ML17299A655

Forwards Util Specific Limiting Large Break LOCA Reanalysis Results,Per NRC 850830 Request.Reanalysis Demonstrates Compliance w/10CFR50.46.Administrative Limit on Linear Heat Rate for Unit 1 Will Be Lifted Upon NRC Approval

ML17299A655
Person / Time
Site:	Palo Verde
Issue date:	10/03/1985
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Knighton G Office of Nuclear Reactor Regulation
References
ANPP-33650-EEVB, NUDOCS 8510080495
Download: ML17299A655 (16)

Text

REGULATOR 'NFORMATION DISTRIBUTION TEM (RIDS)

ACCESSION NBR;8510080495 DOC,DATE'5/10/03 NOTARIZED: NO . DOCKET FACIL':STN 50 528 Palo Ver de Nuclear Station~ Unit i~ Arizona Publ i 05000528 STN 50 529 Palo Verde Nuclear Stations Uni.t 2g Arizona Publi 05000529 STN 50-530 Palo Verde Nuclear Station~ Unit 3~ Arizona Publi 05000530 AUTH, NAME AUTHOR AFFILIATION VAN BRUNTiE-"E; Arizona Nuclear Power Project (formerly Arizona. Public Serv RECIP ~ NAME>> RECIPIENT AFFILIATION KNIGHTON g G, 1'] ~ licensing Branch 3

SUBJECT:

Forwards util specific limiting large break LOCA reanalysis results<per NRC 850830 request. Reanalysis demonstrates compliance .w/10CFR5(f.46 'dministrative limit on- linear>> heat" rate for Unit 1 will be lifted upon NRC approval.

DISTRIBUTION CODE! A001D COPIES RECEIVED'TR ENCL. SIZE':

TITLE: OR Submittal: General Distribution NOTES:Standardized plant, 05000528 OLe12'/31/84 Standardized plant. 05000529 Standardized plant ~ 05000530 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME; LTTR ENCL>> ID CODE/NAME LTTR ENCL" NRR LB3. BC 01 7 INTERNAL: ACRS 09 ADM/LFMB 0, ELD/HDS3 1 0 NRR/DE/MTEB 1 NRR/DL DIR 1 1 NRR/DL/DRAB 0 NRR/DL/TSRG 1 1 /METH 1 NRR/DSI/RAB 1 1 REG FIL 04 1 RGN5 1 1 EXTERNAL; 24X 1 1 EGEG BRUSKEgS LPDR 03 1 1 NRC PDR 02' 1>>

NSIC 05 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 28 ENCL 25

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Arizona Nuclear Power Project P.o. SOX 52034 ~ PHOENIX, ARIZONA85072-2034 Director of Nuclear Reactor Regulation ANPP-33650-EEVB/KLM Attention: Mr. George W. Knighton, Chief October 3, 1985 Licensing Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Palo Verde Nuclear Generating Station Units 1, 2 and 3 Docket Nos. STN 50-528 (License NO. NPF-41)/529/530 Large Break LOCA Evaluation Model Reanalysis Results File: 85-056-026 G.1.01.10

References:

1) Letter to G. W. Knighton, NRC, from E. E. Van Brunt, Jr.,

ANPP, dated September 27, 1985 (ANPP-33583);

Subject:

Large Break LOCA Evaluation Model

2) Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,

ANPP, dated August 30, 1985;

Subject:

Large Break LOCA Evaluation Model Palo Verde Units 2 and 3

3) Letter to G. W. Knighton, NRC, from E. E. Van Brunt, Jr.,

ANPP, dated July 19, 1985 (ANPP-33068);

Subject:

Voluntary Reduction of Linear Heat Rate Limit

Dear Mr. Knighton:

Attached, as discussed in Reference 1), is the PVNGS specific limiting Large Break I,OCA (LBLOCA) analysis. This reanalysis was performed as requested by Reference 2).

ANPP believes the results of the reanalysis demonstrate PVNGS Units 1, 2 and 3 compliance with 10CFR50.46. Upon staff review and approval of this PVNGS specific analysis of the limiting LBLOCA, ANPP will limit on linear heat rate for Unit 1, as discussed in Reference 3).

lift the administrative If you should have any questions concerning this matter, contact Mr. W. F. +inn, of my staff.

Very truly urs

, 85i0080495 851003

..PDR

'P ADQCK 05000528 I .

E. E. Van Brunt, Jr.

PDR Executive Vice President Project Director EEVB/KLM/dim Attachment cc: E. A. Licitra gt R. P. Zimmerman A. C. Gehr

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ATTACHMENT PVNGS PLANT SPECIFIC LARGE BREAK LOCA ANALYSIS INTRODUCTION The limiting large break LOCA from the CESSAR-FSAR (1) , is the 1.0xDEG/PD.* This limiting case was reanalyzed for the PVNGS plants using the current NRC approved Evaluation Model (EM), except for the use of a more conservative axial power shape, and the following PVNGS plant specific characteristics which differ from CESSAR:

The PVNGS containment parameters (e.g., volume and passive heat sinks)','nd (2) PVNGS Safety Injection Pump flow performance n

These changes produce a net reduction in peak clad temperature (PCT) of approximately 80'F (PCT of 2091'F for the PVNGS specific analysis versus 2169'F for the CESSAR-FSAR at the same peak linear heat generation rate of 14.0 kw/ft).

II'ETHOD OF ANALYSIS The calculations reported in this section were performed using the C-E large break EM which is described in Reference 2, except for the selection of axial power shape. In the C-E EM, the CEFLASH-4A(3) computer program is used to determine the primary system flow parameters (4) computer program is during the blowdown phase, and the COMPERC-II used to determine the system behavior during the refill and reflood phases. The core flow and thermodynamic parameters from these two codes are used in the STRIKIN-II (5) computer program to computer calculate the hot rod clad temperature transient. The peak clad temperature and peak local clad oxidation percentage are therefore obtained from the STRIKIN-II calculation. The core-wide clad oxidation percentage is obtained from the results of both the STRIKIN-II and COMZIRC (4, Supplement 1) computer programs.

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The supplemental CESSAR-FSAR analysis (8) used the hot rod rupture criteria of NUREG-0630 and an improved steam cooling model. This analysis resulted in a net reduction in peak clad temperature of 18'F due to improved steam cooling at and above the rupture node and slight improvement in the two-phase (FLECHT) cooled clad temperatures below the rupture node. As in the CESSAR-FSAR analysis, the PVNGS plant specific large break LOCA analysis remains FLECHT cooled limited (the peak clad temperature occurring below the rupture node). Therefore, equivalent results can be expected with the NUREG-0630 rupture criteria model when applied to the, PVNGS plant specific analysis. The revised C-E evaluation model (7) which is currently under NRC review documents these and other model changes.

III. MAJOR ASSUMPTIONS The only model change from the'urrent EM is the selection of the limiting power axial shape. Instead of the axial shape used by the (2) current EM , this analysis uses a 1.52 axial peak located above the core mid-plane with a corresponding increase in the radial peaking to maintain the hot rod power peak at the current 14.0 kw/ft limit of the PVNGS Technical Specifications. (The radial peaking factor was actually increased beyond what is predicted for first cycle of the PVNGS units).

The axial power shape selection considered varying both the magnitude and position of the axial peak while maintaining the peak linear heat rate. With respect to large break LOCA analysis the selected shape bounds the most adverse power distribution allowed (by the PVNGS Technical Specifications limit on axial shape index) during the first cycle of the PVNGS units.

The containment parameters for this analysis are those of the PVNGS plants as detailed in Section 6.2.1.5 of the PVNGS FSAR. In addition, the minimum performance characteristics of the PVNGS safety injection system (SIS) pump flows were assumed for this analysis, as discussed in (9) . All other SIS, core, and system License NPF-34 Condition 2.C.(21) parameter assumptions of .the CESSAR-FSAR analysis (Sections 6.3.3.2.2 and 6.3.3.2.3) remain the same.

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The limiting break from the CESSAR-FSAR analysis (the 1.0xDEG/PD) should remain the limiting break with the more adverse axial shape and the imposed PVNGS specific plant characteristics. The small differences in the peak clad temperature (PCT) for the large break spectrum (ranging from 2149'F to 2169'F for 0.6 and 1.0xDEG/PD, xespectively) are due to a different core flow and thermodynamic values which yield slight variations in residual stored energy in the hot rod at the end of the blowdown phase. The axial shape and PVNGS specific plant characteristics imposed in this analysis have a negligble effect on the hot rod temperature calculation during the blowdown phase of the transient see Figure 1 which compares the hot rod temperature transients of the CESSAR-FSAR analysis and the PVNGS plant specific analysis. The blowdown portion of the transient is essentially unaffected by the axial shape and the Palo Verde plant specific characteristics. The core thermal-hydraulic conditions during the refill and reflood periods are the same for the spectrum of large breaks in the pump discharge leg. Based on this argument, the limiting large break determined in the CESSAR-FSAR analysis will therefore remain the same. In addition, the net reduction in PCT for this analysis is greater than the variation in PCT for various break sizes of the CESSAR-FSAR (approximately 80'F versus 20'F).

IV. RESULTS OF THE ANALYSIS The plant specific PVNGS large break analysis results in a peak clad temperature of 2091'F, peak local oxidation rate of 9.02X, and core wide oxidation rate of less than 0.799X at a peak linear heat generation rate of 14.0 kw/ft. These'esults are well within the 10CFR50.46 Acceptance Criteria for Emerge'ncy Co're Cooling Systems for Light-Water-Cooled Reactors (6) ; peak clad temperatures ~< 2200'F, peak local oxidation rates ~< 17X of 'the total cladding thickness before oxidation, and core-wide oxidation rates ( 1X.

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REFERENCES

1. System 80 Combustion Engineering Standard Safety Analysis Report (CESSAR), Final Safety Analysis Report, and Amendment 4 to the CESSAR-FSAR, dated July 16, 1981.

2. 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," February 1975 (Proprietary).

CENPD-132, Supplement 2, "Calculational Methods for the C-E Large Break LOCA Evaluation Model," July 1975 (Proprietary).

3. CENPD-133, "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis," April 1974 (Proprietary).

CENPD-133, Supplement 2, "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis (Modification)", December 1974 (Proprietary).

4. CENPD-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 (Mo'dification)," December 1974 (Proprietary).

5. CENPD-135, "STRIKIN-II,' 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)," December 1974 (Proprietary).

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REFERENCES (Continued) tt J,l

6. "Acceptance Criteria for Emergency )ICore Cooling Systems for Light-Water Cooled Nuclear Power Reactors'," Federal Register, Vol. 39, No. 3, Friday, January 4, 1974.

7. LD-85-032, "Revision to C-E Evaluation Model for Large Break LOCA Analysis," A. E. Scherer (C-E) to C. 0. Thomas (NRC), July 3, 1985.

8. LD-81-096, "Supplemental Analysis for NUREG-0630," A. E. Scherer (C-E) to R. L. Tedesco (NRC), December 21, 1981.

9. License NPF-34 Condition 2.C.(21)

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