Application of the adjoint function methodology for neutron fluence determination
Conference
·
· Transactions of the American Nuclear Society; (United States)
OSTI ID:5982309
- Pennsylvania State Univ., University Park (United States)
- 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
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
ADJOINT DIFFERENCE METHOD
BUILDING MATERIALS
BWR TYPE REACTORS
CALCULATION METHODS
CONCRETES
CONTAINERS
DOSIMETRY
ENRICHED URANIUM REACTORS
IN CORE INSTRUMENTS
MATERIALS
MESH GENERATION
NEUTRAL-PARTICLE TRANSPORT
NEUTRON DOSIMETRY
NEUTRON FLUENCE
NEUTRON TRANSPORT
POWER REACTORS
PRESSURE VESSELS
PWR TYPE REACTORS
RADIATION TRANSPORT
REACTOR INSTRUMENTATION
REACTORS
SENSITIVITY ANALYSIS
SHIELDS
THERMAL REACTORS
THERMAL SHIELDS
THREE MILE ISLAND-1 REACTOR
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
ADJOINT DIFFERENCE METHOD
BUILDING MATERIALS
BWR TYPE REACTORS
CALCULATION METHODS
CONCRETES
CONTAINERS
DOSIMETRY
ENRICHED URANIUM REACTORS
IN CORE INSTRUMENTS
MATERIALS
MESH GENERATION
NEUTRAL-PARTICLE TRANSPORT
NEUTRON DOSIMETRY
NEUTRON FLUENCE
NEUTRON TRANSPORT
POWER REACTORS
PRESSURE VESSELS
PWR TYPE REACTORS
RADIATION TRANSPORT
REACTOR INSTRUMENTATION
REACTORS
SENSITIVITY ANALYSIS
SHIELDS
THERMAL REACTORS
THERMAL SHIELDS
THREE MILE ISLAND-1 REACTOR
WATER COOLED REACTORS
WATER MODERATED REACTORS