Source: https://www.law.cornell.edu/cfr/text/10/appendix-K_to_part_50
Timestamp: 2017-02-23 03:34:51
Document Index: 213169358

Matched Legal Cases: ['art 50', 'art 50', 'art_50', 'art 50', '§ 50', '§ 50', '§ 50']

10 CFR Appendix K to Part 50, ECCS Evaluation Models | US Law | LII / Legal Information Institute
CFR › Title 10 › Chapter I › Part 50 › Appendix K_to_part_50 10 CFR Appendix K to Part 50, ECCS Evaluation Models
4. Fission Product Decay. The heat generation rates from radioactive decay of fission products shall be assumed to be equal to 1.2 times the values for infinite operating time in the ANS Standard (Proposed American Nuclear Society Standards - “Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors.” Approved by Subcommittee ANS-5, ANS Standards Committee, October 1971). This standard has been approved for incorporation by reference by the Director of the Federal Register. A copy of the standard is available for inspection at the NRC Library, 11545 Rockville Pike, Rockville, Maryland 20852-2738. The fraction of the locally generated gamma energy that is deposited in the fuel (including the cladding) may be different from 1.0; the value used shall be justified by a suitable calculation.
5. Metal - Water Reaction Rate. The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation (Baker, L., Just, L.C., “Studies of Metal Water Reactions at High Temperatures, III. Experimental and Theoretical Studies of the Zirconium-Water Reaction,” ANL-6548, page 7, May 1962). This publication has been approved for incorporation by reference by the Director of the Federal Register. A copy of the publication is available for inspection at the NRC Library, 11545 Rockville Pike, Two White Flint North, Rockville, Maryland 20852-2738. The reaction shall be assumed not to be steam limited. For rods whose cladding is calculated to rupture during the LOCA, the inside of the cladding shall be assumed to react after the rupture. The calculation of the reaction rate on the inside of the cladding shall also follow the Baker-Just equation, starting at the time when the cladding is calculated to rupture, and extending around the cladding inner circumference and axially no less that 1.5 inches each way from the location of the rupture, with the reaction assumed not to be steam limited.
(1)W 3. L. S. Tong, “Prediction of Departure from Nucleate Boiling for an Axially Non-uniform Heat Flux Distribution,” Journal of Nuclear Energy, Vol. 21, 241-248, 1967.
(2)B&W-2. J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, L. J. Stanek, “Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water,” Two-Phase Flow and Heat Transfer in Rod Bundles, ASME, New York, 1969.
(3)Hench-Levy. J. M. Healzer, J. E. Hench, E. Janssen, S. Levy, “Design Basis for Critical Heat Flux Condition in Boiling Water Reactors,” APED-5186, GE Company Private report, July 1966.
(4)Macbeth. R. V. Macbeth, “An Appraisal of Forced Convection Burnout Data,” Proceedings of the Institute of Mechanical Engineers, 1965-1966.
(5)Barnett. P. G. Barnett, “A Correlation of Burnout Data for Uniformly Heated Annuli and Its Uses for Predicting Burnout in Uniformly Heated Rod Bundles,” AEEW-R 463, 1966.
(6)Hughes. E. D. Hughes, “A Correlation of Rod Bundle Critical Heat Flux for Water in the Pressure Range 150 to 725 psia,” IN-1412, Idaho Nuclear Corporation, July 1970.
(1)GE transient CHF. B. C. Slifer, J. E. Hench, “Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors,” NEDO-10329, General Electric Company, Equation C-32, April 1971.
6. Convective Heat Transfer Coefficients for Boiling Water Reactor Fuel Rods Under Spray Cooling. Following the blowdown period, convective heat transfer shall be calculated using coefficients based on appropriate experimental data. For reactors with jet pumps and having fuel rods in a 7 × 7 fuel assembly array, the following convective coefficients are acceptable:
b. During the period after core spray reaches rated flow but prior to reflooding, convective heat transfer coefficients of 3.0, 3.5, 1.5, and 1.5 Btu-hr −1-ft −2 °F −1 shall be applied to the fuel rods in the outer corners, outer row, next to outer row, and to those remaining in the interior, respectively, of the assembly.
c. After the two-phase reflooding fluid reaches the level under consideration, a convective heat transfer coefficient of 25 Btu-hr −1-ft −2 °F −1 shall be applied to all fuel rods.
7. The Boiling Water Reactor Channel Box Under Spray Cooling. Following the blowdown period, heat transfer from, and wetting of, the channel box shall be based on appropriate experimental data. For reactors with jet pumps and fuel rods in a 7 × 7 fuel assembly array, the following heat transfer coefficients and wetting time correlation are acceptable.
b. During the period after core spray reaches rated flow, but prior to wetting of the channel, a convective heat transfer coefficient of 5 Btu-hr −1-ft −2- °F −1 shall be applied to both sides of the channel box.
5. General Standards for Acceptability - Elements of evaluation models reviewed will include technical adequacy of the calculational methods, including: For models covered by § 50.46(a)(1)(ii), compliance with required features of section I of this appendix K; and, for models covered by § 50.46(a)(1)(i), assurance of a high level of probability that the performance criteria of § 50.46(b) would not be exceeded.
[ 39 FR 1003, Jan. 4, 1974, as amended at 51 FR 40311, Nov. 6, 1986; 53 FR 36005, Sept. 16, 1988; 57 FR 61786, Dec. 29, 1992; 59 FR 50689, Oct. 5, 1994; 60 FR 24552, May 9, 1995; 65 FR 34921, June 1, 2000]