Compendium of benchmark neutron fields for reactor dosimetry :: standard neutron field entries

[1]  J. Grundl Examination of 10B (n, He) and 6Li (n, He) cross section measurements in reactor physics benchmarks , 1986 .

[2]  W. Mannhart Spectrum-Averaged Neutron Cross Sections Measured in the U-235 Fission-Neutron Field in Mol , 1985 .

[3]  G. P. Lamaze,et al.  THE U.S. U-235 FISSION SPECTRUM STANDARD NEUTRON FIELD REVISITED , 1985 .

[4]  W. Mannhart Recent Experiments on Cf-252 Spectrum-Averaged Neutron Cross Sections , 1985 .

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[6]  R. A. Forster,et al.  MCNP - a general Monte Carlo code for neutron and photon transport , 1985 .

[7]  A. Carlson Standard cross-section data , 1984 .

[8]  Li Linpei Reflection of 252Cf Fission Neutrons from a Concrete Floor , 1983 .

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[10]  G. P. Lamaze,et al.  Integral Reaction Rate Measurements In 252Cf and 235U Fission Spectra , 1983 .

[11]  D. Madland,et al.  Prompt fission neutron spectra and average prompt neutron multiplicities , 1982 .

[12]  G. Winkler,et al.  Measurement of the Average Activation Cross Section for the Reaction 63 Cu( n, α ) 60 Co in the Spontaneous Fission Neutron Field of Californium-252 , 1981 .

[13]  W. Mannhart Progress in Integral Data and Their Accuracy: Average Neutron Cross Sections in the Californium-252 Benchmark Field , 1981 .

[14]  C. O. Cogburn,et al.  Wall return of 252 Cf neutrons at the SEFOR Calibration Center , 1981 .

[15]  J. Wagschal,et al.  Evaluation of the new ISNF one-dimensional model , 1981 .

[16]  R. Labauve,et al.  Neutronic analysis of the NBS Intermediate-Energy Standard Neutron Field (ISNF) , 1979 .

[17]  J. Wagschal,et al.  NBS intermediate-energy standard neutron field (ISNF) revisited , 1979 .

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[19]  G. Knoll,et al.  Fission cross-sections of 235U and 239Pu averaged over 252Cf neutron spectrum , 1978 .

[20]  C. Eisenhauer,et al.  Neutron transport calculations for the intermediate-energy standard neutron field (ISNF) at the National Bureau of Standards , 1977 .

[21]  D. Gilliam Integral measurement results in standard fields , 1977 .

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[23]  L. S. Kellogg,et al.  Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields , 1976 .

[24]  P. Young,et al.  Light Element Standard Cross Sections for ENDF/B Version IV , 1976 .

[25]  L. M. Petrie,et al.  AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B , 1976 .

[26]  C. Eisenhauer,et al.  Fundamental integral cross section ratio measurements in the thermal-neutron-induced uranium-235 fission neutron spectrum , 1975 .

[27]  C. Eisenhauer,et al.  Fission spectrum neutrons for cross section validation and neutron flux transfer , 1975 .

[28]  T. Hill ONETRAN: a discrete ordinates finite element code for the solution of the one-dimensional multigroup transport equation , 1975 .

[29]  R. B. Schwartz,et al.  Total neutron cross section of carbon from 1 keV to 15 MeV , 1975 .

[30]  J. Jenkins,et al.  WALL RETURN NEUTRON FLUXES FOR HIGH- AND INTERMEDIATE-ENERGY CAVITY NEUTRON SOURCES. , 1972 .

[31]  Grundl FISSION-NEUTRON SPECTRA: MACROSCOPIC AND INTEGRAL RESULTS. , 1971 .

[32]  L. Lesca,et al.  THRESHOLD REACTION EXCITATION FUNCTIONS INTERCALIBRATED IN A PURE FISSION SPECTURM. , 1970 .

[33]  J. Grundl CAVITY FISSION SPECTRA , 1962 .