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

Abstract

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.

Authors:
;  [1];  [2];  [3]
  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)
Publication Date:
Research Org.:
US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
Nuclear Regulatory Commission, Washington, DC (United States)
OSTI Identifier:
10168039
Report Number(s):
NUREG/CR-6206; ORNL/TM-12693
ON: TI94015398; TRN: 94:015448
DOE Contract Number:  
AC05-84OR21400
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Jul 1994
Country of Publication:
United States
Language:
English
Subject:
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

Citation Formats

Asgari, M, Williams, M L, Kam, F B.K., and McGarry, E D. Transport calculations of radiation exposure to vessel support structures in the Trojan Reactor. United States: N. p., 1994. Web. doi:10.2172/10168039.
Asgari, M, Williams, M L, Kam, F B.K., & McGarry, E D. Transport calculations of radiation exposure to vessel support structures in the Trojan Reactor. United States. https://doi.org/10.2172/10168039
Asgari, M, Williams, M L, Kam, F B.K., and McGarry, E D. 1994. "Transport calculations of radiation exposure to vessel support structures in the Trojan Reactor". United States. https://doi.org/10.2172/10168039. https://www.osti.gov/servlets/purl/10168039.
@article{osti_10168039,
title = {Transport calculations of radiation exposure to vessel support structures in the Trojan Reactor},
author = {Asgari, M and Williams, M L and Kam, F B.K. and McGarry, E D},
abstractNote = {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.},
doi = {10.2172/10168039},
url = {https://www.osti.gov/biblio/10168039}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jul 01 00:00:00 EDT 1994},
month = {Fri Jul 01 00:00:00 EDT 1994}
}