Verification of the multi-group generation capability of FRENDY nuclear data processing code for recent nuclear data through comparison of one-group reaction rates

Yamamoto, Akio*; Tada, Kenichi ; Chiba, Go*; Endo, Tomohiro*

Verification calculations for the capability of multi-group cross section generation in FRENDY (FRENDY/MG) are carried out through the comparison of one-group reaction rates using the multi-group cross sections obtained by FRENDY/MG and NJOY2016. Three different neutron spectra (LWR, FR, and 1/E) are used to calculate one-group reaction rates. The discrepancies of one-group reaction rates are small for most cases, showing the validity of FRENDY/MG. The FRENDY/MG will be released as the part of FRENDY nuclear data processing system in the near future.

Language	:	English
Journal	:	Transactions of the American Nuclear Society
Volume	:	124
Number	:	1
Pages	:	p.544 - 547
Publication Year/Month	:	2021/06
Meeting title	:	2021 American Nuclear Society Virtual Annual Meeting
Held date	:	2021/06
Location (city)	:	online
Location (country)	:	U.S.A.
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Keywords	:	FRENDY; JENDL; Verification
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Registration No. : BB20201122
JAEA Abstracts No. : 50000737
Paper Submission No. : 25467

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