ANALYSIS OF FUEL LONGTERM EFFECTS USING SUBCHANNEL CODES
Main Article Content
Abstract
Thermohydraulic computer codes such as COBRA3C are now becoming in practical use as part of power plant licensing process. One of the areas of interest to reactor designer is the effect due to longterm operations and the changes that might occur either due to irradiation (burnup) effects or thermal stresses in the reactor. The present study is an attempt to predict coolant channel temperature profiles at the core End Of Life. In addition deeper insight is given to the individual single partial contribution to the changes in Begin Of Life and End Of Life profiles. The effects due to fuel swelling, rod bowing, and changes in physical properties are studied in details using COBRA3C code. The codes COBRA3C and the kinetic code AIREKMOD together a structure routine are joined in a modular system. As a result coupling between the different transient parameters in normal and abnormal operation conditions could be performed. Results of such coupling are shown to improve the prediction of DNBR.
Article Details
How to Cite
Publication Dates
References
AL AHMADY, ( 988), ‘transient thermal hydraulic Calculation for PWP Reactor A Diss. Nucl Eng Dept Baghdad Uni.
AMIN .E.A. et al, ( 99 ) ‘irraadiation induced changes in fuel elements and their effect on nucleus reactor power performance. submitted to the present procceding
BOURING, R.W, HAMBO, (1967), "A Computer Program for the subchannel Analysis of Hydraulic and Burnout Characteristics of Rod Clusters ' AEEW-R 524 .
C.L.WHEELER. et.al, ( 976), ’COBRA-IV: An Interim Version of COBRA for Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Co ’ BNWL-1962 MARS.
E AMIN, ( 99 ) ‘A Mo l Sy t m o Th mohy l , K n t n St t o pl n ‘to b p bl h
FAJEAU, M., (1969), " Programme FLICA Etude Thermodynamique d'un Reacteur ou ’ n Bo ; ’E ’ CEA-R- 7 6, C nt ’Et N l , S l y F n
HOFMAN et.al ( 98 ) ‘I t on B h v o o th Exp m nt l M k-II EBR II D v F l’ N l T hnol 47, 7
MA B M ( 98 ), N l R to M t l n Appl t on ’ V n No t n R nhol .
MILLER C S ‘LOFT Co D n R po t’ TID -4500.
PAUL R, BETTEN ( 986), ‘Bow n OF El m nt n h n n R l Cool nt Temperature Profile Over Subassemb1y' Trans .Am .Nucl, SOC. 52,532.
ROWE.D.S. (1973), " COBRA-3C: A Digital Computer Program for Steady State and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Element .BNWL-1695 MARCH.