ML17216A422
| ML17216A422 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/31/1986 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| L-86-37, NUDOCS 8602070009 | |
| Download: ML17216A422 (18) | |
Text
r REGULATORY " ORMATION DISTRIBUTION BYWM (RIDB)
DOCKET ¹ 05000389
SUBJECT:
Forwards results oF C-E large break LOG* EECS perFormance analysis which demonstrates compliance w/10CFR50. 46 criteriai per 850930 commitment.
DISTRIBUTION CODE:
A001D COPIES RECEIVED: LTR ENCL SIZE:
TITLE:
OR Submittal:
General Distribution ACCESSION NBR: 8602070009 DOC. DATE: 86/01/31 NOTARIZED:
NO FACIL: 50-389 St.
Lucie Planti Unit 2i Florida Power 5 Light Co.
AUTH. NAME AUTHOR AFFILIATION WOODY'. O.
Florida Power 5 Light Co.
RECIP. NAME RECIPIENT AFFILIATION MIRAGLIAgF. J.
Division oF Pressurized Water Reactor Licensing 8 (post 8
NOTES'L'4/06/83 05000389 RECIPIENT ID CODE/NAME PWR-8 ADTS PNR-8 EB PNR-8 FOB PWR-8 PEICSB COPIES LTTR ENCL 1
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x 14000, JUNO BEACH, FL 33408 iy9 Ilz FLORIDAPOWER & LIGHT COMPANY mS t tIIsa L-86-37 Office of Nuclear Reactor Regulation Attention:
Mr. Frank J. Miraglia, Director Division of PWR Licensing - B U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Miraglia:
Re:
St. Lucie Unit No. 2 Docket No. 50-389 C-E Lar e Break LOCA Anal sis Florida Power and Light Company Letter L-85-373, dated September 30, l985, stated that Combustion Engineering was redoing the St. Lucie Unit 2 Large Break LOCA Analysis, and that the results were expected to be available by January 3I, l986.
Attached are these results which demonstrate compliance with 10 CFR 50.46 criteria.
Very truly yours, C. O. Wood Group Vi resident Nuclear Energy CO W/RJS/gp Attachment 8602070009 860i3i PDR ADQCK 05000389 PDR I
RJS4/008/
I PEOPLE... SERVING PEOPLE
Large Break LOCA ECCS Performance Introduction and Summar On July 3, 1985, Florida Power 5 Light Company (FPL) received notification from Combustion Engineering (C-E) of a potential non-conservatism in one element of C-E's large break loss-of-coolant accident (LOCA) evaluation model.
That notification was received by copy of a letter to NRC from C-E (A. E.
Scherer to G.
W. Knighton, LD-85-031, July 2, 1985).
The particular element in question was the treatment of axial power distribution and peaking factor.
In response to NRC concerns raised relating to the selection of the power distribution used in the C-E large break ECCS analysis, a sensitivity study was performed to determine the limiting power distribution for use in the large break ECCS analysis of C-E NSSS's with the currently approved C-E Evaluation Model.
The results of analyses performed determined the most adverse axial shape to be a 1.55 peak located at 65K of core height (Reference 12).
With respect to the Cycle 2 license (Reference
- 2) the limiting large break LOCA was reanalyzed with the limiting axial power shape (Reference 12), i250 plugged tubes per steam generator, fuel parameters which bound current and expected conditions, and a lower augmentation penalty to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water cooled reactors (Reference 1).
The analysis justifies an allowable peak linear heat generation rate (PLHGR) of -13.0 kw/ft.
This PLHGR is equal to the existing limit for St. Lucie Unit 2.
The method of analysis and detailed results which support this value are presented herein.
Method of Anal sis The method of analysis is based upon the NRC approved C-E large break LOCA
ECCS evaluation model which is described in References 3 through 8.
- Blowdown, and refill/reflood hydraulics, and hot rod temperature calculations were performed with the fuel parameters which bound the current fuel cycle and expected conditions for future cycles at a reactor power level of 2754 Mwt.
The blowdown hydraulic calculations were performed with the CEFLASH-4A (Reference
- 5) code while the refill/reflood hydraulic calculations were performed with the COMPERC-II (Reference
- 6) code.
The hot rod clad temperature and clad oxidation calculations were performed with the STRIKIN-II (Reference
- 7) and PARCH (Reference
- 8) codes.
Fuel performance calculations were performed using the FATES-3A Version of the C-E's NRC approved fuel performance code (References 9 and
- 10) with the fuel grain size restriction as approved by the NRC (Reference 11).
Results Table 1 presents the analysis results for the 1.0 OEG/PD break which produces the highest peak clad temperature.
The results of the evaluation confirm that 13.0 kw/ft is an acceptable value for the PLHGR in the present analysis.
The peak clad temperature and maximum local and core wide clad oxidation values, as shown in Table 1, are below the 10CFR50.46 acceptance limits of 2200'F, 17%
and 1% respectively.
Table 2 presents a list of the significant parameters displayed graphically for the limiting 1.0 OEG/PO break.
The significant parameters and initial conditions for this analysis are shown in Table 3.
This analysis accounts for steam generator U-tube plugging of 1250 average length tubes per steam generator.
Steam generator U-tube plugging increases system resistance to flow and hence the ability of the Reactor Coolant System (RCS) to vent steam during reflood.
The analysis also accoun'ts for the decreased heat transfer area and primary side coolant volume caused by the tube plugging.
DEG/PD = Double-Ended Guillotine at Pump Discharge
and the average rod of the hot assembly.
Higher power of the average rod the hot assembly results in reduced heat transfer from the hot rod to its surrounding rods.
The reduction in the augmentation penalty results in an increase of the hot assembly average channel PLHGR.
The hot assembly average channel PLHGR influences the radiation heat transfer between the hot rod of the hot assembly of To determine the limiting power distribution for use in the large break ECCS'nalysis, a sensitivity study was performed and is documented in Reference 12.
Based on this study the limiting power shape was used for the present analy-si 5.
The 1.0 DEG/PD produced the highest peak clad temperature of 2106'F and a peak local oxidation of 16. 12% compared to the acceptance criteria of 2200'F and 17%, respectively.
The 1.0 DEG/PD also resulted in the highest core wide oxidation of less than 0.70% which is below the 1% acceptance criterion.
Conclusions The results of the ECCS performance evaluation for the present analysis for St. Lucie Unit 2, demonstrated a peak clad temperature of 2106'F, a peak local clad oxidation percentage of 16. 12%,
and a peak core wide clad oxidation percentage of less than 0.70% compared to the acceptance criteria of 2200'F, 17%,
and 1%, respectively.
Therefore, operation of St. Lucie Unit 2 at a core power level of 2754 Mwth ( 102% of 2700 Mwth) and a
PLKGR of 13.0 kw/ft is in conformance with 10CFR50.46.
REFERENCES l.
Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Reactors, Federal
- Register, Vol. 39, No. 3, Friday, January 4,
1974.
2.
Letter from J.
W. Williams, Jr.,
(FPSL), to E.
G. Eisenhut (US NRC),
St. Lucie Unit 2, Docket No. 50-389, "Proposed License Amendment Cycle 2
Reload," L-84-148, June 4, '1984.
3.
CENPD-132, "Calculative Methods for the C-E Large Break LOCA Evaluation Model," August 1974 (Proprietary).
CENPD-132, Supplement 1, "Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model," December 1974 (Proprietary).
4.
CENPD-132, Supplement 2, "Calculational Methods for the C-E Large Break LOCA Evaluation Model, " July 1975 (Proprietary) 5.
CENPD-133, "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis", April 1974 (Proprietary).
CENP0-133, Supplement 2,
"CEFLASM-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis (Modification)", December 1974 (Proprietary).
6.
CENP0-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," April 1974 (Proprietary).
CENPD-134, Supplement 1, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)", December 1974 (Proprietary).
7.
CENPD-135, "STRIKIN, A Cylindrical Geometry Fuel Rod Heat Transfer Program, April 1974 (Proprietary).
CENPD-135, Supplement 2, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)", February 1975.
CENPD-135, Supplement 4, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program",
August 1976 (Proprietary).
8.
CENP0-138, and Supplement 1 "PARCH, A FORTRAN IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup,
" February 1975.
CENPD-138 Supplement 2 (P),
"PARCH - A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", January 1977 (Proprietary).
9.
CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report", July 1974.
10.
CEN-161(B)-P, "Improvements to Fuel Evaluation Model Topical Report," July 1981.
11.
Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr.
(BG&E), dated March 31, 1983.
12.
Letter <F2-CE-R-043, "St. Lucie 2 Large Break LOCA Axial Shape Sensitivity Study Results Based on the August 1974 Evaluation Medel",
November 14, 1985.
Table 1
St. Lucie - Unit 2 Limitin Break Size (1.0 DEG/PD)
Parameter Value Peak Linear Heat Generation Rate (kw/ft)
Peak Clad Temperature
('F)
Time of Peak Clad Temperature (Seconds)
Time of Clad Rupture (Seconds)
Peak Local Cl ad Oxidati on (I)
Total Core-Wide Clad Oxidation (X) 13.0 2106 259 55.85 c 16.12
~ 0.70
Table 2
St. Lucie - Unit 2 Anal sis Plots Variables Figure Desi nation Peak Clad Temperature Hot Spot Gap Conductance Peak Local Clad Oxidation
- Clad, Fuel Centerline, Average Fuel, and Coolant Temperatures for Hottest Node Hot Spot Heat Transfer Coefficient Hot Rod Internal Gas Pressure
Table 3
St. Lucie - Unit 2 Si nificant Parameters and Initial Conditions Parameter Core Power Level at 102% of. Nominal, MWt Core Average Linear Heat Rate (1025 of Nominal) (kw/ft)
Peak Linear Heat Generation Rate (PLHGR)
Hot Assembly, Hot Channel (kw/ft)
Peak Linear Heat Generation Rate (PLHGR)
Hot Assembly, Average Channel (kw/ft)
Core Inlet Temperature
('F)
Core Outlet Temperature
('F)
System Flow Rate (ibm/hr)
Core Fl ow Ra te
( 1 bm/hr )
Gap Conductance at PLHGR (Btu/hr-ft -'F)
Fuel Centerline Temperature at PLHGR
('F)
Fuel Average Temperature at PLHGR
('F)
(2)
Hot Rod Gas Pressure (psia)
~
(2)
Hot Rod Burnup (MWD/MTU)
Number. of Tubes Plugged Per Steam Generator Augmentation Factor Value 2754 4.90 13.0 11.45 552 603.8 136.1xlO 131.1x10 1416 3228 2078 1118 1038 1250 1.01 System flowrate consistent with 363,000 gpm.
STRIKIN-II values at hot rod burnup which yields highest peak clad temperature.
ZZCC FIGURE l PEAK CLAD TH'iPERPTURE i,0 x DOUFLE ENDED 64ILLOTIl~E FREAK Iil PUtiP DISCIIARGE LEG 2GGC it QQ ty
/ i
/
////
////
UJ~:2GG I
C uJ~ ~GGG
PEAK CLAD TEf'PERATURE NODE
--- PEAK CLAD OXIDATION NODE GGG 400 100 2GG 3GG Tli'iL, SECOHBS 400 500 600 70G
V 4.
I. ICI'RE 2 HOT SPOT GAP COI>DUCTAilCE 1,0 x DOUBLE biDED GUILLOT IhE BREAK Ii~ FUIIF DISCI-;ARGE LEG a GGG
'i 4GG
~ iZGG
~GGG
=
8CG S'G 4GG 2GG iGO 200 3GO TIf'iL, SECOf,'DS 400 SGO GGG 700
F IGLiRF 3 PEAK LOCAL CLAD OXII3ATIOh j.,Q x DOLfELE EhDED CL ILL01INE BREAK Iii PLiffP D,ISCl-iARGE LEG 8
CD I
OC CD Ch 0
/
/
/
/
/
///
/
////
/I/
PEAK CLAD TEf+ERATURE NODE
-- PEAK CLAD OXIDATIOfu fKDE
////
/
/
/
<00 20G 3GG TIf'fE, SECONDS 400 500 6GG 7GG
2 7GC FI64RE 4 CLAD, CEfiTERLIIIE, AVERACE FIDEL ANL'OOLAIIT Tkf'iPERATURE FOR HOTTEST NODE i,G x DOI'FLE EtiDED CliILLOTIi~E BREAK Ih I'Uiip DISC}iARGE LEG 2 EGG
~FUEL CB<TERLINE AVERAGE FUEL CG
=GG COOLANT 200 300 TII'iL, SECONDS 400 SG0 6GG 700
F I6LIRE 5 liOT SPOT liEAT TRANSFER COEFFICIENT 1,0 x DOUFLE ENDED CUILLOTIHE I'BEAK Ii'3 PliYiP DISCliARGE LEG
- BC I4-I
- 2G k
L Cu gr J
"C sJ iGO 2GG 3GG 1 IfiE, SE.COi]I'S 4GG SGG GGG 7GG
FIGURE 6 HOT ROD INTERNAL GAS PRESSURE 1,0 x DOUBLE ENDED GUILLOTINE BREAK I[0 PUf'1P DISCHARGE LEG 120',N,T,~
1118, LI PSIA 1000 RUPTURE AT 55,S5 SEC-COO 600 400 200 0
60
- TIrIE, SECOt~OS SO 100