Filters

Results

**1**-**10**of**180** Results

**1**-**10**of**180**. Search took:**0.023**secondsSort by: date | relevance |

Saltelli, A.; Bertozzi, G.; Stanners, D.A.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] Description of program or function: LISA is used in the analysis of hazards due to the disposal of nuclear waste in geological formations. The risk linked to pre-established release scenarios is assessed in terms of dose rate to a maximum exposed individual. The various sub-models in the code simulate the system of barriers - both natural and man made - which are interposed between the contaminants and man

Primary Subject

Secondary Subject

Source

12 Oct 1995; [html]; Available on-line: http://www.nea.fr/abs/html/nea-0860.html; Country of input: International Atomic Energy Agency (IAEA); 7 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Rieffe, Henk Ch.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: FITOCO converts neutron spectrum and group cross section data from a fine group structure to a coarser group structure. The program uses a SAND-II type fine- group structure as input. The output from FITOCO can directly be applied as input for the spectrum unfolding code STAY'SL or similar codes. FITOCO can also calculate the effective coarse group cross values with detector cover materials, e.g. B-10 or Cd. 2 - Restrictions on the complexity of the problem: The lowest energy bound of the coarse group structure is fixed at 10

^{-10}MeVPrimary Subject

Secondary Subject

Source

4 Sep 1992; [html]; Available on-line: http://www.nea.fr/abs/html/nea-0894.html; Country of input: International Atomic Energy Agency (IAEA); 1 ref.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Patrakka, E.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: TRANSHEX is a multigroup integral transport program that determines the thermal scalar flux distribution arising from a known epithermal flux in two- dimensional hexagonal geometry. 2 - Method of solution: The program solves the isotropic collision probability equations for a region-averaged scalar flux by an iterative method. Either a successive over-relaxation or an inner-outer iteration technique is applied. Flat flux collision probabilities between trigonal space regions with white boundary condition are utilized. The effect of epithermal flux is taken into consideration as a slowing-down source that is calculated for a given spatial distribution and 1/E energy dependence of the epithermal flux

Primary Subject

Secondary Subject

Source

23 Mar 1994; [html]; Available on-line: http://www.nea.fr/abs/html/nea-0953.html; Country of input: International Atomic Energy Agency (IAEA); refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

BOUNDARY CONDITIONS, COMPUTER PROGRAM DOCUMENTATION, ENERGY DEPENDENCE, EPITHERMAL NEUTRONS, EQUATIONS, GEOMETRY, HEXAGONAL CONFIGURATION, INTEGRALS, ITERATIVE METHODS, MULTIGROUP THEORY, PROBABILITY, SLOWING-DOWN, SPATIAL DISTRIBUTION, T CODES, THERMAL NEUTRONS, TWO-DIMENSIONAL CALCULATIONS, WEBSITES

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Thompson, B.G.J.; Broyd, T.; Summerling, T.; Dalrymple, G.; Gralewski, A; Rae, I.C.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: The SYVAC program simulates the ground water mediated movement of radionuclides from underground facilities for the disposal of low and intermediate level wastes to the accessible environment, and provides an estimate of the subsequent radiological risk to man. The simulated timescales are usually within the range 1.0 E+03 to 1.0 E+07 years. SYVAC is capable of modelling both shallow disposal facilities (located in argillaceous media and overlaying an aquifer) and deep disposal facilities (in a saturated environment). It is not suitable for considering high level wastes as it does not allow for the heat produced by such wastes. SYVAC has been developed for use within the DOE Radioactive Waste Management Programme, as one tool in the DOE Assessment Methodology. 2 - Method of solution: - The Physical System: A simplistic description of the physical simulated by SYVAC is given. This simulation requires data to describe the physical structure and transport characteristics of the vault and geosphere. The necessary data is supplied to SYVAC in the form of parameters that represent hydraulic conductivities, diffusion coefficients, porosities etc for the vault and geosphere. Thirteen basic vault parameters are used to model deep disposal sites, and fifteen for shallow disposals. For both types of site, an additional six parameters are needed for each geological layer modelled in the geosphere. The values of some of the parameters are not precisely known, and will vary over the timescale of the transport to the biosphere. SYVAC allows for this with a probabilistic approach; the parameter values are sampled from distributions chosen to cover the expected range of values that may be encountered during the post-closure phase of a vault. For each parameter, SYVAC allows the user to select one of five distributions: constant, uniform, log uniform, normal and log normal. Usually, for each parameter, the user will select the distribution that best fits the observed parameter values. The calculation of risk to man requires knowledge of both the dose from a particular run, and the probability of that run. Clearly, the dose to man calculated for a particular simulation run will depend on the parameter values chosen for that run. Two different sampling methods can be chosen for SYVAC; Random Sampling and Deterministic Generator Sampling. Random Sampling chooses values (randomly) from the entire range of the distribution chosen for each parameter. Deterministic Generator sampling selects one of 11 possible (discrete) sample points for each parameter for a run; the method of sampling is such that all eleven points are chosen in the course of 11 consecutive runs. The eleven points span the range of the distribution chosen for each parameter. The Executive is the framework that controls the simulation program as a whole, and specifically calls the Vault and Geosphere sub-models. The Vault sub-model simulates the release of radionuclides from non heat-producing radioactive waste placed within an engineered vault in either deep or shallow land disposal facilities. The model includes representations of the waste matrix, waste canister, vault liner, the vault backfill material and the region of host rock around the vault disturbed during construction. The migration model includes the processes of solubility-limited or mass-limited leaching of radionuclides from the waste matrix, linear equilibrium sorption of radionuclides onto the barrier materials, radioactive chain decay, dispersion and advection. The Geosphere sub-model is a 1-dimensional model that deals with the migration of radionuclides from the vault to the biosphere through the geosphere. This is through the calculation of a response function for each geosphere layer. Up to five distinct layers can be modelled. The biosphere code ECOS calculates the rate of accumulation of committed effective dose equivalent (dose) arising from the flux of radionuclides entering the biosphere. Dose to a maximally exposed individual and collective dose are calculated deterministically. 3 - Restrictions on the complexity of the problem: - for the Vault Sub-model, they concern the representations of the flow, the time invariant chemistry, the equations for flow and transport, the water table, the interaction between adjacent Vault structures, the limitations on number of canisters, nuclides per chain, etc.; for the Geosphere Sub-model, they concern the time-steps in the model, the number of nuclides and geosphere layers; for the Biosphere Model, they concern the estimates of activity and dose to man, the production or transport of gaseous radionuclide species and doses

Primary Subject

Secondary Subject

Source

10 Oct 1991; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1023.html; Country of input: International Atomic Energy Agency (IAEA); 29 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

ADVECTION, BIOSPHERE, COMPUTER PROGRAM DOCUMENTATION, COMPUTERIZED SIMULATION, CONTAINERS, DISTRIBUTION FUNCTIONS, DOSE EQUIVALENTS, ENVIRONMENTAL TRANSPORT, GEOLOGIC STRATA, GROUND WATER, HAZARDS, HIGH-LEVEL RADIOACTIVE WASTES, HYDRAULIC CONDUCTIVITY, INTERMEDIATE-LEVEL RADIOACTIVE WASTES, LEACHING, PROBABILISTIC ESTIMATION, RADIOACTIVE WASTE DISPOSAL, RADIOACTIVE WASTE STORAGE, RADIOISOTOPES, RESPONSE FUNCTIONS, RISK ASSESSMENT, S CODES, SAMPLING, UNDERGROUND FACILITIES, UNDERGROUND STORAGE, WATER TABLES, WEBSITES

CALCULATION METHODS, COMPUTER CODES, DISSOLUTION, DOCUMENT TYPES, FUNCTIONS, GEOLOGIC STRUCTURES, HYDROGEN COMPOUNDS, ISOTOPES, MANAGEMENT, MASS TRANSFER, MATERIALS, OXYGEN COMPOUNDS, RADIOACTIVE MATERIALS, RADIOACTIVE WASTE MANAGEMENT, RADIOACTIVE WASTES, SEPARATION PROCESSES, SIMULATION, STORAGE, WASTE DISPOSAL, WASTE MANAGEMENT, WASTE STORAGE, WASTES, WATER

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Nishimura, Tatsuo; Baba, Yumiko; Kinjo, Hidehito

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of problem or function: This modular package will calculate neutron flux in a geometry containing several void regions by using the removal diffusion method coupled with albedo-transport equation. 2 - Method of solution: Solution details are discussed in Ref.(1). 3 - Restrictions on the complexity of the problem: The principle restriction is the availability of adequate core storage. All large modules are variably dimensioned which means that array sizes are set for the particular problem being run at execution

Primary Subject

Source

20 Apr 1994; [html]; Available on-line: http://www.nea.fr/abs/html/nea-0836.html; Country of input: International Atomic Energy Agency (IAEA); 4 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Fowler, T.B.; Vondy, D.R.; Cunningham, G.W.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of problem or function: CITATION is designed to solve problems using the finite-difference representation of neutron diffusion theory, treating up to three space dimensions with arbitrary group-to-group scattering. X-y-z, theta-r-z, hexagonal-z, and trigonal-z geometries may be treated. Depletion problems may be solved and fuel managed for multi-cycle analysis. Extensive first-order perturbation results may be obtained given microscopic data and nuclide concentrations. Statics problems may be solved and perturbation results obtained with microscopic data. CITATION-2-3-VP2 is a vectorized version for FACOM VP-100 and VP-200 vector computers. 2 - Method of solution: Explicit, finite-difference approximations in space and time have been implemented. The neutron-flux-eigenvalue problems are solved by direct iteration to determine the multiplication factor or the nuclide densities required for a critical system. CITATION-2-3-VP2: Algorithms for the inner-outer iterative calculations are adapted to vector computers. The SLOR method, which is used in the original CITATION code, and the SOR method, which is adopted in the revised code, are vectorized by odd-even mesh ordering. 3 - Restrictions on the complexity of the problem: CITATION has been designed to attack problems which can be run in a reasonable amount of time. Storage of data is allocated dynamically to give the user flexibility in dimensioning. Typically, a finite-difference diffusion problem could have 200 depleting zones, 10,000 nuclide densities, and 30,000 space-energy point flux values

Primary Subject

Secondary Subject

Source

21 Apr 1995; [html]; Available on-line: http://www.nea.fr/abs/html/nesc0387.html; Country of input: International Atomic Energy Agency (IAEA); 13 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

ALGORITHMS, APPROXIMATIONS, C CODES, COMPUTER PROGRAM DOCUMENTATION, CRITICALITY, EIGENVALUES, FLEXIBILITY, FUEL MANAGEMENT, GEOMETRY, HEXAGONAL CONFIGURATION, ITERATIVE METHODS, MULTIGROUP THEORY, MULTIPLICATION FACTORS, NEUTRON DIFFUSION EQUATION, NEUTRON FLUX, ONE-DIMENSIONAL CALCULATIONS, PERTURBATION THEORY, THREE-DIMENSIONAL CALCULATIONS, TWO-DIMENSIONAL CALCULATIONS, WEBSITES

CALCULATION METHODS, COMPUTER CODES, CONFIGURATION, DIFFERENTIAL EQUATIONS, DIFFUSION EQUATIONS, DIMENSIONLESS NUMBERS, DOCUMENT TYPES, EQUATIONS, MANAGEMENT, MATHEMATICAL LOGIC, MATHEMATICS, MECHANICAL PROPERTIES, NEUTRON TRANSPORT THEORY, NUCLEAR MATERIALS MANAGEMENT, PARTIAL DIFFERENTIAL EQUATIONS, RADIATION FLUX, TENSILE PROPERTIES, TRANSPORT THEORY

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Bowers, H.I.; Gratteau, J.E.; Zielsinki, T.J.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of problem or function: The CONCEPT computer code system was developed to provide conceptual capital cost estimates for nuclear and coal-fired power plants. Cost estimates can be made as a function of plant type, size, location, and date of initial operation. The output includes a detailed breakdown of the estimate into direct and indirect costs similar to the accounting system described in document NUS-531. Cost models are provided in CONCEPT5, the fifth generation in the development of the CONCEPT package, for single-unit coal-fired plants, pressurized-water reactors, boiling- water reactors, liquid-metal-cooled reactors, and multi-unit coal- fired plants based on today's average or best operating experience. Costs may be obtained for any of twenty U.S. cities, a hypothetical Middletown site, and two Canadian cities. CONCEPT5 models are updated models of those available in CONCEPT3 and, in addition, this edition contains historical factory equipment cost data for the generation of cost indices and escalation rates; indirect costs are calculated as a function of unit size rather than a function of direct costs; and an indirect cost account for owner's costs and an improved time-dependent escalation feature are included. The CONCEPT3 models and cost data are outdated; the package is being retained in the library since it is the only UNIVAC1108 machine version of CONCEPT available and could prove helpful in converting the latest IBM release. 2 - Method of solution: CONCEPT is based on the premise that any central station power plant involves approximately the same major cost components regardless of location or date of initial operation. The program has detailed cost models for each plant type at a reference condition. Through use of size, time, and location- dependent cost adjustments, a reference cost model is modified to produce a specific capital cost estimate. CONCEPT is supported by two auxiliary programs--CONTAC, which generates and maintains the cost-model data file, and CONLAM, which generates and maintains the equipment, labor, and materials historical cost data file. 3 - Restrictions on the complexity of the problem: CONCEPT5 accepts power levels of 500 to 1500 MW(e) for nuclear plants and 300 to 1000 MW(e) for coal-fired plants, while CONCEPT3 accepts levels of 500 to 2000 MW(e). The auxiliary program CONLAM is limited to 30 time periods of historical cost data for 23 locations without program alterations

Primary Subject

Secondary Subject

Source

18 Feb 1992; [html]; Available on-line: http://www.nea.fr/abs/html/nesc0498.html; Country of input: International Atomic Energy Agency (IAEA); 8 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

CARBONACEOUS MATERIALS, COMPUTER CODES, COST, DOCUMENT TYPES, ENERGY SOURCES, ENRICHED URANIUM REACTORS, FOSSIL FUELS, FUELS, GAS COOLED REACTORS, GRAPHITE MODERATED REACTORS, MATERIALS, NUCLEAR FACILITIES, POWER PLANTS, POWER REACTORS, REACTOR MATERIALS, REACTORS, THERMAL POWER PLANTS, THERMAL REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Bennet, D.G.; Liew, S.K.; Mawbey, C.S.; Read, D.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: CHEMTARD is a computer program capable of simulating the migration of chemical species through saturated porous media. 2 - Method of solution: The code employs the direct method of coupling, whereby mass action terms are incorporated within transport equations, leading implicitly to strong interaction between equilibrium chemistry and hydrodynamic transport. Component models permit the treatment of aqueous complexation, transport by advection and/or diffusion, reversible precipitation-dissolution, sorption by ion-exchange or surface complexation, radioactive decay, redox reactions, multiple-layered transport and a wide range of boundary conditions

Primary Subject

Secondary Subject

Source

11 Aug 1993; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1346.html; Country of input: International Atomic Energy Agency (IAEA); 4 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Bennett, D.E.; Chanin, D.I.; Shiver, A.W.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: TOXRISK is an interactive program developed to aid in the evaluation of nuclear power plant control room habitability in the event of a nearby toxic material release. The program uses a model which is consistent with the approach described in the NRC Regulatory Guide 1.78. Release of the gas is treated as an initial puff followed by a continuous plume. The relative proportions of these as well as the plume release rate are supplied by the user. Transport of the gas is modeled as a Gaussian distribution and occurs through the action of a constant velocity, constant direction wind. Great flexibility is afforded the user in specifying the release description, meteorological conditions, relative geometry of the accident and plant, and the plant ventilation system characteristics. Two types of simulation can be performed: multiple case (parametric) studies and probabilistic analyses. Upon execution, TOXRISK presents a menu, and the user chooses between the Data Base Manager, the Multiple Case program, and the Probabilistic Study Program. The Data Base Manager provides a convenient means of storing, retrieving, and modifying blocks of data required by the analysis programs. The Multiple Case program calculates resultant gas concentrations inside the control room and presents a summary of information that describes the event for each set of conditions given. Optimally, a time history profile of inside and outside concentrations can also be produced. The Probabilistic Study program provides a means for estimating the annual probability of operator incapacitation due to toxic gas accidents on surrounding transportation routes and storage sites. 2 - Method of solution: Dispersion or diffusion of the gas during transport is described by modified Pasquill-Gifford dispersion coefficients

Primary Subject

Secondary Subject

Source

25 Feb 1993; [html]; Available on-line: http://www.nea.fr/abs/html/nesc9710.html; Country of input: International Atomic Energy Agency (IAEA); 2 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

COMPUTER PROGRAM DOCUMENTATION, COMPUTERIZED SIMULATION, CONTROL ROOMS, DIFFUSION, EVALUATION, GAUSS FUNCTION, METEOROLOGY, NUCLEAR POWER PLANTS, PARAMETRIC ANALYSIS, PLUMES, PROBABILISTIC ESTIMATION, PROBABILITY, REACTOR ACCIDENT SIMULATION, REACTOR ACCIDENTS, REGULATORY GUIDES, T CODES, TOXIC MATERIALS, TOXICITY, VENTILATION SYSTEMS, WEBSITES

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

Wong, R.L.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] Description of program or function: CNVUFAC, the General Dynamics heat-transfer radiation view program, was adapted for use on the LLNL computer system. The input and output were modified, and a node incrementing logic added for compatibility with TRUMP (NESC 771) thermal analyzer and related codes. The program performs the multiple integration necessary to evaluate the geometric black-body radiation node to node view factors. CNVUFAC uses an elemental area summation scheme to evaluate the multiple integrals. The program permits shadowing and self-shadowing. The basic configuration shapes that can be considered are cylinders, cones, spheres, ellipsoids, flat plates, disks, toroids, and polynomials of revolution. Portions of these shapes can also be considered. Card-image output containing node number and view factor information is generated for input to GRAY, a related code. GRAY performs the matrix manipulations necessary to convert black-body radiation heat-transfer view factors to gray-body view factors as required by thermal analyzer codes. The black-body view factors contain only geometric relationships. GRAY allows the effects of multiple gray-body reflections to be included. The resulting effective gray-body view factors can then be used with the corresponding fourth-power temperature differences to obtain the net radiative heat flux. GRAY accepts a matrix input or the card-image output generated by CNVUFAC. The resulting card-image GRAY output is in a form usable by TRUMP

Primary Subject

Secondary Subject

Source

2 May 1991; [html]; Available on-line: http://www.nea.fr/abs/html/nesc9911.html; Country of input: International Atomic Energy Agency (IAEA); 2 refs.

Record Type

Miscellaneous

Literature Type

Software

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

External URLExternal URL

1 | 2 | 3 | Next |