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ASTM Selected Technical Papers
Reactor Dosimetry
By
H Farrar, IV IV,
H Farrar, IV IV
1
Symposium General Chairman Consultant Bell Canyon, California
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EP Lippincott,
EP Lippincott
2
Program Committee Chairman
?Westinghouse
?Pittsburgh, Pennsylvania
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JG Williams,
JG Williams
3
Program Committee Vice Chairman
?University of Arizona
?Tucson, Arizona
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DW Vehar
DW Vehar
4
Scientific Secretary
?Sandia National Laboratories
?Albuquerque, New Mexico
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ISBN-10:
0-8031-1899-6
ISBN:
978-0-8031-1899-7
No. of Pages:
131
Publisher:
ASTM International
Publication date:
1994
eBook Chapter
Maine Yankee Dosimetry Capsule and Pressure Vessel Neutron Fluence Calculations
By
JR White
,
JR White
1
Professor
, Chemical and Nuclear Engineering Department, University of Massachusetts-Lowell
, Lowell, MA 01854
.
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KB Spinney
,
KB Spinney
2
Senior Engineer, Senior Nuclear Engineer and Manager
respectively, Yankee Atomic Electric Co.
, 580 Main St., Bolton, MA 01740
.
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KJ Morrissey
,
KJ Morrissey
2
Senior Engineer, Senior Nuclear Engineer and Manager
respectively, Yankee Atomic Electric Co.
, 580 Main St., Bolton, MA 01740
.
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RJ Cacciapouti
RJ Cacciapouti
2
Senior Engineer, Senior Nuclear Engineer and Manager
respectively, Yankee Atomic Electric Co.
, 580 Main St., Bolton, MA 01740
.
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Page Count:
9
-
Published:1994
Citation
White, J, Spinney, K, Morrissey, K, & Cacciapouti, R. "Maine Yankee Dosimetry Capsule and Pressure Vessel Neutron Fluence Calculations." Reactor Dosimetry. Ed. Farrar, H, IV, Lippincott, E, Williams, J, & Vehar, D. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959 : ASTM International, 1994.
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In-house capability for deterministic neutron and gamma transport analyses has been implemented at Yankee Atomic Electric Company (YAEC). A detailed R-Theta (R-θ) calculational model of Maine 4ankee was developed to help in validation of the methods and to establish appropriate models for support of the ongoing Maine Yankee pressure vessel surveillance program. Several data and modeling sensitivity studies were performed and comparisons to measured dosimetry capsule data were emphasized. The calculated results establish confidence in the YAEC in-house computational methodology for general pressure vessel fluence analyses.
References
1.
“
TORT - A Three Dimensional Discrete Ordinates Transport Code
,” Radiation Shielding Information Center Computer Code Collection, CCC-543, Oak Ridge National Laboratory (1990
). This also contains the DORT code and documentation.2.
Anderson
S. L.
, “Summary of Fast Neutron Exposure Evaluations for the Maine Yankee Reactor Pressure Vessel Through Fuel Cycle 11
,” Westinghouse Electric Corporation, WCAP-11335, Rev.l (1991
).3.
“
SAILOR - A Coupled Self-Shielded 47 Neutron 20 Gamma Ray P3 Cross Section Library for Light Water Reactors
,” Radiation Shielding Information Center Data Library Collection, DLC-76, Oak Ridge National Laboratory
(1987
).4.
“
BUGLE-80 - A Coupled 47 Neutron 20 Gamma Ray P3 Cross Section Library for LWR Shielding Calculations
,” Radiation Shielding Information Center Data Library Collection, DLC-75, Oak Ridge National Laboratory
(1981
).5.
“
LEPRICON - PWR Pressure Vessel Surveillance Dosimetry Analysis System
,” Radiation Shielding Information Center Peripheral Shielding Routine Collection, PSR-277, Oak Ridge National Laboratory
(1990
).6.
“
SIMULATE-3P Advanced Three Dimensional Two-Group Reactor Analysis Code
,” Studsvik of America, Studsvik/SOA-92/01 (1992
).7.
“
CASMO-3, A Fuel Assembly Burnup Program
,” Studsvik Energitenik, Studsvik/NFA-86/7 (1986
).8.
Basha
H. S.
and Manahan
M. P.
, “A Comparison of the BUGLE-80, SAILOR, and ELXSIR Neutron Cross-Section Libraries for PWR Pressure Vessel Surveillance Dosimetry and Shielding Applications
,” Nuclear Technology
, 100
, 79 (1992
).9.
Haghighat
A.
and Veerasingam
R.
, “Transport Analysis of Several Cross-Section Libraries Used for Reactor Pressure Vessel Fluence Calculations
,” Nuclear Technology
, 101
, 237 (1993
).
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