W3P87-1367, Forwards Reanalysis of Large Break LOCA for Cycle 2.Response to Request for Addl Info Re Methodology Used for Cycle 2 Large Break Analysis Submitted in 861028 & 1126 Ltrs

Forwards Reanalysis of Large Break LOCA for Cycle 2.Response to Request for Addl Info Re Methodology Used for Cycle 2 Large Break Analysis Submitted in 861028 & 1126 Ltrs

ML20214K645
Person / Time
Site:	Waterford
Issue date:	05/18/1987
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
W3P87-1367, NUDOCS 8705290027
Download: ML20214K645 (26)

Text

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fO LOUISIANA POWER S LIGHT COMPANY e Post Office Box 6008

• New Orleans. Louisiana 70174 L

wum NEW ORLEANS PUBLIC SERVICE INC.

• Post Office Box 60340. New Orleans. Louisiana 70160 9: May 18, 1987 W3P87-1367 A4.05 QA a U.S. Nuclear Aegulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 SUBJECT. Waterford SES Unit 3 Docket No. 50-382 Reanalysis of the Large Break LOCA for Cycle 2

REFERENCES:

1. W3P86-3328 dated October 1, 1986

2. W3P86-3384 dated October 28, 1986

3. W3P86-2941 dated November 26, 1986

4. W3P86-2982 dated December 23, 1986

5. NRC letter dated January 16, 1987 Gentlemen:

In the Reference 1 letter LP&L submitted Part B of the Cycle 2 Reload Analysis Report (RAR). This submittal included those sections of the RAR

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that described both the LOCA and non-LOCA safety analyses. In response 9 to requests for additional information regarding the methodology used for the Cycle 2 large break LOCA analysis, LP&L submitted References 2 and 3.

In subsequent discussions with the NRC staff, additional questions were raised concerning changes that were made to the large break LOCA analytical models since'the Cycle 1 break spectrum was performed and why, in light of

. these changes, the 0.8 double-ended guillotine (DEG) break remained the most limiting break size. LP&L, in Reference 4, agreed to reanalyze the large break LOCA (including a new break spectrum) using the latest NRC approw d analytical models. The NRC concluded, in Reference 5, that there was suf ficient justification for relying on the previous break spectrum

' analyses until the updated LOCA analysis was completed.

• LP&L has completed the reanalysis of the large break LOCA for Cycle 2.

/ Eacl j >ame~osed

,/r format please as the find ECCSdetailed Analysisresults section of this of thereanalysis RAR. Thepresented in the most significant of these results is that the limiting break size (0.8 DEG) and the peak clad temperature (2150*F) are unchanged from the LOCA analysis presented in Reference 1.

8705290027 870518 2 PDR ADOCK 050 P

+h "AN EQUAL OPPORTUNITY EMPLOYER"

(- - .

Page 2 W3P87-1367 If there are any questians or if you require further information please contact me or R.J. Murillo at (504) 595-2831.

"7

/W( . I K.W. Cook Nuclear Safety &

Regulatory Affairs Manager KWC/DPS/pim Enc 1csure cc: E.L. Blake, W.M. Stevenson, J.A. Calvo, J.H. Wilson, N. Lauben (NRC-NRR),

R.D. Martin, NRC Resident Inspector's Of fice (W3)

ECCS ANALYSIS f

INTRODUCTION AND SUfeARY The NRC Safety Evaluation Report (SER) for Waterford Unit 3 Cycle 2 approved Cycle 2 operation with the requirement that LP&L reanalyze the large break loss of coolant accident (LOCA) including a new break spectrum by May 1987 (Reference 1).

In response to this requirement, a large break LOCA (including a new break spectrum) ECCS performance analysis was performed for Waterford Unit 3 Cycle 2 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Eme*gency Core Cooling Systems for light water-cooled reactors (Reference 2). The analysis utilizes the recently approved June 1985 version of the C-E Large Break LOCA Evaluation Model (Reference 3). The analysis justifies an allowable peak Linear Heat Generation Rate (PLHGR) of 13.4 kw/ft.

This PLHGR is equal to the existing limit for Waterford Unit 3. The method of analysis and detailed results which support this value are presented herein.

METHODS OF ANALYSIS The reanalysis utilized the C-E June 1985 Large Break LOCA Evaluation Model which is described in References 4 through 11 and was approved by the NRC in Reference 3. A complete six break spectrum analysis was performed to determine the limiting large break. The original Cycle 2 analysis (Reference 15) performed the ECCS performance analysis for the limiting break obtained from Cycle 1.

In addition, the original Cycle 2 analysis utilized the blowdown hydraulic calculations of Cycle 1 which used the pre-1985 evaluation model. However, the original Cycle 2 analysis utilized the June 1985 evaluation mooel for the refill /reflood and hot rod calculations.

i

l METHODS OF ANALYSIS - continued Except for the differences discussed above, the method used for the present reanalysis is identical to the original Cycle 2 analysis.

Blowdown hydraulics, refill /reflood hydraulics and hot rod temperature calculations were performed with fuel parameters which bound the current fuel cycle and expected conditions for future cycles at a reactor power level of 3458 Mwt. The blowdown hydraulics calculations were performed with the 'CEFLASH-4A code (Reference 7) while the refill /reflood hydraulics calculations were performed with the COMPERC-II code (Reference 8). The hot rod clad temperature and clad oxidation calculations were performed with the STRIKIN-II and PARCH codes (References 10 and 11, respectively). Fuel performance calculations were performed using the FATES-3A version of C-E's NRC approved fuel performance code (References 12 and 13) with the grain size restriction as required by the NRC (Reference 14).

Significant core and system parameters for the present reanalysis are tabulated in Table 1. They are identical to those used in the original Cycle 2 analysis. The core-wide clad oxidation percentages were conservatively estimated based on the pin census of Cycle 2.

I RESULTS It was determined from the analysis that the allowable PLHGR is 13.4 kw/ft with the limiting break size identified as the 0.8 DEG/PD break.

0EG/PD = Double-Ended Guillotine at pump discharge. ,

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RESULTS - continued The 0.8 DEG/PD break' produced the highest peak clad temperature and the maximum peak local oxidation of 2150 0 F and 7.93%,

respectively. The 0.8 DEG/PD break also resulted in the highest core-wide oxjdation which was less than 0.805%. The peak clad temperature and the peak local and core-wide oxidation values are well below the 10CFR50.46 acceptance limits of 2200 F,17%, and 1%, respectively.

The times of interest for each of the breaks are presented in Table 2. Table 3 contains a summary of the peak clad temperatures and oxidation percentages for the break spectrum.

Table 4 provides the detailed results of the limiting break. For comparison purposes the results of the limiting break of - the original Cycle 2 analysis are also tabulated in Table 4. Table 5 contains a list of the pertinent variables plotted for the limiting break.

A review of the effects of initial operating conditions on these results was performed. It was determined that over the ranges of initial operating conditions as specified in the Technical Specifications, the current PLHGR Technical Specification (allowing up to 13.4 kw/ft) remains acceptable for Cycle 2 operation.

CONCLUSION The ECCS performance evaluation for Waterford Unit 3 Cycle 2 i results in a peak clad temperature of 2150 F, a peak local clad oxidation percentage of 7.93% and a peak core-wide clad oxidation percentage of less than 0.805% compared to the acceptance '

criteria of 2200 F, 17%, and 1%, respectively. Therefore,.

operation of Waterford Unit 3 Cycle 2 at a core power level of 3458 Nt (102% of 3390 Nt) and a PLHGR of 13.4 kw/ft is in conformance with 10CFR50.46.

]

TABLE 1 Waterford Unit 3 Cycle 2 Core and System Parameters Parameter (Units) Value Reactor Power Level 9 102% of Nominal (Mwt) 3458 Average Linear Heat Rate 9 102% of 5.6 Nominal (kw/ft)

Peak Linear Heat Generation Rate (kw/ft) 13.4 Core Inlet Temperature ( F) 557.5 Core Outlet Temperature ( F) 618.6 6

System Flow Rate (1bm/hr) 148x10 6

Core Flow Rate (1bm/hr) 144x10 Gap Conductance (1) (Btu /hr ft2 F) 1534 Fuel Centerline Temperature (1)(oF) 3321.6 Fuel Average Temperature (1)( F) 2111.3 Hot Rod Gas Pressure (1)(psia) 1113.3 Hot Rod Burnup (Mwd /Mtu) 1000 Number of Steam Generator Tubes 250 Plugged per Steam Generator Augmentation Factor 1.0 Minimum Initial Containment 14.4 Pressure (psia)

Containment Heat Sink Data (2) 3 6 Containment Free Volume (ft ) 2.684x10 Axial Peaking Factor 1.53 III Initial values at the limiting hot rod burnup as calculated by STRIKIN-II at 13.4 kw/ft.

(2) Same as those used in the original Cycle 2 analysis.

TABLE 2 Waterford Unit 3 Cycle 2 Significant Events Summary 8reak Spectrum (seconds after break) l Time Hot Rod Safety Start of Safety End of Contact Rupture Injection Break Injection Tanks 8ypass Time Time Tanks Emoty 1.0 DES /PD II) 10.00 20.38 40.74 45.40 111.24 0.8 DES /PD 11.01 21.58 41.95 46.05 112.38 0.6 DES /PD 12.80 23.96 44.41 49.29 114.52 1.0 DEG/PD(2) 11.01 21.77 42.15 43.48 112.50 0.8 DEG/PD 12.21 23.35 43.81 43.70 113.89 0.6 DEG/PD 14.20 25.75 46.24 47.83 116.15 1

I (1) DES /PD = Double-Ended Slot at Pump Discharge (2) DEG/PD = Double-Ended Guillotine at Pump Discharge 1

TABLE 3 Waterford Unit 3 Cycle 2 Fuel Rod Performance Summary Large Break Spectrum Peak -

Core-Wide Peak Clad Local Clad Clad Break Temperature ( F) Oxidation (%) Oxidation (%)

1.0 DES /PD 2117 7.09 <.762 0.8 DES /PD 2113 7.06 <.769 0.6 DES /PD 2097 6.76 <.718 1.0 DEG/PC 2146 7.83 <.804 0.8 DEG/PD 2150 7.93 <.805 0.6 DEG/PD 2122 7.31 <.691 3183-2 dir/SA

TABLE 4 Waterford Unit 3 Cycle 2 Limiting Break Size (0.8 DEG/PD)

Cycle 2 Cycle 2 Original Analysis Break Spectrum Peak Linear Heat Generation 13.4 . 13.4 Rate (kw/ft)

Peak Clad Temperature ( F) 2149.8 2150.0 Time of Peak Clad Temperature 259.1 265.4 (seconds)

Time of Clad Rupture (Seconds) 41.9 43.7 Peak Local Clad Oxidation (%) 7.80 7.93 Total Core-Wide Clad Oxidation (%) <0.805 <0.805

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TABLE 5 Waterford Unit 3 Cycle 2 Variables Plotted as a Function of Time for the Limiting large Break Variables Figure Number 1

Core Power l' Pressure in Center Hot Assembly Node 2 Leak Flow 3 Hot Assembly Flow (below hot spot) 4 Hot Assembly Flow (above hot spot) 5 Hot Assembly Quality 6 Containment Pressure 7 Mass Added to Core During Reflood 8 Peak Clad Temperature 9 Hot Spot Gap Conductance 10 Peak Local Clad Oxidation 11 Temperature of Fuel Centerline 12 Fuel Average, Clad and Coolant at Hottest Node Hot Spot Heat Transfer Coefficient 13 Hot Rod Internal Gas Pressure 14 l

References:

1. Letter, J. H. Wilson (NRC) to J. G. Dewease (LP&L), dated January 16, 1987.

2. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39. No. 3, January 4, 1974.

3. Letter, D. M. Crutchfield (NRC) to A. E. Scherer (C-E), " Safety Eval-uation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports",

July 31, 1986.

4. Letter, A. E. Scherer (C-E) to J. R. Miller (NRC) LD-81-095, Enclosure 1-P-A, "C-E ECCS Evaluation Model Flow Blockage Analysis", (Proprietary),

December 15, 1981.

5. Letter, A. E. Scherer (C-E) to C. O. Thomas (NRC), LD-86-027. " Responses to Questions on C-E's Revised Evaluation Model for Large Break LOCA Analysis", (Proprietary), June 17, 1986.

6. Letter, A. E. Scherer (C-E) to C. O. Thomas (NRC). LD-85-032,

" Revision to C-E Model for Large Break LOCA Analysis", July 3, 1985.

7. CENPD-133, Supplement 5-A, "CEFLASH-4A. A FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis", June 1985.

8. CENPD-134, Supplement 2-A, "COMPERC-II. A Program for Emergency Refill-Reflood of the Core", June 1985.

9. CENPD-132-P, Supplement 3-P-A, " Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS",

June 1985.

10. CENPD-135, Supplement 2-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications)", February 1975.

CENPD-135-P, Supplement 4-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", August 1976.

CENPD-135-P, Supplement 5-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1977.

11. CENPD-138-P, and Supplement 1-P, " 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.

12. CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report", July 1974.

13. CEN-161(B)-P, " Improvements to Fuel Evaluation Model Topical Report",

July 1981.

14 Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E), dated March 31, 1983.

15. LP&L Letter, W3P86-3328, dated October 1, 1986.

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Rgure 1 WATERFORD UNIT 3 CYCLE 2 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER 1 2001 4

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- Figure 3 WATERFORD UNIT 3 CYCLE 2 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG LEAK FLOW PUMP SIDE

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Figure 6 WATERFORD UNIT 3 CYCLE 2 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY QUAUTY NODE 13, BELOW HOTTEST REGION

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Rgure 8 WATERFORD UNIT 3 CYCLE 2 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD 150000 125000 TIME (SEC) REFLOOD RATE 0.00 - 14.00 1.68 IN/SEC 100000- 14.0 - 58.M 1.23 IN/SEC -

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Figure 9 WATERFORD UNIT 3 CYCLE 2 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE 2200 2000 's 1800 7 1600 1400 m

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Figure 10 WATERFORD UNIT 3 CYCLE 2 0.B x DOUBLE ENDED GUILLOTINE BREAK IN PUMP D HOT SPOT GAP CONDUCTANCE 1800 1600 1400

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WATERFORD UNIT 3 CYCLE 2 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG TEMPERATURE OF FUEL CENTERUNE, FUEL AVERAGE, CLAD AND COOLANT AT HOTTEST N0DE 2700 2400 x \ FbEL CENTERLINE 2100 y CLAD / x 1800 e \

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