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in creator/contributor: "Chung, K.S."

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Fuel and control rod failure behavior during degraded core accidents. [PWR; BWR]

As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accid… more

Date: January 1, 1984

Creator: Chung, K.S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Study of turbulent natural-circulation flow and low-Prandtl-number forced-convection flow. [LMFBR]

Calculational methods and results are discussed for the coupled energy and momentum equations of turbulent natural circulation flow and low Prandtl number forced convection flow. The objective of this paper is to develop a calculational method for the study of the thermal-hydraulic behavior of coolant flowing in a liquid metal fast breeder reactor channel under natural circulation conditions. The two-equation turbulence model is used to evaluate the turbulent momentum transport property. Becaus… more

Date: January 1, 1980

Creator: Chung, K. S. & Thompson, D. H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Calculational method for combined natural circulation and forced-convection flow in a channel. [LMFBR]

This paper presents a finite element solution for combined natural circulation and forced convection flow in a channel. Because the buoyancy force plays an important role in a mixed convection flow, an iteration scheme was used in solving the coupled energy-momentum equations. The momentum equations and the pressure equation are solved to calculate velocity profiles instead of solving the momentum equations with the continuity equation. Though the pressure equation is obtained by using the cont… more

Date: January 1, 1980

Creator: Chung, K. S. & Thompson, D. H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Audit calculation of the limiting CESSAR feedwater-line-break transient with RELAP5/MOD1. [PWR]

Argonne National Laboratory (ANL) performed a series of audit calculations of the limiting FLB transient presented in Appendix 15B to the CESSAR FSAR, supported by a limited number of additional calculations to investigate the sensitivity of the results (in terms of peak primary reactor system pressure) to break area and reactor trip time. The latter calculations were performed to quantify potential benefits in crediting reactor tip on low steam generator downcomer water level, which occurs ear… more

Date: January 1, 1983

Creator: Chung, K. S.; Kennedy, M. F. & Guttmann, J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Calculation of the limiting CESSAR feedwater line break transient

Argonne National Laboratory (ANL), under contract to the Nuclear Regulation Commission, performed audit calculations of the limiting Feedwater Line Break (FLB) transient presented in the CESSAR FSAR. These calculations were performed to investigate the sensitivity of peak reactor system pressure to break area and reactor trip time. The latter calculations were performed to quantify the effect of a reactor trip on low steam generator downcomer water level, versus the limiting FSAR transient assu… more

Date: January 1, 1984

Creator: Peeler, G. B.; Kennedy, M. F.; Guttmann, J. & Chung, K. S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Performance of large LWR system codes in calculating the steam-generator heat-transfer behavior

This paper presents a series of modeling experiences and problems in simulating the thermal-hydraulic behavior of large PWR steam generators using the RELAP4 and RELAP5 computer codes. Sensitivity studies investigating the heat transfer characteristics of both once-through and U-tube steam generators are discussed. Suggestions and recommendations are given for effective use and potential future improvements of these codes.

Date: January 1, 1982

Creator: Chung, K. S.; Abramson, P. B.; Kennedy, M. F. & Kim, J. S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Calculation of the limiting CESSAR Feedwater Line-Break and Steam Line-Break transients

Argonne National Laboratory (ANL), under contract to the Nuclear Regulatory Commission, performed audit calculations of the limiting Feedwater Line Break (FLB) and Steam Line Break (SLB) transients presented in the CESSAR FSAR. The results of the FLB and SLB calculations are discussed.

Date: January 1, 1983

Creator: Peeler, G. B.; Kennedy, M. F.; Caraher, D. L.; Guttmann, J. & Chung, K. S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

TREAT light water reactor source term experiments program

Four experiments are being conducted in the TREAT facility to investigate the behavior of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients are used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe accidents. The experiments are integral in nature and are aimed at the physicochemical characterization, near the point of origin, of the biolo… more

Date: July 1, 1984

Creator: Herceg, J. E.; Blomquist, C. A.; Chung, K. S.; Dunn, P. F.; Johnson, C. E.; Kraft, D. A. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department