Transport calculations of radiation exposure to vessel support structures in the Trojan Reactor

Technical Report · Fri Jul 01 00:00:00 EDT 1994

DOI:https://doi.org/10.2172/10168039· OSTI ID:10168039

Asgari, M; Williams, M L ^[1]; Kam, F B.K. ^[2]; McGarry, E D ^[3]

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1. Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center
2. Oak Ridge National Lab., TN (United States)
3. National Inst. of Standards and Technology, Gaithersburg, MD (United States)

Comparison of transport calculations of the dosimeter activities with the experimental measurements shows that the values obtained with ENDF/B-VI cross-section data overestimate the measured results for high-energy-threshold reactions in the cavity by up to 41%, and thermal reactions by up to a factor of 3.0. The transport calculations performed with the original SAILOR cross-section library (based on ENDF/B-VI data) overestimate measured threshold reactions by only 15% and the thermal reactions by about a factor of 2.50. These results are inconsistent with those obtained in earlier studies that compared transport calculations done with SAILOR vs ENDF/B-VI, which indicate that SAILOR tends to underestimate cavity dosimeter activities for threshold reactions, while the ENDF/B-VI values usually agree better with experimental results. One factor that probably contributes to the rather large discrepancy between the computed and measured activities is the core power distribution used in the transport calculations. Because of unavailability of plant-specific data, a generic power distribution provided by Westinghouse was used. Since the calculated cavity flux levels appear to be over-estimated, the results estimated for the exposure to the support structure should be conservative.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10168039
• Report Number(s): NUREG/CR-6206; ORNL/TM-12693; ON: TI94015398; TRN: 94:015448
• Resource Relation: Other Information: PBD: Jul 1994
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
TROJAN REACTOR
RADIATION TRANSPORT
NUCLEAR DATA COLLECTIONS
CAVITIES
POWER DISTRIBUTION
CALCULATION METHODS
NEUTRON FLUX
DOSIMETRY
COMPUTER CALCULATIONS
210200
220100
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
THEORY AND CALCULATION