Online Catalog
Click on Package Name to get detailed information.
Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.

Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
FEDC -- US Government Agencies and Their Contractors Only
OECD -- Restricted/See Abstract
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages starting with N
Package NameAbstractRSICC TapelistTitle
NAAPROAbstractC00722 PC586 00Neutron Activation Analysis PRognosis and Optimization Code System.
NABAbstractD00018 I0360 00100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum.
NACAbstractC00164 C0000 00Neutron Activation Analysis and Product Isotope Inventory Code System.
NACAbstractC00164 IBMMF 00Neutron Activation Analysis and Product Isotope Inventory Code System.
NAC-PCAbstractC00164 IBMPC 00Neutron Activation Analysis and Product Isotope Inventory Code System.
NACTAbstractC00502 U1100 00Screening Program for Neutron Activation Products.
NAISAPAbstractP00085 F2306 00Theory and Use of Gamma-Ray Spectrum Analysis Codes for NaI(Tl) Detectors.
NANICKAbstractP00120 I0360 00Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B.
NAPAbstractC00101 I7090 00Multigroup Time-Dependent Neutron Activation Prediction Code.
NASIF-NARESAbstractP00121 I0360 00A Code System for Computing Shielding Factors from ENDF/B Tapes.
NAUA-MOD5 NAUA-MOD5/MAbstractP00556 MNYCP 00Aerosols in Reactor Containment During Meltdown.
NCRP49AbstractC00462 IBMPC 00X-Ray Shield Calculation System.
NCSP-DATAbstractM00002 MNYCP 01Nuclear Data in Support of the Nuclear Criticality Safety Program.
NEACRP-H2O-LATTICESAbstractD00265 MNYCP 00Compilation of Reactor Physics Measurements in LWRs Lattices.
NE-SPECAbstractP00150 F2307 00A Code System for Unfolding a Pulse Height Distribution of Neutrons Measured by an NE-213 Organic Scintillator.
NESTLE 5.2.1AbstractC00641 MNYCP 04Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source
NEUPACAbstractP00177 FM200 00Neutron Unfolding Code System for Calculating Neutron Flux Spectra from Activation Data of Dosimeter Foils.
NEVEMORAbstractP00026 I3675 00Multigroup-Multiregion Calculation of Flux Spectra and Energy Deposition for Fast Neutrons.
NITRANAbstractC00582 FM380 00Neutron Transport Code System Based On Anisotropic Scattering.
NJOY91.119AbstractP00171 MFMWS 04Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY94.61AbstractP00355 MFMWS 03Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY97.0AbstractP00368 MNYCP 00Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY99.0AbstractP00480 MNYCP 00Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY-UTIL-EIRAbstractP00296 C0825 00Utilities For the NJOY (6/83) Nuclear Data Processing System.
NMTC/JAERI97AbstractC00694 SUN05 00Monte Carlo Nucleon Meson Transport Code System.
NMTC/JAMAbstractC00717 PC586 00High Energy Particle Transport Code System.
NONSAP-CAbstractP00458 C7600 00Code System for Analysis of 3-D Reinforced Concrete Structures.
NORMAAbstractP00471 PC586 00Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.
NORMA-FPAbstractP00470 PC586 00Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.
NOXAbstractD00017 I0360 00199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen.
NPCSL-81AbstractD00082 I0370 00Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II.
NPTXSAbstractP00090 I0360 00Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters.
NRCDOSE 2.3.20AbstractC00684 PC586 14Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with Windows Interface.
NRCDOSE72V1.2.3AbstractC00768 PCX86 03Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with Windows Interface.
NRCPAGEAbstractP00491 DVX11 00Code System to Detect Recurring Loss of Special Nuclear Materials.
NRCPIPES 2.0AAbstractP00429 IBMPC 00Code System for Fracture Mechanics Analysis of Circumferential Surface Cracks in Pipes.
NRNAbstractC00054 C6600 00Multigroup Removal-Diffusion Code System for Planes, Cylinders and Spheres.
NSLINKAbstractP00314 D0VAX 00NJOY SCALE LINK.
NUCCONAbstractC00439 S7800 00A Code System for Calculation of Time-Dependent Nuclide Concentrations, Activity, Gamma-Ray Dose Rate and Biological Hazard Potential of Fusion Reactor Materials Due to Neutron Irradiation.
NUCDECAYAbstractD00172 PC386 01Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP and MIRD.
NUCDECAYCALCAbstractD00202 PC586 00Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP. See newer version in RASCAL (CCC-553).
NUCHARTAbstractP00545 IBMPC 00Nuclear Properties and Decay Data Chart of Nuclides.
NUFACEAbstractP00284 CYXMP 00An Interface Code For The Calculation of Nuclear Responses.
NUGAM 2&3 SSLABAbstractC00210 I0360 00Monte Carlo Prediction of Photon Transport Distributions.
NUTRANAbstractC00675 I0370 00Code System for Long-Term Repository Safety Analysis.
NX1-NX2AbstractP00310 D0VAX 00Code System to Calculate Excitation Functions for (n,charged particle) Reactions.
The Radiation Safety Information Computational Center (RSICC) is a Department of Energy Specialized Information Analysis Center (SIAC) authorized to collect, analyze, maintain, and distribute computer software and data sets in the areas of radiation transport and safety. RSICC resides in the Reactor and Nuclear Systems Division (RNSD) at Oak Ridge National Laboratory.