218-group neutron cross-section library in the AMPX master interface format for criticality safety studies

Ford, William E.; Webster, C.C.; Westfall, R.M.

doi:10.2172/7155403

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Gross Section D E V E L O P M E N T F R O M Ehdf/b To 25 Groups2

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Cited by 5 publications

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“…These weighting spectra were then used to collapse the 70 group cross section set into a 25 group ANISN library tape. (Ford et al, 1976). (Ford et al, 1976 (Honeck, 1961).…”

Section: Gross Section D E V E L O P M E N T F R O M Ehdf/b To 25 Groupsmentioning

confidence: 99%

“…(Ford et al, 1976). (Ford et al, 1976 (Honeck, 1961). This step was necessary since no File 7 data was available for Na-23.…”

Section: Gross Section D E V E L O P M E N T F R O M Ehdf/b To 25 Groupsmentioning

confidence: 99%

“…ENBF/B-IV to 218 GroupsThe 218 group master cross section library used in this study is a composite of non-resonance nuclei drawn from the 218 group OENL library(Ford, Webster, and Westfall, 1976), and resonance nuclei processed from ENDF/B-IV by XLACS. The 218 group energy structure, which includes 140 epithermal and 78 thermal groups, is included inAppendix A.…”

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confidence: 99%

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Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

Ramsower¹

1979

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“…These weighting spectra were then used to collapse the 70 group cross section set into a 25 group ANISN library tape. (Ford et al, 1976). (Ford et al, 1976 (Honeck, 1961).…”

Section: Gross Section D E V E L O P M E N T F R O M Ehdf/b To 25 Groupsmentioning

confidence: 99%

“…(Ford et al, 1976). (Ford et al, 1976 (Honeck, 1961). This step was necessary since no File 7 data was available for Na-23.…”

Section: Gross Section D E V E L O P M E N T F R O M Ehdf/b To 25 Groupsmentioning

confidence: 99%

mentioning

confidence: 99%

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Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

Ramsower¹

1979

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“…Ratio of New '''Pu Data to ENDF/B-IV '''Pu Data ENDF/B-IV Capture Data for '''Pu 3. ENDF/B-IV Absorption Cross Section for '''Pu 1 .…”

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confidence: 99%

Integral data evaluation of stainless steel, /sup 239/Pu, /sup 240/Pu, and H/sub 2/O for homogeneous plutonium systems

Jenquin¹,

Thompson²,

Trapp³

et al. 1979

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Calculation of the Neutron Source Distribution in the Venus PWR Mockup Experiment

et al. 1985

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218-group neutron cross-section library in the AMPX master interface format for criticality safety studies

Cited by 5 publications

References 0 publications

Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

Integral data evaluation of stainless steel, /sup 239/Pu, /sup 240/Pu, and H/sub 2/O for homogeneous plutonium systems

Calculation of the Neutron Source Distribution in the Venus PWR Mockup Experiment

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