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Application of the adjoint function methodology for neutron fluence determination

Conference · · Transactions of the American Nuclear Society; (United States)
OSTI ID:5982309
; ;  [1]; ;  [2]
  1. Pennsylvania State Univ., University Park (United States)
  2. GPU Nuclear Corp., Parsippany, NJ (United States)
In previous studies, the neutron fluence at a reactor pressure vessel has been estimated based on consolidation of transport theory calculations and experimental data obtained from in-vessel capsules and/or cavity dosimeters. Normally, a forward neutron transport calculation is performed for each fuel cycle and the neutron fluxes are integrated over the reactor operating time to estimate the neutron fluence. Such calculations are performed for a geometrical model which is composed of one-eighth (0 to 45 deg) of the reactor core and its surroundings; i.e., core barrel, thermal shield, downcomer, reactor vessel, cavity region, concrete wall, and instrumentation well. Because the model is large, transport theory calculations generally require a significant amount of computer memory and time; hence, more efficient methodologies such as the adjoint transport approach have been proposed. These studies, however, do not address the necessary sensitivity studies needed for adjoint function calculations. The adjoint methodology has been employed to estimate the activity of a cavity dosimeter and that of an in-vessel capsule. A sensitivity study has been performed on the mesh distribution used in and around the cavity dosimeter and the in-vessel capsule. Further, since a major portion of the detector response is due to the neutrons originated in the peripheral fuel assemblies, a study on the use of a smaller calculational model has been performed.
OSTI ID:
5982309
Report Number(s):
CONF-910603--
Conference Information:
Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 63
Country of Publication:
United States
Language:
English