ML20023E060
| ML20023E060 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/25/1983 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20023E061 | List: |
| References | |
| NUDOCS 8306070156 | |
| Download: ML20023E060 (5) | |
Text
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BALTIMORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475. BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL JR.
v cc p,.c.oc~r May 25,1983 SUPPLY Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 ATTENTION:
Mr. R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 and 2, Docket Nos. 50-317 and 50-318 Response to Questions Concerning the MacBeth CIIF Correlation Used in Unit 1 Cycle 6 and Unit 2 Cycle 5 Steam Line Break Analysis
REFERENCE:
D.
A.
Jaffe to J.
A.
Mihalcik telecopy dated April 15,1983 Gentlemen:
Attached is the response to the questions concerning the use of the MacBeth Critical liest Flux Correlation requested in the referenced telecopy.
Should you have any questions, please contact us.
Very truly yours, BALTIMORE GAS AND ELECTRIC COMPANY l
k dwt&yx A. E; Lundvall, Jr.
Vice P' esident - Suppl ['
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AEL/ JAM /Imt Attachment (40 copies) cc:
J. A. Biddison, Esquire G. F. Trowbridge, Esquire l
D.11. Jaffe - NRC R. R. Mills - CE R. E. Architzel-NRC/CC
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8306070156 830525 PDR ADOCK 05000317
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p ATTACHMENT Question The MacBeth CHF correlation is approved in 10 CFR Part 50 Appendix K for LOCA analysis; the acceptance of the correlation for steam line break (SLB) analysis requires verification of the ranges of applicability of the correlation with respect to the transient ranges of the SLB. In addition, the minimum DNBR limit of'the correlations must be established for the determination of fuel pin failure analysis. We will need the test data as well as the analysis result.
Response
The MacBeth correlation (Reference (1)) is used only for the determination of DNBR during the post-trip return to power portion of the main steam line break (MSLB) transients ' presented in References (2) and (3). For all other transients (presented in References (2) and (3)), the CE-1 correlation is used in the TORC code (References (4) and (5)) to calculate DNBR. For determination of DNBR during the post-trip return to power portion of MSLB transients the methods applied by Lee (Reference (6)) are employed in order to use the MacBeth correlation for rod bundles to predict burnout as a function of axial height, accounting for non-uniform axial heat flux.
MacBeth demonstrated that five parameters were accessary for correlation of critical heat flux (CHF) data for rod bundles with vertical upflow:
mass flux, G; inlet subcooling, A H; pressure, P; heated diameter, d ; and channel length,1. However, the h
channel length is climinated from the correlation by application of Lee's method to predict CHF as a function of axial height. To determine applicability, the height at which minimum DNBR occurs is compared with the channellengths for the experiments from which MacBeth drew his data.
The data used for the MacBeth rod bundle corregations with vertical upflow has a range of values for G from 0.18 to 4.1 million Ibm /hr ft (Reference (7)). The uniformity of the correlation of the CHF data for rod bundles as a function of G over this range and the data and correlations for CHF in heated tubes give confidence to extend thejower end gf the range of applicability of the rod bundle correlation to at least 0.09x10 lbm/hr ft.
(The MacBetg(correlations for heated tubes are based on data for 0.01 (Gx10-7.8 lbm/hr ft ).
I The ran e of values of A H upon which the rod bundle correlations is based is
-150 < AH < 380 BTU /lbm (Reference (7)). The rod bundles from which the CHF data was obtained has values of d between 0.113 and 0.902 inches and lengths of from 17 to h
72 inches (Reference (1)).
The data used for the rod bundle correlations is for 1000 psia _. However, application of MacBeth's correlations for heated tubes indicates that using the correlation developed for 1000 psia at lower pressures produces values for DNBR which are conservative.
Further, other correlations for rod bundles, such as those of Bowring (Reference (8)),
yield a variation in DNBR of the order of 10% as pressure varies from 500 to 1000 psia for the range of G and A H of interest. This is of the same order as the unecrtainty in MacBeth's correlation.
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t Table 1 summarizes the above applicability ranges and compares them with values of the parameters obtained for the MSLB transients presented in References (2) and (3). The applicability ranges cover all the expected conditions for the post trip return to power portion of these MSLB transients.
C-E has historically used the MacBeth correlation to demonstrate that critical heat fluxes are not e teeeded during the MSLB transient (e.g., References (9), (10), and (11)).
A DNBR limit of 1.30 is used as an acceptance limit for the MacBeth correlation. This limit was adopted for consistency with the SRPs (Reference (12)), and is conservative with respect to Reference (13) which establishes that the MacBeth correlations bounds
, 97% of the data at a DNBR of 1.25 (95% of data bounded at a DNBR of 1.20). Reference (13) (attached) was provided to the NRC (then the USAEC) in response to a similar question in Reference (14)(attached). Since the results presented in References (2) and (3) do not violate the 1.30 limit, critical heat fluxes are not exceeded and hence no fuel failure is [sredicted. Since no fuel failure is predicted, it has not been necessary to t:se the MacBeth correlation to determine the number of fuel pin failures.
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TABLE 1
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COMPARISON OF APPLICABILITY RANGES FOR MACBETH CHF CORRELATION FOR VERTICAL UPFLOW IN ROD BUNDLES WITH VALUES OBTAINED FOR THE MSLB TRANSIE!.7S PRESENTED IN CALVERT CLIFFS UNIT I CYCLE 6 AND UNIT II CYCLE 5 LICENSING SUBMITTALS (REFERENCES 2 AND 3)
Range of Values
~
Obtained for MSLBs During Post-Trip Return Parameter Range of Applicability to Power 6 1mb/hr ft 0.09 4.1 0.18 - 0.21 2
G (10 4LH GTU /lbm)
-150 - 380 80 - 108 P (psia) 500 - 1000 700 - 793 dh (inches)
.113
.902 0.471 1 (inches) 17 - 72 22 - 33 I
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REFERENCES (1) MacBeth, R.
V.,
"An Appraisal of Forced Convection Burn-out Data," Proc.
Instn. Mech. Engrs., Vol. 180, Pt3c, pp 37-50, 1965-66.
(2) Letter, P. W. Kruse (C-E) to W. J. Lippold (BG&E), "Calvert Cliffs Unit 1 Cycle 6 Reload License Submittal," BG&E-9676-693, April 26,1982.
(3) Letter, R.
R.
Mills (C-E) to W.
J.
Lippold (BG&E), " Supplement 1.to Calvert Cliffs Unit 2 Cycle 5 Refueling Amendment," BG&E-83-16, January 14, 1983 D'termining the Thermal' Margin of a (4)
" TORC Code A Computer Code for e
Reactor Core," CENPD-161-P; July 1975 Proprietary Information.
(5) " TORC Code - Verification and Simplified Modeling Methods," CENPD-206-P, January 1977, Proprietary Information.
(6) Lee, D. H., "An Experimental Investigation of Forced Convection Burn-out in High Pressure Water-Part IV, Large Diameter Tubes at About 1600 psia."
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A. E. E. W. Report R479, 1966.
(7) MacBeth, R.
V., " Burn-out Analysis Part 5: Examination of Published World Data for Rod Bundles," A. E. E. W. Report R358, 1964.
(8) Bowring, R. W., "A New Mixed Fjow Cluster Dryout Correlation for Pressures For Use in a Transient in the Range 0.6 - 15.5 MN/m (90-2250 psla)
Blowdown Code," I Mech E_ Conference Publications 1977-8, pp 175-182,1977.-
(9) Maine Yankee Atomic Power Station, Final Safety Analysis Report, Volume II, Chapter 14.11, Steam Line Rupture Incident, Rev. June 1971.
(10) Millstone Nuclear Power Station Unit 2,
Final Safety Analysis Report, Volume II, Chapter 14.12, Steam Line Rupture Incident.
(11) Baltimore Gas and Electric Company, Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Final Safety Analysis Report, Volume III, Chapter 14.12 Steam Line Rupture Incident, Rev. August 15, 1972.
(12) USNRC Standard Review Plan, Section 4.4,
" Thermal and Hydraulic Design,"
NUREG 0800, Rcvision 1, July 1981.
(13) NY-123-69, Idaho Nuclear Corporation, Nuclear Safety Program Division, Monthly Report, October 1969 (14) Letter, C. L. Klin.g (C-E) to S. Salah (USAEC), August 1, 1973.
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