Modifying scoping codes to accurately calculate TMI-cores with lifetimes greater than 500 effective full-power days
- Pennsylvania State Univ., University Park (United States)
- GPU Nuclear Corp., Parsippany, NJ (United States)
The Three Mile Island unit 1 core reloads have been designed using fast but accurate scoping codes, PSUI-LEOPARD and ADMARC. PSUI-LEOPARD has been normalized to EPRI-CPM2 results and used to calculate the two-group constants, whereas ADMARC is a modern two-dimensional, two-group diffusion theory nodal code. Problems in accuracy were encountered for cycles 8 and higher as the core lifetime was increased beyond 500 effective full-power days. This is because the heavier loaded cores in both {sup 235}U and {sup 10}B have harder neutron spectra, which produces a change in the transport effect in the baffle reflector region, and the burnable poison (BP) simulations were not accurate enough for the cores containing the increased amount of {sup 10}B required in the BP rods. In the authors study, a technique has been developed to take into account the change in the transport effect in the baffle region by modifying the fast neutron diffusion coefficient as a function of cycle length and core exposure or burnup. A more accurate BP simulation method is also developed, using integral transport theory and CPM2 data, to calculate the BP contribution to the equivalent fuel assembly (supercell) two-group constants. The net result is that the accuracy of the scoping codes is as good as that produced by CASMO/SIMULATE or CPM2/SIMULATE when comparing with measured data.
- OSTI ID:
- 6966406
- Report Number(s):
- CONF-920606-; CODEN: TANSA
- Journal Information:
- Transactions of the American Nuclear Society; (United States), Vol. 65; Conference: American Nuclear Society annual meeting, Boston, MA (United States), 7-12 Jun 1992; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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THREE MILE ISLAND-1 REACTOR
FUEL MANAGEMENT
A CODES
ACCURACY
BURNUP
E CODES
FUEL ELEMENTS
GROUP CONSTANTS
L CODES
MODERATORS
MULTIGROUP THEORY
MULTIPLICATION FACTORS
NEUTRON FLUX
P CODES
REACTOR CORES
REACTOR PHYSICS
TWO-DIMENSIONAL CALCULATIONS
COMPUTER CODES
CROSS SECTIONS
ENRICHED URANIUM REACTORS
NEUTRON TRANSPORT THEORY
PHYSICS
POWER REACTORS
PWR TYPE REACTORS
RADIATION FLUX
REACTOR COMPONENTS
REACTORS
THERMAL REACTORS
TRANSPORT THEORY
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200* - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled