computer, compare favorably to results of manual computations in both speed and accuracy.« less

Numerical simulation of tornado wind loading on structures

NASA Technical Reports Server (NTRS)

Maiden, D. E.

1976-01-01

A numerical simulation of a tornado interacting with a building was undertaken in order to compare the pressures due to a rotational unsteady wind with that due to steady straight winds used in design of nuclear facilities. The numerical simulations were performed on a two-dimensional compressible hydrodynamics code. Calculated pressure profiles for a typical building were then subjected to a tornado wind field and the results were compared with current quasisteady design calculations. The analysis indicates that current design practices are conservative.

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

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

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

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...

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.

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

A Demonstration of Concrete Structural Health Monitoring Framework for Degradation due to Alkali-Silica Reaction

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

Mahadevan, Sankaran; Agarwal, Vivek; Neal, Kyle

Assessment and management of aging concrete structures in nuclear power plants require a more systematic approach than simple reliance on existing code margins of safety. Structural health monitoring of concrete structures aims to understand the current health condition of a structure based on heterogeneous measurements to produce high-confidence actionable information regarding structural integrity that supports operational and maintenance decisions. This ongoing research project is seeking to develop a probabilistic framework for health diagnosis and prognosis of aging concrete structures in a nuclear power plant that is subjected to physical, chemical, environment, and mechanical degradation. The proposed framework consists of fourmore » elements: monitoring, data analytics, uncertainty quantification and prognosis. This report focuses on degradation caused by ASR (alkali-silica reaction). Controlled specimens were prepared to develop accelerated ASR degradation. Different monitoring techniques – thermography, digital image correlation (DIC), mechanical deformation measurements, nonlinear impact resonance acoustic spectroscopy (NIRAS), and vibro-acoustic modulation (VAM) -- were used to detect the damage caused by ASR. Heterogeneous data from the multiple techniques was used for damage diagnosis and prognosis, and quantification of the associated uncertainty using a Bayesian network approach. Additionally, MapReduce technique has been demonstrated with synthetic data. This technique can be used in future to handle large amounts of observation data obtained from the online monitoring of realistic structures.« 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.

Identification of a chemical inhibitor for nuclear speckle formation: Implications for the function of nuclear speckles in regulation of alternative pre-mRNA splicing

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

Kurogi, Yutaro; Matsuo, Yota; Mihara, Yuki

2014-03-28

Highlights: • We identified tubercidin as a compound inducing aberrant formation of the speckles. • Tubercidin causes delocalization of poly (A){sup +}RNAs from nuclear speckles. • Tubercidin induces dispersion of splicing factors from nuclear speckles. • Tubercidin affects alternative pre-mRNA splicing. • Nuclear speckles play a role in regulation of alternative pre-mRNA splicing. - Abstract: Nuclear speckles are subnuclear structures enriched with RNA processing factors and poly (A){sup +} RNAs comprising mRNAs and poly (A){sup +} non-coding RNAs (ncRNAs). Nuclear speckles are thought to be involved in post-transcriptional regulation of gene expression, such as pre-mRNA splicing. By screening 3585 culturemore » extracts of actinomycetes with in situ hybridization using an oligo dT probe, we identified tubercidin, an analogue of adenosine, as an inhibitor of speckle formation, which induces the delocalization of poly (A){sup +} RNA and dispersion of splicing factor SRSF1/SF2 from nuclear speckles in HeLa cells. Treatment with tubercidin also decreased steady-state MALAT1 long ncRNA, thought to be involved in the retention of SRSF1/SF2 in nuclear speckles. In addition, we found that tubercidin treatment promoted exon skipping in the alternative splicing of Clk1 pre-mRNA. These results suggest that nuclear speckles play a role in modulating the concentration of splicing factors in the nucleoplasm to regulate alternative pre-mRNA splicing.« less

Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

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

Dr. F. W. Brust; Dr. G. M. Wilkowski; Dr. P. Krishnaswamy

2010-01-27

The subsection ASME NH high temperature design procedure does not admit crack-like defects into the structural components. The US NRC identified the lack of treatment of crack growth within NH as a limitation of the code and thus this effort was undertaken. This effort is broken into two parts. Part 1, summarized here, involved examining all high temperature creep-fatigue crack growth codes being used today and from these, the task objective was to choose a methodology that is appropriate for possible implementation within NH. The second part of this task, which has just started, is to develop design rules formore » possible implementation within NH. This second part is a challenge since all codes require step-by-step analysis procedures to be undertaken in order to assess the crack growth and life of the component. Simple rules for design do not exist in any code at present. The codes examined in this effort included R5, RCC-MR (A16), BS 7910, API 579, and ATK (and some lesser known codes). There are several reasons that the capability for assessing cracks in high temperature nuclear components is desirable. These include: (1) Some components that are part of GEN IV reactors may have geometries that have sharp corners - which are essentially cracks. Design of these components within the traditional ASME NH procedure is quite challenging. It is natural to ensure adequate life design by modeling these features as cracks within a creep-fatigue crack growth procedure. (2) Workmanship flaws in welds sometimes occur and are accepted in some ASME code sections. It can be convenient to consider these as flaws when making a design life assessment. (3) Non-destructive Evaluation (NDE) and inspection methods after fabrication are limited in the size of the crack or flaw that can be detected. It is often convenient to perform a life assessment using a flaw of a size that represents the maximum size that can elude detection. (4) Flaws that are observed using in-service detection methods often need to be addressed as plants age. Shutdown inspection intervals can only be designed using creep and creep-fatigue crack growth techniques. (5) The use of crack growth procedures can aid in examining the seriousness of creep damage in structural components. How cracks grow can be used to assess margins on components and lead to further safe operation. After examining the pros and cons of all these methods, the R5 code was chosen as the most up-to-date and validated high temperature creep and creep fatigue code currently used in the world at present. R5 is considered the leader because the code: (1) has well established and validated rules, (2) has a team of experts continually improving and updating it, (3) has software that can be used by designers, (4) extensive validation in many parts with available data from BE resources as well as input from Imperial college's database, and (5) was specifically developed for use in nuclear plants. R5 was specifically developed for use in gas cooled nuclear reactors which operate in the UK and much of the experience is based on materials and temperatures which are experienced in these reactors. If the next generation advanced reactors to be built in the US used these same materials within the same temperature ranges as these reactors, then R5 may be appropriate for consideration of direct implementation within ASME code NH or Section XI. However, until more verification and validation of these creep/fatigue crack growth rules for the specific materials and temperatures to be used in the GEN IV reactors is complete, ASME should consider delaying this implementation. With this in mind, it is this authors opinion that R5 methods are the best available for code use today. The focus of this work was to examine the literature for creep and creep-fatigue crack growth procedures that are well established in codes in other countries and choose a procedure to consider implementation into ASME NH. It is very important to recognize that all creep and creep fatigue crack growth procedures that are part of high temperature design codes are related and very similar. This effort made no attempt to develop a new creep-fatigue crack growth predictive methodology. Rather examination of current procedures was the only goal. The uncertainties in the R5 crack growth methods and recommendations for more work are summarized here also.« 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

Gene pools in wild Lima bean (Phaseolus lunatus L.) from the Americas: evidences for an Andean origin and past migrations.

PubMed

Serrano-Serrano, Martha L; Hernández-Torres, Jorge; Castillo-Villamizar, Genis; Debouck, Daniel G; Sánchez, María I Chacón

2010-01-01

The aims of this research were to assess the genetic structure of wild Phaseolus lunatus L. in the Americas and the hypothesis of a relatively recent Andean origin of the species. For this purpose, nuclear and non-coding chloroplast DNA markers were analyzed in a collection of 59 wild Lima bean accessions and six allied species. Twenty-three chloroplast and 28 nuclear DNA haplotypes were identified and shown to be geographically structured. Three highly divergent wild Lima bean gene pools, AI, MI, and MII, with mostly non-overlapping geographic ranges, are proposed. The results support an Andean origin of wild Lima beans during Pleistocene times and an early divergence of the three gene pools at an age that is posterior to completion of the Isthmus of Panama and major Andean orogeny. Gene pools would have evolved and reached their current geographic distribution mainly in isolation and therefore are of high priority for conservation and breeding programs.

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

Euglena gracilis chloroplast DNA: analysis of a 1.6 kb intron of the psb C gene containing an open reading frame of 458 codons.

PubMed

Montandon, P E; Vasserot, A; Stutz, E

1986-01-01

We retrieved a 1.6 kbp intron separating two exons of the psb C gene which codes for the 44 kDa reaction center protein of photosystem II. This intron is 3 to 4 times the size of all previously sequenced Euglena gracilis chloroplast introns. It contains an open reading frame of 458 codons potentially coding for a basic protein of 54 kDa of yet unknown function. The intron boundaries follow consensus sequences established for chloroplast introns related to class II and nuclear pre-mRNA introns. Its 3'-terminal segment has structural features similar to class II mitochondrial introns with an invariant base A as possible branch point for lariat formation.

The role of configuration interaction in the LTE opacity of Fe

NASA Astrophysics Data System (ADS)

Colgan, James; Kilcrease, David; Magee, Norm; Armstrong, Gregory; Abdallah, Joe; Sherrill, Manolo; Fontes, Christopher; Zhang, Honglin; Hakel, Peter

2013-05-01

The Los Alamos National Laboratory code ATOMIC has been recently used to generate a series of local-thermodynamic-equilibrium (LTE) light element opacities for the elements H through Ne. Our calculations, which include fine-structure detail, represent a systematic improvement over previous Los Alamos opacity calculations using the LEDCOP legacy code. Recent efforts have resulted in comprehensive new calculations of the opacity of Fe. In this presentation we explore the role of configuration interaction (CI) in the Fe opacity, and show where CI influences the monochromatic opacity. We present such comparisons for conditions of astrophysical interest. The Los Alamos National Laboratory is operated by Los Alamos National Security, LLC for the National Nuclear Security Administration of the U.S. Department of Energy under Contract No. DE-AC5206NA25396.

DAGON: a 3D Maxwell-Bloch code

NASA Astrophysics Data System (ADS)

Oliva, Eduardo; Cotelo, Manuel; Escudero, Juan Carlos; González-Fernández, Agustín.; Sanchís, Alberto; Vera, Javier; Vicéns, Sergio; Velarde, Pedro

2017-05-01

The amplification of UV radiation and high order harmonics (HOH) in plasmas is a subject of raising interest due to its different potential applications in several fields like environment and security (detection at distance), biology, materials science and industry (3D imaging) and atomic and plasma physics (pump-probe experiments). In order to develop these sources, it is necessary to properly understand the amplification process. Being the plasma an inhomogeneous medium which changes with time, it is desirable to have a full time-dependent 3D description of the interaction of UV and XUV radiation with plasmas. For these reasons, at the Instituto de Fusíon Nuclear we have developed DAGON, a 3D Maxwell-Bloch code capable of studying the full spationtemporal structure of the amplification process abovementioned.

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.

Three-dimensional Monte Carlo calculation of some nuclear parameters

NASA Astrophysics Data System (ADS)

Günay, Mehtap; Şeker, Gökmen

2017-09-01

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

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

Module-oriented modeling of reactive transport with HYTEC

NASA Astrophysics Data System (ADS)

van der Lee, Jan; De Windt, Laurent; Lagneau, Vincent; Goblet, Patrick

2003-04-01

The paper introduces HYTEC, a coupled reactive transport code currently used for groundwater pollution studies, safety assessment of nuclear waste disposals, geochemical studies and interpretation of laboratory column experiments. Based on a known permeability field, HYTEC evaluates the groundwater flow paths, and simulates the migration of mobile matter (ions, organics, colloids) subject to geochemical reactions. The code forms part of a module-oriented structure which facilitates maintenance and improves coding flexibility. In particular, using the geochemical module CHESS as a common denominator for several reactive transport models significantly facilitates the development of new geochemical features which become automatically available to all models. A first example shows how the model can be used to assess migration of uranium from a sub-surface source under the effect of an oxidation front. The model also accounts for alteration of hydrodynamic parameters (local porosity, permeability) due to precipitation and dissolution of mineral phases, which potentially modifies the migration properties in general. The second example illustrates this feature.

Non-coding RNAs' partitioning in the evolution of photosynthetic organisms via energy transduction and redox signaling.

PubMed

Kotakis, Christos

2015-01-01

Ars longa, vita brevis -Hippocrates Chloroplasts and mitochondria are genetically semi-autonomous organelles inside the plant cell. These constructions formed after endosymbiosis and keep evolving throughout the history of life. Experimental evidence is provided for active non-coding RNAs (ncRNAs) in these prokaryote-like structures, and a possible functional imprinting on cellular electrophysiology by those RNA entities is described. Furthermore, updated knowledge on RNA metabolism of organellar genomes uncovers novel inter-communication bridges with the nucleus. This class of RNA molecules is considered as a unique ontogeny which transforms their biological role as a genetic rheostat into a synchronous biochemical one that can affect the energetic charge and redox homeostasis inside cells. A hypothesis is proposed where such modulation by non-coding RNAs is integrated with genetic signals regulating gene transfer. The implications of this working hypothesis are discussed, with particular reference to ncRNAs involvement in the organellar and nuclear genomes evolution since their integrity is functionally coupled with redox signals in photosynthetic organisms.

Hepatitis B virus nuclear export elements: RNA stem-loop α and β, key parts of the HBV post-transcriptional regulatory element.

PubMed

Lim, Chun Shen; Brown, Chris M

2016-09-01

Many viruses contain RNA elements that modulate splicing and/or promote nuclear export of their RNAs. The RNAs of the major human pathogen, hepatitis B virus (HBV) contain a large (~600 bases) composite cis-acting 'post-transcriptional regulatory element' (PRE). This element promotes expression from these naturally intronless transcripts. Indeed, the related woodchuck hepadnavirus PRE (WPRE) is used to enhance expression in gene therapy and other expression vectors. These PRE are likely to act through a combination of mechanisms, including promotion of RNA nuclear export. Functional components of both the HBV PRE and WPRE are 2 conserved RNA cis-acting stem-loop (SL) structures, SLα and SLβ. They are within the coding regions of polymerase (P) gene, and both P and X genes, respectively. Based on previous studies using mutagenesis and/or nuclear magnetic resonance (NMR), here we propose 2 covariance models for SLα and SLβ. The model for the 30-nucleotide SLα contains a G-bulge and a CNGG(U) apical loop of which the first and the fourth loop residues form a CG pair and the fifth loop residue is bulged out, as observed in the NMR structure. The model for the 23-nucleotide SLβ contains a 7-base-pair stem and a 9-nucleotide loop. Comparison of the models with other RNA structural elements, as well as similarity searches of human transcriptome and viral genomes demonstrate that SLα and SLβ are specific to HBV transcripts. However, they are well conserved among the hepadnaviruses of non-human primates, the woodchuck and ground squirrel.

Hepatitis B virus nuclear export elements: RNA stem-loop α and β, key parts of the HBV post-transcriptional regulatory element

PubMed Central

Lim, Chun Shen; Brown, Chris M.

2016-01-01

ABSTRACT Many viruses contain RNA elements that modulate splicing and/or promote nuclear export of their RNAs. The RNAs of the major human pathogen, hepatitis B virus (HBV) contain a large (~600 bases) composite cis-acting 'post-transcriptional regulatory element' (PRE). This element promotes expression from these naturally intronless transcripts. Indeed, the related woodchuck hepadnavirus PRE (WPRE) is used to enhance expression in gene therapy and other expression vectors. These PRE are likely to act through a combination of mechanisms, including promotion of RNA nuclear export. Functional components of both the HBV PRE and WPRE are 2 conserved RNA cis-acting stem-loop (SL) structures, SLα and SLβ. They are within the coding regions of polymerase (P) gene, and both P and X genes, respectively. Based on previous studies using mutagenesis and/or nuclear magnetic resonance (NMR), here we propose 2 covariance models for SLα and SLβ. The model for the 30-nucleotide SLα contains a G-bulge and a CNGG(U) apical loop of which the first and the fourth loop residues form a CG pair and the fifth loop residue is bulged out, as observed in the NMR structure. The model for the 23-nucleotide SLβ contains a 7-base-pair stem and a 9-nucleotide loop. Comparison of the models with other RNA structural elements, as well as similarity searches of human transcriptome and viral genomes demonstrate that SLα and SLβ are specific to HBV transcripts. However, they are well conserved among the hepadnaviruses of non-human primates, the woodchuck and ground squirrel. PMID:27031749

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 ...

A dose assessment method for arbitrary geometries with virtual reality in the nuclear facilities decommissioning

NASA Astrophysics Data System (ADS)

Chao, Nan; Liu, Yong-kuo; Xia, Hong; Ayodeji, Abiodun; Bai, Lu

2018-03-01

During the decommissioning of nuclear facilities, a large number of cutting and demolition activities are performed, which results in a frequent change in the structure and produce many irregular objects. In order to assess dose rates during the cutting and demolition process, a flexible dose assessment method for arbitrary geometries and radiation sources was proposed based on virtual reality technology and Point-Kernel method. The initial geometry is designed with the three-dimensional computer-aided design tools. An approximate model is built automatically in the process of geometric modeling via three procedures namely: space division, rough modeling of the body and fine modeling of the surface, all in combination with collision detection of virtual reality technology. Then point kernels are generated by sampling within the approximate model, and when the material and radiometric attributes are inputted, dose rates can be calculated with the Point-Kernel method. To account for radiation scattering effects, buildup factors are calculated with the Geometric-Progression formula in the fitting function. The effectiveness and accuracy of the proposed method was verified by means of simulations using different geometries and the dose rate results were compared with that derived from CIDEC code, MCNP code and experimental measurements.

Common radiation analysis model for 75,000 pound thrust NERVA engine (1137400E)

NASA Technical Reports Server (NTRS)

Warman, E. A.; Lindsey, B. A.

1972-01-01

The mathematical model and sources of radiation used for the radiation analysis and shielding activities in support of the design of the 1137400E version of the 75,000 lbs thrust NERVA engine are presented. The nuclear subsystem (NSS) and non-nuclear components are discussed. The geometrical model for the NSS is two dimensional as required for the DOT discrete ordinates computer code or for an azimuthally symetrical three dimensional Point Kernel or Monte Carlo code. The geometrical model for the non-nuclear components is three dimensional in the FASTER geometry format. This geometry routine is inherent in the ANSC versions of the QAD and GGG Point Kernal programs and the COHORT Monte Carlo program. Data are included pertaining to a pressure vessel surface radiation source data tape which has been used as the basis for starting ANSC analyses with the DASH code to bridge into the COHORT Monte Carlo code using the WANL supplied DOT angular flux leakage data. In addition to the model descriptions and sources of radiation, the methods of analyses are briefly described.

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.

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

Comparative genetic structure of two mangrove species in Caribbean and Pacific estuaries of Panama

PubMed Central

2012-01-01

Background Mangroves are ecologically important and highly threatened forest communities. Observational and genetic evidence has confirmed the long distance dispersal capacity of water-dispersed mangrove seeds, but less is known about the relative importance of pollen vs. seed gene flow in connecting populations. We analyzed 980 Avicennia germinans for 11 microsatellite loci and 940 Rhizophora mangle for six microsatellite loci and subsampled two non-coding cpDNA regions in order to understand population structure, and gene flow within and among four major estuaries on the Caribbean and Pacific coasts of Panama. Results Both species showed similar rates of outcrossing (t= 0.7 in A. germinans and 0.8 in R. mangle) and strong patterns of spatial genetic structure within estuaries, although A. germinans had greater genetic structure in nuclear and cpDNA markers (7 demes > 4 demes and Sp= 0.02 > 0.002), and much greater cpDNA diversity (Hd= 0.8 > 0.2) than R. mangle. The Central American Isthmus serves as an exceptionally strong barrier to gene flow, with high levels nuclear (FST= 0.3-0.5) and plastid (FST= 0.5-0.8) genetic differentiation observed within each species between coasts and no shared cpDNA haplotypes between species on each coast. Finally, evidence of low ratios of pollen to seed dispersal (r = −0.6 in A. germinans and 7.7 in R. mangle), coupled with the strong observed structure in nuclear and plastid DNA among most estuaries, suggests low levels of gene flow in these mangrove species. Conclusions We conclude that gene dispersal in mangroves is usually limited within estuaries and that coastal geomorphology and rare long distance dispersal events could also influence levels of structure. PMID:23078287

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

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...

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.

Radioactive decay data tables: A handbook of decay data for application to radiation dosimetry and radiological assessments

NASA Astrophysics Data System (ADS)

Kocher, D. C.; Smith, J. S.

Decay data are presented for approximately 500 radionuclides including those occurring naturally in the environment, those of potential importance in routine or accidental releases from the nuclear fuel cycle, those of current interest in nuclear medicine and fusion reactor technology, and some of those of interest to Committee 2 of the International Commission on Radiological Protection for the estimation of annual limits on intake via inhalation and ingestion for occupationally exposed individuals. Physical processes involved in radioactive decay which produce the different types of radiation observed, methods used to prepare the decay data sets for each radionuclide in the format of the computerized evaluated nuclear structure data file, the tables of radioactive decay data, and the computer code MEDLIST used to produce the tables are described. Applications of the data to problems of interest in radiation dosimetry and radiological assessments are considered as well as the calculations of the activity of a daughter radionuclide relative to the activity of its parent in a radioactive decay chain.

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

Generalized Nuclear Data

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

Conlin, Jeremy

2017-03-15

This software is code related to reading/writing/manipulating nuclear data in the Generalized Nuclear Data (GND) format, a new format for sharing nuclear data among institutions. In addition to the software and its documentation, notes and documentation from the WPEC Subgroup 43 will be included. WPEC Subgroup 43 is an international committee charged with creating the API for the GND format.

76 FR 26320 - Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance of Amendment to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-06

... Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone (301..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-11107 Filed 5-5-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0071; Docket No. 50-382] Entergy Operations, Inc...

76 FR 13676 - Exelon Generation Company, LLC; Notice of Withdrawal of Application for Amendment to Facility...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-14

.... Brown, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-5756 Filed 3-11-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-373 and 50-374; NRC-2011-0051] Exelon Generation...

76 FR 35922 - Interim Staff Guidance Regarding the Environmental Report for Applications To Construct and/or...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-20

... Nuclear Reactor Regulation on the information that should be included in the Environmental Report, which...: Mr. Scott Sloan, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-15227 Filed 6-17-11; 8:45 am] BILLING CODE 7590-01-P ...

49 CFR 176.2 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-10-01

..., spillage, or other accident. INF cargo means packaged irradiated nuclear fuel, plutonium or high-level... Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes on Board Ships” (INF Code) contained in...

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.

High-Energy Activation Simulation Coupling TENDL and SPACS with FISPACT-II

NASA Astrophysics Data System (ADS)

Fleming, Michael; Sublet, Jean-Christophe; Gilbert, Mark

2018-06-01

To address the needs of activation-transmutation simulation in incident-particle fields with energies above a few hundred MeV, the FISPACT-II code has been extended to splice TENDL standard ENDF-6 nuclear data with extended nuclear data forms. The JENDL-2007/HE and HEAD-2009 libraries were processed for FISPACT-II and used to demonstrate the capabilities of the new code version. Tests of the libraries and comparisons against both experimental yield data and the most recent intra-nuclear cascade model results demonstrate that there is need for improved nuclear data libraries up to and above 1 GeV. Simulations on lead targets show that important radionuclides, such as 148Gd, can vary by more than an order of magnitude where more advanced models find agreement within the experimental uncertainties.

A novel nuclear genetic code alteration in yeasts and the evolution of codon reassignment in eukaryotes.

PubMed

Mühlhausen, Stefanie; Findeisen, Peggy; Plessmann, Uwe; Urlaub, Henning; Kollmar, Martin

2016-07-01

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 tRNA(Ala) 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. © 2016 Mühlhausen et al.; Published by Cold Spring Harbor Laboratory Press.

MitoNuc: a database of nuclear genes coding for mitochondrial proteins. Update 2002.

PubMed

Attimonelli, Marcella; Catalano, Domenico; Gissi, Carmela; Grillo, Giorgio; Licciulli, Flavio; Liuni, Sabino; Santamaria, Monica; Pesole, Graziano; Saccone, Cecilia

2002-01-01

Mitochondria, besides their central role in energy metabolism, have recently been found to be involved in a number of basic processes of cell life and to contribute to the pathogenesis of many degenerative diseases. All functions of mitochondria depend on the interaction of nuclear and organelle genomes. Mitochondrial genomes have been extensively sequenced and analysed and data have been collected in several specialised databases. In order to collect information on nuclear coded mitochondrial proteins we developed MitoNuc, a database containing detailed information on sequenced nuclear genes coding for mitochondrial proteins in Metazoa. The MitoNuc database can be retrieved through SRS and is available via the web site http://bighost.area.ba.cnr.it/mitochondriome where other mitochondrial databases developed by our group, the complete list of the sequenced mitochondrial genomes, links to other mitochondrial sites and related information, are available. The MitoAln database, related to MitoNuc in the previous release, reporting the multiple alignments of the relevant homologous protein coding regions, is no longer supported in the present release. In order to keep the links among entries in MitoNuc from homologous proteins, a new field in the database has been defined: the cluster identifier, an alpha numeric code used to identify each cluster of homologous proteins. A comment field derived from the corresponding SWISS-PROT entry has been introduced; this reports clinical data related to dysfunction of the protein. The logic scheme of MitoNuc database has been implemented in the ORACLE DBMS. This will allow the end-users to retrieve data through a friendly interface that will be soon implemented.

User input verification and test driven development in the NJOY21 nuclear data processing code

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

Trainer, Amelia Jo; Conlin, Jeremy Lloyd; McCartney, Austin Paul

Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, andmore » capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.« less

MD simulations of papillomavirus DNA-E2 protein complexes hints at a protein structural code for DNA deformation.

PubMed

Falconi, M; Oteri, F; Eliseo, T; Cicero, D O; Desideri, A

2008-08-01

The structural dynamics of the DNA binding domains of the human papillomavirus strain 16 and the bovine papillomavirus strain 1, complexed with their DNA targets, has been investigated by modeling, molecular dynamics simulations, and nuclear magnetic resonance analysis. The simulations underline different dynamical features of the protein scaffolds and a different mechanical interaction of the two proteins with DNA. The two protein structures, although very similar, show differences in the relative mobility of secondary structure elements. Protein structural analyses, principal component analysis, and geometrical and energetic DNA analyses indicate that the two transcription factors utilize a different strategy in DNA recognition and deformation. Results show that the protein indirect DNA readout is not only addressable to the DNA molecule flexibility but it is finely tuned by the mechanical and dynamical properties of the protein scaffold involved in the interaction.

78 FR 37721 - Approval of American Society of Mechanical Engineers' Code Cases

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-24

...-0359] RIN 3150-AI72 Approval of American Society of Mechanical Engineers' Code Cases AGENCY: Nuclear... mandatory American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code and... Guide'' series. In a notice of proposed rulemaking, ``Approval of American Society of Mechanical...

Analysis Code - Data Analysis in 'Leveraging Multiple Statistical Methods for Inverse Prediction in Nuclear Forensics Applications' (LMSMIPNFA) v. 1.0

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

Lewis, John R

R code that performs the analysis of a data set presented in the paper ‘Leveraging Multiple Statistical Methods for Inverse Prediction in Nuclear Forensics Applications’ by Lewis, J., Zhang, A., Anderson-Cook, C. It provides functions for doing inverse predictions in this setting using several different statistical methods. The data set is a publicly available data set from a historical Plutonium production experiment.

Nuclear Fuel Depletion Analysis Using Matlab Software

NASA Astrophysics Data System (ADS)

Faghihi, F.; Nematollahi, M. R.

Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.

Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

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

Camous, F.; Jacq, F.; Chatelard, P.

1997-07-01

In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2013 CFR

2013-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2012 CFR

2012-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2014 CFR

2014-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

Computer optimization of reactor-thermoelectric space power systems

NASA Technical Reports Server (NTRS)

Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

1973-01-01

A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

Linear and Nonlinear Statistical Characterization of DNA

NASA Astrophysics Data System (ADS)

Norio Oiwa, Nestor; Goldman, Carla; Glazier, James

2002-03-01

We find spatial order in the distribution of protein-coding (including RNAs) and control segments of GenBank genomic sequences, irrespective of ATCG content. This is achieved by correlations, histograms, fractal dimensions and singularity spectra. Estimates of these quantities in complete nuclear genome indicate that coding sequences are long-range correlated and their disposition are self-similar (multifractal) for eukaryotes. These characteristics are absent in prokaryotes, where there are few noncoding sequences, suggesting the `junk' DNA play a relevant role to the genome structure and function. Concerning the genetic message of ATCG sequences, we build a random walk (Levy flight), using DNA symmetry arguments, where we associate A, T, C and G as left, right, down and up steps, respectively. Nonlinear analysis of mitochondrial DNA walks reveal multifractal pattern based on palindromic sequences, which fold in hairpins and loops.

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.

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

Digital Image Correlation of Concrete Slab at University of Tennessee, Knoxville

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

Mahadevan, Sankaran; Agarwal, Vivek; Pham, Binh T.

Assessment and management of aging concrete structures in nuclear power plants require a more systematic approach than simple reliance on existing code margins of safety. Some degradation mechanisms of concrete manifest themselves via swelling or by other shape deformation of the concrete. Specifically, degradation of concrete structure damaged by ASR is viewed as one of the dominant factors impacting the structural integrity of aging nuclear power plants. Structural health monitoring of concrete structures aims to understand the current health condition of a structure based on heterogeneous measurements to produce high-confidence actionable information regarding structural integrity that supports operational and maintenancemore » decisions. Number of nondestructive examination techniques (i.e., thermography, digital image correlation, mechanical deformation measurements, nonlinear impact resonance (DIC) acoustic spectroscopy, and vibro-acoustic modulation) is used to detect the damage caused by ASR. DIC techniques have been increasing in popularity, especially in micro- and nano-scale mechanical testing applications due to its relative ease of implementation and use. Advances in computer technology and digital cameras help this method moving forward. To ensure the best outcome of the DIC system, important factors in the experiment are identified. They include standoff distance, speckle size, speckle pattern, and durable paint. These optimal experimental options are selected basing on a thorough investigation. The resulting DIC deformation map indicates that this technique can be used to generate data related to degradation assessment of concrete structure damaged by the impact of ASR.« less

Using FLUKA to Calculate Spacecraft: Single Event Environments: A Practical Approach

NASA Technical Reports Server (NTRS)

Koontz, Steve; Boeder, Paul; Reddell, Brandon

2009-01-01

The FLUKA nuclear transport and reaction code can be developed into a practical tool for calculation of spacecraft and planetary surface asset SEE and TID environments. Nuclear reactions and secondary particle shower effects can be estimated with acceptable accuracy both in-flight and in test. More detailed electronic device and/or spacecraft geometries than are reported here are possible using standard FLUKA geometry utilities. Spacecraft structure and shielding mass. Effects of high Z elements in microelectronic structure as reported previously. Median shielding mass in a generic slab or concentric sphere target geometry are at least approximately applicable to more complex spacecraft shapes. Need the spacecraft shielding mass distribution function applicable to the microelectronic system of interest. SEE environment effects can be calculated for a wide range of spacecraft and microelectronic materials with complete nuclear physics. Evaluate benefits of low Z shielding mass can be evaluated relative to aluminum. Evaluate effects of high Z elements as constituents of microelectronic devices. The principal limitation on the accuracy of the FLUKA based method reported here are found in the limited accuracy and incomplete character of affordable heavy ion test data. To support accurate rate estimates with any calculation method, the aspect ratio of the sensitive volume(s) and the dependence must be better characterized.

Towards a supported common NEAMS software stack

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

Cormac Garvey

2012-04-01

The NEAMS IPSC's are developing multidimensional, multiphysics, multiscale simulation codes based on first principles that will be capable of predicting all aspects of current and future nuclear reactor systems. These new breeds of simulation codes will include rigorous verification, validation and uncertainty quantification checks to quantify the accuracy and quality of the simulation results. The resulting NEAMS IPSC simulation codes will be an invaluable tool in designing the next generation of Nuclear Reactors and also contribute to a more speedy process in the acquisition of licenses from the NRC for new Reactor designs. Due to the high resolution of themore » models, the complexity of the physics and the added computational resources to quantify the accuracy/quality of the results, the NEAMS IPSC codes will require large HPC resources to carry out the production simulation runs.« less

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

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

Tome, Carlos N; Caro, J A; Lebensohn, R A

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating themore » phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.« less

A MODEL BUILDING CODE ARTICLE ON FALLOUT SHELTERS WITH RECOMMENDATIONS FOR INCLUSION OF REQUIREMENTS FOR FALLOUT SHELTER CONSTRUCTION IN FOUR NATIONAL MODEL BUILDING CODES.

ERIC Educational Resources Information Center

American Inst. of Architects, Washington, DC.

A MODEL BUILDING CODE FOR FALLOUT SHELTERS WAS DRAWN UP FOR INCLUSION IN FOUR NATIONAL MODEL BUILDING CODES. DISCUSSION IS GIVEN OF FALLOUT SHELTERS WITH RESPECT TO--(1) NUCLEAR RADIATION, (2) NATIONAL POLICIES, AND (3) COMMUNITY PLANNING. FALLOUT SHELTER REQUIREMENTS FOR SHIELDING, SPACE, VENTILATION, CONSTRUCTION, AND SERVICES SUCH AS ELECTRICAL…

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... specified in § 50.55, except that each combined license for a boiling or pressurized water-cooled nuclear... boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f... performed. (2) Systems and components of boiling and pressurized water-cooled nuclear power reactors must...

78 FR 9745 - Kewaunee Power Station; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-11

... FURTHER INFORMATION CONTACT: Karl Feintuch, Project Manager, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2013-03037 Filed 2-8-13; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket No. 50-305; NRC-2013-0028] Kewaunee Power Station...

Nuclear fuel management optimization using genetic algorithms

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

DeChaine, M.D.; Feltus, M.A.

1995-07-01

The code independent genetic algorithm reactor optimization (CIGARO) system has been developed to optimize nuclear reactor loading patterns. It uses genetic algorithms (GAs) and a code-independent interface, so any reactor physics code (e.g., CASMO-3/SIMULATE-3) can be used to evaluate the loading patterns. The system is compared to other GA-based loading pattern optimizers. Tests were carried out to maximize the beginning of cycle k{sub eff} for a pressurized water reactor core loading with a penalty function to limit power peaking. The CIGARO system performed well, increasing the k{sub eff} after lowering the peak power. Tests of a prototype parallel evaluation methodmore » showed the potential for a significant speedup.« less

Hadron-rich cosmic-ray families detected by emulsion chamber.

NASA Astrophysics Data System (ADS)

Navia, C. E.; Augusto, C. R. K.; Pinto, F. A.; Shibuya, H.

1995-11-01

Observed hadrons in excess, larger-than-expected charged mesons (pions) in cosmic-ray families detected in emulsion chamber experiment at mountain altitude and produced in a cosmic-ray hadronic interaction not far from the PeV energy region are studied. The hypothesis that these extra hadrons could be a bundle of surviving nuclear fragments (nucleons) is verified through a simulation method using a hybrid code composed of a superposition model to describe the number of interacting nucleon-nucleon pairs in a nucleus-nucleus collision. Together with the UA5 algorithm to describe a nucleon-nucleon collision, atmospheric propagation structure is also considered. A comparison between simulation output with experimental data shows that the surviving-nuclear-fragments hypothesis is not enough to explain the non-pionic hadron excess, even if a heavy dominance composition in the primary flux is considered.

DYNSYL: a general-purpose dynamic simulator for chemical processes

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

Patterson, G.K.; Rozsa, R.B.

1978-09-05

Lawrence Livermore Laboratory is conducting a safeguards program for the Nuclear Regulatory Commission. The goal of the Material Control Project of this program is to evaluate material control and accounting (MCA) methods in plants that handle special nuclear material (SNM). To this end we designed and implemented the dynamic chemical plant simulation program DYNSYL. This program can be used to generate process data or to provide estimates of process performance; it simulates both steady-state and dynamic behavior. The MCA methods that may have to be evaluated range from sophisticated on-line material trackers such as Kalman filter estimators, to relatively simplemore » material balance procedures. This report describes the overall structure of DYNSYL and includes some example problems. The code is still in the experimental stage and revision is continuing.« less

Track structure based modelling of light ion radiation effects on nuclear and mitochondrial DNA

NASA Astrophysics Data System (ADS)

Schmitt, Elke; Ottolenghi, Andrea; Dingfelder, Michael; Friedland, Werner; Kundrat, Pavel; Baiocco, Giorgio

2016-07-01

Space radiation risk assessment is of great importance for manned spaceflights in order to estimate risks and to develop counter-measures to reduce them. Biophysical simulations with PARTRAC can help greatly to improve the understanding of initial biological response to ionizing radiation. Results from modelling radiation quality dependent DNA damage and repair mechanisms up to chromosomal aberrations (e.g. dicentrics) can be used to predict radiation effects depending on the kind of mixed radiation field exposure. Especially dicentric yields can serve as a biomarker for an increased risk due to radiation and hence as an indicator for the effectiveness of the used shielding. PARTRAC [1] is a multi-scale biophysical research MC code for track structure based initial DNA damage and damage response modelling. It integrates physics, radiochemistry, detailed nuclear DNA structure and molecular biology of DNA repair by NHEJ-pathway to assess radiation effects on cellular level [2]. Ongoing experiments with quasi-homogeneously distributed compared to sub-micrometre focused bunches of protons, lithium and carbon ions allow a separation of effects due to DNA damage complexity on nanometre scale from damage clustering on (sub-) micrometre scale [3, 4]. These data provide an unprecedented benchmark for the DNA damage response model in PARTRAC and help understand the mechanisms leading to cell killing and chromosomal aberrations (e.g. dicentrics) induction. A large part of space radiation is due to a mixed ion field of high energy protons and few heavier ions that can be only partly absorbed by the shielding. Radiation damage induced by low-energy ions significantly contributes to the high relative biological efficiency (RBE) of ion beams around Bragg peak regions. For slow light ions the physical cross section data basis in PARTRAC has been extended to investigate radiation quality effects in the Bragg peak region [5]. The resulting range and LET values agree with ICRU data and SRIM calculations. Preliminary studies regarding the biological endpoints DSB (cluster) and chromosomal aberrations have been performed for selected light ions up to neon. Validation with experimental data as well as further calculations are underway and final results will be presented at the meeting. Mitochondrial alterations have been implicated in radiation-induced cardiovascular effects. To extend the applicability of PARTRAC biophysical tool towards effects on mitochondria, the nuclear DNA and chromatin as the primary target of radiation has been complemented by a model of mitochondrial DNA (mtDNA) to mimic a coronary cell with thousand mitochondria contained in the cytoplasm. Induced mtDNA damage (SSB, DSB) has been scored for 60Co photons and 5 MeV alpha-particle irradiation, assuming alternative radical scavenging capacities within the mitochondria. While direct radiation effects in mtDNA are identical to nuclear DNA, indirect effects in mtDNA are in general larger due to lower scavenging and the lack of DNA-protecting histones. These simulations complement the scarce experimental data on radiation-induced mtDNA damage and help elucidate the relative roles of initial mtDNA versus nuclear DNA damage and of pathways that amplify their respective effects. Ongoing and planned developments of PARTRAC include coupling with a radiation transport code and track-structure based calculations of cell killing for RBE studies on macroscopic scales within a mixed ion field. [1] Friedland, Dingfelder et al. (2011): "Track structures, DNA targets and radiation effects in the biophysical Monte Carlo simulation code PARTRAC", Mutat. Res. 711, 28-40 [2] Friedland et al. (2013): "Track structure based modelling of chromosome aberrations after photon and alpha-particle irradiation", Mutat. Res. 756, 213-223 [3] Schmid, Friedland et al. (2015): "Sub-micrometer 20 MeV protons or 45 MeV lithium spot irradiation enhances yields of dicentric chromosomes due to clustering of DNA double-strand breaks", Mutat. Res. 793, 30-40 [4] Friedland, Schmitt, Kundrat (2015): "Modelling Proton bunches focussed to submicrometre scales: Low-LET Radiation damage in high-LET-like spatial structure", Radiat. Prot. Dosim. 166, 34-37 [5] Schmitt, Friedland, Kundrat, Dingfelder, Ottolenghi (2015): "Cross section scaling for track structure simulations of low-energy ions in liquid water", Radiat. Prot. Dosim. 166, 15-18} Supported by the European Atomic Energy Community's Seventh Framework Programme (FP7/2007-2011) under grant agreement no 249689 "DoReMi" and the German Federal Ministry on Education and Research (KVSF-Projekt "LET-Verbund").

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.

SASS-1--SUBASSEMBLY STRESS SURVEY CODE

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

Friedrich, C.M.

1960-01-01

SASS-1, an IBM-704 FORTRAN code, calculates pressure, thermal, and combined stresses in a nuclear reactor core subassembly. In addition to cross- section stresses, the code calculates axial shear stresses needed to keep plane cross sections plane under axial variations of temperature. The input and output nomenclature, arrangement, and formats are described. (B.O.G.)

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.

Stochastic Modeling of Radioactive Material Releases

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

Andrus, Jason; Pope, Chad

2015-09-01

Nonreactor nuclear facilities operated under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculatesmore » the radiation dose associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA was developed using the MATLAB coding framework. The software application has a graphical user input. SODA can be installed on both Windows and Mac computers and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC, rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The work was funded through a grant from the DOE Nuclear Safety Research and Development Program.« less

Barrier distributions and signatures of transfer channels in the Ca40+Ni58,64 fusion reactions at energies around and below the Coulomb barrier

NASA Astrophysics Data System (ADS)

Bourgin, D.; Courtin, S.; Haas, F.; Stefanini, A. M.; Montagnoli, G.; Goasduff, A.; Montanari, D.; Corradi, L.; Fioretto, E.; Huiming, J.; Scarlassara, F.; Rowley, N.; Szilner, S.; Mijatović, T.

2014-10-01

Background: The nuclear structure of colliding nuclei is known to influence the fusion process. Couplings of the relative motion to nuclear shape deformations and vibrations lead to an enhancement of the sub-barrier fusion cross section in comparison with the predictions of one-dimensional barrier penetration models. This enhancement is explained by coupled-channels calculations including these couplings. The sub-barrier fusion cross section is also affected by nucleon transfer channels between the colliding nuclei. Purpose: The aim of the present experiment is to investigate the influence of the projectile and target nuclear structures on the fusion cross sections in the Ca40+Ni58 and Ca40+Ni64 systems. Methods: The experimental and theoretical fusion excitation functions as well as the barrier distributions were compared for these two systems. Coupled-channels calculations were performed using the ccfull code. Results: Good agreement was found between the measured and calculated fusion cross sections for the Ca40+Ni58 system. The situation is different for the Ca40+Ni64 system where the coupled-channels calculations with no nucleon transfer clearly underestimate the fusion cross sections below the Coulomb barrier. The fusion excitation function was, however, well reproduced at low and high energies by including the coupling to the neutron pair-transfer channel in the calculations. Conclusions: The nuclear structure of the colliding nuclei influences the fusion cross sections below the Coulomb barrier for both Ca40+Ni58,64 systems. Moreover, we highlighted the effect of the neutron pair-transfer channel on the fusion cross sections in Ca40+Ni64.

Chromosomal localization and partial genomic structure of the human peroxisome proliferator activated receptor-gamma (hPPAR gamma) gene.

PubMed

Beamer, B A; Negri, C; Yen, C J; Gavrilova, O; Rumberger, J M; Durcan, M J; Yarnall, D P; Hawkins, A L; Griffin, C A; Burns, D K; Roth, J; Reitman, M; Shuldiner, A R

1997-04-28

We determined the chromosomal localization and partial genomic structure of the coding region of the human PPAR gamma gene (hPPAR gamma), a nuclear receptor important for adipocyte differentiation and function. Sequence analysis and long PCR of human genomic DNA with primers that span putative introns revealed that intron positions and sizes of hPPAR gamma are similar to those previously determined for the mouse PPAR gamma gene[13]. Fluorescent in situ hybridization localized hPPAR gamma to chromosome 3, band 3p25. Radiation hybrid mapping with two independent primer pairs was consistent with hPPAR gamma being within 1.5 Mb of marker D3S1263 on 3p25-p24.2. These sequences of the intron/exon junctions of the 6 coding exons shared by hPPAR gamma 1 and hPPAR gamma 2 will facilitate screening for possible mutations. Furthermore, D3S1263 is a suitable polymorphic marker for linkage analysis to evaluate PPAR gamma's potential contribution to genetic susceptibility to obesity, lipoatrophy, insulin resistance, and diabetes.

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

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

BRISC is a developmental prototype for a nextgeneration “systems-level” integrated performance and safety code (IPSC) for nuclear reactors. Its development served to demonstrate how a lightweight multi-physics coupling approach can be used to tightly couple the physics models in several different physics codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled “burner” nuclear reactor. For example, the RIO Fluid Flow and Heat transfer code developed at Sandia (SNL: Chris Moen, Dept. 08005) is used in BRISC to model fluid flow and heat transfer, as well as conduction heat transfermore » in solids. Because BRISC is a prototype, its most practical application is as a foundation or starting point for developing a true production code. The sub-codes and the associated models and correlations currently employed within BRISC were chosen to cover the required application space and demonstrate feasibility, but were not optimized or validated against experimental data within the context of their use in BRISC.« less

A new code for modelling the near field diffusion releases from the final disposal of nuclear waste

NASA Astrophysics Data System (ADS)

Vopálka, D.; Vokál, A.

2003-01-01

The canisters with spent nuclear fuel produced during the operation of WWER reactors at the Czech power plants are planned, like in other countries, to be disposed of in an underground repository. Canisters will be surrounded by compacted bentonite that will retard the migration of safety-relevant radionuclides into the host rock. A new code that enables the modelling of the critical radionuclides transport from the canister through the bentonite layer in the cylindrical geometry was developed. The code enables to solve the diffusion equation for various types of initial and boundary conditions by means of the finite difference method and to take into account the non-linear shape of the sorption isotherm. A comparison of the code reported here with code PAGODA, which is based on analytical solution of the transport equation, was made for the actinide chain 4N+3 that includes 239Pu. A simple parametric study of the releases of 239Pu, 129I, and 14C into geosphere is discussed.

MCNP capabilities for nuclear well logging calculations

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

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

RELAP-7 Code Assessment Plan and Requirement Traceability Matrix

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

Yoo, Junsoo; Choi, Yong-joon; Smith, Curtis L.

2016-10-01

The RELAP-7, a safety analysis code for nuclear reactor system, is under development at Idaho National Laboratory (INL). Overall, the code development is directed towards leveraging the advancements in computer science technology, numerical solution methods and physical models over the last decades. Recently, INL has also been putting an effort to establish the code assessment plan, which aims to ensure an improved final product quality through the RELAP-7 development process. The ultimate goal of this plan is to propose a suitable way to systematically assess the wide range of software requirements for RELAP-7, including the software design, user interface, andmore » technical requirements, etc. To this end, we first survey the literature (i.e., international/domestic reports, research articles) addressing the desirable features generally required for advanced nuclear system safety analysis codes. In addition, the V&V (verification and validation) efforts as well as the legacy issues of several recently-developed codes (e.g., RELAP5-3D, TRACE V5.0) are investigated. Lastly, this paper outlines the Requirement Traceability Matrix (RTM) for RELAP-7 which can be used to systematically evaluate and identify the code development process and its present capability.« less

A Programmable Liquid Collimator for Both Coded Aperture Adaptive Imaging and Multiplexed Compton Scatter Tomography

DTIC Science & Technology

2012-03-01

environments where a source is either weak or shielded. A vehicle of this type could survey large areas after a nuclear attack or a nuclear reactor accident...to prevent its detection by γ-rays. The best application for unmanned vehicles is the detection of radioactive material after a nuclear reactor ...accident or a nuclear weapon detonation [70]. Whether by a nuclear detonation or a nuclear reactor accident, highly radioactive substances could be dis

A Simple Demonstration of Concrete Structural Health Monitoring Framework

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

Mahadevan, Sankaran; Agarwal, Vivek; Cai, Guowei

Assessment and management of aging concrete structures in nuclear power plants require a more systematic approach than simple reliance on existing code margins of safety. Structural health monitoring of concrete structures aims to understand the current health condition of a structure based on heterogeneous measurements to produce high confidence actionable information regarding structural integrity that supports operational and maintenance decisions. This ongoing research project is seeking to develop a probabilistic framework for health diagnosis and prognosis of aging concrete structures in a nuclear power plant subjected to physical, chemical, environment, and mechanical degradation. The proposed framework consists of four elements—damagemore » modeling, monitoring, data analytics, and uncertainty quantification. This report describes a proof-of-concept example on a small concrete slab subjected to a freeze-thaw experiment that explores techniques in each of the four elements of the framework and their integration. An experimental set-up at Vanderbilt University’s Laboratory for Systems Integrity and Reliability is used to research effective combination of full-field techniques that include infrared thermography, digital image correlation, and ultrasonic measurement. The measured data are linked to the probabilistic framework: the thermography, digital image correlation data, and ultrasonic measurement data are used for Bayesian calibration of model parameters, for diagnosis of damage, and for prognosis of future damage. The proof-of-concept demonstration presented in this report highlights the significance of each element of the framework and their integration.« less

Establishment of Imaging Spectroscopy of Nuclear Gamma-Rays based on Geometrical Optics

PubMed Central

Tanimori, Toru; Mizumura, Yoshitaka; Takada, Atsushi; Miyamoto, Shohei; Takemura, Taito; Kishimoto, Tetsuro; Komura, Shotaro; Kubo, Hidetoshi; Kurosawa, Shunsuke; Matsuoka, Yoshihiro; Miuchi, Kentaro; Mizumoto, Tetsuya; Nakamasu, Yuma; Nakamura, Kiseki; Parker, Joseph D.; Sawano, Tatsuya; Sonoda, Shinya; Tomono, Dai; Yoshikawa, Kei

2017-01-01

Since the discovery of nuclear gamma-rays, its imaging has been limited to pseudo imaging, such as Compton Camera (CC) and coded mask. Pseudo imaging does not keep physical information (intensity, or brightness in Optics) along a ray, and thus is capable of no more than qualitative imaging of bright objects. To attain quantitative imaging, cameras that realize geometrical optics is essential, which would be, for nuclear MeV gammas, only possible via complete reconstruction of the Compton process. Recently we have revealed that “Electron Tracking Compton Camera” (ETCC) provides a well-defined Point Spread Function (PSF). The information of an incoming gamma is kept along a ray with the PSF and that is equivalent to geometrical optics. Here we present an imaging-spectroscopic measurement with the ETCC. Our results highlight the intrinsic difficulty with CCs in performing accurate imaging, and show that the ETCC surmounts this problem. The imaging capability also helps the ETCC suppress the noise level dramatically by ~3 orders of magnitude without a shielding structure. Furthermore, full reconstruction of Compton process with the ETCC provides spectra free of Compton edges. These results mark the first proper imaging of nuclear gammas based on the genuine geometrical optics. PMID:28155870

Establishment of Imaging Spectroscopy of Nuclear Gamma-Rays based on Geometrical Optics.

PubMed

Tanimori, Toru; Mizumura, Yoshitaka; Takada, Atsushi; Miyamoto, Shohei; Takemura, Taito; Kishimoto, Tetsuro; Komura, Shotaro; Kubo, Hidetoshi; Kurosawa, Shunsuke; Matsuoka, Yoshihiro; Miuchi, Kentaro; Mizumoto, Tetsuya; Nakamasu, Yuma; Nakamura, Kiseki; Parker, Joseph D; Sawano, Tatsuya; Sonoda, Shinya; Tomono, Dai; Yoshikawa, Kei

2017-02-03

Since the discovery of nuclear gamma-rays, its imaging has been limited to pseudo imaging, such as Compton Camera (CC) and coded mask. Pseudo imaging does not keep physical information (intensity, or brightness in Optics) along a ray, and thus is capable of no more than qualitative imaging of bright objects. To attain quantitative imaging, cameras that realize geometrical optics is essential, which would be, for nuclear MeV gammas, only possible via complete reconstruction of the Compton process. Recently we have revealed that "Electron Tracking Compton Camera" (ETCC) provides a well-defined Point Spread Function (PSF). The information of an incoming gamma is kept along a ray with the PSF and that is equivalent to geometrical optics. Here we present an imaging-spectroscopic measurement with the ETCC. Our results highlight the intrinsic difficulty with CCs in performing accurate imaging, and show that the ETCC surmounts this problem. The imaging capability also helps the ETCC suppress the noise level dramatically by ~3 orders of magnitude without a shielding structure. Furthermore, full reconstruction of Compton process with the ETCC provides spectra free of Compton edges. These results mark the first proper imaging of nuclear gammas based on the genuine geometrical optics.

Deflection by Kinetic Impact or Nuclear Ablation: Sensitivity to Asteroid Properties

NASA Astrophysics Data System (ADS)

Bruck Syal, M.

2015-12-01

Impulsive deflection of a threatening asteroid can be achieved by deploying either a kinetic impactor or a standoff nuclear device to impart a modest velocity change to the body. Response to each of these methods is sensitive to the individual asteroid's characteristics, some of which may not be well constrained before an actual deflection mission. Numerical simulations of asteroid deflection, using both hypervelocity impacts and nuclear ablation of the asteroid's surface, provide detailed information on asteroid response under a range of initial conditions. Here we present numerical results for the deflection of asteroids by both kinetic and nuclear methods, focusing on the roles of target body composition, strength, porosity, rotational state, shape, and internal structure. These results provide a framework for evaluating the planetary defense-related value of future asteroid characterization missions and capture some of the uncertainty that may be present in a real threat scenario. Part of this work was funded by the Laboratory Directed Research and Development Program at LLNL under project tracking code 12-ERD-005, performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. LLNL-ABS-675914.

The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

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

Fischer, U., E-mail: ulrich.fischer@kit.edu; Avrigeanu, M.; Avrigeanu, V.

This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

75 FR 3946 - License Nos. DPR-42 and DPR-60; Northern States Power Company; Prairie Island Nuclear Generating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

..., Office of Nuclear Reactor Regulation. [FR Doc. 2010-1309 Filed 1-22-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0022] License Nos. DPR-42 and DPR-60; Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2...

Facility Targeting, Protection and Mission Decision Making Using the VISAC Code

NASA Technical Reports Server (NTRS)

Morris, Robert H.; Sulfredge, C. David

2011-01-01

The Visual Interactive Site Analysis Code (VISAC) has been used by DTRA and several other agencies to aid in targeting facilities and to predict the associated collateral effects for the go, no go mission decision making process. VISAC integrates the three concepts of target geometric modeling, damage assessment capabilities, and an event/fault tree methodology for evaluating accident/incident consequences. It can analyze a variety of accidents/incidents at nuclear or industrial facilities, ranging from simple component sabotage to an attack with military or terrorist weapons. For nuclear facilities, VISAC predicts the facility damage, estimated downtime, amount and timing of any radionuclides released. Used in conjunction with DTRA's HPAC code, VISAC also can analyze transport and dispersion of the radionuclides, levels of contamination of the surrounding area, and the population at risk. VISAC has also been used by the NRC to aid in the development of protective measures for nuclear facilities that may be subjected to attacks by car/truck bombs.

Photoneutron Reaction Data for Nuclear Physics and Astrophysics

NASA Astrophysics Data System (ADS)

Utsunomiya, Hiroaki; Renstrøm, Therese; Tveten, Gry Merete; Gheorghe, Ioana; Filipescu, Dan Mihai; Belyshev, Sergey; Stopani, Konstantin; Wang, Hongwei; Fan, Gongtao; Lui, Yiu-Wing; Symochko, Dmytro; Goriely, Stephane; Larsen, Ann-Cecilie; Siem, Sunniva; Varlamov, Vladimir; Ishkhanov, Boris; Glodariu, Tudor; Krzysiek, Mateusz; Takenaka, Daiki; Ari-izumi, Takashi; Amano, Sho; Miyamoto, Shuji

2018-05-01

We discuss the role of photoneutron reaction data in nuclear physics and astrophysics in conjunction with the Coordinated Research Project of the International Atomic Energy Agency with the code F41032 (IAEA-CRP F41032).

SINE_scan: an efficient tool to discover short interspersed nuclear elements (SINEs) in large-scale genomic datasets.

PubMed

Mao, Hongliang; Wang, Hao

2017-03-01

Short Interspersed Nuclear Elements (SINEs) are transposable elements (TEs) that amplify through a copy-and-paste mode via RNA intermediates. The computational identification of new SINEs are challenging because of their weak structural signals and rapid diversification in sequences. Here we report SINE_Scan, a highly efficient program to predict SINE elements in genomic DNA sequences. SINE_Scan integrates hallmark of SINE transposition, copy number and structural signals to identify a SINE element. SINE_Scan outperforms the previously published de novo SINE discovery program. It shows high sensitivity and specificity in 19 plant and animal genome assemblies, of which sizes vary from 120 Mb to 3.5 Gb. It identifies numerous new families and substantially increases the estimation of the abundance of SINEs in these genomes. The code of SINE_Scan is freely available at http://github.com/maohlzj/SINE_Scan , implemented in PERL and supported on Linux. wangh8@fudan.edu.cn. Supplementary data are available at Bioinformatics online. © The Author 2016. Published by Oxford University Press.

SINE_scan: an efficient tool to discover short interspersed nuclear elements (SINEs) in large-scale genomic datasets

PubMed Central

Mao, Hongliang

2017-01-01

Abstract Motivation: Short Interspersed Nuclear Elements (SINEs) are transposable elements (TEs) that amplify through a copy-and-paste mode via RNA intermediates. The computational identification of new SINEs are challenging because of their weak structural signals and rapid diversification in sequences. Results: Here we report SINE_Scan, a highly efficient program to predict SINE elements in genomic DNA sequences. SINE_Scan integrates hallmark of SINE transposition, copy number and structural signals to identify a SINE element. SINE_Scan outperforms the previously published de novo SINE discovery program. It shows high sensitivity and specificity in 19 plant and animal genome assemblies, of which sizes vary from 120 Mb to 3.5 Gb. It identifies numerous new families and substantially increases the estimation of the abundance of SINEs in these genomes. Availability and Implementation: The code of SINE_Scan is freely available at http://github.com/maohlzj/SINE_Scan, implemented in PERL and supported on Linux. Contact: wangh8@fudan.edu.cn Supplementary information: Supplementary data are available at Bioinformatics online. PMID:28062442

Examination of total cross section resonance structure of niobium and silicon in neutron transmission experiments

NASA Astrophysics Data System (ADS)

Andrianova, Olga; Lomakov, Gleb; Manturov, Gennady

2017-09-01

The neutron transmission experiments are one of the main sources of information about the neutron cross section resonance structure and effect in the self-shielding. Such kind of data for niobium and silicon nuclides in energy range 7 keV to 3 MeV can be obtained from low-resolution transmission measurements performed earlier in Russia (with samples of 0.027 to 0.871 atom/barn for niobium and 0.076 to 1.803 atom/barn for silicon). A significant calculation-to-experiment discrepancy in energy range 100 to 600 keV and 300 to 800 keV for niobium and silicon, respectively, obtained using the evaluated nuclear data library ROSFOND, were found. The EVPAR code was used for estimation the average resonance parameters in energy range 7 to 600 keV for niobium. For silicon a stochastic optimization method was used to modify the resolved resonance parameters in energy range 300 to 800 keV. The improved ROSFOND evaluated nuclear data files were tested in calculation of ICSBEP integral benchmark experiments.

76 FR 14699 - Southern Nuclear Operating Company; Notice of Availability of Application for a Combined License

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-17

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-025 and 52-026; NRC-2008-0252] Southern Nuclear... Atomic Energy Act and Title 10 of the Code of Federal Regulations (10 CFR) Part 52, ``Licenses... of 10 CFR Part 52. The information submitted by the applicant includes certain administrative...

76 FR 16645 - Southern Nuclear Operating Company; Notice of Availability of Application for a Combined License

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-24

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-025 and 52-026; NRC-2008-0252] Southern Nuclear... Atomic Energy Act and Title 10 of the Code of Federal Regulations (10 CFR) Part 52, ``Licenses... of 10 CFR Part 52. The information submitted by the applicant includes certain administrative...

The Science of Little Boy

ERIC Educational Resources Information Center

Askew, Jennifer; Gray, Ron

2017-01-01

Near the end of World War II, the United States dropped the first nuclear bomb ever used in warfare. The bomb was code named "Little Boy." The fission-type nuclear bomb exploded with the energy equivalent of approximately 13 kilotons of TNT. This article describes a 16 day model-based inquiry (MBI) unit on nuclear chemistry that…

Nonlinear Time Domain Seismic Soil-Structure Interaction (SSI) Deep Soil Site Methodology Development

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

Spears, Robert Edward; Coleman, Justin Leigh

Currently the Department of Energy (DOE) and the nuclear industry perform seismic soil-structure interaction (SSI) analysis using equivalent linear numerical analysis tools. For lower levels of ground motion, these tools should produce reasonable in-structure response values for evaluation of existing and new facilities. For larger levels of ground motion these tools likely overestimate the in-structure response (and therefore structural demand) since they do not consider geometric nonlinearities (such as gaping and sliding between the soil and structure) and are limited in the ability to model nonlinear soil behavior. The current equivalent linear SSI (SASSI) analysis approach either joins the soilmore » and structure together in both tension and compression or releases the soil from the structure for both tension and compression. It also makes linear approximations for material nonlinearities and generalizes energy absorption with viscous damping. This produces the potential for inaccurately establishing where the structural concerns exist and/or inaccurately establishing the amplitude of the in-structure responses. Seismic hazard curves at nuclear facilities have continued to increase over the years as more information has been developed on seismic sources (i.e. faults), additional information gathered on seismic events, and additional research performed to determine local site effects. Seismic hazard curves are used to develop design basis earthquakes (DBE) that are used to evaluate nuclear facility response. As the seismic hazard curves increase, the input ground motions (DBE’s) used to numerically evaluation nuclear facility response increase causing larger in-structure response. As ground motions increase so does the importance of including nonlinear effects in numerical SSI models. To include material nonlinearity in the soil and geometric nonlinearity using contact (gaping and sliding) it is necessary to develop a nonlinear time domain methodology. This methodology will be known as, NonLinear Soil-Structure Interaction (NLSSI). In general NLSSI analysis should provide a more accurate representation of the seismic demands on nuclear facilities their systems and components. INL, in collaboration with a Nuclear Power Plant Vender (NPP-V), will develop a generic Nuclear Power Plant (NPP) structural design to be used in development of the methodology and for comparison with SASSI. This generic NPP design has been evaluated for the INL soil site because of the ease of access and quality of the site specific data. It is now being evaluated for a second site at Vogtle which is located approximately 15 miles East-Northeast of Waynesboro, Georgia and adjacent to Savanna River. The Vogtle site consists of many soil layers spanning down to a depth of 1058 feet. The reason that two soil sites are chosen is to demonstrate the methodology across multiple soil sites. The project will drive the models (soil and structure) using successively increasing acceleration time histories with amplitudes. The models will be run in time domain codes such as ABAQUS, LS-DYNA, and/or ESSI and compared with the same models run in SASSI. The project is focused on developing and documenting a method for performing time domain, non-linear seismic soil structure interaction (SSI) analysis. Development of this method will provide the Department of Energy (DOE) and industry with another tool to perform seismic SSI analysis.« less

Uncertainty evaluation of nuclear reaction model parameters using integral and microscopic measurements. Covariances evaluation with CONRAD code

NASA Astrophysics Data System (ADS)

de Saint Jean, C.; Habert, B.; Archier, P.; Noguere, G.; Bernard, D.; Tommasi, J.; Blaise, P.

2010-10-01

In the [eV;MeV] energy range, modelling of the neutron induced reactions are based on nuclear reaction models having parameters. Estimation of co-variances on cross sections or on nuclear reaction model parameters is a recurrent puzzle in nuclear data evaluation. Major breakthroughs were asked by nuclear reactor physicists to assess proper uncertainties to be used in applications. In this paper, mathematical methods developped in the CONRAD code[2] will be presented to explain the treatment of all type of uncertainties, including experimental ones (statistical and systematic) and propagate them to nuclear reaction model parameters or cross sections. Marginalization procedure will thus be exposed using analytical or Monte-Carlo solutions. Furthermore, one major drawback found by reactor physicist is the fact that integral or analytical experiments (reactor mock-up or simple integral experiment, e.g. ICSBEP, …) were not taken into account sufficiently soon in the evaluation process to remove discrepancies. In this paper, we will describe a mathematical framework to take into account properly this kind of information.

NDEC: A NEA platform for nuclear data testing, verification and benchmarking

NASA Astrophysics Data System (ADS)

Díez, C. J.; Michel-Sendis, F.; Cabellos, O.; Bossant, M.; Soppera, N.

2017-09-01

The selection, testing, verification and benchmarking of evaluated nuclear data consists, in practice, in putting an evaluated file through a number of checking steps where different computational codes verify that the file and the data it contains complies with different requirements. These requirements range from format compliance to good performance in application cases, while at the same time physical constraints and the agreement with experimental data are verified. At NEA, the NDEC (Nuclear Data Evaluation Cycle) platform aims at providing, in a user friendly interface, a thorough diagnose of the quality of a submitted evaluated nuclear data file. Such diagnose is based on the results of different computational codes and routines which carry out the mentioned verifications, tests and checks. NDEC also searches synergies with other existing NEA tools and databases, such as JANIS, DICE or NDaST, including them into its working scheme. Hence, this paper presents NDEC, its current development status and its usage in the JEFF nuclear data project.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

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

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spentmore » fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.« less

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

SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 1, Part 2: Control modules S1--H1; Revision 5

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

NONE

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2010-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2011-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2012-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

Probabilistic approach for decay heat uncertainty estimation using URANIE platform and MENDEL depletion code

NASA Astrophysics Data System (ADS)

Tsilanizara, A.; Gilardi, N.; Huynh, T. D.; Jouanne, C.; Lahaye, S.; Martinez, J. M.; Diop, C. M.

2014-06-01

The knowledge of the decay heat quantity and the associated uncertainties are important issues for the safety of nuclear facilities. Many codes are available to estimate the decay heat. ORIGEN, FISPACT, DARWIN/PEPIN2 are part of them. MENDEL is a new depletion code developed at CEA, with new software architecture, devoted to the calculation of physical quantities related to fuel cycle studies, in particular decay heat. The purpose of this paper is to present a probabilistic approach to assess decay heat uncertainty due to the decay data uncertainties from nuclear data evaluation like JEFF-3.1.1 or ENDF/B-VII.1. This probabilistic approach is based both on MENDEL code and URANIE software which is a CEA uncertainty analysis platform. As preliminary applications, single thermal fission of uranium 235, plutonium 239 and PWR UOx spent fuel cell are investigated.

Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

DOE PAGES

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; ...

2016-01-01

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Effects of Cluster Porosity on the Tensile Properties of Butt-Weldments in T-1 Steel

DTIC Science & Technology

1974-11-01

i 12 Boiler and Pressure Vessel Code .19 In this code, the algebraic difference between the largest and smallest principal stresses is defined...Report U1LU- HN(J 7l-2()24 (University ot Illinois. 1971). "Nuclear Power Components.’* ASME Boiler and Pressure Vessel Code . Section HI. Subsections

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

Regulated Formation of lncRNA-DNA Hybrids Enables Faster Transcriptional Induction and Environmental Adaptation.

PubMed

Cloutier, Sara C; Wang, Siwen; Ma, Wai Kit; Al Husini, Nadra; Dhoondia, Zuzer; Ansari, Athar; Pascuzzi, Pete E; Tran, Elizabeth J

2016-02-04

Long non-coding (lnc)RNAs, once thought to merely represent noise from imprecise transcription initiation, have now emerged as major regulatory entities in all eukaryotes. In contrast to the rapidly expanding identification of individual lncRNAs, mechanistic characterization has lagged behind. Here we provide evidence that the GAL lncRNAs in the budding yeast S. cerevisiae promote transcriptional induction in trans by formation of lncRNA-DNA hybrids or R-loops. The evolutionarily conserved RNA helicase Dbp2 regulates formation of these R-loops as genomic deletion or nuclear depletion results in accumulation of these structures across the GAL cluster gene promoters and coding regions. Enhanced transcriptional induction is manifested by lncRNA-dependent displacement of the Cyc8 co-repressor and subsequent gene looping, suggesting that these lncRNAs promote induction by altering chromatin architecture. Moreover, the GAL lncRNAs confer a competitive fitness advantage to yeast cells because expression of these non-coding molecules correlates with faster adaptation in response to an environmental switch. Copyright © 2016 Elsevier Inc. All rights reserved.

Two-dimensional over-all neutronics analysis of the ITER device

NASA Astrophysics Data System (ADS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Cloud prediction of protein structure and function with PredictProtein for Debian.

PubMed

Kaján, László; Yachdav, Guy; Vicedo, Esmeralda; Steinegger, Martin; Mirdita, Milot; Angermüller, Christof; Böhm, Ariane; Domke, Simon; Ertl, Julia; Mertes, Christian; Reisinger, Eva; Staniewski, Cedric; Rost, Burkhard

2013-01-01

We report the release of PredictProtein for the Debian operating system and derivatives, such as Ubuntu, Bio-Linux, and Cloud BioLinux. The PredictProtein suite is available as a standard set of open source Debian packages. The release covers the most popular prediction methods from the Rost Lab, including methods for the prediction of secondary structure and solvent accessibility (profphd), nuclear localization signals (predictnls), and intrinsically disordered regions (norsnet). We also present two case studies that successfully utilize PredictProtein packages for high performance computing in the cloud: the first analyzes protein disorder for whole organisms, and the second analyzes the effect of all possible single sequence variants in protein coding regions of the human genome.

Cloud Prediction of Protein Structure and Function with PredictProtein for Debian

PubMed Central

Kaján, László; Yachdav, Guy; Vicedo, Esmeralda; Steinegger, Martin; Mirdita, Milot; Angermüller, Christof; Böhm, Ariane; Domke, Simon; Ertl, Julia; Mertes, Christian; Reisinger, Eva; Rost, Burkhard

2013-01-01

We report the release of PredictProtein for the Debian operating system and derivatives, such as Ubuntu, Bio-Linux, and Cloud BioLinux. The PredictProtein suite is available as a standard set of open source Debian packages. The release covers the most popular prediction methods from the Rost Lab, including methods for the prediction of secondary structure and solvent accessibility (profphd), nuclear localization signals (predictnls), and intrinsically disordered regions (norsnet). We also present two case studies that successfully utilize PredictProtein packages for high performance computing in the cloud: the first analyzes protein disorder for whole organisms, and the second analyzes the effect of all possible single sequence variants in protein coding regions of the human genome. PMID:23971032

A first principles study of the electronic structure, elastic and thermal properties of UB2

NASA Astrophysics Data System (ADS)

Jossou, Ericmoore; Malakkal, Linu; Szpunar, Barbara; Oladimeji, Dotun; Szpunar, Jerzy A.

2017-07-01

Uranium diboride (UB2) has been widely deployed for refractory use and is a proposed material for Accident Tolerant Fuel (ATF) due to its high thermal conductivity. However, the applicability of UB2 towards high temperature usage in a nuclear reactor requires the need to investigate the thermomechanical properties, and recent studies have failed in highlighting applicable properties. In this work, we present an in-depth theoretical outlook of the structural and thermophysical properties of UB2, including but not limited to elastic, electronic and thermal transport properties. These calculations were performed within the framework of Density Functional Theory (DFT) + U approach, using Quantum ESPRESSO (QE) code considering the addition of Coulomb correlations on the uranium atom. The phonon spectra and elastic constant analysis show the dynamic and mechanical stability of UB2 structure respectively. The electronic structure of UB2 was investigated using full potential linear augmented plane waves plus local orbitals method (FP-LAPW+lo) as implemented in WIEN2k code. The absence of a band gap in the total and partial density of states confirms the metallic nature while the valence electron density plot reveals the presence of covalent bond between adjacent B-B atoms. We predicted the lattice thermal conductivity (kL) by solving Boltzmann Transport Equation (BTE) using ShengBTE. The second order harmonic and third-order anharmonic interatomic force constants required as input to ShengBTE was calculated using the Density-functional perturbation theory (DFPT). However, we predicted the electronic thermal conductivity (kel) using Wiedemann-Franz law as implemented in Boltztrap code. We also show that the sound velocity along 'a' and 'c' axes exhibit high anisotropy, which accounts for the anisotropic thermal conductivity of UB2.

Advances in Monte-Carlo code TRIPOLI-4®'s treatment of the electromagnetic cascade

NASA Astrophysics Data System (ADS)

Mancusi, Davide; Bonin, Alice; Hugot, François-Xavier; Malouch, Fadhel

2018-01-01

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

Development of single-copy nuclear intron markers for species-level phylogenetics: Case study with Paullinieae (Sapindaceae).

PubMed

Chery, Joyce G; Sass, Chodon; Specht, Chelsea D

2017-09-01

We developed a bioinformatic pipeline that leverages a publicly available genome and published transcriptomes to design primers in conserved coding sequences flanking targeted introns of single-copy nuclear loci. Paullinieae (Sapindaceae) is used to demonstrate the pipeline. Transcriptome reads phylogenetically closer to the lineage of interest are aligned to the closest genome. Single-nucleotide polymorphisms are called, generating a "pseudoreference" closer to the lineage of interest. Several filters are applied to meet the criteria of single-copy nuclear loci with introns of a desired size. Primers are designed in conserved coding sequences flanking introns. Using this pipeline, we developed nine single-copy nuclear intron markers for Paullinieae. This pipeline is highly flexible and can be used for any group with available genomic and transcriptomic resources. This pipeline led to the development of nine variable markers for phylogenetic study without generating sequence data de novo.

Extension of the energy range of the experimental activation cross-sections data of longer-lived products of proton induced nuclear reactions on dysprosium up to 65MeV.

PubMed

Tárkányi, F; Ditrói, F; Takács, S; Hermanne, A; Ignatyuk, A V

2015-04-01

Activation cross-sections data of longer-lived products of proton induced nuclear reactions on dysprosium were extended up to 65MeV by using stacked foil irradiation and gamma spectrometry experimental methods. Experimental cross-sections data for the formation of the radionuclides (159)Dy, (157)Dy, (155)Dy, (161)Tb, (160)Tb, (156)Tb, (155)Tb, (154m2)Tb, (154m1)Tb, (154g)Tb, (153)Tb, (152)Tb and (151)Tb are reported in the 36-65MeV energy range, and compared with an old dataset from 1964. The experimental data were also compared with the results of cross section calculations of the ALICE and EMPIRE nuclear model codes and of the TALYS nuclear reaction model code as listed in the latest on-line libraries TENDL 2013. Copyright © 2015. Published by Elsevier Ltd.

Development of a New 47-Group Library for the CASL Neutronics Simulators

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

Kim, Kang Seog; Williams, Mark L; Wiarda, Dorothea

The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACTmore » library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.« less

Hydrodynamic Studies of Turbulent AGN Tori

NASA Astrophysics Data System (ADS)

Schartmann, M.; Meisenheimer, K.; Klahr, H.; Camenzind, M.; Wolf, S.; Henning, Th.; Burkert, A.; Krause, M.

Recently, the MID-infrared Interferometric instrument (MIDI) at the VLTI has shown that dust tori in the two nearby Seyfert galaxies NGC 1068 and the Circinus galaxy are geometrically thick and can be well described by a thin, warm central disk, surrounded by a colder and fluffy torus component. By carrying out hydrodynamical simulations with the help of the TRAMP code (Klahr et al. 1999), we follow the evolution of a young nuclear star cluster in terms of discrete mass-loss and energy injection from stellar processes. This naturally leads to a filamentary large scale torus component, where cold gas is able to flow radially inwards. The filaments join into a dense and very turbulent disk structure. In a post-processing step, we calculate spectral energy distributions and images with the 3D radiative transfer code MC3D Wolf (2003) and compare them to observations. Turbulence in the dense disk component is investigated in a separate project.

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

Steiner, J.L.; Lime, J.F.; Elson, J.S.

One dimensional TRAC transient calculations of the process inherent ultimate safety (PIUS) advanced reactor design were performed for a pump-trip SCRAM. The TRAC calculations showed that the reactor power response and shutdown were in qualitative agreement with the one-dimensional analyses presented in the PIUS Preliminary Safety Information Document (PSID) submitted by Asea Brown Boveri (ABB) to the US Nuclear Regulatory Commission for preapplication safety review. The PSID analyses were performed with the ABB-developed RIGEL code. The TRAC-calculated phenomena and trends were also similar to those calculated with another one-dimensional PIUS model, the Brookhaven National Laboratory developed PIPA code. A TRACmore » pump-trip SCRAM transient has also been calculated with a TRAC model containing a multi-dimensional representation of the PIUS intemal flow structures and core region. The results obtained using the TRAC fully one-dimensional PIUS model are compared to the RIGEL, PIPA, and TRAC multi-dimensional results.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE PAGES

Chopra, O. K.; Stevens, G. L.; Tregoning, R.; ...

2017-10-06

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

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

Chopra, O. K.; Stevens, G. L.; Tregoning, R.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

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

SKYSINE-II procedure: calculation of the effects of structure design on neutron, primary gamma-ray and secondary gamma-ray dose rates in air

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

Lampley, C.M.

1979-01-01

An updated version of the SKYSHINE Monte Carlo procedure has been developed. The new computer code, SKYSHINE-II, provides a substantial increase in versatility in that the program possesses the ability to address three types of point-isotropic radiation sources: (1) primary gamma rays, (2) neutrons, and (3) secondary gamma rays. In addition, the emitted radiation may now be characterized by an energy emission spectrum product of a new energy-dependent atmospheric transmission data base developed by Radiation Research Associates, Inc. for each of the three source types described above. Most of the computational options present in the original program have been retainedmore » in the new version. Hence, the SKYSHINE-II computer code provides a versatile and viable tool for the analysis of the radiation environment in the vicinity of a building structure containing radiation sources, situated within the confines of a nuclear power plant. This report describes many of the calculational methods employed within the SKYSHINE-II program. A brief description of the new data base is included. Utilization instructions for the program are provided for operation of the SKYSHINE-II code on the Brookhaven National Laboratory Central Scientific Computing Facility. A listing of the source decks, block data routines, and the new atmospheric transmission data base are provided in the appendices of the report.« less

Parallelization of TWOPORFLOW, a Cartesian Grid based Two-phase Porous Media Code for Transient Thermo-hydraulic Simulations

NASA Astrophysics Data System (ADS)

Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor

2014-06-01

TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.

Mediator directs co-transcriptional heterochromatin assembly by RNA interference-dependent and -independent pathways.

PubMed

Oya, Eriko; Kato, Hiroaki; Chikashige, Yuji; Tsutsumi, Chihiro; Hiraoka, Yasushi; Murakami, Yota

2013-01-01

Heterochromatin at the pericentromeric repeats in fission yeast is assembled and spread by an RNAi-dependent mechanism, which is coupled with the transcription of non-coding RNA from the repeats by RNA polymerase II. In addition, Rrp6, a component of the nuclear exosome, also contributes to heterochromatin assembly and is coupled with non-coding RNA transcription. The multi-subunit complex Mediator, which directs initiation of RNA polymerase II-dependent transcription, has recently been suggested to function after initiation in processes such as elongation of transcription and splicing. However, the role of Mediator in the regulation of chromatin structure is not well understood. We investigated the role of Mediator in pericentromeric heterochromatin formation and found that deletion of specific subunits of the head domain of Mediator compromised heterochromatin structure. The Mediator head domain was required for Rrp6-dependent heterochromatin nucleation at the pericentromere and for RNAi-dependent spreading of heterochromatin into the neighboring region. In the latter process, Mediator appeared to contribute to efficient processing of siRNA from transcribed non-coding RNA, which was required for efficient spreading of heterochromatin. Furthermore, the head domain directed efficient transcription in heterochromatin. These results reveal a pivotal role for Mediator in multiple steps of transcription-coupled formation of pericentromeric heterochromatin. This observation further extends the role of Mediator to co-transcriptional chromatin regulation.

Potential Effects of Leak-Before-Break on Light Water Reactor Design.

DTIC Science & Technology

1985-08-26

Boiler and Pressure Vessel Code . In fact, section 3 of that code was created for nuclear applications. This... Boiler and Pressure Vessel Code . The only major change which leak-before-break would require in these analyses would be that all piping to be considered...XI of the ASME Boiler and Pressure Vessel Code , and is already required for all Class I piping systems in the plant. Class I systems are those

Integrated System Modeling for Nuclear Thermal Propulsion (NTP)

NASA Technical Reports Server (NTRS)

Ryan, Stephen W.; Borowski, Stanley K.

2014-01-01

Nuclear thermal propulsion (NTP) has long been identified as a key enabling technology for space exploration beyond LEO. From Wernher Von Braun's early concepts for crewed missions to the Moon and Mars to the current Mars Design Reference Architecture (DRA) 5.0 and recent lunar and asteroid mission studies, the high thrust and specific impulse of NTP opens up possibilities such as reusability that are just not feasible with competing approaches. Although NTP technology was proven in the Rover / NERVA projects in the early days of the space program, an integrated spacecraft using NTP has never been developed. Such a spacecraft presents a challenging multidisciplinary systems integration problem. The disciplines that must come together include not only nuclear propulsion and power, but also thermal management, power, structures, orbital dynamics, etc. Some of this integration logic was incorporated into a vehicle sizing code developed at NASA's Glenn Research Center (GRC) in the early 1990s called MOMMA, and later into an Excel-based tool called SIZER. Recently, a team at GRC has developed an open source framework for solving Multidisciplinary Design, Analysis and Optimization (MDAO) problems called OpenMDAO. A modeling approach is presented that builds on previous work in NTP vehicle sizing and mission analysis by making use of the OpenMDAO framework to enable modular and reconfigurable representations of various NTP vehicle configurations and mission scenarios. This approach is currently applied to vehicle sizing, but is extensible to optimization of vehicle and mission designs. The key features of the code will be discussed and examples of NTP transfer vehicles and candidate missions will be presented.

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

Assessment of WWMCCS Performance in a Post-Nuclear Attack Environment. Sanitized

DTIC Science & Technology

1975-11-01

34cONTRACT No. 001-75-C-0217 * 1 THIS WORK SPONSOQRED BY THE DEFENSE NUCLEAR AGENC" UNDER RDT&E RMSS CODE X36075669 Q94OAXDDO01O1 H2590D. Pr p r d 1 rTHIS DOCUM...8217 < : " ! , . .I/• I V...... .... . . . . . . 1 "- 7 . . . • .tEPORT DOCJMENTATION PAGE "rAr, . Irt’. ±~ATSSIS9.C.7NF VMC EPPINEI 21 Jan 75-31 Oc- 75 IOST-I.CLLEA...U .. 4II4 .i %Iy•11 .4 @$ . .~ *󈧯.I . 1 . S $P.IL%-V ka"I 1 "- This work sponsored by the Defe.xse Nuclear Agency under %DT&E WISS Code X360075469 Q9

Coulomb effects in low-energy nuclear fragmentation

NASA Technical Reports Server (NTRS)

Wilson, John W.; Chun, Sang Y.; Badavi, Francis F.; John, Sarah

1993-01-01

Early versions of the Langley nuclear fragmentation code NUCFRAG (and a publicly released version called HZEFRG1) assumed straight-line trajectories throughout the interaction. As a consequence, NUCFRAG and HZEFRG1 give unrealistic cross sections for large mass removal from the projectile and target at low energies. A correction for the distortion of the trajectory by the nuclear Coulomb fields is used to derive fragmentation cross sections. A simple energy-loss term is applied to estimate the energy downshifts that greatly alter the Coulomb trajectory at low energy. The results, which are far more realistic than prior versions of the code, should provide the data base for future transport calculations. The systematic behavior of charge-removal cross sections compares favorably with results from low-energy experiments.

The new interactive CESAR

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

Fox, P.B.; Yatabe, M.

1987-01-01

In this report the Nuclear Criticality Safety Analytical Methods Resource Center describes a new interactive version of CESAR, a critical experiments storage and retrieval program available on the Nuclear Criticality Information System (NCIS) database at Lawrence Livermore National Laboratory. The original version of CESAR did not include interactive search capabilities. The CESAR database was developed to provide a convenient, readily accessible means of storing and retrieving code input data for the SCALE Criticality Safety Analytical Sequences and the codes comprising those sequences. The database includes data for both cross section preparation and criticality safety calculations. 3 refs., 1 tab.

New interactive CESAR

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

Fox, P.B.; Yatabe, M.

1987-01-01

The Nuclear Criticality Safety Analytical Methods Resource Center announces the availability of a new interactive version of CESAR, a critical experiments storage and retrieval program available on the Nuclear Criticality Information System (NCIS) data base at Lawrence Livermore National Laboratory. The original version of CESAR did not include interactive search capabilities. The CESAR data base was developed to provide a convenient, readily accessible means of storing and retrieving code input data for the SCALE criticality safety analytical sequences and the codes comprising those sequences. The data base includes data for both cross-section preparation and criticality safety calculations.

Comparison of MELCOR and SCDAP/RELAP5 results for a low-pressure, short-term station blackout at Browns Ferry

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

Carbajo, J.J.

1995-12-31

This study compares results obtained with two U.S. Nuclear Regulatory Commission (NRC)-sponsored codes, MELCOR version 1.8.3 (1.8PQ) and SCDAP/RELAP5 Mod3.1 release C, for the same transient - a low-pressure, short-term station blackout accident at the Browns Ferry nuclear plant. This work is part of MELCOR assessment activities to compare core damage progression calculations of MELCOR against SCDAP/RELAP5 since the two codes model core damage progression very differently.

78 FR 15009 - Consideration of Withdrawal From Commercial Production and Distribution of the Radioisotope...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-08

... may be addressed to: Dr. Marc Garland, Program Manager, Office of Nuclear Physics, Office of Science... Management Division, Office of Nuclear Physics, Office of Science, U.S. Department of Energy, Germantown..., Office of Nuclear Physics, Office of Science. [FR Doc. 2013-05444 Filed 3-7-13; 8:45 am] BILLING CODE...

75 FR 57535 - Northern States Power Company-Minnesota Notice of Issuance of Amendments to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-21

... issuance. FOR FURTHER INFORMATION CONTACT: Thomas J. Wengert, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2010-23516 Filed 9-20-10; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0290] Northern States Power...

The Oncogenic Fusion Proteins SET-Nup214 and Sequestosome-1 (SQSTM1)-Nup214 Form Dynamic Nuclear Bodies and Differentially Affect Nuclear Protein and Poly(A)+ RNA Export.

PubMed

Port, Sarah A; Mendes, Adélia; Valkova, Christina; Spillner, Christiane; Fahrenkrog, Birthe; Kaether, Christoph; Kehlenbach, Ralph H

2016-10-28

Genetic rearrangements are a hallmark of several forms of leukemia and can lead to oncogenic fusion proteins. One example of an affected chromosomal region is the gene coding for Nup214, a nucleoporin that localizes to the cytoplasmic side of the nuclear pore complex (NPC). We investigated two such fusion proteins, SET-Nup214 and SQSTM1 (sequestosome)-Nup214, both containing C-terminal portions of Nup214. SET-Nup214 nuclear bodies containing the nuclear export receptor CRM1 were observed in the leukemia cell lines LOUCY and MEGAL. Overexpression of SET-Nup214 in HeLa cells leads to the formation of similar nuclear bodies that recruit CRM1, export cargo proteins, and certain nucleoporins and concomitantly affect nuclear protein and poly(A) + RNA export. SQSTM1-Nup214, although mostly cytoplasmic, also forms nuclear bodies and inhibits nuclear protein but not poly(A) + RNA export. The interaction of the fusion proteins with CRM1 is RanGTP-dependent, as shown in co-immunoprecipitation experiments and binding assays. Further analysis revealed that the Nup214 parts mediate the inhibition of nuclear export, whereas the SET or SQSTM1 part determines the localization of the fusion protein and therefore the extent of the effect. SET-Nup214 nuclear bodies are highly mobile structures, which are in equilibrium with the nucleoplasm in interphase and disassemble during mitosis or upon treatment of cells with the CRM1-inhibitor leptomycin B. Strikingly, we found that nucleoporins can be released from nuclear bodies and reintegrated into existing NPC. Our results point to nuclear bodies as a means of preventing the formation of potentially insoluble and harmful protein aggregates that also may serve as storage compartments for nuclear transport factors. © 2016 by The American Society for Biochemistry and Molecular Biology, Inc.

The Oncogenic Fusion Proteins SET-Nup214 and Sequestosome-1 (SQSTM1)-Nup214 Form Dynamic Nuclear Bodies and Differentially Affect Nuclear Protein and Poly(A)+ RNA Export*

PubMed Central

Port, Sarah A.; Mendes, Adélia; Valkova, Christina; Spillner, Christiane; Fahrenkrog, Birthe; Kaether, Christoph; Kehlenbach, Ralph H.

2016-01-01

Genetic rearrangements are a hallmark of several forms of leukemia and can lead to oncogenic fusion proteins. One example of an affected chromosomal region is the gene coding for Nup214, a nucleoporin that localizes to the cytoplasmic side of the nuclear pore complex (NPC). We investigated two such fusion proteins, SET-Nup214 and SQSTM1 (sequestosome)-Nup214, both containing C-terminal portions of Nup214. SET-Nup214 nuclear bodies containing the nuclear export receptor CRM1 were observed in the leukemia cell lines LOUCY and MEGAL. Overexpression of SET-Nup214 in HeLa cells leads to the formation of similar nuclear bodies that recruit CRM1, export cargo proteins, and certain nucleoporins and concomitantly affect nuclear protein and poly(A)+ RNA export. SQSTM1-Nup214, although mostly cytoplasmic, also forms nuclear bodies and inhibits nuclear protein but not poly(A)+ RNA export. The interaction of the fusion proteins with CRM1 is RanGTP-dependent, as shown in co-immunoprecipitation experiments and binding assays. Further analysis revealed that the Nup214 parts mediate the inhibition of nuclear export, whereas the SET or SQSTM1 part determines the localization of the fusion protein and therefore the extent of the effect. SET-Nup214 nuclear bodies are highly mobile structures, which are in equilibrium with the nucleoplasm in interphase and disassemble during mitosis or upon treatment of cells with the CRM1-inhibitor leptomycin B. Strikingly, we found that nucleoporins can be released from nuclear bodies and reintegrated into existing NPC. Our results point to nuclear bodies as a means of preventing the formation of potentially insoluble and harmful protein aggregates that also may serve as storage compartments for nuclear transport factors. PMID:27613868

SkyNet: A Modular Nuclear Reaction Network Library

NASA Astrophysics Data System (ADS)

Lippuner, Jonas; Roberts, Luke F.

2017-12-01

Almost all of the elements heavier than hydrogen that are present in our solar system were produced by nuclear burning processes either in the early universe or at some point in the life cycle of stars. In all of these environments, there are dozens to thousands of nuclear species that interact with each other to produce successively heavier elements. In this paper, we present SkyNet, a new general-purpose nuclear reaction network that evolves the abundances of nuclear species under the influence of nuclear reactions. SkyNet can be used to compute the nucleosynthesis evolution in all astrophysical scenarios where nucleosynthesis occurs. SkyNet is free and open source, and aims to be easy to use and flexible. Any list of isotopes can be evolved, and SkyNet supports different types of nuclear reactions. SkyNet is modular so that new or existing physics, like nuclear reactions or equations of state, can easily be added or modified. Here, we present in detail the physics implemented in SkyNet with a focus on a self-consistent transition to and from nuclear statistical equilibrium to non-equilibrium nuclear burning, our implementation of electron screening, and coupling of the network to an equation of state. We also present comprehensive code tests and comparisons with existing nuclear reaction networks. We find that SkyNet agrees with published results and other codes to an accuracy of a few percent. Discrepancies, where they exist, can be traced to differences in the physics implementations.

Prospective scenarios of nuclear energy evolution over the 21. century

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

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

Thermodynamic Effects of Noncoded and Coded Methionine Substitutions in Calmodulin

PubMed Central

Yamniuk, Aaron P.; Ishida, Hiroaki; Lippert, Dustin; Vogel, Hans J.

2009-01-01

The methionine residues in the calcium (Ca2+) regulatory protein calmodulin (CaM) are structurally and functionally important. They are buried within the N- and C-domains of apo-CaM but become solvent-exposed in Ca2+-CaM, where they interact with numerous target proteins. Previous structural studies have shown that methionine substitutions to the noncoded amino acids selenomethionine, ethionine, or norleucine, or mutation to leucine do not impact the main chain structure of CaM. Here we used differential scanning calorimetry to show that these substitutions enhance the stability of both domains, with the largest increase in melting temperature (19–26°C) achieved with leucine or norleucine in the apo-C-domain. Nuclear magnetic resonance spectroscopy experiments also revealed the loss of a slow conformational exchange process in the Leu-substituted apo-C-domain. In addition, isothermal titration calorimetry experiments revealed considerable changes in the enthalpy and entropy of target binding to apo-CaM and Ca2+-CaM, but the free energy of binding was largely unaffected due to enthalpy-entropy compensation. Collectively, these results demonstrate that noncoded and coded methionine substitutions can be accommodated in CaM because of the structural plasticity of the protein. However, adjustments in side-chain packing and dynamics lead to significant differences in protein stability and the thermodynamics of target binding. PMID:19217866

Microstructured silicon neutron detectors for security applications

NASA Astrophysics Data System (ADS)

Esteban, S.; Fleta, C.; Guardiola, C.; Jumilla, C.; Pellegrini, G.; Quirion, D.; Rodriguez, J.; Lozano, M.

2014-12-01

In this paper we present the design and performance of a perforated thermal neutron silicon detector with a 6LiF neutron converter. This device was manufactured within the REWARD project workplace whose aim is to develop and enhance technologies for the detection of nuclear and radiological materials. The sensor perforated structure results in a higher efficiency than that obtained with an equivalent planar sensor. The detectors were tested in a thermal neutron beam at the nuclear reactor at the Instituto Superior Técnico in Lisbon and the intrinsic detection efficiency for thermal neutrons and the gamma sensitivity were obtained. The Geant4 Monte Carlo code was used to simulate the experimental conditions, i.e. thermal neutron beam and the whole detector geometry. An intrinsic thermal neutron detection efficiency of 8.6%±0.4% with a discrimination setting of 450 keV was measured.

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.

Extremely accurate sequential verification of RELAP5-3D

DOE PAGES

Mesina, George L.; Aumiller, David L.; Buschman, Francis X.

2015-11-19

Large computer programs like RELAP5-3D solve complex systems of governing, closure and special process equations to model the underlying physics of nuclear power plants. Further, these programs incorporate many other features for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. For RELAP5-3D, verification and validation are restricted to nuclear power plant applications. Verification means ensuring that the program is built right by checking that it meets its design specifications, comparing coding to algorithms and equations and comparing calculations against analytical solutions and method ofmore » manufactured solutions. Sequential verification performs these comparisons initially, but thereafter only compares code calculations between consecutive code versions to demonstrate that no unintended changes have been introduced. Recently, an automated, highly accurate sequential verification method has been developed for RELAP5-3D. The method also provides to test that no unintended consequences result from code development in the following code capabilities: repeating a timestep advancement, continuing a run from a restart file, multiple cases in a single code execution, and modes of coupled/uncoupled operation. In conclusion, mathematical analyses of the adequacy of the checks used in the comparisons are provided.« less

Demonstration of emulator-based Bayesian calibration of safety analysis codes: Theory and formulation

DOE PAGES

Yurko, Joseph P.; Buongiorno, Jacopo; Youngblood, Robert

2015-05-28

System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here withmore » Markov Chain Monte Carlo (MCMC) sampling feasible. This study uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.« less

Modeling Potential Tephra Dispersal at Yucca Mountain, Nevada

NASA Astrophysics Data System (ADS)

Hooper, D.; Franklin, N.; Adams, N.; Basu, D.

2006-12-01

Quaternary basaltic volcanoes exist within 20 km [12 mi] of the potential radioactive waste repository at Yucca Mountain, Nevada, and future basaltic volcanism at the repository is considered a low-probability, potentially high-consequence event. If radioactive waste was entrained in the conduit of a future volcanic event, tephra and waste could be transported in the resulting eruption plume. During an eruption, basaltic tephra would be dispersed primarily according to the height of the eruption column, particle-size distribution, and structure of the winds aloft. Following an eruption, contaminated tephra-fall deposits would be affected by surface redistribution processes. The Center for Nuclear Waste Regulatory Analyses developed the computer code TEPHRA to calculate atmospheric dispersion and subsequent deposition of tephra and spent nuclear fuel from a potential eruption at Yucca Mountain and to help prepare the U.S. Nuclear Regulatory Commission to review a potential U.S. Department of Energy license application. The TEPHRA transport code uses the Suzuki model to simulate the thermo-fluid dynamics of atmospheric tephra dispersion. TEPHRA models the transport of airborne pyroclasts based on particle diffusion from an eruption column, horizontal diffusion of particles by atmospheric and plume turbulence, horizontal advection by atmospheric circulation, and particle settling by gravity. More recently, TEPHRA was modified to calculate potential tephra deposit distributions using stratified wind fields based on upper atmosphere data from the Nevada Test Site. Wind data are binned into 1-km [0.62-mi]-high intervals with coupled distributions of wind speed and direction produced for each interval. Using this stratified wind field and discretization with respect to height, TEPHRA calculates particle fall and lateral displacement for each interval. This implementation permits modeling of split wind fields. We use a parallel version of the code to calculate expected tephra and high-level waste accumulation at specified points on a two-dimensional spatial grid, thereby simulating a three- dimensional initial deposit. To assess subsequent tephra and high-level waste redistribution and resuspension, modeling grids were devised to measure deposition in eolian and fluvial source regions. The eolian grid covers an area of 2,600 km2 [1,000 mi2] and the fluvial grid encompasses 318 km2 [123 mi2] of the southernmost portion of the Fortymile Wash catchment basin. Because each realization is independent, distributions of tephra and high-level waste reflect anticipated variations in source-term and transport characteristics. This abstract is an independent product of the Center for Nuclear Waste Regulatory Analyses and does not necessarily reflect the view or regulatory position of the U.S. Nuclear Regulatory Commission.

CEM2k and LAQGSM Codes as Event-Generators for Space Radiation Shield and Cosmic Rays Propagation Applications

NASA Technical Reports Server (NTRS)

Mashnik, S. G.; Gudima, K. K.; Sierk, A. J.; Moskalenko, I. V.

2002-01-01

Space radiation shield applications and studies of cosmic ray propagation in the Galaxy require reliable cross sections to calculate spectra of secondary particles and yields of the isotopes produced in nuclear reactions induced both by particles and nuclei at energies from threshold to hundreds of GeV per nucleon. Since the data often exist in a very limited energy range or sometimes not at all, the only way to obtain an estimate of the production cross sections is to use theoretical models and codes. Recently, we have developed improved versions of the Cascade-Exciton Model (CEM) of nuclear reactions: the codes CEM97 and CEM2k for description of particle-nucleus reactions at energies up to about 5 GeV. In addition, we have developed a LANL version of the Quark-Gluon String Model (LAQGSM) to describe reactions induced both by particles and nuclei at energies up to hundreds of GeVhucleon. We have tested and benchmarked the CEM and LAQGSM codes against a large variety of experimental data and have compared their results with predictions by other currently available models and codes. Our benchmarks show that CEM and LAQGSM codes have predictive powers no worse than other currently used codes and describe many reactions better than other codes; therefore both our codes can be used as reliable event-generators for space radiation shield and cosmic ray propagation applications. The CEM2k code is being incorporated into the transport code MCNPX (and several other transport codes), and we plan to incorporate LAQGSM into MCNPX in the near future. Here, we present the current status of the CEM2k and LAQGSM codes, and show results and applications to studies of cosmic ray propagation in the Galaxy.

Method and apparatus for reading lased bar codes on shiny-finished fuel rod cladding tubes

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

Goldenfield, M.P.; Lambert, D.V.

1990-10-02

This patent describes, in a nuclear fuel rod identification system, a method of reading a bar code etched directly on a surface of a nuclear fuel rod. It comprises: defining a pair of light diffuser surfaces adjacent one another but in oppositely inclined relation to a beam of light emitted from a light reader; positioning a fuel rod, having a cylindrical surface portion with a bar code etched directly thereon, relative to the light diffuser surfaces such that the surfaces are disposed adjacent to and in oppositely inclined relation along opposite sides of the fuel rod surface portion and themore » fuel rod surface portion is aligned with the beam of light emitted from the light reader; directing the beam of light on the bar code on fuel rod cylindrical surface portion such that the light is reflected therefrom onto one of the light diffuser surfaces; and receiving and reading the reflected light from the bar code via the one of the light diffuser surfaces to the light reader.« less

Assessment of current atomic scale modelling methods for the investigation of nuclear fuels under irradiation: Example of uranium dioxide

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

Bertolus, Marjorie; Krack, Matthias; Freyss, Michel

Multiscale approaches are developed to build more physically based kinetic and mechanical mesoscale models to enhance the predictive capability of fuel performance codes and increase the efficiency of the development of the safer and more innovative nuclear materials needed in the future. Atomic scale methods, and in particular electronic structure and empirical potential methods, form the basis of this multiscale approach. It is therefore essential to know the accuracy of the results computed at this scale if we want to feed them into higher scale models. We focus here on the assessment of the description of interatomic interactions in uraniummore » dioxide using on the one hand electronic structure methods, in particular in the density functional theory (DFT) framework and on the other hand empirical potential methods. These two types of methods are complementary, the former enabling to get results from a minimal amount of input data and further insight into the electronic and magnetic properties, while the latter are irreplaceable for studies where a large number of atoms needs to be considered. We consider basic properties as well as specific ones, which are important for the description of nuclear fuel under irradiation. These are especially energies, which are the main data passed to higher scale models. We limit ourselves to uranium dioxide.« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

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

Hurst, Aaron M.

A data structure based on an eXtensible Markup Language (XML) hierarchy according to experimental nuclear structure data in the Evaluated Nuclear Structure Data File (ENSDF) is presented. A Python-coded translator has been developed to interpret the standard one-card records of the ENSDF datasets, together with their associated quantities defined according to field position, and generate corresponding representative XML output. The quantities belonging to this mixed-record format are described in the ENSDF manual. Of the 16 ENSDF records in total, XML output has been successfully generated for 15 records. An XML-translation for the Comment Record is yet to be implemented; thismore » will be considered in a separate phase of the overall translation effort. Continuation records, not yet implemented, will also be treated in a future phase of this work. Several examples are presented in this document to illustrate the XML schema and methods for handling the various ENSDF data types. However, the proposed nomenclature for the XML elements and attributes need not necessarily be considered as a fixed set of constructs. Indeed, better conventions may be suggested and a consensus can be achieved amongst the various groups of people interested in this project. The main purpose here is to present an initial phase of the translation effort to demonstrate the feasibility of interpreting ENSDF datasets and creating a representative XML-structured hierarchy for data storage.« less

Comparative analysis of structural concrete quality assurance practices on nine nuclear power plant construction projects. Final report

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

Willenbrock, J.H.; Thomas, H.R. Jr.; Burati, J.L. Jr.

1978-06-01

The basic objective of this research effort was to perform a comparative analysis of the Quality Assurance practices related to the structural concrete phase on nine nuclear power plant projects which are (or have been) under construction in the United States in the past ten years. This analysis identified the response of each Quality Assurance program to the applicable criteria of 10 CFR Part 50, Appendix B as well as to the pertinent regulatory requirements and industry standards. The major emphasis was placed on the construction aspects of the structural concrete phase of each project. The engineering and design aspectsmore » were examined whenever they interfaced with the construction aspects. For those aspects of the Quality Assurance system which can be considered managerial in nature (i.e., organizational relationships, types of Quality Assurance programs, corrective action procedures, etc.) an attempt has been made to present the alternative approaches that were identified. For those aspects of the Quality Assurance system which are technical in nature (i.e., the frequency of testing for slump, compressive strength, etc.) an attempt has been made to present a comparative analysis between projects and in relation to the recommended or mandated practices presented in the appropriate industry codes and standards.« less

A Freirean Approach to Peacemaking.

ERIC Educational Resources Information Center

Moriarty, Pia

1989-01-01

The Nuclear Disarmament Project involving Catholic churches and schools in San Francisco used Freirean codifications, with photographs as codes, to develop discussions on the moral issues of nuclear arms. Group discussions led to concrete action in the cause of peace and social justice. (SK)

Examples of Use of SINBAD Database for Nuclear Data and Code Validation

NASA Astrophysics Data System (ADS)

Kodeli, Ivan; Žerovnik, Gašper; Milocco, Alberto

2017-09-01

The SINBAD database currently contains compilations and evaluations of over 100 shielding benchmark experiments. The SINBAD database is widely used for code and data validation. Materials covered include: Air, N. O, H2O, Al, Be, Cu, graphite, concrete, Fe, stainless steel, Pb, Li, Ni, Nb, SiC, Na, W, V and mixtures thereof. Over 40 organisations from 14 countries and 2 international organisations have contributed data and work in support of SINBAD. Examples of the use of the database in the scope of different international projects, such as the Working Party on Evaluation Cooperation of the OECD and the European Fusion Programme demonstrate the merit and possible usage of the database for the validation of modern nuclear data evaluations and new computer codes.

Activation assessment of the soil around the ESS accelerator tunnel

NASA Astrophysics Data System (ADS)

Rakhno, I. L.; Mokhov, N. V.; Tropin, I. S.; Ene, D.

2018-06-01

Activation of the soil surrounding the ESS accelerator tunnel calculated by the MARS15 code is presented. A detailed composition of the soil, that comprises about 30 chemical elements, is considered. Spatial distributions of the produced activity are provided in both transverse and longitudinal directions. A realistic irradiation profile for the entire planned lifetime of the facility is used. The nuclear transmutation and decay of the produced radionuclides is calculated with the DeTra code which is a built-in tool for the MARS15 code. Radionuclide production by low-energy neutrons is calculated using the ENDF/B-VII evaluated nuclear data library. In order to estimate quality of this activation assessment, a comparison between calculated and measured activation of various foils in a similar radiation environment is presented.

History of the Nuclei Important for Cosmochemistry

NASA Technical Reports Server (NTRS)

Meyer, Bradley S.

2004-01-01

An essential aspect of studying the nuclei important for cosmochemistry is their production in stars. Over the grant period, we have further developed the Clemson/American University of Beirut stellar evolution code. Through use of a biconjugate-gradient matrix solver, we now routinely solve l0(exp 6) x l0(exp 6) sparse matrices on our desktop computers. This has allowed us to couple nucleosynthesis and convection fully in the 1-D star, which, in turn, provides better estimates of nuclear yields when the mixing and nuclear burning timescales are comparable. We also have incorporated radiation transport into our 1-D supernova explosion code. We used the stellar evolution and explosion codes to compute iron abundances in a 25 Solar mass star and compared the results to data from RIMS.

Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

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

Ghoos, K., E-mail: kristel.ghoos@kuleuven.be; Dekeyser, W.; Samaey, G.

2016-10-01

The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracymore » by making use of averaging in the Random Noise coupling technique.« less

Determination of the melon chloroplast and mitochondrial genome sequences reveals that the largest reported mitochondrial genome in plants contains a significant amount of DNA having a nuclear origin

PubMed Central

2011-01-01

Background The melon belongs to the Cucurbitaceae family, whose economic importance among vegetable crops is second only to Solanaceae. The melon has a small genome size (454 Mb), which makes it suitable for molecular and genetic studies. Despite similar nuclear and chloroplast genome sizes, cucurbits show great variation when their mitochondrial genomes are compared. The melon possesses the largest plant mitochondrial genome, as much as eight times larger than that of other cucurbits. Results The nucleotide sequences of the melon chloroplast and mitochondrial genomes were determined. The chloroplast genome (156,017 bp) included 132 genes, with 98 single-copy genes dispersed between the small (SSC) and large (LSC) single-copy regions and 17 duplicated genes in the inverted repeat regions (IRa and IRb). A comparison of the cucumber and melon chloroplast genomes showed differences in only approximately 5% of nucleotides, mainly due to short indels and SNPs. Additionally, 2.74 Mb of mitochondrial sequence, accounting for 95% of the estimated mitochondrial genome size, were assembled into five scaffolds and four additional unscaffolded contigs. An 84% of the mitochondrial genome is contained in a single scaffold. The gene-coding region accounted for 1.7% (45,926 bp) of the total sequence, including 51 protein-coding genes, 4 conserved ORFs, 3 rRNA genes and 24 tRNA genes. Despite the differences observed in the mitochondrial genome sizes of cucurbit species, Citrullus lanatus (379 kb), Cucurbita pepo (983 kb) and Cucumis melo (2,740 kb) share 120 kb of sequence, including the predicted protein-coding regions. Nevertheless, melon contained a high number of repetitive sequences and a high content of DNA of nuclear origin, which represented 42% and 47% of the total sequence, respectively. Conclusions Whereas the size and gene organisation of chloroplast genomes are similar among the cucurbit species, mitochondrial genomes show a wide variety of sizes, with a non-conserved structure both in gene number and organisation, as well as in the features of the noncoding DNA. The transfer of nuclear DNA to the melon mitochondrial genome and the high proportion of repetitive DNA appear to explain the size of the largest mitochondrial genome reported so far. PMID:21854637

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

The Italian experience on T/H best estimate codes: Achievements and perspectives

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

Alemberti, A.; D`Auria, F.; Fiorino, E.

1997-07-01

Themalhydraulic system codes are complex tools developed to simulate the power plants behavior during off-normal conditions. Among the objectives of the code calculations the evaluation of safety margins, the operator training, the optimization of the plant design and of the emergency operating procedures, are mostly considered in the field of the nuclear safety. The first generation of codes was developed in the United States at the end of `60s. Since that time, different research groups all over the world started the development of their own codes. At the beginning of the `80s, the second generation codes were proposed; these differmore » from the first generation codes owing to the number of balance equations solved (six instead of three), the sophistication of the constitutive models and of the adopted numerics. The capabilities of available computers have been fully exploited during the years. The authors then summarize some of the major steps in the process of developing, modifying, and advancing the capabilities of the codes. They touch on the fact that Italian, and for that matter non-American, researchers have not been intimately involved in much of this work. They then describe the application of these codes in Italy, even though there are no operating or under construction nuclear power plants at this time. Much of this effort is directed at the general question of plant safety in the face of transient type events.« less

Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2) multi-group (199 n + 42 γ) cross section library in AMPX format for nuclear fission applications

NASA Astrophysics Data System (ADS)

Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

2017-09-01

The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

Photonuclear reactions in astrophysical p-process: Theoretical calculations and experiment simulation based on ELI-NP

NASA Astrophysics Data System (ADS)

Xu, Yi; Luo, Wen; Balabanski, Dimiter; Goriely, Stephane; Matei, Catalin; Tesileanu, Ovidiu

2017-09-01

The astrophysical p-process is an important way of nucleosynthesis to produce the stable and proton-rich nuclei beyond Fe which can not be reached by the s- and r-processes. In the present study, the astrophysical reaction rates of (γ,n), (γ,p), and (γ,α) reactions are computed within the modern reaction code TALYS for about 3000 stable and proton-rich nuclei with 12 < Z < 110. The nuclear structure ingredients involved in the calculation are determined from experimental data whenever available and, if not, from global microscopic nuclear models. In particular, both of the Wood-Saxon potential and the double folding potential with density dependent M3Y (DDM3Y) effective interaction are used for the calculations. It is found that the photonuclear reaction rates are very sensitive to the nuclear potential, and the better determination of nuclear potential would be important to reduce the uncertainties of reaction rates. Meanwhile, the Extreme Light Infrastructure-Nuclear Physics (ELI-NP) facility is being developed, which will provide the great opportunity to experimentally study the photonuclear reactions in p-process. Simulations of the experimental setup for the measurements of the photonuclear reactions 96Ru(γ,p) and 96Ru(γ,α) are performed. It is shown that the experiments of photonuclear reactions in p-process based on ELI-NP are quite promising.

Investigation of HZETRN 2010 as a Tool for Single Event Effect Qualification of Avionics Systems

NASA Technical Reports Server (NTRS)

Rojdev, Kristina; Koontz, Steve; Atwell, William; Boeder, Paul

2014-01-01

NASA's future missions are focused on long-duration deep space missions for human exploration which offers no options for a quick emergency return to Earth. The combination of long mission duration with no quick emergency return option leads to unprecedented spacecraft system safety and reliability requirements. It is important that spacecraft avionics systems for human deep space missions are not susceptible to Single Event Effect (SEE) failures caused by space radiation (primarily the continuous galactic cosmic ray background and the occasional solar particle event) interactions with electronic components and systems. SEE effects are typically managed during the design, development, and test (DD&T) phase of spacecraft development by using heritage hardware (if possible) and through extensive component level testing, followed by system level failure analysis tasks that are both time consuming and costly. The ultimate product of the SEE DD&T program is a prediction of spacecraft avionics reliability in the flight environment produced using various nuclear reaction and transport codes in combination with the component and subsystem level radiation test data. Previous work by Koontz, et al.1 utilized FLUKA, a Monte Carlo nuclear reaction and transport code, to calculate SEE and single event upset (SEU) rates. This code was then validated against in-flight data for a variety of spacecraft and space flight environments. However, FLUKA has a long run-time (on the order of days). CREME962, an easy to use deterministic code offering short run times, was also compared with FLUKA predictions and in-flight data. CREME96, though fast and easy to use, has not been updated in several years and underestimates secondary particle shower effects in spacecraft structural shielding mass. Thus, this paper will investigate the use of HZETRN 20103, a fast and easy to use deterministic transport code, similar to CREME96, that was developed at NASA Langley Research Center primarily for flight crew ionizing radiation dose assessments. HZETRN 2010 includes updates to address secondary particle shower effects more accurately, and might be used as another tool to verify spacecraft avionics system reliability in space flight SEE environments.

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.

Methods for nuclear air-cleaning-system accident-consequence assessment

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

Andrae, R.W.; Bolstad, J.W.; Gregory, W.S.

1982-01-01

This paper describes a multilaboratory research program that is directed toward addressing many questions that analysts face when performing air cleaning accident consequence assessments. The program involves developing analytical tools and supportive experimental data that will be useful in making more realistic assessments of accident source terms within and up to the atmospheric boundaries of nuclear fuel cycle facilities. The types of accidents considered in this study includes fires, explosions, spills, tornadoes, criticalities, and equipment failures. The main focus of the program is developing an accident analysis handbook (AAH). We will describe the contents of the AAH, which include descriptionsmore » of selected nuclear fuel cycle facilities, process unit operations, source-term development, and accident consequence analyses. Three computer codes designed to predict gas and material propagation through facility air cleaning systems are described. These computer codes address accidents involving fires (FIRAC), explosions (EXPAC), and tornadoes (TORAC). The handbook relies on many illustrative examples to show the analyst how to approach accident consequence assessments. We will use the FIRAC code and a hypothetical fire scenario to illustrate the accident analysis capability.« less

Thermal and thermomechanical calculations of deep-rock nuclear waste disposal with the enhanced SANGRE code

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

Heuze, F.E.

1983-03-01

An attempt to model the complex thermal and mechanical phenomena occurring in the disposal of high-level nuclear wastes in rock at high power loading is described. Such processes include melting of the rock, convection of the molten material, and very high stressing of the rock mass, leading to new fracturing. Because of the phase changes and the wide temperature ranges considered, realistic models must provide for coupling of the thermal and mechanical calculations, for large deformations, and for steady-state temperature-depenent creep of the rock mass. Explicit representation of convection would be desirable, as would the ability to show fracture developmentmore » and migration of fluids in cracks. Enhancements to SNAGRE consisted of: array modifications to accommodate complex variations of thermal and mechanical properties with temperature; introduction of the ability of calculate thermally induced stresses; improved management of the minimum time step and minimum temperature step to increase code efficiency; introduction of a variable heat-generation algorithm to accommodate heat decay of the nuclear materials; streamlining of the code by general editing and extensive deletion of coding used in mesh generation; and updating of the program users' manual. The enhanced LLNL version of the code was renamed LSANGRE. Phase changes were handled by introducing sharp variations in the specific heat of the rock in a narrow range about the melting point. The accuracy of this procedure was tested successfully on a melting slab problem. LSANGRE replicated the results of both the analytical solution and calculations with the finite difference TRUMP code. Following enhancement and verification, a purely thermal calculation was carried to 105 years. It went beyond the extent of maximum melt and into the beginning of the cooling phase.« less

16th International Conference on Nuclear Structure: NS2016

DOE PAGES

Galindo-Uribarri, Alfredo

2016-10-28

Every two years the Nuclear Structure (NS) conference series brings together researchers from an international community of experimental and theoretical nuclear physicists to present and discuss their latest results in nuclear structure. This biennial conference covered the latest results on experimental and theoretical research into the structure of nuclei at the extremes of isospin, excitation energy, mass, and angular momentum. Topics included many of the most exciting areas of modern nuclear structure research such as transitional behavior, nuclear structure and its evolution across the nuclear landscape, shell structure, collectivity, nuclear structure with radioactive beams, and macroscopic and microscopic approaches tomore » nuclear structure.« less

16th International Conference on Nuclear Structure: NS2016

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

Galindo-Uribarri, Alfredo

Every two years the Nuclear Structure (NS) conference series brings together researchers from an international community of experimental and theoretical nuclear physicists to present and discuss their latest results in nuclear structure. This biennial conference covered the latest results on experimental and theoretical research into the structure of nuclei at the extremes of isospin, excitation energy, mass, and angular momentum. Topics included many of the most exciting areas of modern nuclear structure research such as transitional behavior, nuclear structure and its evolution across the nuclear landscape, shell structure, collectivity, nuclear structure with radioactive beams, and macroscopic and microscopic approaches tomore » nuclear structure.« less

Monte Carlo Modeling of the Initial Radiation Emitted by a Nuclear Device in the National Capital Region

DTIC Science & Technology

2013-07-01

also simulated in the models. Data was derived from calculations using the three-dimensional Monte Carlo radiation transport code MCNP (Monte Carlo N...32  B.  MCNP PHYSICS OPTIONS ......................................................................................... 33  C.  HAZUS...input deck’) for the MCNP , Monte Carlo N-Particle, radiation transport code. MCNP is a general-purpose code designed to simulate neutron, photon

Integrated system for production of neutronics and photonics calculational constants. Program SIGMA1 (Version 77-1): Doppler broaden evaluated cross sections in the Evaluated Nuclear Data File/Version B (ENDF/B) format

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

Cullen, D.E.

1977-01-12

A code, SIGMA1, has been designed to Doppler broaden evaluated cross sections in the ENDF/B format. The code can only be applied to tabulated data that vary linearly in energy and cross section between tabulated points. This report describes the methods used in the code and serves as a user's guide to the code.

Safety engineering: KTA code of practice. Lifting mechanisms in nuclear plant

NASA Astrophysics Data System (ADS)

Lifting mechanisms safety requirements are discussed in accordance with the present state of development of science and engineering for the protection of life, health, and assets against the dangers of nuclear energy and the ill effects of ionizing radiation.

Acoustic emission non-destructive testing of structures using source location techniques.

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

Beattie, Alan G.

2013-09-01

The technology of acoustic emission (AE) testing has been advanced and used at Sandia for the past 40 years. AE has been used on structures including pressure vessels, fire bottles, wind turbines, gas wells, nuclear weapons, and solar collectors. This monograph begins with background topics in acoustics and instrumentation and then focuses on current acoustic emission technology. It covers the overall design and system setups for a test, with a wind turbine blade as the object. Test analysis is discussed with an emphasis on source location. Three test examples are presented, two on experimental wind turbine blades and one onmore » aircraft fire extinguisher bottles. Finally, the code for a FORTRAN source location program is given as an example of a working analysis program. Throughout the document, the stress is on actual testing of real structures, not on laboratory experiments.« less

The Dilemmas of Developing an Indigenous Advanced Arms Industry for Developing Countries: The Case of India and China

DTIC Science & Technology

2006-12-01

of providing nuclear power. Once you have the nuclear weapons, they require a delivery system resulting in a missile program. It is afforded higher...out that some domestic advancements may be made in certain sectors, such as nuclear bombs and missiles, because resources may be spent on narrowly...capital, fighter, aviation, nuclear weapons, missiles 16. PRICE CODE 17. SECURITY CLASSIFICATION OF REPORT Unclassified 18. SECURITY CLASSIFICATION

15 CFR 740.7 - Computers (APP).

Code of Federal Regulations, 2010 CFR

2010-01-01

... 4A003. (2) Technology and software. License Exception APP authorizes exports of technology and software... programmability. (ii) Technology and source code. Technology and source code eligible for License Exception APP..., reexports and transfers (in-country) for nuclear, chemical, biological, or missile end-users and end-uses...

Nuclear envelopathies: a complex LINC between nuclear envelope and pathology.

PubMed

Janin, Alexandre; Bauer, Delphine; Ratti, Francesca; Millat, Gilles; Méjat, Alexandre

2017-08-30

Since the identification of the first disease causing mutation in the gene coding for emerin, a transmembrane protein of the inner nuclear membrane, hundreds of mutations and variants have been found in genes encoding for nuclear envelope components. These proteins can be part of the inner nuclear membrane (INM), such as emerin or SUN proteins, outer nuclear membrane (ONM), such as Nesprins, or the nuclear lamina, such as lamins A and C. However, they physically interact with each other to insure the nuclear envelope integrity and mediate the interactions of the nuclear envelope with both the genome, on the inner side, and the cytoskeleton, on the outer side. The core of this complex, called LINC (LInker of Nucleoskeleton to Cytoskeleton) is composed of KASH and SUN homology domain proteins. SUN proteins are INM proteins which interact with lamins by their N-terminal domain and with the KASH domain of nesprins located in the ONM by their C-terminal domain.Although most of these proteins are ubiquitously expressed, their mutations have been associated with a large number of clinically unrelated pathologies affecting specific tissues. Moreover, variants in SUN proteins have been found to modulate the severity of diseases induced by mutations in other LINC components or interactors. For these reasons, the diagnosis and the identification of the molecular explanation of "nuclear envelopathies" is currently challenging.The aim of this review is to summarize the human diseases caused by mutations in genes coding for INM proteins, nuclear lamina, and ONM proteins, and to discuss their potential physiopathological mechanisms that could explain the large spectrum of observed symptoms.

New developments and prospects on COSI, the simulation software for fuel cycle analysis

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

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis (LDRD 17-0958) FY17 Report

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

Wheeler, Timothy A.; Denman, Matthew R.; Williams, R. A.

Instrumentation and control of nuclear power is transforming from analog to modern digital assets. These control systems perform key safety and security functions. This transformation is occurring in new plant designs as well as in the existing fleet of plants as the operation of those plants is extended to 60 years. This transformation introduces new and unknown issues involving both digital asset induced safety issues and security issues. Traditional nuclear power risk assessment tools and cyber security assessment methods have not been modified or developed to address the unique nature of cyber failure modes and of cyber security threat vulnerabilities.more » iii This Lab-Directed Research and Development project has developed a dynamic cyber-risk in- formed tool to facilitate the analysis of unique cyber failure modes and the time sequencing of cyber faults, both malicious and non-malicious, and impose those cyber exploits and cyber faults onto a nuclear power plant accident sequence simulator code to assess how cyber exploits and cyber faults could interact with a plants digital instrumentation and control (DI&C) system and defeat or circumvent a plants cyber security controls. This was achieved by coupling an existing Sandia National Laboratories nuclear accident dynamic simulator code with a cyber emulytics code to demonstrate real-time simulation of cyber exploits and their impact on automatic DI&C responses. Studying such potential time-sequenced cyber-attacks and their risks (i.e., the associated impact and the associated degree of difficulty to achieve the attack vector) on accident management establishes a technical risk informed framework for developing effective cyber security controls for nuclear power.« 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.

Nuclear export of RNA: Different sizes, shapes and functions.

PubMed

Williams, Tobias; Ngo, Linh H; Wickramasinghe, Vihandha O

2018-03-01

Export of protein-coding and non-coding RNA molecules from the nucleus to the cytoplasm is critical for gene expression. This necessitates the continuous transport of RNA species of different size, shape and function through nuclear pore complexes via export receptors and adaptor proteins. Here, we provide an overview of the major RNA export pathways in humans, highlighting the similarities and differences between each. Its importance is underscored by the growing appreciation that deregulation of RNA export pathways is associated with human diseases like cancer. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

Extension of the BRYNTRN code to monoenergetic light ion beams

NASA Technical Reports Server (NTRS)

Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.

1994-01-01

A monoenergetic version of the BRYNTRN transport code is extended to beam transport of light ions (H-2, H-3, He-3, and He-4) in shielding materials (thick targets). The redistribution of energy in nuclear reactions is included in transport solutions that use nuclear fragmentation models. We also consider an equilibrium target-fragment spectrum for nuclei with mass number greater than four to include target fragmentation effects in the linear energy transfer (LET) spectrum. Illustrative results for water and aluminum shielding, including energy and LET spectra, are discussed for high-energy beams of H-2 and He-4.

EMPIRE: Nuclear Reaction Model Code System for Data Evaluation

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

Herman, M.; Capote, R.; Carlson, B.V.

EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including heavy-ions) or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions ({approx} keV) and goes up to several hundred MeV for heavy-ion induced reactions. The code accounts for the major nuclear reaction mechanisms, including direct, pre-equilibrium and compound nucleus ones. Direct reactions are described by a generalized optical model (ECIS03) or by the simplified coupled-channels approachmore » (CCFUS). The pre-equilibrium mechanism can be treated by a deformation dependent multi-step direct (ORION + TRISTAN) model, by a NVWY multi-step compound one or by either a pre-equilibrium exciton model with cluster emission (PCROSS) or by another with full angular momentum coupling (DEGAS). Finally, the compound nucleus decay is described by the full featured Hauser-Feshbach model with {gamma}-cascade and width-fluctuations. Advanced treatment of the fission channel takes into account transmission through a multiple-humped fission barrier with absorption in the wells. The fission probability is derived in the WKB approximation within the optical model of fission. Several options for nuclear level densities include the EMPIRE-specific approach, which accounts for the effects of the dynamic deformation of a fast rotating nucleus, the classical Gilbert-Cameron approach and pre-calculated tables obtained with a microscopic model based on HFB single-particle level schemes with collective enhancement. A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers, moments of inertia and {gamma}-ray strength functions. The results can be converted into ENDF-6 formatted files using the accompanying code EMPEND and completed with neutron resonances extracted from the existing evaluations. The package contains the full EXFOR (CSISRS) library of experimental reaction data that are automatically retrieved during the calculations. Publication quality graphs can be obtained using the powerful and flexible plotting package ZVView. The graphic user interface, written in Tcl/Tk, provides for easy operation of the system. This paper describes the capabilities of the code, outlines physical models and indicates parameter libraries used by EMPIRE to predict reaction cross sections and spectra, mainly for nucleon-induced reactions. Selected applications of EMPIRE are discussed, the most important being an extensive use of the code in evaluations of neutron reactions for the new US library ENDF/B-VII.0. Future extensions of the system are outlined, including neutron resonance module as well as capabilities of generating covariances, using both KALMAN and Monte-Carlo methods, that are still being advanced and refined.« less

Overview of the Graphical User Interface for the GERMcode (GCR Event-Based Risk Model)

NASA Technical Reports Server (NTRS)

Kim, Myung-Hee Y.; 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 GERMcode 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 GERMcode 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 GERMcode 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 GERMcode for application to thick target experiments. The GERMcode 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 GERMcode GUI, as well as providing training applications.

Frameworks for Assessing the Quality of Modeling and Simulation Capabilities

NASA Astrophysics Data System (ADS)

Rider, W. J.

2012-12-01

The importance of assuring quality in modeling and simulation has spawned several frameworks for structuring the examination of quality. The format and content of these frameworks provides an emphasis, completeness and flow to assessment activities. I will examine four frameworks that have been developed and describe how they can be improved and applied to a broader set of high consequence applications. Perhaps the first of these frameworks was known as CSAU [Boyack] (code scaling, applicability and uncertainty) used for nuclear reactor safety and endorsed the United States' Nuclear Regulatory Commission (USNRC). This framework was shaped by nuclear safety practice, and the practical structure needed after the Three Mile Island accident. It incorporated the dominant experimental program, the dominant analysis approach, and concerns about the quality of modeling. The USNRC gave it the force of law that made the nuclear industry take it seriously. After the cessation of nuclear weapons' testing the United States began a program of examining the reliability of these weapons without testing. This program utilizes science including theory, modeling, simulation and experimentation to replace the underground testing. The emphasis on modeling and simulation necessitated attention on the quality of these simulations. Sandia developed the PCMM (predictive capability maturity model) to structure this attention [Oberkampf]. PCMM divides simulation into six core activities to be examined and graded relative to the needs of the modeling activity. NASA [NASA] has built yet another framework in response to the tragedy of the space shuttle accidents. Finally, Ben-Haim and Hemez focus upon modeling robustness and predictive fidelity in another approach. These frameworks are similar, and applied in a similar fashion. The adoption of these frameworks at Sandia and NASA has been slow and arduous because the force of law has not assisted acceptance. All existing frameworks are incomplete and need to be extended incorporating elements from the other as well as new elements related to how models are solved, and how the model will be applied. I will describe this merger of approach and how it should be applied. The problems in adoption are related to basic human nature in that no one likes to be graded, or told they are not sufficiently quality oriented. Rather than engage in an adversarial role, I suggest that the frameworks be viewed as a collaborative tool. Instead these frameworks should be used to structure collaborations that can be used to assist the modeling and simulation efforts to be high quality. The framework provides a comprehensive setting of modeling and simulation themes that should be explored in providing high quality. W. Oberkampf, M. Pilch, and T. Trucano, Predictive Capability Maturity Model for Computational Modeling and Simulation, SAND2007-5948, 2007. B. Boyack, Quantifying Reactor Safety Margins Part 1: An Overview of the Code Scaling, Applicability, and Uncertainty Evaluation Methodology, Nuc. Eng. Design, 119, pp. 1-15, 1990. National Aeronautics and Space Administration, STANDARD FOR MODELS AND SIMULATIONS, NASA-STD-7009, 2008. Y. Ben-Haim and F. Hemez, Robustness, fidelity and prediction-looseness of models, Proc. R. Soc. A (2012) 468, 227-244.

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Extension of the energy range of experimental activation cross-sections data of deuteron induced nuclear reactions on indium up to 50MeV.

PubMed

Tárkányi, F; Ditrói, F; Takács, S; Hermanne, A; Ignatyuk, A V

2015-11-01

The energy range of our earlier measured activation cross-sections data of longer-lived products of deuteron induced nuclear reactions on indium were extended from 40MeV up to 50MeV. The traditional stacked foil irradiation technique and non-destructive gamma spectrometry were used. No experimental data were found in literature for this higher energy range. Experimental cross-sections for the formation of the radionuclides (113,110)Sn, (116m,115m,114m,113m,111,110g,109)In and (115)Cd are reported in the 37-50MeV energy range, for production of (110)Sn and (110g,109)In these are the first measurements ever. The experimental data were compared with the results of cross section calculations of the ALICE and EMPIRE nuclear model codes and of the TALYS 1.6 nuclear model code as listed in the on-line library TENDL-2014. Copyright © 2015 Elsevier Ltd. All rights reserved.

The radiation fields around a proton therapy facility: A comparison of Monte Carlo simulations

NASA Astrophysics Data System (ADS)

Ottaviano, G.; Picardi, L.; Pillon, M.; Ronsivalle, C.; Sandri, S.

2014-02-01

A proton therapy test facility with a beam current lower than 10 nA in average, and an energy up to 150 MeV, is planned to be sited at the Frascati ENEA Research Center, in Italy. The accelerator is composed of a sequence of linear sections. The first one is a commercial 7 MeV proton linac, from which the beam is injected in a SCDTL (Side Coupled Drift Tube Linac) structure reaching the energy of 52 MeV. Then a conventional CCL (coupled Cavity Linac) with side coupling cavities completes the accelerator. The linear structure has the important advantage that the main radiation losses during the acceleration process occur to protons with energy below 20 MeV, with a consequent low production of neutrons and secondary radiation. From the radiation protection point of view the source of radiation for this facility is then almost completely located at the final target. Physical and geometrical models of the device have been developed and implemented into radiation transport computer codes based on the Monte Carlo method. The scope is the assessment of the radiation field around the main source for supporting the safety analysis. For the assessment independent researchers used two different Monte Carlo computer codes named FLUKA (FLUktuierende KAskade) and MCNPX (Monte Carlo N-Particle eXtended) respectively. Both are general purpose tools for calculations of particle transport and interactions with matter, covering an extended range of applications including proton beam analysis. Nevertheless each one utilizes its own nuclear cross section libraries and uses specific physics models for particle types and energies. The models implemented into the codes are described and the results are presented. The differences between the two calculations are reported and discussed pointing out disadvantages and advantages of each code in the specific application.

Radiation Transport Tools for Space Applications: A Review

NASA Technical Reports Server (NTRS)

Jun, Insoo; Evans, Robin; Cherng, Michael; Kang, Shawn

2008-01-01

This slide presentation contains a brief discussion of nuclear transport codes widely used in the space radiation community for shielding and scientific analyses. Seven radiation transport codes that are addressed. The two general methods (i.e., Monte Carlo Method, and the Deterministic Method) are briefly reviewed.

Fatigue Behavior of HY-130 Steel Weldments Containing Fabrication Discontinuities.

DTIC Science & Technology

1985-04-18

discontinuities to solutions for elliptical discontinuities. One such approach has been formalized in the ASME Section XI Boiler and Pressure Vessel Code [1... Boiler and Pressure Vessel Code , Section XI, "Rules for Inservice Inspection of Nuclear Reactor Coolant Systems," American Society of Mechanical

Data Documentation for Navy Civilian Manpower Study,

DTIC Science & Technology

1986-09-01

Engineering 0830 Mechanical Engineer 0840 Nuclear Engineering 0850 Electrical Engineering 0855 Electronics Engineering 0856 Electronics ...OCCUPATIONAL LEVEL (DONOL) CODES DONOL code Title 1060 Engineering Drafting 1061 Electronics Technician w 1062 Engineering Technician 1063 Industrial...Architect 2314 Electrical Engineer 2315 Electronic Engineer 2316 Industrial Engineer 2317 Mechanical Engineer 2318

Implementation of radiation shielding calculation methods. Volume 1: Synopsis of methods and summary of results

NASA Technical Reports Server (NTRS)

Capo, M. A.; Disney, R. K.

1971-01-01

The work performed in the following areas is summarized: (1) Analysis of Realistic nuclear-propelled vehicle was analyzed using the Marshall Space Flight Center computer code package. This code package includes one and two dimensional discrete ordinate transport, point kernel, and single scatter techniques, as well as cross section preparation and data processing codes, (2) Techniques were developed to improve the automated data transfer in the coupled computation method of the computer code package and improve the utilization of this code package on the Univac-1108 computer system. (3) The MSFC master data libraries were updated.

ASTEC—the Aarhus STellar Evolution Code

NASA Astrophysics Data System (ADS)

Christensen-Dalsgaard, Jørgen

2008-08-01

The Aarhus code is the result of a long development, starting in 1974, and still ongoing. A novel feature is the integration of the computation of adiabatic oscillations for specified models as part of the code. It offers substantial flexibility in terms of microphysics and has been carefully tested for the computation of solar models. However, considerable development is still required in the treatment of nuclear reactions, diffusion and convective mixing.

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.

Molecular interplay of pro-inflammatory transcription factors and non-coding RNAs in esophageal squamous cell carcinoma.

PubMed

Sundaram, Gopinath M; Veera Bramhachari, Pallaval

2017-06-01

Esophageal squamous cell carcinoma is the sixth most common cancer in the developing world. The aggressive nature of esophageal squamous cell carcinoma, its tendency for relapse, and the poor survival prospects of patients diagnosed at advanced stages, represent a pressing need for the development of new therapies for this disease. Chronic inflammation is known to have a causal link to cancer pre-disposition. Nuclear factor kappa B and signal transducer and activator of transcription 3 are transcription factors which regulate immunity and inflammation and are emerging as key regulators of tumor initiation, progression, and metastasis. Although these pro-inflammatory factors in esophageal squamous cell carcinoma have been well-characterized with reference to protein-coding targets, their functional interactions with non-coding RNAs have only recently been gaining attention. Non-coding RNAs, especially microRNAs and long non-coding RNAs demonstrate potential as biomarkers and alternative therapeutic targets. In this review, we summarize the recent literature and concepts on non-coding RNAs that are regulated by/regulate nuclear factor kappa B and signal transducer and activator of transcription 3 in esophageal cancer progression. We also discuss how these recent discoveries can pave way for future therapeutic options to treat esophageal squamous cell carcinoma.

Activation Assessment of the Soil Around the ESS Accelerator Tunnel

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

Rakhno, I. L.; Mokhov, N. V.; Tropin, I. S.

Activation of the soil surrounding the ESS accelerator tunnel calculated by the MARS15 code is presented. A detailed composition of the soil, that comprises about 30 different chemical elements, is considered. Spatial distributions of the produced activity are provided in both transverse and longitudinal direction. A realistic irradiation profile for the entire planned lifetime of the facility is used. The nuclear transmutation and decay of the produced radionuclides is calculated with the DeTra code which is a built-in tool for the MARS15 code. Radionuclide production by low-energy neutrons is calculated using the ENDF/B-VII evaluated nuclear data library. In order tomore » estimate quality of this activation assessment, a comparison between calculated and measured activation of various foils in a similar radiation environment is presented.« less

Advanced Technology and Mitigation (ATDM) SPARC Re-Entry Code Fiscal Year 2017 Progress and Accomplishments for ECP.

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

Crozier, Paul; Howard, Micah; Rider, William J.

The SPARC (Sandia Parallel Aerodynamics and Reentry Code) will provide nuclear weapon qualification evidence for the random vibration and thermal environments created by re-entry of a warhead into the earth’s atmosphere. SPARC incorporates the innovative approaches of ATDM projects on several fronts including: effective harnessing of heterogeneous compute nodes using Kokkos, exascale-ready parallel scalability through asynchronous multi-tasking, uncertainty quantification through Sacado integration, implementation of state-of-the-art reentry physics and multiscale models, use of advanced verification and validation methods, and enabling of improved workflows for users. SPARC is being developed primarily for the Department of Energy nuclear weapon program, with additional developmentmore » and use of the code is being supported by the Department of Defense for conventional weapons programs.« less

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

Januszyk, Kurt; Liu, Quansheng; Lima, Christopher D.

The eukaryotic RNA exosome is a highly conserved multi-subunit complex that catalyzes degradation and processing of coding and noncoding RNA. A noncatalytic nine-subunit exosome core interacts with Rrp44 and Rrp6, two subunits that possess processive and distributive 3'-to-5' exoribonuclease activity, respectively. While both Rrp6 and Rrp44 are responsible for RNA processing in budding yeast, Rrp6 may play a more prominent role in processing, as it has been demonstrated to be inhibited by stable RNA secondary structure in vitro and because the null allele in budding yeast leads to the buildup of specific structured RNA substrates. Human RRP6, otherwise known asmore » PM/SCL-100 or EXOSC10, shares sequence similarity to budding yeast Rrp6 and is proposed to catalyze 3'-to-5' exoribonuclease activity on a variety of nuclear transcripts including ribosomal RNA subunits, RNA that has been poly-adenylated by TRAMP, as well as other nuclear RNA transcripts destined for processing and/or destruction. To characterize human RRP6, we expressed the full-length enzyme as well as truncation mutants that retain catalytic activity, compared their activities to analogous constructs for Saccharomyces cerevisiae Rrp6, and determined the X-ray structure of a human construct containing the exoribonuclease and HRDC domains that retains catalytic activity. Structural data show that the human active site is more exposed when compared to the yeast structure, and biochemical data suggest that this feature may play a role in the ability of human RRP6 to productively engage and degrade structured RNA substrates more effectively than the analogous budding yeast enzyme.« less

Lack of genetic structure in the jellyfish Pelagia noctiluca (Cnidaria: Scyphozoa: Semaeostomeae) across European seas.

PubMed

Stopar, Katja; Ramsak, Andreja; Trontelj, Peter; Malej, Alenka

2010-10-01

The genetic structure of the holopelagic scyphozoan Pelagia noctiluca was inferred based on the study of 144 adult medusae. The areas of study were five geographic regions in two European seas (Eastern Atlantic and Mediterranean Sea). A 655-bp sequence of mitochondrial cytochrome c oxidase subunit I (COI), and a 645-bp sequence of two nuclear internal transcribed spacers (ITS1 and ITS2) were analyzed. The protein coding COI gene showed a higher level of divergence than the combined nuclear ITS fragment (haplotype diversity 0.962 vs. 0.723, nucleotide diversity 1.16% vs. 0.31%). Phylogeographic analysis on COI gene revealed two clades, the larger consisting of specimens from all sampling sites, and the smaller mostly formed of specimens from the Mediterranean Sea. Haplotype diversity was very high throughout the sampled area, and within sample diversity was higher than diversity among geographical regions. No strongly supported genetically or geographically distinct groups of P. noctiluca were found. The results - long distance dispersal, insignificant F(ST) values, lack of isolation by distance - pointed toward an admixture among Mediterranean and East Atlantic populations. Copyright 2010 Elsevier Inc. All rights reserved.

Independent Qualification of the CIAU Tool Based on the Uncertainty Estimate in the Prediction of Angra 1 NPP Inadvertent Load Rejection Transient

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

Borges, Ronaldo C.; D'Auria, Francesco; Alvim, Antonio Carlos M.

2002-07-01

The Code with - the capability of - Internal Assessment of Uncertainty (CIAU) is a tool proposed by the 'Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP)' of the University of Pisa. Other Institutions including the nuclear regulatory body from Brazil, 'Comissao Nacional de Energia Nuclear', contributed to the development of the tool. The CIAU aims at providing the currently available Relap5/Mod3.2 system code with the integrated capability of performing not only relevant transient calculations but also the related estimates of uncertainty bands. The Uncertainty Methodology based on Accuracy Extrapolation (UMAE) is used to characterize the uncertainty in themore » prediction of system code calculations for light water reactors and is internally coupled with the above system code. Following an overview of the CIAU development, the present paper deals with the independent qualification of the tool. The qualification test is performed by estimating the uncertainty bands that should envelope the prediction of the Angra 1 NPP transient RES-11. 99 originated by an inadvertent complete load rejection that caused the reactor scram when the unit was operating at 99% of nominal power. The current limitation of the 'error' database, implemented into the CIAU prevented a final demonstration of the qualification. However, all the steps for the qualification process are demonstrated. (authors)« less

Reliability assessment of MVP-BURN and JENDL-4.0 related to nuclear transmutation of light platinum group elements

NASA Astrophysics Data System (ADS)

Terashima, Atsunori; Nilsson, Mikael; Ozawa, Masaki; Chiba, Satoshi

2017-09-01

The Aprés ORIENT research program, as a concept of advanced nuclear fuel cycle, was initiated in FY2011 aiming at creating stable, highly-valuable elements by nuclear transmutation from ↓ssion products. In order to simulate creation of such elements by (n, γ) reaction succeeded by β- decay in reactors, a continuous-energy Monte Carlo burnup calculation code MVP-BURN was employed. Then, it is one of the most important tasks to con↓rm the reliability of MVP-BURN code and evaluated neutron cross section library. In this study, both an experiment of neutron activation analysis in TRIGA Mark I reactor at University of California, Irvine and the corresponding burnup calculation using MVP-BURN code were performed for validation of the simulation on transmutation of light platinum group elements. Especially, some neutron capture reactions such as 102Ru(n, γ)103Ru, 104Ru(n, γ)105Ru, and 108Pd(n, γ)109Pd were dealt with in this study. From a comparison between the calculation (C) and the experiment (E) about 102Ru(n, γ)103Ru, the deviation (C/E-1) was signi↓cantly large. Then, it is strongly suspected that not MVP-BURN code but the neutron capture cross section of 102Ru belonging to JENDL-4.0 used in this simulation have made the big di↑erence as (C/E-1) >20%.

Update and evaluation of decay data for spent nuclear fuel analyses

NASA Astrophysics Data System (ADS)

Simeonov, Teodosi; Wemple, Charles

2017-09-01

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

A Dynamic/Anisotropic Low Earth Orbit (LEO) Ionizing Radiation Model

NASA Technical Reports Server (NTRS)

Badavi, Francis F.; West, Katie J.; Nealy, John E.; Wilson, John W.; Abrahms, Briana L.; Luetke, Nathan J.

2006-01-01

The International Space Station (ISS) provides the proving ground for future long duration human activities in space. Ionizing radiation measurements in ISS form the ideal tool for the experimental validation of ionizing radiation environmental models, nuclear transport code algorithms, and nuclear reaction cross sections. Indeed, prior measurements on the Space Transportation System (STS; Shuttle) have provided vital information impacting both the environmental models and the nuclear transport code development by requiring dynamic models of the Low Earth Orbit (LEO) environment. Previous studies using Computer Aided Design (CAD) models of the evolving ISS configurations with Thermo Luminescent Detector (TLD) area monitors, demonstrated that computational dosimetry requires environmental models with accurate non-isotropic as well as dynamic behavior, detailed information on rack loading, and an accurate 6 degree of freedom (DOF) description of ISS trajectory and orientation.

Long Non-Coding RNAs in Haematological Malignancies

PubMed Central

Garitano-Trojaola, Andoni; Agirre, Xabier; Prósper, Felipe; Fortes, Puri

2013-01-01

Long non-coding RNAs (lncRNAs) are functional RNAs longer than 200 nucleotides in length. LncRNAs are as diverse as mRNAs and they normally share the same biosynthetic machinery based on RNA polymerase II, splicing and polyadenylation. However, lncRNAs have low coding potential. Compared to mRNAs, lncRNAs are preferentially nuclear, more tissue specific and expressed at lower levels. Most of the lncRNAs described to date modulate the expression of specific genes by guiding chromatin remodelling factors; inducing chromosomal loopings; affecting transcription, splicing, translation or mRNA stability; or serving as scaffolds for the organization of cellular structures. They can function in cis, cotranscriptionally, or in trans, acting as decoys, scaffolds or guides. These functions seem essential to allow cell differentiation and growth. In fact, many lncRNAs have been shown to exert oncogenic or tumor suppressor properties in several cancers including haematological malignancies. In this review, we summarize what is known about lncRNAs, the mechanisms for their regulation in cancer and their role in leukemogenesis, lymphomagenesis and hematopoiesis. Furthermore, we discuss the potential of lncRNAs in diagnosis, prognosis and therapy in cancer, with special attention to haematological malignancies. PMID:23887658

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

NASA Astrophysics Data System (ADS)

Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

2012-10-01

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

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

Gougar, Hans

This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s majormore » emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code with high fidelity simulations that would allow investigation of multi-dimensional, multi-phase containment phenomena that are only treated approximately in established codes.« less

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

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...

Criticality Calculations with MCNP6 - Practical Lectures

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

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2016-11-29

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input modelmore » for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.« less

NSR&D Program Fiscal Year 2015 Funded Research Stochastic Modeling of Radioactive Material Releases Final Report

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

Andrus, Jason P.; Pope, Chad; Toston, Mary

2016-12-01

Nonreactor nuclear facilities operating under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculatesmore » the radiation dose distribution associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. Users can also specify custom distributions through a user defined distribution option. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA, developed using the MATLAB coding framework, has a graphical user interface and can be installed on both Windows and Mac computers. SODA is a standalone software application and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC; rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The SODA development project was funded through a grant from the DOE Nuclear Safety Research and Development Program.« less

NSR&D Program Fiscal Year 2015 Funded Research Stochastic Modeling of Radioactive Material Releases Final Report

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

Andrus, Jason P.; Pope, Chad; Toston, Mary

Nonreactor nuclear facilities operating under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculatesmore » the radiation dose distribution associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. Users can also specify custom distributions through a user defined distribution option. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA, developed using the MATLAB coding framework, has a graphical user interface and can be installed on both Windows and Mac computers. SODA is a standalone software application and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC; rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The SODA development project was funded through a grant from the DOE Nuclear Safety Research and Development Program.« less

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE PAGES

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson; ...

2018-06-14

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

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

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

SIGACE Code for Generating High-Temperature ACE Files; Validation and Benchmarking

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

Sharma, Amit R.; Ganesan, S.; Trkov, A.

2005-05-24

A code named SIGACE has been developed as a tool for MCNP users within the scope of a research contract awarded by the Nuclear Data Section of the International Atomic Energy Agency (IAEA) (Ref: 302-F4-IND-11566 B5-IND-29641). A new recipe has been evolved for generating high-temperature ACE files for use with the MCNP code. Under this scheme the low-temperature ACE file is first converted to an ENDF formatted file using the ACELST code and then Doppler broadened, essentially limited to the data in the resolved resonance region, to any desired higher temperature using SIGMA1. The SIGACE code then generates a high-temperaturemore » ACE file for use with the MCNP code. A thinning routine has also been introduced in the SIGACE code for reducing the size of the ACE files. The SIGACE code and the recipe for generating ACE files at higher temperatures has been applied to the SEFOR fast reactor benchmark problem (sodium-cooled fast reactor benchmark described in ENDF-202/BNL-19302, 1974 document). The calculated Doppler coefficient is in good agreement with the experimental value. A similar calculation using ACE files generated directly with the NJOY system also agrees with our SIGACE computed results. The SIGACE code and the recipe is further applied to study the numerical benchmark configuration of selected idealized PWR pin cell configurations with five different fuel enrichments as reported by Mosteller and Eisenhart. The SIGACE code that has been tested with several FENDL/MC files will be available, free of cost, upon request, from the Nuclear Data Section of the IAEA.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

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

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

Resonance Parameter Adjustment Based on Integral Experiments

DOE PAGES

Sobes, Vladimir; Leal, Luiz; Arbanas, Goran; ...

2016-06-02

Our project seeks to allow coupling of differential and integral data evaluation in a continuous-energy framework and to use the generalized linear least-squares (GLLS) methodology in the TSURFER module of the SCALE code package to update the parameters of a resolved resonance region evaluation. We recognize that the GLLS methodology in TSURFER is identical to the mathematical description of a Bayesian update in SAMMY, the SAMINT code was created to use the mathematical machinery of SAMMY to update resolved resonance parameters based on integral data. Traditionally, SAMMY used differential experimental data to adjust nuclear data parameters. Integral experimental data, suchmore » as in the International Criticality Safety Benchmark Experiments Project, remain a tool for validation of completed nuclear data evaluations. SAMINT extracts information from integral benchmarks to aid the nuclear data evaluation process. Later, integral data can be used to resolve any remaining ambiguity between differential data sets, highlight troublesome energy regions, determine key nuclear data parameters for integral benchmark calculations, and improve the nuclear data covariance matrix evaluation. Moreover, SAMINT is not intended to bias nuclear data toward specific integral experiments but should be used to supplement the evaluation of differential experimental data. Using GLLS ensures proper weight is given to the differential data.« less

EPRI Guide to Managing Nuclear Utility Protective Clothing Programs. PCEVAL User`s Manual, A computer code for evaluating the economics of nuclear plant protective clothing programs: Final report

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

Kelly, J.J.; Kelly, D.M.

1993-10-01

The Electric Power Research Institute (EPRI) commissioned a radioactive waste related project (RP2414-34) in 1989 to produce a guide for developing and managing nuclear plant protective clothing programs. Every nuclear facility must coordinate some type of protective clothing program for its radiation workers to ensure proper and safe protection for the wearer and to maintain control over the spread of contamination. Yet, every nuclear facility has developed its own unique program for managing such clothing. Accordingly, a need existed for a reference guide to assist with standardizing protective clothing programs and in controlling the potentially escalating economics of such programs.more » The initial Guide to Managing Nuclear Utility Protective Clothing Programs, NP-7309, was published in May 1991. Since that time, a number of utilities have reviewed and/or used the report to enhance their protective clothing programs. Some of these utilities requested that a computer program be developed to assist utilities in evaluating the economics of protective clothing programs consistent with the guidance in NP-7309. The PCEVAL computer code responds to that industry need. This report, the PCEVAL User`s Manual, provides detailed instruction on use of the software.« less

Integrated system for production of neutronics and photonics calculational constants. Volume 17, Part B, Rev. 1. Program SIGMA 1 (Version 78-1): Doppler broadened evaluated cross sections in the evaluated nuclear data file/Version B (ENDF/B) format. [For CDC-7600

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

Cullen, D.E.

1978-07-04

The code SIGMA1 Doppler broadens evaluated cross sections in the ENDF/B format. The code can be applied only to data that vary as a linear function of energy and cross section between tabulated points. This report describes the methods used in the code and serves as a user's guide to the code. 6 figures, 2 tables.

Evaluation of the DRAGON code for VHTR design analysis.

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

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-12

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by themore » IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.« less

Impacts of DNAPL Source Treatment: Experimental and Modeling Assessment of the Benefits of Partial DNAPL Source Removal

DTIC Science & Technology

2009-09-01

nuclear industry for conducting performance assessment calculations. The analytical FORTRAN code for the DNAPL source function, REMChlor, was...project. The first was to apply existing deterministic codes , such as T2VOC and UTCHEM, to the DNAPL source zone to simulate the remediation processes...but describe the spatial variability of source zones unlike one-dimensional flow and transport codes that assume homogeneity. The Lagrangian models

Features of cues and processes during chloroplast-mediated retrograde signaling in the alga Chlamydomonas

USDA-ARS?s Scientific Manuscript database

Retrograde signalling is a selective process defined by cues generated in chloroplast/mitochondria which traverse membranes and end up regulating nuclear gene expression and protein synthesis. The coding and encoding of organellar message(s) that alter nuclear gene expression and/or cellular metabo...

Quantification of the Spatial Organization of the Nuclear Lamina as a Tool for Cell Classification

PubMed Central

Righolt, Christiaan H.; Zatreanu, Diana A.; Raz, Vered

2013-01-01

The nuclear lamina is the structural scaffold of the nuclear envelope that plays multiple regulatory roles in chromatin organization and gene expression as well as a structural role in nuclear stability. The lamina proteins, also referred to as lamins, determine nuclear lamina organization and define the nuclear shape and the structural integrity of the cell nucleus. In addition, lamins are connected with both nuclear and cytoplasmic structures forming a dynamic cellular structure whose shape changes upon external and internal signals. When bound to the nuclear lamina, the lamins are mobile, have an impact on the nuclear envelop structure, and may induce changes in their regulatory functions. Changes in the nuclear lamina shape cause changes in cellular functions. A quantitative description of these structural changes could provide an unbiased description of changes in cellular function. In this review, we describe how changes in the nuclear lamina can be measured from three-dimensional images of lamins at the nuclear envelope, and we discuss how structural changes of the nuclear lamina can be used for cell classification. PMID:27335676

Quantification of the Spatial Organization of the Nuclear Lamina as a Tool for Cell Classification.

PubMed

Righolt, Christiaan H; Zatreanu, Diana A; Raz, Vered

2013-01-01

The nuclear lamina is the structural scaffold of the nuclear envelope that plays multiple regulatory roles in chromatin organization and gene expression as well as a structural role in nuclear stability. The lamina proteins, also referred to as lamins, determine nuclear lamina organization and define the nuclear shape and the structural integrity of the cell nucleus. In addition, lamins are connected with both nuclear and cytoplasmic structures forming a dynamic cellular structure whose shape changes upon external and internal signals. When bound to the nuclear lamina, the lamins are mobile, have an impact on the nuclear envelop structure, and may induce changes in their regulatory functions. Changes in the nuclear lamina shape cause changes in cellular functions. A quantitative description of these structural changes could provide an unbiased description of changes in cellular function. In this review, we describe how changes in the nuclear lamina can be measured from three-dimensional images of lamins at the nuclear envelope, and we discuss how structural changes of the nuclear lamina can be used for cell classification.

Understanding Lustre Internals

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

Wang, Feiyi; Oral, H Sarp; Shipman, Galen M

2009-04-01

Lustre was initiated and funded, almost a decade ago, by the U.S. Department of Energy (DoE) Office of Science and National Nuclear Security Administration laboratories to address the need for an open source, highly-scalable, high-performance parallel filesystem on by then present and future supercomputing platforms. Throughout the last decade, it was deployed over numerous medium-to-large-scale supercomputing platforms and clusters, and it performed and met the expectations of the Lustre user community. As it stands at the time of writing this document, according to the Top500 list, 15 of the top 30 supercomputers in the world use Lustre filesystem. This reportmore » aims to present a streamlined overview on how Lustre works internally at reasonable details including relevant data structures, APIs, protocols and algorithms involved for Lustre version 1.6 source code base. More importantly, it tries to explain how various components interconnect with each other and function as a system. Portions of this report are based on discussions with Oak Ridge National Laboratory Lustre Center of Excellence team members and portions of it are based on our own understanding of how the code works. We, as the authors team bare all responsibilities for all errors and omissions in this document. We can only hope it helps current and future Lustre users and Lustre code developers as much as it helped us understanding the Lustre source code and its internal workings.« less

ANITA-IEAF activation code package - updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

NASA Astrophysics Data System (ADS)

Frisoni, Manuela

2017-09-01

ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

Theoretical modeling of laser-induced plasmas using the ATOMIC code

NASA Astrophysics Data System (ADS)

Colgan, James; Johns, Heather; Kilcrease, David; Judge, Elizabeth; Barefield, James, II; Clegg, Samuel; Hartig, Kyle

2014-10-01

We report on efforts to model the emission spectra generated from laser-induced breakdown spectroscopy (LIBS). LIBS is a popular and powerful method of quickly and accurately characterizing unknown samples in a remote manner. In particular, LIBS is utilized by the ChemCam instrument on the Mars Science Laboratory. We model the LIBS plasma using the Los Alamos suite of atomic physics codes. Since LIBS plasmas generally have temperatures of somewhere between 3000 K and 12000 K, the emission spectra typically result from the neutral and singly ionized stages of the target atoms. We use the Los Alamos atomic structure and collision codes to generate sets of atomic data and use the plasma kinetics code ATOMIC to perform LTE or non-LTE calculations that generate level populations and an emission spectrum for the element of interest. In this presentation we compare the emission spectrum from ATOMIC with an Fe LIBS laboratory-generated plasma as well as spectra from the ChemCam instrument. We also discuss various physics aspects of the modeling of LIBS plasmas that are necessary for accurate characterization of the plasma, such as multi-element target composition effects, radiation transport effects, and accurate line shape treatments. The Los Alamos National Laboratory is operated by Los Alamos National Security, LLC for the National Nuclear Security Administration of the U.S. Department of Energy under Contract No. DE-AC5206NA25396.

Overview of Particle and Heavy Ion Transport Code System PHITS

NASA Astrophysics Data System (ADS)

Sato, Tatsuhiko; Niita, Koji; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Furuta, Takuya; Noda, Shusaku; Ogawa, Tatsuhiko; Iwase, Hiroshi; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Sihver, Lembit

2014-06-01

A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.

Stochastic analog neutron transport with TRIPOLI-4 and FREYA: Bayesian uncertainty quantification for neutron multiplicity counting

DOE PAGES

Verbeke, J. M.; Petit, O.

2016-06-01

From nuclear safeguards to homeland security applications, the need for the better modeling of nuclear interactions has grown over the past decades. Current Monte Carlo radiation transport codes compute average quantities with great accuracy and performance; however, performance and averaging come at the price of limited interaction-by-interaction modeling. These codes often lack the capability of modeling interactions exactly: for a given collision, energy is not conserved, energies of emitted particles are uncorrelated, and multiplicities of prompt fission neutrons and photons are uncorrelated. Many modern applications require more exclusive quantities than averages, such as the fluctuations in certain observables (e.g., themore » neutron multiplicity) and correlations between neutrons and photons. In an effort to meet this need, the radiation transport Monte Carlo code TRIPOLI-4® was modified to provide a specific mode that models nuclear interactions in a full analog way, replicating as much as possible the underlying physical process. Furthermore, the computational model FREYA (Fission Reaction Event Yield Algorithm) was coupled with TRIPOLI-4 to model complete fission events. As a result, FREYA automatically includes fluctuations as well as correlations resulting from conservation of energy and momentum.« less

Group I introns are inherited through common ancestry in the nuclear-encoded rRNA of Zygnematales (Charophyceae).

PubMed Central

Bhattacharya, D; Surek, B; Rüsing, M; Damberger, S; Melkonian, M

1994-01-01

Group I introns are found in organellar genomes, in the genomes of eubacteria and phages, and in nuclear-encoded rRNAs. The origin and distribution of nuclear-encoded rRNA group I introns are not understood. To elucidate their evolutionary relationships, we analyzed diverse nuclear-encoded small-subunit rRNA group I introns including nine sequences from the green-algal order Zygnematales (Charophyceae). Phylogenetic analyses of group I introns and rRNA coding regions suggest that lateral transfers have occurred in the evolutionary history of group I introns and that, after transfer, some of these elements may form stable components of the host-cell nuclear genomes. The Zygnematales introns, which share a common insertion site (position 1506 relative to the Escherichia coli small-subunit rRNA), form one subfamily of group I introns that has, after its origin, been inherited through common ancestry. Since the first Zygnematales appear in the middle Devonian within the fossil record, the "1506" group I intron presumably has been a stable component of the Zygnematales small-subunit rRNA coding region for 350-400 million years. PMID:7937917

Overview and evaluation of different nuclear level density models for the 123I radionuclide production.

PubMed

Nikjou, A; Sadeghi, M

2018-06-01

The 123 I radionuclide (T 1/2 = 13.22 h, β+ = 100%) is one of the most potent gamma emitters for nuclear medicine. In this study, the cyclotron production of this radionuclide via different nuclear reactions namely, the 121 Sb(α,2n), 122 Te(d,n), 123 Te(p,n), 124 Te(p,2n), 124 Xe(p,2n), 127 I(p,5n) and 127 I(d,6n) were investigated. The effect of the various phenomenological nuclear level density models such as Fermi gas model (FGM), Back-shifted Fermi gas model (BSFGM), Generalized superfluid model (GSM) and Enhanced generalized superfluid model (EGSM) moreover, the three microscopic level density models were evaluated for predicting of cross sections and production yield predictions. The SRIM code was used to obtain the target thickness. The 123 I excitation function of reactions were calculated by using of the TALYS-1.8, EMPIRE-3.2 nuclear codes and with data which taken from TENDL-2015 database, and finally the theoretical calculations were compared with reported experimental measurements in which taken from EXFOR database. Copyright © 2018 Elsevier Ltd. All rights reserved.

{sup 45}Sc Solid State NMR studies of the silicides ScTSi (T=Co, Ni, Cu, Ru, Rh, Pd, Ir, Pt)

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

Harmening, Thomas; Eckert, Hellmut, E-mail: eckerth@uni-muenster.de; Fehse, Constanze M.

The silicides ScTSi (T=Fe, Co, Ni, Cu, Ru, Rh, Pd, Ir, Pt) were synthesized by arc-melting and characterized by X-ray powder diffraction. The structures of ScCoSi, ScRuSi, ScPdSi, and ScIrSi were refined from single crystal diffractometer data. These silicides crystallize with the TiNiSi type, space group Pnma. No systematic influences of the {sup 45}Sc isotropic magnetic shift and nuclear electric quadrupolar coupling parameters on various structural distortion parameters calculated from the crystal structure data can be detected. {sup 45}Sc MAS-NMR data suggest systematic trends in the local electronic structure probed by the scandium atoms: both the electric field gradients andmore » the isotropic magnetic shifts relative to a 0.2 M aqueous Sc(NO{sub 3}){sub 3} solution decrease with increasing valence electron concentration and within each T group the isotropic magnetic shift decreases monotonically with increasing atomic number. The {sup 45}Sc nuclear electric quadrupolar coupling constants are generally well reproduced by quantum mechanical electric field gradient calculations using the WIEN2k code. Highlights: Black-Right-Pointing-Pointer Arc-melting synthesis of silicides ScTSi. Black-Right-Pointing-Pointer Single crystal X-ray data of ScCoSi, ScRuSi, ScPdSi, and ScIrSi. Black-Right-Pointing-Pointer {sup 45}Sc solid state NMR of silicides ScTSi.« less

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

The RIB production target for the SPES project

NASA Astrophysics Data System (ADS)

Monetti, Alberto; Andrighetto, Alberto; Petrovich, Carlo; Manzolaro, Mattia; Corradetti, Stefano; Scarpa, Daniele; Rossetto, Francesco; Martinez Dominguez, Fernando; Vasquez, Jesus; Rossignoli, Massimo; Calderolla, Michele; Silingardi, Roberto; Mozzi, Aldo; Borgna, Francesca; Vivian, Gianluca; Boratto, Enrico; Ballan, Michele; Prete, Gianfranco; Meneghetti, Giovanni

2015-10-01

Facilities making use of the Isotope Separator On-Line (ISOL) method for the production of Radioactive Ion Beams (RIB) attract interest because they can be used for nuclear structure and reaction studies, astrophysics research and interdisciplinary applications. The ISOL technique is based on the fast release of the nuclear reaction products from the chosen target material together with their ionization into short-lived nuclei beams. Within this context, the SPES (Selective Production of Exotic Species) facility is now under construction in Italy at INFN-LNL (Istituto Nazionale di Fisica Nucleare — Laboratori Nazionali di Legnaro). The SPES facility will produce RIBs mainly from n-rich isotopes obtained by a 40 MeV cyclotron proton beam (200 μA) directly impinging on a uranium carbide multi-foil fission target. The aim of this work is to describe and update, from a comprehensive point of view, the most important results obtained by the analysis of the on-line behavior of the SPES production target assembly. In particular an improved target configuration has been studied by comparing different codes and physics models: the thermal analyses and the isotope production are re-evaluated. Then some consequent radioprotection aspects, which are essential for the installation and operation of the facility, are presented.

Multidimensional Modeling of Atmospheric Effects and Surface Heterogeneities on Remote Sensing

NASA Technical Reports Server (NTRS)

Gerstl, S. A. W.; Simmer, C.; Zardecki, A. (Principal Investigator)

1985-01-01

The overall goal of this project is to establish a modeling capability that allows a quantitative determination of atmospheric effects on remote sensing including the effects of surface heterogeneities. This includes an improved understanding of aerosol and haze effects in connection with structural, angular, and spatial surface heterogeneities. One important objective of the research is the possible identification of intrinsic surface or canopy characteristics that might be invariant to atmospheric perturbations so that they could be used for scene identification. Conversely, an equally important objective is to find a correction algorithm for atmospheric effects in satellite-sensed surface reflectances. The technical approach is centered around a systematic model and code development effort based on existing, highly advanced computer codes that were originally developed for nuclear radiation shielding applications. Computational techniques for the numerical solution of the radiative transfer equation are adapted on the basis of the discrete-ordinates finite-element method which proved highly successful for one and two-dimensional radiative transfer problems with fully resolved angular representation of the radiation field.

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.

Monte Carlo simulations in Nuclear Medicine

NASA Astrophysics Data System (ADS)

Loudos, George K.

2007-11-01

Molecular imaging technologies provide unique abilities to localise signs of disease before symptoms appear, assist in drug testing, optimize and personalize therapy, and assess the efficacy of treatment regimes for different types of cancer. Monte Carlo simulation packages are used as an important tool for the optimal design of detector systems. In addition they have demonstrated potential to improve image quality and acquisition protocols. Many general purpose (MCNP, Geant4, etc) or dedicated codes (SimSET etc) have been developed aiming to provide accurate and fast results. Special emphasis will be given to GATE toolkit. The GATE code currently under development by the OpenGATE collaboration is the most accurate and promising code for performing realistic simulations. The purpose of this article is to introduce the non expert reader to the current status of MC simulations in nuclear medicine and briefly provide examples of current simulated systems, and present future challenges that include simulation of clinical studies and dosimetry applications.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

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

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« 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

Nuclear quantum many-body dynamics. From collective vibrations to heavy-ion collisions

NASA Astrophysics Data System (ADS)

Simenel, Cédric

2012-11-01

A summary of recent researches on nuclear dynamics with realistic microscopic quantum approaches is presented. The Balian-Vénéroni variational principle is used to derive the time-dependent Hartree-Fock (TDHF) equation describing the dynamics at the mean-field level, as well as an extension including small-amplitude quantum fluctuations which is equivalent to the time-dependent random-phase approximation (TDRPA). Such formalisms as well as their practical implementation in the nuclear physics framework with modern three-dimensional codes are discussed. Recent applications to nuclear dynamics, from collective vibrations to heavy-ion collisions are presented. Particular attention is devoted to the interplay between collective motions and internal degrees of freedom. For instance, the harmonic nature of collective vibrations is questioned. Nuclei are also known to exhibit superfluidity due to pairing residual interaction. Extensions of the theoretical approach to study such pairing vibrations are now available. Large amplitude collective motions are investigated in the framework of heavy-ion collisions leading, for instance, to the formation of a compound system. How fusion is affected by the internal structure of the collision partners, such as their deformation, is discussed. Other mechanisms in competition with fusion, and responsible for the formation of fragments which differ from the entrance channel (transfer reactions, deep-inelastic collisions, and quasi-fission) are investigated. Finally, studies of actinide collisions forming, during very short times of few zeptoseconds, the heaviest nuclear systems available on Earth, are presented.

Current and anticipated uses of thermal-hydraulic codes in Germany

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

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

The FLUKA Code: An Overview

NASA Technical Reports Server (NTRS)

Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M. V.;

Radiation Environment Inside Spacecraft

NASA Technical Reports Server (NTRS)

O'Neill, Patrick

2015-01-01

Dr. Patrick O'Neill, NASA Johnson Space Center, will present a detailed description of the radiation environment inside spacecraft. The free space (outside) solar and galactic cosmic ray and trapped Van Allen belt proton spectra are significantly modified as these ions propagate through various thicknesses of spacecraft structure and shielding material. In addition to energy loss, secondary ions are created as the ions interact with the structure materials. Nuclear interaction codes (FLUKA, GEANT4, HZTRAN, MCNPX, CEM03, and PHITS) transport free space spectra through different thicknesses of various materials. These "inside" energy spectra are then converted to Linear Energy Transfer (LET) spectra and dose rate - that's what's needed by electronics systems designers. Model predictions are compared to radiation measurements made by instruments such as the Intra-Vehicular Charged Particle Directional Spectrometer (IV-CPDS) used inside the Space Station, Orion, and Space Shuttle.

The crystal structure of the Split End protein SHARP adds a new layer of complexity to proteins containing RNA recognition motifs

PubMed Central

Arieti, Fabiana; Gabus, Caroline; Tambalo, Margherita; Huet, Tiphaine; Round, Adam; Thore, Stéphane

2014-01-01

The Split Ends (SPEN) protein was originally discovered in Drosophila in the late 1990s. Since then, homologous proteins have been identified in eukaryotic species ranging from plants to humans. Every family member contains three predicted RNA recognition motifs (RRMs) in the N-terminal region of the protein. We have determined the crystal structure of the region of the human SPEN homolog that contains these RRMs—the SMRT/HDAC1 Associated Repressor Protein (SHARP), at 2.0 Å resolution. SHARP is a co-regulator of the nuclear receptors. We demonstrate that two of the three RRMs, namely RRM3 and RRM4, interact via a highly conserved interface. Furthermore, we show that the RRM3–RRM4 block is the main platform mediating the stable association with the H12–H13 substructure found in the steroid receptor RNA activator (SRA), a long, non-coding RNA previously shown to play a crucial role in nuclear receptor transcriptional regulation. We determine that SHARP association with SRA relies on both single- and double-stranded RNA sequences. The crystal structure of the SHARP–RRM fragment, together with the associated RNA-binding studies, extend the repertoire of nucleic acid binding properties of RRM domains suggesting a new hypothesis for a better understanding of SPEN protein functions. PMID:24748666

The Effect of Weld Metal Strength Mismatch on the Deformation and Fracture Behavior of Steel Butt Weldments

DTIC Science & Technology

1991-01-01

Society 6 of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [ 1980]. Their results are similar to those of Satoh and Toyoda, and are...E813-89. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code , Section III, Nuclear Power Plant Components, 1980. American

Multi-Constraint Multi-Variable Optimization of Source-Driven Nuclear Systems

NASA Astrophysics Data System (ADS)

Watkins, Edward Francis

1995-01-01

A novel approach to the search for optimal designs of source-driven nuclear systems is investigated. Such systems include radiation shields, fusion reactor blankets and various neutron spectrum-shaping assemblies. The novel approach involves the replacement of the steepest-descents optimization algorithm incorporated in the code SWAN by a significantly more general and efficient sequential quadratic programming optimization algorithm provided by the code NPSOL. The resulting SWAN/NPSOL code system can be applied to more general, multi-variable, multi-constraint shield optimization problems. The constraints it accounts for may include simple bounds on variables, linear constraints, and smooth nonlinear constraints. It may also be applied to unconstrained, bound-constrained and linearly constrained optimization. The shield optimization capabilities of the SWAN/NPSOL code system is tested and verified in a variety of optimization problems: dose minimization at constant cost, cost minimization at constant dose, and multiple-nonlinear constraint optimization. The replacement of the optimization part of SWAN with NPSOL is found feasible and leads to a very substantial improvement in the complexity of optimization problems which can be efficiently handled.

The COPERNIC3 project: how AREVA is successfully developing an advanced global fuel rod performance code

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

Garnier, Ch.; Mailhe, P.; Sontheimer, F.

2007-07-01

Fuel performance is a key factor for minimizing operating costs in nuclear plants. One of the important aspects of fuel performance is fuel rod design, based upon reliable tools able to verify the safety of current fuel solutions, prevent potential issues in new core managements and guide the invention of tomorrow's fuels. AREVA is developing its future global fuel rod code COPERNIC3, which is able to calculate the thermal-mechanical behavior of advanced fuel rods in nuclear plants. Some of the best practices to achieve this goal are described, by reviewing the three pillars of a fuel rod code: the database,more » the modelling and the computer and numerical aspects. At first, the COPERNIC3 database content is described, accompanied by the tools developed to effectively exploit the data. Then is given an overview of the main modelling aspects, by emphasizing the thermal, fission gas release and mechanical sub-models. In the last part, numerical solutions are detailed in order to increase the computational performance of the code, with a presentation of software configuration management solutions. (authors)« less

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Functional modules F1--F8 -- Volume 2, Part 1, Revision 4

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

Greene, N.M.; Petrie, L.M.; Westfall, R.M.

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. The manual is divided into three volumes: Volume 1--for the control module documentation; Volume 2--for functional module documentation; and Volume 3--for documentation of the data libraries and subroutine libraries.« less

Shutdown Dose Rate Analysis for the long-pulse D-D Operation Phase in KSTAR

NASA Astrophysics Data System (ADS)

Park, Jin Hun; Han, Jung-Hoon; Kim, D. H.; Joo, K. S.; Hwang, Y. S.

2017-09-01

KSTAR is a medium size fully superconducting tokamak. The deuterium-deuterium (D-D) reaction in the KSTAR tokamak generates neutrons with a peak yield of 3.5x1016 per second through a pulse operation of 100 seconds. The effect of neutron generation from full D-D high power KSTAR operation mode to the machine, such as activation, shutdown dose rate, and nuclear heating, are estimated for an assurance of safety during operation, maintenance, and machine upgrade. The nuclear heating of the in-vessel components, and neutron activation of the surrounding materials have been investigated. The dose rates during operation and after shutdown of KSTAR have been calculated by a 3D CAD model of KSTAR with the Monte Carlo code MCNP5 (neutron flux and decay photon), the inventory code FISPACT (activation and decay photon) and the FENDL 2.1 nuclear data library.

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

Anhydrobiosis-Associated Nuclear DNA Damage and Repair in the Sleeping Chironomid: Linkage with Radioresistance

PubMed Central

Vanyagina, Veronica; Malutina, Ludmila; Cornette, Richard; Sakashita, Tetsuya; Hamada, Nobuyuki; Kikawada, Takahiro; Kobayashi, Yasuhiko; Okuda, Takashi

2010-01-01

Anhydrobiotic chironomid larvae can withstand prolonged complete desiccation as well as other external stresses including ionizing radiation. To understand the cross-tolerance mechanism, we have analyzed the structural changes in the nuclear DNA using transmission electron microscopy and DNA comet assays in relation to anhydrobiosis and radiation. We found that dehydration causes alterations in chromatin structure and a severe fragmentation of nuclear DNA in the cells of the larvae despite successful anhydrobiosis. Furthermore, while the larvae had restored physiological activity within an hour following rehydration, nuclear DNA restoration typically took 72 to 96 h. The DNA fragmentation level and the recovery of DNA integrity in the rehydrated larvae after anhydrobiosis were similar to those of hydrated larvae irradiated with 70 Gy of high-linear energy transfer (LET) ions (4He). In contrast, low-LET radiation (gamma-rays) of the same dose caused less initial damage to the larvae, and DNA was completely repaired within within 24 h. The expression of genes encoding the DNA repair enzymes occurred upon entering anhydrobiosis and exposure to high- and low-LET radiations, indicative of DNA damage that includes double-strand breaks and their subsequent repair. The expression of antioxidant enzymes-coding genes was also elevated in the anhydrobiotic and the gamma-ray-irradiated larvae that probably functions to reduce the negative effect of reactive oxygen species upon exposure to these stresses. Indeed the mature antioxidant proteins accumulated in the dry larvae and the total activity of antioxidants increased by a 3–4 fold in association with anhydrobiosis. We conclude that one of the factors explaining the relationship between radioresistance and the ability to undergo anhydrobiosis in the sleeping chironomid could be an adaptation to desiccation-inflicted nuclear DNA damage. There were also similarities in the molecular response of the larvae to damage caused by desiccation and ionizing radiation. PMID:21103355

Evaluating nuclear physics inputs in core-collapse supernova models

NASA Astrophysics Data System (ADS)

Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

Accurate Structural Correlations from Maximum Likelihood Superpositions

PubMed Central

Theobald, Douglas L; Wuttke, Deborah S

2008-01-01

The cores of globular proteins are densely packed, resulting in complicated networks of structural interactions. These interactions in turn give rise to dynamic structural correlations over a wide range of time scales. Accurate analysis of these complex correlations is crucial for understanding biomolecular mechanisms and for relating structure to function. Here we report a highly accurate technique for inferring the major modes of structural correlation in macromolecules using likelihood-based statistical analysis of sets of structures. This method is generally applicable to any ensemble of related molecules, including families of nuclear magnetic resonance (NMR) models, different crystal forms of a protein, and structural alignments of homologous proteins, as well as molecular dynamics trajectories. Dominant modes of structural correlation are determined using principal components analysis (PCA) of the maximum likelihood estimate of the correlation matrix. The correlations we identify are inherently independent of the statistical uncertainty and dynamic heterogeneity associated with the structural coordinates. We additionally present an easily interpretable method (“PCA plots”) for displaying these positional correlations by color-coding them onto a macromolecular structure. Maximum likelihood PCA of structural superpositions, and the structural PCA plots that illustrate the results, will facilitate the accurate determination of dynamic structural correlations analyzed in diverse fields of structural biology. PMID:18282091

Flow Instability Tests for a Particle Bed Reactor Nuclear Thermal Rocket Fuel Element

DTIC Science & Technology

1993-05-01

2.0 with GWBASIC or higher (DOS 5.0 was installed on the machine). Since the source code was written in BASIC, it was easy to make modifications...8217 AVAILABILITY STATEMENT 12b. DISTRIBUTION CODE Approved for Public Release IAW 190-1 Distribution Unlimited MICHAEL M. BRICKER, SMSgt, USAF Chief...Administration 13. ABSTRACT (Maximum 200 words) i.14. SUBJECT TERMS 15. NUMBER OF PAGES 339 16. PRICE CODE 󈧕. SECURITY CLASSIFICATION 18. SECURITY

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.

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

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

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

The FLUKA code for space applications: recent developments

NASA Technical Reports Server (NTRS)

Andersen, V.; Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.;

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

NASA Astrophysics Data System (ADS)

Barani, T.; Bruschi, E.; Pizzocri, D.; Pastore, G.; Van Uffelen, P.; Williamson, R. L.; Luzzi, L.

2017-04-01

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

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

Barani, T.; Bruschi, E.; Pizzocri, D.

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which ismore » applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.« less

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

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

Andrews, Nathan C.; Gauntt, Randall O.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of themore » of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.« less

WIX: statistical nuclear multifragmentation with collective expansion and Coulomb forces

NASA Astrophysics Data System (ADS)

Randrup, J.∅rgen

1993-10-01

By suitable augmentation of the event generator FREESCO, a code WIX has been constructed with which it is possible to simulate the statistical multifragmentation of a specified nuclear source, which may be both hollow and deformed, in the presence of a collective expansion and with the interfragment Coulomb forces included.

SCALE Code System

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Inventory Uncertainty Quantification using TENDL Covariance Data in Fispact-II

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

Eastwood, J.W.; Morgan, J.G.; Sublet, J.-Ch., E-mail: jean-christophe.sublet@ccfe.ac.uk

2015-01-15

The new inventory code Fispact-II provides predictions of inventory, radiological quantities and their uncertainties using nuclear data covariance information. Central to the method is a novel fast pathways search algorithm using directed graphs. The pathways output provides (1) an aid to identifying important reactions, (2) fast estimates of uncertainties, (3) reduced models that retain important nuclides and reactions for use in the code's Monte Carlo sensitivity analysis module. Described are the methods that are being implemented for improving uncertainty predictions, quantification and propagation using the covariance data that the recent nuclear data libraries contain. In the TENDL library, above themore » upper energy of the resolved resonance range, a Monte Carlo method in which the covariance data come from uncertainties of the nuclear model calculations is used. The nuclear data files are read directly by FISPACT-II without any further intermediate processing. Variance and covariance data are processed and used by FISPACT-II to compute uncertainties in collapsed cross sections, and these are in turn used to predict uncertainties in inventories and all derived radiological data.« less

SCALE Code System 6.2.1

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Visualization of nuclear particle trajectories in nuclear oil-well logging

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

Case, C.R.; Chiaramonte, J.M.

Nuclear oil-well logging measures specific properties of subsurface geological formations as a function of depth in the well. The knowledge gained is used to evaluate the hydrocarbon potential of the surrounding oil field. The measurements are made by lowering an instrument package into an oil well and slowly extracting it at a constant speed. During the extraction phase, neutrons or gamma rays are emitted from the tool, interact with the formation, and scatter back to the detectors located within the tool. Even though only a small percentage of the emitted particles ever reach the detectors, mathematical modeling has been verymore » successful in the accurate prediction of these detector responses. The two dominant methods used to model these devices have been the two-dimensional discrete ordinates method and the three-dimensional Monte Carlo method has routinely been used to investigate the response characteristics of nuclear tools. A special Los Alamos National Laboratory version of their standard MCNP Monte carlo code retains the details of each particle history of later viewing within SABRINA, a companion three-dimensional geometry modeling and debugging code.« less

Random sampling and validation of covariance matrices of resonance parameters

NASA Astrophysics Data System (ADS)

Plevnik, Lucijan; Zerovnik, Gašper

2017-09-01

Analytically exact methods for random sampling of arbitrary correlated parameters are presented. Emphasis is given on one hand on the possible inconsistencies in the covariance data, concentrating on the positive semi-definiteness and consistent sampling of correlated inherently positive parameters, and on the other hand on optimization of the implementation of the methods itself. The methods have been applied in the program ENDSAM, written in the Fortran language, which from a file from a nuclear data library of a chosen isotope in ENDF-6 format produces an arbitrary number of new files in ENDF-6 format which contain values of random samples of resonance parameters (in accordance with corresponding covariance matrices) in places of original values. The source code for the program ENDSAM is available from the OECD/NEA Data Bank. The program works in the following steps: reads resonance parameters and their covariance data from nuclear data library, checks whether the covariance data is consistent, and produces random samples of resonance parameters. The code has been validated with both realistic and artificial data to show that the produced samples are statistically consistent. Additionally, the code was used to validate covariance data in existing nuclear data libraries. A list of inconsistencies, observed in covariance data of resonance parameters in ENDF-VII.1, JEFF-3.2 and JENDL-4.0 is presented. For now, the work has been limited to resonance parameters, however the methods presented are general and can in principle be extended to sampling and validation of any nuclear data.

A Flexible and Non-instrusive Approach for Computing Complex Structural Coverage Metrics

NASA Technical Reports Server (NTRS)

Whalen, Michael W.; Person, Suzette J.; Rungta, Neha; Staats, Matt; Grijincu, Daniela

2015-01-01

Software analysis tools and techniques often leverage structural code coverage information to reason about the dynamic behavior of software. Existing techniques instrument the code with the required structural obligations and then monitor the execution of the compiled code to report coverage. Instrumentation based approaches often incur considerable runtime overhead for complex structural coverage metrics such as Modified Condition/Decision (MC/DC). Code instrumentation, in general, has to be approached with great care to ensure it does not modify the behavior of the original code. Furthermore, instrumented code cannot be used in conjunction with other analyses that reason about the structure and semantics of the code under test. In this work, we introduce a non-intrusive preprocessing approach for computing structural coverage information. It uses a static partial evaluation of the decisions in the source code and a source-to-bytecode mapping to generate the information necessary to efficiently track structural coverage metrics during execution. Our technique is flexible; the results of the preprocessing can be used by a variety of coverage-driven software analysis tasks, including automated analyses that are not possible for instrumented code. Experimental results in the context of symbolic execution show the efficiency and flexibility of our nonintrusive approach for computing code coverage information

Rate heterogeneity in six protein-coding genes from the holoparasite Balanophora (Balanophoraceae) and other taxa of Santalales

PubMed Central

Su, Huei-Jiun; Hu, Jer-Ming

2012-01-01

Background and Aims The holoparasitic flowering plant Balanophora displays extreme floral reduction and was previously found to have enormous rate acceleration in the nuclear 18S rDNA region. So far, it remains unclear whether non-ribosomal, protein-coding genes of Balanophora also evolve in an accelerated fashion and whether the genes with high substitution rates retain their functionality. To tackle these issues, six different genes were sequenced from two Balanophora species and their rate variation and expression patterns were examined. Methods Sequences including nuclear PI, euAP3, TM6, LFY and RPB2 and mitochondrial matR were determined from two Balanophora spp. and compared with selected hemiparasitic species of Santalales and autotrophic core eudicots. Gene expression was detected for the six protein-coding genes and the expression patterns of the three B-class genes (PI, AP3 and TM6) were further examined across different organs of B. laxiflora using RT-PCR. Key Results Balanophora mitochondrial matR is highly accelerated in both nonsynonymous (dN) and synonymous (dS) substitution rates, whereas the rate variation of nuclear genes LFY, PI, euAP3, TM6 and RPB2 are less dramatic. Significant dS increases were detected in Balanophora PI, TM6, RPB2 and dN accelerations in euAP3. All of the protein-coding genes are expressed in inflorescences, indicative of their functionality. PI is restrictively expressed in tepals, synandria and floral bracts, whereas AP3 and TM6 are widely expressed in both male and female inflorescences. Conclusions Despite the observation that rates of sequence evolution are generally higher in Balanophora than in hemiparasitic species of Santalales and autotrophic core eudicots, the five nuclear protein-coding genes are functional and are evolving at a much slower rate than 18S rDNA. The mechanism or mechanisms responsible for rapid sequence evolution and concomitant rate acceleration for 18S rDNA and matR are currently not well understood and require further study in Balanophora and other holoparasites. PMID:23041381

JASMIN: Japanese-American study of muon interactions and neutron detection

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

Nakashima, Hiroshi; /JAEA, Ibaraki; Mokhov, N.V.

Experimental studies of shielding and radiation effects at Fermi National Accelerator Laboratory (FNAL) have been carried out under collaboration between FNAL and Japan, aiming at benchmarking of simulation codes and study of irradiation effects for upgrade and design of new high-energy accelerator facilities. The purposes of this collaboration are (1) acquisition of shielding data in a proton beam energy domain above 100GeV; (2) further evaluation of predictive accuracy of the PHITS and MARS codes; (3) modification of physics models and data in these codes if needed; (4) establishment of irradiation field for radiation effect tests; and (5) development of amore » code module for improved description of radiation effects. A series of experiments has been performed at the Pbar target station and NuMI facility, using irradiation of targets with 120 GeV protons for antiproton and neutrino production, as well as the M-test beam line (M-test) for measuring nuclear data and detector responses. Various nuclear and shielding data have been measured by activation methods with chemical separation techniques as well as by other detectors such as a Bonner ball counter. Analyses with the experimental data are in progress for benchmarking the PHITS and MARS15 codes. In this presentation recent activities and results are reviewed.« less

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

The amplitude effects of sedimentary basins on through-passing surface waves

NASA Astrophysics Data System (ADS)

Feng, L.; Ritzwoller, M. H.; Pasyanos, M.

2016-12-01

Understanding the effect of sedimentary basins on through-passing surface waves is essential in many aspects of seismology, including the estimation of the magnitude of natural and anthropogenic events, the study of the attenuation properties of Earth's interior, and the analysis of ground motion as part of seismic hazard assessment. In particular, knowledge of the physical causes of amplitude variations is important in the application of the Ms:mb discriminant of nuclear monitoring. Our work addresses two principal questions, both in the period range between 10 s and 20 s. The first question is: In what respects can surface wave propagation through 3D structures be simulated as 2D membrane waves? This question is motivated by our belief that surface wave amplitude effects down-stream from sedimentary basins result predominantly from elastic focusing and defocusing, which we understand as analogous to the effect of a lens. To the extent that this understanding is correct, 2D membrane waves will approximately capture the amplitude effects of focusing and defocusing. We address this question by applying the 3D simulation code SW4 (a node-based finite-difference code for 3D seismic wave simulation) and the 2D code SPECFEM2D (a spectral element code for 2D seismic wave simulation). Our results show that for surface waves propagating downstream from 3D sedimentary basins, amplitude effects are mostly caused by elastic focusing and defocusing which is modeled accurately as a 2D effect. However, if the epicentral distance is small, higher modes may contaminate the fundamental mode, which may result in large errors in the 2D membrane wave approximation. The second question is: Are observations of amplitude variations across East Asia following North Korean nuclear tests consistent with simulations of amplitude variations caused by elastic focusing/defocusing through a crustal reference model of China (Shen et al., A seismic reference model for the crust and uppermost mantle beneath China from surface wave dispersion, Geophys. J. Int., 206(2), 2015)? We simulate surface wave propagation across Eastern Asia with SES3D (a spectral element code for 3D seismic wave simulation) and observe significant amplitude variations caused by focusing and defocusing with a magnitude that is consistent with the observations.

Computing the cross sections of nuclear reactions with nuclear clusters emission for proton energies between 30 MeV and 2.6 GeV

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

Korovin, Yu. A.; Maksimushkina, A. V., E-mail: AVMaksimushkina@mephi.ru; Frolova, T. A.

2016-12-15

The cross sections of nuclear reactions involving emission of clusters of light nuclei in proton collisions with a heavy-metal target are computed for incident-proton energies between 30 MeV and 2.6 GeV. The calculation relies on the ALICE/ASH and CASCADE/INPE computer codes. The parameters determining the pre-equilibrium cluster emission are varied in the computation.

Current Research on Non-Coding Ribonucleic Acid (RNA).

PubMed

Wang, Jing; Samuels, David C; Zhao, Shilin; Xiang, Yu; Zhao, Ying-Yong; Guo, Yan

2017-12-05

Non-coding ribonucleic acid (RNA) has without a doubt captured the interest of biomedical researchers. The ability to screen the entire human genome with high-throughput sequencing technology has greatly enhanced the identification, annotation and prediction of the functionality of non-coding RNAs. In this review, we discuss the current landscape of non-coding RNA research and quantitative analysis. Non-coding RNA will be categorized into two major groups by size: long non-coding RNAs and small RNAs. In long non-coding RNA, we discuss regular long non-coding RNA, pseudogenes and circular RNA. In small RNA, we discuss miRNA, transfer RNA, piwi-interacting RNA, small nucleolar RNA, small nuclear RNA, Y RNA, single recognition particle RNA, and 7SK RNA. We elaborate on the origin, detection method, and potential association with disease, putative functional mechanisms, and public resources for these non-coding RNAs. We aim to provide readers with a complete overview of non-coding RNAs and incite additional interest in non-coding RNA research.

Determination and Fabrication of New Shield Super Alloys Materials for Nuclear Reactor Safety by Experiments and Cern-Fluka Monte Carlo Simulation Code, Geant4 and WinXCom

NASA Astrophysics Data System (ADS)

Aygun, Bünyamin; Korkut, Turgay; Karabulut, Abdulhalik

2016-05-01

Despite the possibility of depletion of fossil fuels increasing energy needs the use of radiation tends to increase. Recently the security-focused debate about planned nuclear power plants still continues. The objective of this thesis is to prevent the radiation spread from nuclear reactors into the environment. In order to do this, we produced higher performanced of new shielding materials which are high radiation holders in reactors operation. Some additives used in new shielding materials; some of iron (Fe), rhenium (Re), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta), boron carbide (B4C). The results of this experiments indicated that these materials are good shields against gamma and neutrons. The powder metallurgy technique was used to produce new shielding materials. CERN - FLUKA Geant4 Monte Carlo simulation code and WinXCom were used for determination of the percentages of high temperature resistant and high-level fast neutron and gamma shielding materials participated components. Super alloys was produced and then the experimental fast neutron dose equivalent measurements and gamma radiation absorpsion of the new shielding materials were carried out. The produced products to be used safely reactors not only in nuclear medicine, in the treatment room, for the storage of nuclear waste, nuclear research laboratories, against cosmic radiation in space vehicles and has the qualities.

NUCLEAR HEATING IN LIF DOSEMETERS IN A FUSION NEUTRON FIELD, TRIAL OF DIRECT COMPARISON OF EXPERIMENTAL AND SIMULATED RESULTS.

PubMed

Pohorecki, Wladyslaw; Obryk, Barbara

2017-09-29

The results of nuclear heating measured by means of thermoluminescent dosemeters (TLD-LiF) in a Cu block irradiated by 14 MeV neutrons are presented. The integral Cu experiment relevant for verification of copper nuclear data at neutron energies characteristic for fusion facilities was performed in the ENEA FNG Laboratory at Frascati. Five types of TLDs were used: highly photon sensitive LiF:Mg,Cu,P (MCP-N), 7LiF:Mg,Cu,P (MCP-7) and standard, lower sensitivity LiF:Mg,Ti (MTS-N), 7LiF:Mg,Ti (MTS-7) and 6LiF:Mg,Ti (MTS-6). Calibration of the detectors was performed with gamma rays in terms of air-kerma (10 mGy of 137Cs air-kerma). Nuclear heating in the Cu block was also calculated with the use of MCNP transport code Nuclear heating in Cu and air in TLD's positions was calculated as well. The nuclear heating contribution from all simulated by MCNP6 code particles including protons, deuterons, alphas tritons and heavier ions produced by the neutron interactions were calculated. A trial of the direct comparison between experimental results and results of simulation was performed. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Flash Galaxy Cluster Merger, Simulated using the Flash Code, Mass Ratio 1:1

ScienceCinema

None

2018-05-11

Since structure in the universe forms in a bottom-up fashion, with smaller structures merging to form larger ones, modeling the merging process in detail is crucial to our understanding of cosmology. At the current epoch, we observe clusters of galaxies undergoing mergers. It is seen that the two major components of galaxy clusters, the hot intracluster gas and the dark matter, behave very differently during the course of a merger. Using the N-body and hydrodynamics capabilities in the FLASH code, we have simulated a suite of representative galaxy cluster mergers, including the dynamics of both the dark matter, which is collisionless, and the gas, which has the properties of a fluid. 3-D visualizations such as these demonstrate clearly the different behavior of these two components over time. Credits: Science: John Zuhone (Harvard-Smithsonian Center for Astrophysics Visualization: Jonathan Gallagher (Flash Center, University of Chicago)

 This research used resources of the Argonne Leadership Computing Facility at Argonne National Laboratory, which is supported by the Office of Science of the U.S. Dept. of Energy (DOE) under contract DE-AC02-06CH11357. This research was supported by the National Nuclear Security Administration's (NNSA) Advanced Simulation and Computing (ASC) Academic Strategic Alliance Program (ASAP).

Flash Galaxy Cluster Merger, Simulated using the Flash Code, Mass Ratio 1:1

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

None

2010-08-09

Since structure in the universe forms in a bottom-up fashion, with smaller structures merging to form larger ones, modeling the merging process in detail is crucial to our understanding of cosmology. At the current epoch, we observe clusters of galaxies undergoing mergers. It is seen that the two major components of galaxy clusters, the hot intracluster gas and the dark matter, behave very differently during the course of a merger. Using the N-body and hydrodynamics capabilities in the FLASH code, we have simulated a suite of representative galaxy cluster mergers, including the dynamics of both the dark matter, which ismore » collisionless, and the gas, which has the properties of a fluid. 3-D visualizations such as these demonstrate clearly the different behavior of these two components over time. Credits: Science: John Zuhone (Harvard-Smithsonian Center for Astrophysics Visualization: Jonathan Gallagher (Flash Center, University of Chicago)

 This research used resources of the Argonne Leadership Computing Facility at Argonne National Laboratory, which is supported by the Office of Science of the U.S. Dept. of Energy (DOE) under contract DE-AC02-06CH11357. This research was supported by the National Nuclear Security Administration's (NNSA) Advanced Simulation and Computing (ASC) Academic Strategic Alliance Program (ASAP).« less