157 results on '"group constants"'
Search Results
2. PARTI; optimal group or mesh collapsing. [CDC6500; FORTRAN IV]
3. ETOG1; ENDF/B to MUFT, GAM, ANISN corss sections format. [CDC6600; FORTRAN IV]
4. ENDRUN2; multigroup constants from ENDF/B data. [GE635; FORTRAN IV]
5. HAMMER; LITHE; HELP; critical analysis system. [IBM360; FORTRAN IV and BAL]
6. GROUSE; space-dependent cross section generation. [GE625,635; FORTRAN IV]
7. LASL group-averaged cross-sections; SN 18- 24- and 25-group sets. [CDC6600; FORTRAN IV]
8. GAMTEC2; multigroup constant calculations for 0 to 10 MeV. [UNIVAC1107; GE625; CDC6400; FORTRAN IV]
9. MC-2; fast neutron spectra and multigroup cross section. [IBM360,370; CDC7600; FORTRAN IV (99nd Assembly language (1IBM370), FORTRAN IV (99) and COMPASS (1) (CDC7600)]
10. GLEN; group constant calculations from TOR output data. [CDC6600; FORTRAN IV]
11. ETOX3; multigroup constants from ENDF/B for one dimension. [UNIVAC1108; IBM360,370; FORTRAN IV]
12. ETOM1; ENDF/B format to MUFT format cross sections. [CDC6600; FORTRAN IV]
13. APRFX1; 99-group DLC-2B library group-collapsing. [CDC6600; FORTRAN IV]
14. PHROG; multi-group constant and fast spectra calculations. [IBM360; FORTRAN IV]
15. EXTERMINATOR2; two-dimensional, multigroup diffusion program diffusion program. [IBM360; GE625; CDC6600; FORTRAN IV]
16. AILMOE; cross section calculation of elastic scattering resonancess. [IBM360; FORTRAN IV (H)]
17. SUPERTOG; ENDF/B fine-group constants generation. [IBM360; FORTRAN IV (ANSI FORTRAN)]
18. ARC-XSEC1; microscopic cross section manipulation. [IBM360; FORTRAN IV]
19. Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model''
20. Final Technical Report
21. An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model. Nuclear Energy Research Initiative (NERI) Program. Quarterly Technical Progress Report, July 1 - September 30, 2000
22. An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model. Quarterly Technical Progress Report, April 1 - June 30, 2000
23. An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model. Nuclear Energy Research Initiative (NERI) Program. Quarterly Technical Progress Report
24. Generation of few-group constants by Monte Carlo code cosRMC.
25. Analysis of the uncertainties in the physical calculations of water-moderated power reactors of the VVER type by the parameters of models of preparing few-group constants
26. Generation of SFR few-group constants using the Monte Carlo code Serpent
27. Development status of the lattice physics code in COSINE project
28. A multi-group Monte Carlo core analysis method and its application in SCWR design
29. Interpolations of nuclide-specific scattering kernels generated with Serpent
30. Revised methods for few-group cross sections generation in the Serpent Monte Carlo code
31. Uncertainty quantification of few group diffusion theory constants generated by the B1 theory-augmented Monte Carlo method
32. Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code.
33. Use and Impact of Covariance Data in the Japanese Latest Adjusted Library ADJ2010 Based on JENDL-4.0
34. Determination of delayed neutrons source in the frequency domain based on in-pile oscillation measurements
35. Homogenized group cross sections by Monte Carlo
36. Production of Multigroup data covariance in the resonance range by Monte-Carlo calculations
37. Current status of the PSG Monte Carlo neutron transport code
38. RZ calculations for self shielded multigroup cross sections
39. Development of two-step procedure for the prismatic VHTR physics analysis
40. Comparison of 3-D deterministic parallel and Monte Carlo transport computations for special nuclear materials assessments
41. Group constants generation by Monte Carlo code MCS for LWR analysis.
42. Calculation of reactivities using ionization chamber currents with different sets of kinetic parameters for reduced scram system efficiency in the VVER-1000 of the third unit of the Kalinin nuclear power plant at the stage of physical start-up
43. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors
44. Application of Covariances to Fast Reactor Core Analysis
45. Universal bounds for SU(3) low energy constants
46. Data processing for power reactor fuel cycle codes
47. Group structure and weighting function effects on neutron penetration through thick sodium-iron shields
48. Background cross section method as a general tool for reactor analysis
49. Estimated uncertainties in nuclear data: an approach
50. Fine structure effects and the central worth discrepancy
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