Proceedings of the Korean Nuclear Society Conference (한국원자력학회:학술대회논문집)
- 1998.05b
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- Pages.771-776
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- 1998
Calculation of Reactor Pressure Vessel Fluence Using TORT Code
- Shin, Chul-Ho (Hanyang University) ;
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Kim, Jong kyung
(Hanyang University)
- Published : 1998.05.01
Abstract
TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Vnit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) far all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library. BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The makimum fast neutron nun calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effgctive full power days is 1.784x10
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