TY - GEN
T1 - Thermal-hydraulic correlations for the advanced neutron source reactor fuel element design and analysis
AU - Siman-Tov, M.
AU - Gambill, W. R.
AU - Nelson, W. R.
AU - Ruggles, A. E.
AU - Yoder, G. L.
PY - 1991
Y1 - 1991
N2 - The Advanced Neutron Source (ANS) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. The core of the ANS reactor is therefore designed to accommodate both a high power and high-power density using a very high coolant mass flux and subcooling level. A statistical/probabilistic approach for uncertainty analysis is being developed to determine the maximum operating power while simultaneously providing the necessary safety margins. This requires the selection of the most appropriate thermal-hydraulic (T-H) correlations and the development of uncertainty probability distributions based on the best data available. The core is composed of parallel aluminum-clad fuel plates in an involute geometry. In the present reference design, the plates effectively form thin rectangular flow channels with a flow gap of 1.27 mm, a span of 70 or 87 mm, and a heated length of 507 mm. The heavy-water coolant flows vertically upward at a mass flux of 30 Mg/m2s (inlet velocity of 27.4 m/s). All the channels have common inlet and outlet pressures of 3.7 and 1.9 MPa, respectively. The average inlet and outlet coolant temperatures are 49°C and 92°C, respectively (197°C and 118°C subcooled). The average and peak heat fluxes are 6.1 and 16.6 MW/m2, respectively. This paper discusses the T-H correlations currently used by the ANS project for nominal steady-state conditions and the appropriate correlations and data base collected for the determination of the uncertainty probability distributions. The T-H correlations covered in the present paper are critical heat flux (or departure from nucleate boiling), onset of flow-excursion heat flux, incipient boiling heat flux, forced-convection heat-transfer coefficient, and forced-convection friction factor. The T-H correlations currently used are presented in detail with their specific interpretation for the ANS reactor design and analysis and their preliminary comparison to the data base collected to date.
AB - The Advanced Neutron Source (ANS) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. The core of the ANS reactor is therefore designed to accommodate both a high power and high-power density using a very high coolant mass flux and subcooling level. A statistical/probabilistic approach for uncertainty analysis is being developed to determine the maximum operating power while simultaneously providing the necessary safety margins. This requires the selection of the most appropriate thermal-hydraulic (T-H) correlations and the development of uncertainty probability distributions based on the best data available. The core is composed of parallel aluminum-clad fuel plates in an involute geometry. In the present reference design, the plates effectively form thin rectangular flow channels with a flow gap of 1.27 mm, a span of 70 or 87 mm, and a heated length of 507 mm. The heavy-water coolant flows vertically upward at a mass flux of 30 Mg/m2s (inlet velocity of 27.4 m/s). All the channels have common inlet and outlet pressures of 3.7 and 1.9 MPa, respectively. The average inlet and outlet coolant temperatures are 49°C and 92°C, respectively (197°C and 118°C subcooled). The average and peak heat fluxes are 6.1 and 16.6 MW/m2, respectively. This paper discusses the T-H correlations currently used by the ANS project for nominal steady-state conditions and the appropriate correlations and data base collected for the determination of the uncertainty probability distributions. The T-H correlations covered in the present paper are critical heat flux (or departure from nucleate boiling), onset of flow-excursion heat flux, incipient boiling heat flux, forced-convection heat-transfer coefficient, and forced-convection friction factor. The T-H correlations currently used are presented in detail with their specific interpretation for the ANS reactor design and analysis and their preliminary comparison to the data base collected to date.
UR - http://www.scopus.com/inward/record.url?scp=0026378264&partnerID=8YFLogxK
M3 - Conference contribution
AN - SCOPUS:0026378264
SN - 079180884X
T3 - American Society of Mechanical Engineers, Heat Transfer Division, (Publication) HTD
SP - 63
EP - 78
BT - Nuclear Reactor Thermal-Hydraulics
PB - Publ by ASME
T2 - Winter Annual Meeting of the American Society of Mechanical Engineers
Y2 - 1 December 1991 through 6 December 1991
ER -