have been examined using the codes RELAP, TRAC, RAMONA, and SMABRB. Severe accidents have been-of-coolant accident analysis. Pour differ- ent codes have been studied during 1982. The American code RELAP5 has been and to the RELAP5 calcu- lations of Finland and Sweden. The comparison to data ...

APRIL is a fast-running and user-friendly system code for interactive simulations of severe accidents in boiling water reactors (BWRs). The component models in the most recent version, APRIL.MOD3X, include the reactor core and pressure vessel, as well as the primary and secondary containments. Whereas APRIL.MOD3X is a fast-running ...

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss ...

This document has been prepared as a user's guide for the computer program RISKIN developed at Sandia National Laboratories. The RISKIN code generates integrated risk tables and the weighted mean risk associated with a user-selected set of consequences from up to five output files generated by the MELCOR Accident Consequence ...

... Accession Number : AD0620258. Title : FURTHER ATTEMPTS AT CODING AIRCRAFT ACCIDENTS. Descriptive Note : Research rept.,. ...

... AD0076065. Title : CODED FLEET ACCIDENTS OF RECENT GRADUATES. Corporate Author : TULANE UNIV NEW ORLEANS LA. ...

The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to ...

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Steam Intrusion from Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included within.

We introduce a new coupled neutronics/thermal hydraulics code system for analyzing transients of nuclear power plants and research reactors, based on a neutron transport theory approach. For the neutron kinetics, we have developed the code DORT-TD, a time-dependent extension of the well-known discrete ordinates code DORT. DORT-TD uses ...

It is prerequisite for the establishment of nuclear power plant safety and for the maximization of operation efficiency to devlope the accident analysis computer code packages which can predict the results of postulated accidents and evaluate the performa...

The VICTORIA code is designed to be a module of the MELPROG severe accident analysis code. The VICTORIA code is designed to model the in-vessel phase of a severe reactor accident. The functions of VICTORIA in MELPROG are fission product release from fuel,...

The initial activities of a Department of Energy (DOE) Safety Analysis Software Group to establish a Safety Analysis Toolbox of computer models are discussed. The toolbox shall be a DOE Complex repository of verified and validated computer models that are configuration-controlled and made available for specific accident analysis applications. The toolbox concept was ...

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed ...

The capability to perform thermal-hydraulic analyses of an SP-100 space reactor was demonstrated using the ATHENA computer code. The preliminary General Electric SP-100 design was modeled using Athena. The model simulates the fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of this design. Two ATHENA demonstration calculations were performed simulating ...

The primary system of a pressurized water reactor (PWR) operates at a relatively high pressure (15.5 MPa, 2250 psia) and consists of piping and components designed to withstand these pressures. The low-pressure-injection system (LPIS) connects to the primary system but possesses low-pressure piping passing outside the containment. Therefore, a potential exists for a loss-of-coolant ...

... the interfaces and responsibilities for subspeciality coding, counseling, utilization and ... code are analyzed with regard to career paths (graduate ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

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. ...

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. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The MARCH code, written at Battelle's Columbus Laboratories for the U.S. Nuclear Regulatory Commission, describes the response of LWR systems to accidents which can result in core meltdown. The calculations are performed from the start of the accident thr...

An exploratory sensitivity study with the MELCOR Accident Consequence Code System (MACCS) is presented. This study was performed to provide (1) an indication of the possible impact of consequence modeling uncertainties on the results of an integrated prob...

In the analysis of accidents a collision between aircraft is treated as one accident in the overall total. However, a complete analysis and coding is made on each aircraft involved in a collision. This produces two aircraft accident records, one for each ...

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system and the core during a severe accident transient, ...

This report describes a coupled Eulerian-Lagrangian code, ICECO-CEL, for analyzing the response of the primary system during hypothetical core disruptive accidents. The implicit Eulerian method is used to calculate the fluid motion so that large fluid distortion, two-dimensional sliding interface, flow around corners, flow through ...

A reactor safety code that provides faster-than-real-time capabilities greatly enhances the effectiveness of an analyst. We have developed a new version of the Transient Reactor Analysis Code, TRAC-NPA, which is capable of running many one- and three-dimensional transients faster than real time on current computers. TRAC is an Eulerian finite-difference ...

The capability to perform thermal-hydraulic analyses of a TRIGA reactor was demonstrated using the ATHENA computer code. TRIGA is an advanced reactor designed to produce electrical power while being inherently safe during reactivity accidents, loss-of-coolant accidents (LOCAs), and station blackout. The TRIGA system contains a ...

Failure of a containment building to perform its function takes place when a leak path is formed. Bellows expansion joints, that are used at the penetrations where piping systems penetrate the containment structures, are one of the potential points of failure. In the event of a severe accident, they can be simultaneously subjected to internal pressure, temperature and gross ...

A computer code called TRAC is being developed by the Los Alamos Scientific Laboratory for analysis of loss-of-coolant accidents (LOCA's) and other transients in light water reactors. This code differs from existing codes and other codes under development...

... Title : Uncertainty and Sensitivity Analysis of a Fire-Induced Accident Scenario Involving Binary Variables and Mechanistic Codes. ...

This paper is concerned with the application of severe-accident codes to develop improved accident mitigation strategies for nuclear power plants employing boiling water reactors (BWRs). Specifically, the results of analyses using the APRIL.MOD3X code are presented for typical station blackout ...

Researchers at Pacific Northwest Laboratory are attempting to improve the user interface to a number of computational chemistry codes. We used TAE+ to implement a new graphical user interface to one of these codes. From this task we learned that TAE+, in ...

The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe ...

Circuit originally developed for a flight simulator interfaces a computer graphics system with color

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report (FSAR): Steam Intrusion From Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included in the following sections to aid in the understanding of ...

This document includes overviews of the interfaces, and code examples. ...... The following GCTP projection codes are used in the grid API described in ...

includes overviews of the interfaces, and code examples. ..... 8.3 The Metadata Configuration File (MCF) for Codes in Section 8.2. ...

... Title : Code Maintenance and Design for a Visual Programming Language Graphical User Interface. Descriptive Note : Master's thesis. ...

... Abstract : The Absolute Radiometric Code (ARC) is a collection of Matlab functions tied together under a Matlab Graphical User Interface (GUI). ...

It is not feasible to use the existing British Coal accident information collection, storage and retrieval system to identify systematically human factors issues relevant to the accident. The implementation of a human factors code list, developed during t...

A key parameter in the calculation of accident dose-risks by the RADTRAN 4 code is the time assigned for evacuation of the affected area surrounding the accident. Currently, in the interest of assured conservatism, this time is set at 24 hrs. Casual anecd...

This report describes the analytical results of operational transient and accidents for the safety assessment of upgraded JRR-3. The following six items of transients and accidents have been selected and analyzed for the assessment; 1. decrease of primary...

The MARCH and HECTR computer codes are used in this study to examine hydrogen production, transport, and combustion in an ice-condenser containment for a number of hypothesized severe accidents. Both degraded-core and core-meltdown accidents are treated. ...

The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and ...

The Department of Energy has developed new capabilities to predict how an accident event would spread through a facility. These capabilities arose as a part of the development of a future generation accident code - CONACS. They can be used to predict the ...

After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into ...

One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat t...

Cheap and clean energy can be supplied with nuclear power plant whose accident probability is very low comparing with that of the other industrial facilities. However, the consequences of severe accident of nuclear power plant may result in a critical imp...

This report provides documentation for UMTRIs file of Buses Involved in Fatal Accidents (BIFA), 2008, including distributions of the code values for each variable in the file. The 2008 BIFA file is a census of all buses involved in a fatal accident in the...

This report provides documentation for UMTRIs file of Buses Involved in Fatal Accidents (BIFA), 2007, including distributions of the code values for each variable in the file. The 2007 BIFA file is a census of all buses involved in a fatal accident in the...

The publication, containing reports of U. S. Civil aircraft accidents, in brief format, is issued monthly and includes those accidents analyzed and processed during the preceding month. The reports are reproduced directly from the coded record on magnetic...

The publication, contains reports of U. S. Civil aricraft accidents, in brief format, is issued monthly and includes those accidents analyzed and processed during the preceding month. The reports are reproduced directly from the coded record on magnetic t...

manual. CONCLUSIONS. A graphical user interface,. XFLOPS, has been created for the aircraft performance and sizing computer code FLOPS. This interface ...

A computer modeling code, CRIT8, was written to allow prediction of the radiological doses to workers and members of the public resulting from these postulated maximum-effect accidents. The code accounts for the relationships of the initial parent radionuclide inventory at the time of the accident to the growth of ...

In this paper, a simple analytical code, QUASAR, is developed to analyze the phenomena related to severe subassembly accidents, such as a total instantaneous blockage event for a subassembly inlet. The code models failed and neighboring subassemblies, focusing mainly on the thermal consequences and the propagation potential of the ...

Operator actions play an important role in accident sequences. However, design analysis (Safety Analysis Report, SAR) seldom includes consideration of operator actions, although they are required by compulsory Emergency Operating Procedures (EOP) to perform some checks and actions from the very beginning of the accident. The basic aim of the project is to ...

The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reac...

The SAS4A code system is a new tool for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failures of the subassembly walls. The objective in the development of SAS4A is to provide improved analytical mod...

During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident an...

The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accid...

Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cool...

This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and...

The accident at Three Mile Island Unit 2 (M-2) provides an opportunity to benchmark severe accident analysis methods against full-scale, integrated facility data. In collaboration with the US Department of Energy (DOE), the OECD Nuclear Energy Agency established a joint task group to analyze various periods of the accident and ...

The accident at Three Mile Island Unit 2 (M-2) provides an opportunity to benchmark severe accident analysis methods against full-scale, integrated facility data. In collaboration with the US Department of Energy (DOE), the OECD Nuclear Energy Agency established a joint task group to analyze various periods of the accident and ...

Two computer codes are developed for performing the calculations of SUPER-PHENIX integrity when subject to an hypothetical core disruptive accident; both codes are 2D axisymmetric but they use different meshes: Lagrangian for the SIRIUS code, Eulerian and...

Sandia has made considerable progress in the past year on the MELCOR code for integrated severe nuclear reactor accident analysis. Actinities for the past year are presented.

It is believed that the numerical simulation of thermal-hydraulic phenomena of multiphase, multicomponent flows in a reactor core is essential to investigate core disruptive accidents (CDAs) of liquid-metal fast reactors. A new multicomponent vaporization/condensation (V/C) model was developed to provide a generalized model for a fast reactor safety analysis ...

A hazard assessment was conducted on the GRC managed payloads in support of a NASA Headquarters Code

... ACCIDENTS, UNITED STATES, INTEGRATED SYSTEMS, INDUSTRIES ... CASUALTIES, PERSONNEL, WORK, MORTALITY RATE, AGRICULTURE. ...

SECTION 1 - AMICABLE ADJUSTMENT OF DIFFERENCES: In the belief that all ...... may be granted by supervisors for illness, marriage, accident, or critical ...

VIPRE (Versatile Internals and Component Program for Reactors;EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (NDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed ...

The International Atomic Energy Agency has sponsored a program, �Accident Analysis and its Associated Training Programme for RBMK-1000 Kursk-1 NPP (Phase II)�. Under the auspices of this program, Reactor Research Centre �Kurchatov Institute� (RRC-KI) has implemented a Neutron Flux Correction Methodology in Version 1.2.2 of the RELAP5-3D code. The ...

C++ code then compiled and linked into a library; Perl scripts call Perl ... Code Toolkit interfaces and computationally intensive code in C++; Need a better way to ...

ible Face Seals (IFACE); Labyrinth Seals (KTK); Seal Dynamics Analysis Code for Cylindrical and Face Seals. (SEALDY); User Interface for Scientific Code ...

technologies. The Center has developed standardized interfaces to local mesh refinement codes. Existing codes of the Center partners and new codes are supported to create a plug...

This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in ...

The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to in core pressure ...

Fission product release analysis code, RACPAC (Fission Product Release Analysis Code from Fuel Particle in Accident Condition), was developed to calculate fractional release from the core during accident conditions of High Temperature Gas-cooled Reactor. ...

The ORION code is designed to determine very quickly the immediate consequences (such as plume passage time, instantaneous maximum hazards irradiation, inhalation, deposit) due to an accident spreading out radioactive or chemical pollution into the atmosp...

An external code interface is defined which allows other transient applications to communicate with the PRONTO family of finite element programs. This interface is written in ANSI FORTRAN and allows an independent author to specify requirements for an external code to PRONTO. The interface is ...

Modern digital communication techniques often require the generation of pseudorandom numbers or sequences. The ability to quickly and easily produce various codes such as maximal length codes, Gold codes, Jet Propulsion Laboratory ranging codes, syncopate...

The purpose of this paper is to present and document the differences discovered when comparing the two accident dose codes FUSEDOSE and MACCS2. Each code`s methodology is first discussed. With this background, the important comparison parameters are discussed and the resulting differences are presented. It is not the purpose of this ...

This report describes the Application Programming Interface for the PVMEXEC program and the code coupling systems that it implements. The information in the report is intended for programmers wanting to add a new code into the coupling system.

Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included.

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the ...

This paper describes the ALICE-II analysis of and comparison with complex vessel experiments. Tests SM-2 through SM-5 were performed by SRI International in 1978 in studying the structural response of 1/20 scale models of the Clinch River Breeder Reactor to a simulated hypothetical core-disruptive accident. These experiments provided quality data for validating treatments of ...

All simulations require some type of user interface to be functional. The problem is that this interface is sometimes written as a tightly coupled part of the simulation itself, buried in the main body of code. This makes the interface very difficult to c...

A major benefit of a Quality Assurance system in a radiotherapy centre is that it reduces the likelihood of an accident. For over 20 years I have been the interface in the UK between the Institute of Physics and Engineering in Medicine and the media � newspapers, radio and TV � and so I have learned about radiotherapy accidents ...

Major activities included: (1) the vital area analyses of operating reactor facilities (both PWR and BWR); (2) assistance to NRC for evaluation of the Three Mile Island accident; (3) further development and testing of the ADPATH (adversary paths) subroutine for finding single-target theft and multiple-target sabotage paths in a facility; (4) the continued design and ...

The LOFT Engineering Simulator was developed to supply plant equivalent data for evaluating graphic aids and advanced control concepts for nuclear plant operators. The Simulator, a combination of hardware and software, combines some of the features of best estimate (safety analysis) computer codes with reactor operator training simulators. The LOFT Engineering Simulator ...

This paper describes methods, in the form of a handbook and five computer codes, that can be used for air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the ...

The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the en...

The TRAC (Transient Reactor Analysis Code) program is an advanced computer code which is designed to model postulated accidents in light water reactors. The paper discusses some of the methodology employed in modeling large internal flow systems. Emphasis is placed on the numerical and modeling problems inherent in computer ...

A number of computer codes have been written to attempt to predict the consequences of hypothetical accidents. These codes all have one thing in common: they produce reams of paper containing vast quantities of numbers. In an effort to make these numbers ...

CONTAIN is a computer code to analyze physical, chemical and radiological processes inside the reactor containment in the sequence of severe reactor accident. Modelling of the aerosol behavior is included. We have improved the code by implementing a subro...

Several computer codes have been developed to predict the behavior of aerosols in steam, a situation which is expected to occur in a light-water reactor accident. Among the codes with capabilities in this respect are MAEROS (Sandia) AEROMECH (University o...

The VICTORIA code is designed to be a module of the MELPROG severe accident analysis code. The functions of VICTORIA in MELPROG are fission product release from fuel, vapor-phase and aerosol physics, and chemical interactions involving fuel, structures, c...

The objective of this paper is to discuss how Postaccident Heat Removal (PAHR) calculations interface with the earlier accident phases in overall mechanistic accident analyses, to briefly review the evolution of the transition phase as an important path i...

Page 1. DSMC Simulations in Support of the STS-107 Accident ... Entry Interface for STS-107 occurred at 8:44:09 AM (EST) on February 1st, 2003. ...

The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the ...

The purpose of the work described in this paper was to evaluate the adequacy of criticality accident alarm system (CAAS) coverage of several buildings at the Portsmouth Gaseous Diffusion Plant (PORTS) located in Piketon, Ohio. An analysis was performed in which the General Monte Carlo N-Particle Transport Code (MCNP) was used to model several of the ...

The RELAP5/MOD1 computer code was used in the analysis of a loss coolant accident (LOCA), postulated for Angra-2. The power plant was simulated through a division in control volumes and junctions suitable for a small break accident calculation. Initially ...

Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

The paper discusses Nuclear Regulatory Commission (NRC) efforts regarding severe reactor accident management and the Nuclear Management and Resources Council (NUMAEX), activities. (EPRI) Electric Power Research Institute accident management program consists of the two products just mentioned plus one related to severe accident plant ...

. Although analyses of HFACS-coded data reveal that CRM is a factor in aviation accidents, the actual1 Paper Number 2005-xx-yyyy CLASSIFYING CREW PERFORMANCE FAILURES IN COMMERCIAL AVIATION ACCIDENTS University of Illinois at Urbana-Champaign Champaign, IL The purpose of accident investigation is to identify

This report documents the key safety and operational aspects of a Z-pinch Externally Driven Nuclear Assembly (ZEDNA) reactor concept which is envisioned to be built and operated at the Z-machine facility in Technical Area IV. Operating parameters and reactor neutronic conditions are established that would meet the design requirements of the system. Accident and off-normal ...

A broadband data base was established on the energetics of single phase and two-phase discharge into a liquid pool. The experimental setup using simulant fluids was a 1/7 scale model of the core-and-vessel configuration of loop-type liquid metal fast breeder reactor. The condensation-driven entrainment of the pool liquid into the expanding bubble was significant. The impact pressure on the top of ...

A scaled test facility of the Babcock and Wilcox raised loop nuclear steam supply system was used to perform small break loss-of-coolant accident testing, thereby, establishing a data base from which plant predictive system codes could be benchmarked. About 250 instruments were used to record the thermal/hydraulic response of the test facility during the ...

INTRODUCTION: The rate for work related accidents in the Spanish mining sector is notably higher than in other countries such as the United States. It produces a very negative impact on the mining industry. This paper is the report of a study on serious and fatal accidents in Spanish mining from 1982-2006. It is based on the reports of 212 ...

A new code to calculate the effects of intrabeam scattering (IBS) has been developed in MATLAB based on the approximation suggested by K. Bane. It interfaces with the Accelerator Toolbox but can also read in lattice functions from other codes. The code ha...

This viewgraph presentation describes the development of user interfaces for OS/2 versions of computer codes for the analysis of seals. Current status, new features, work in progress, and future plans are discussed.

This viewgraph presentation describes the development of user interfaces for OS/2 versions of computer codes for the analysis of seals. Current status, new features, work in progress, and future plans are discussed.

... User Interface for a Physical Optics Radar Cross Section Prediction Code. ... objects utilizes the scientific computational features of MATLAB and its ...

... In general, the platform interface should not define wrappers for any procedure ... Figure 4 shows an excerpt from the Java API platform interface that ...

From Bill Blackwell/SVERDRUP: Below is the latest set of interface headers for CXORAD. The changes made affect only the inputs to the code. ...

... Contract N00014-87-J-1118 R & T Code 4133016 Technical Report No. 15 Adsorption of Pyridine at the Au( 110) - Solution Interface DTIC by ...

May 31, 2011 ... Therefore, a graphical user interface has been developed to allow the analyst to more quickly and easily apply these codes. ...

Therefore, a graphical user interface has been developed to allow the analyst to more quickly and easily apply these codes. This paper describes the ...