A Broad-Group Cross-Section Library Based on ENDF/B-VII.0 for Fast Neutron Dosimetry Using the CPXSD Methodology

2012 ◽  
pp. 548-560 ◽  
Author(s):  
F. A. Alpan
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Neutron Dosimetry ◽  
Broad Group

Ex-Vessel Neutron Dosimetry Results in the Vicinity of RPV Supports

2007 ◽  
Author(s):  
Daniel N. Hopkins ◽  
Eugene T. Hayes ◽  
Arnold H. Ferro
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Pressure Vessel ◽  
Transport Theory ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Neutron Dosimetry ◽  
Pressurized Water ◽  
Reactor Pressure

Neutron-induced embrittlement of the reactor pressure vessel has been a long standing concern for pressurized water reactors (PWR). To date, the beltline region of the pressure vessel, defined as the portion of the pressure vessel experiencing fast neutron fluence (E > 1.0 MeV) equal to or greater than 1017 n/cm2, has been the primary focus of evaluations assessing this embrittlement. These evaluations typically include a calculation of the neutron flux incident on the reactor pressure vessel beltline region, which is in part validated by direct comparison with dosimetry measurements. Two general types of measurements are commonly used, those being dosimetry sets that are included as part of the in-vessel surveillance capsules, and at some plants, those that are included in supplemental surveillance programs such as Ex-Vessel Neutron Dosimetry. In the context of life extension, the beltline region as defined above is getting larger. Present fluence calculations for a number of plants indicate that beltline region at the end of the 60 years of operation will extend to the bottom of the reactor pressure vessel nozzle welds. This extended beltline creates a new problem in terms of validating the neutron fluence calculations in this region well above the top of the active fuel, in that there are no measurements available to confirm calculated results in this new region of interest. Prior to the start-up of Cycle 11 at Comanche Peak Unit 1, an Ex-Vessel Neutron Dosimetry Program was initiated. This program included placement of neutron dosimetry sensor sets in the vicinity of the reactor pressure vessel supports. At the conclusion of Cycle 11, the first set of dosimetry was replaced and the irradiated set analyzed. The Ex-Vessel Neutron Dosimetry set from Cycle 11 was analyzed using a 2D/1D flux synthesis technique using the two dimensional discrete ordinates transport theory calculations (DORT) along with the BUGLE 96 cross-section library and the SNLRML neutron dosimetry cross-section library. The measurements in the vicinity of the vessel supports compare well with the transport calculations, thus confirming that the expected fast neutron fluence (E > 1.0 MeV) in the vicinity of the reactor vessel supports is below the 1018 n/cm2.

scholarly journals Measurement of Fast Neutron Induced Fission Cross Section for Actinide Nuclides

1989 ◽  
Vol 26 (1) ◽  
pp. 11-14 ◽  
Author(s):  
Mamoru BABA ◽  
Naohiro HIRAKAWA ◽  
Tomohiko IWASAKI ◽  
Kazutaka KANDA
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Fission Cross Section ◽  
Induced Fission

Fast Neutron Induced Fission Cross Section for Pu-239

1983 ◽  
pp. 36-38 ◽  
Author(s):  
Zhou Xianjian ◽  
Yan Wuguang ◽  
Zhou Huiming ◽  
Deng Xinlu ◽  
Rong Chaofan ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Fission Cross Section ◽  
Induced Fission
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