ML20205A350

Forwards 10CFR50.46 Annual Rept for 1998,re Effects of ECCS Evaluation Model Mod on Peak Cladding Temp Results Since 1997 Annual Rept & Most Recent PCT Error Rept Submitted 980910

ML20205A350
Person / Time
Site:	Farley
Issue date:	03/19/1999
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NEL-99-0121, NEL-99-121, NUDOCS 9903300362
Download: ML20205A350 (11)

Text

Dave M: rey Sruthern Nuclear g..

Vice President Op: rating Company Farley Project P.O. Box 1295 Birmingham, Alabama 35201 Tel 205.992.5131 SOUTHERN COMPANY March 19, 1999 p,,,gy,, 3,,,, 7,,, y,,,,.

Docket Nos.:

50-348 NFL-99-0121 50-364 U. S. Nuclear Regulatory Comnussion A'ITN: Document Control Desk Wuhingtan. D. C. 20555-0001 Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1998 and Sinnificant Error Reocg1 i

Ladies and Gentlemen:

1 Provisions in 10 CFR 50.46 require applicants and holders of op: rating licenses or construction

)

permits to annually notify the Nuclear Regulatory Comnussion (NRC) of changes and errors in the Emergency Core Cooling System (ECCS) Evaluation Models. I:n compliance with this requirement, i

enclosed is the Southern Nuclear Operating Company (SNC) retort for Joseph M. Farley Nuclear Plant Units 1 and 2 for the calendar year 1998.

De annual report provides information regarding the effects of the ECCS Evaluation Model modifications on the peak claddmg temperature (PCT) results since the 1997 annual report and the most recent PCT Error Report submitted September 10,1998. Also, the attached annual report provides a summary of the plant changes performed under the provisions of 10 CFR 50.59 that also affect the PCT results. b repost is in accordance with the Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting (WCAP-13451).

j It has been determined that compliance with the requirements of 10 CFR 50.46 continues to be maintained when the effects of plant design changes are combined with the effects of the ECCS I

Evaluation Model changes and errors applicable to Farley Units 1 and 2.

[

This annual report is also serving as a 30 day Significant Error Report for large-break LOCA PCT.

M error is significant due to a correction of greater than 50'F in the vessel flow modeling. He change in increased contamment spray flow is also accounted for. This error anxl containment s, ray flow inccease was reported by Westinghouse to SNC on March 5,1999. %e impacted large-break LOCA has been re-analyzed for Steam Generator Replacement (SGR). He re-anal:$ corrected 00/

these errors and was submitted to the NRC (Enclosure Reference 11). %crefore, no reevaluation schedule is provided.

9903300362 990319 PDR ADOCK 05000348 PDR a R

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_ U. S. Nuclear Regulatory Commission c-i If there are any questions, please advise. Here are no commitments associated with this submittal.

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Respectfully submitted, i

SOUTHERNNUCLEAR OPERATING COMPANY 1

DaveMorey /

EWC:maf pct 98nrc. doc

Enclosure:

10 CFR 50.46 ECCS Evaluation Model 1998 AnnualReport

o Page 3 U. S. Nuclear Regilatary Commicsion cc:

Southern Nuclear Operatina Comoany Mr. L. M. Stinson, General Manager - Farley U. S. Nuclear Renulatory Cammi== ion. Washirmton. D. C.

Mr. J. I. Zimmerman, Licasing Project Manager - Farley U. S. Nuclear Regulatory Commission. Reaion II Mr. L. A. Reyes, Regional Admmistrator Mr. T. P. Johnson, Semor Resident Inspector - Farley i

ENCLOSURE Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model 1998 Annual Report i

I JOSEPH M. FARLEY NUCLEAR PLANT 10 CFR 50.46 ECCS EVALUATION MODEL 1998 ANNUAL REPORT j

L BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operatmg licenses or construction permits 4

to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluntion Models on an annual basis. 10 CFR 50.46 also requires that 1

I signina=at errors or changes in the ECCS Evaluation Model be reported to tk NRC within 30 days with a proposed schedule for providing a reanalysis or takmg other action as may be needed to show compliance with 10 CFR 50.46 requirements 10 CFR 50.46 defines a significant crror or change as one wluch results in a calculated fuel peak claddmg i-widure (PCT) different by more than 50*F from the t-waure calculated for the limiting tranaient using the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50*F.

In Reference 1, information was submitted to the NRC regardmg modi 6 cations to the Wa tia:W=> large-break and small-break Loss-of-Coolant Accident (LOCA) ECCS Evaluation Models as applicable to the Farley Nuclear Plant (FNP) analyses for the entendar year 1997.

W following presents an assessment of the effects of modifications to the Wa tiagb aa ECCS l

Evaluation Models on the Farley LOCA analysis results since the 1997 annual report (Reference 1) for the calendar year 1998. This annual report has been prepared in accordance with the Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting (WCAP-13451, Reference 2).

This annual report is also servmg as a 30 day Significant Error Report for large-break LOCA. This is due to an error in the reactor vessel flow modeling cod an inerease to containment spray flow that resulted in a PCT impact greater than 50*F, as reported to SNC by WwingWse on March 5,1999. h large-break LOCA has been re-analyzed for Steam Generator Rep:acement (SGR). & re-analysis corrected the error and has already been submitted to the NRC. No additional reanalysis schedule is needed IL LARGE-BREAK LOCA Table 1 shows the large-break LOCA PCT rack-ups for both Unit I and Unit 2.

ILA LARGE-BREAK LOCA ANALYSIS-OF-RECORD i

The large-break LOCA analyses fbr Farley Units 1 and 2 were av==sia~i to assess the effects of the changes and errors in the Wa tiagW=> large-break LOCA ECCS Evaluation Model on PCT results.

The large-break LOCA analysis +f-record results for Farley Units I and 2 were calculated using WantingWse's BE-LOCA analysis (Reference 3).

V i

ENCLOSURE-Page 2 h Urdt I and Unit 2 analyses assumed the following information kiireneiht to the large-break LOCA in

' the BE-LOCA analysis (Reference 3). One analysis was used to bound both Fadey Unit I and Unit 2.

Core Power = 2775 MWT 17x17 VANTAGE + Fuel Assembly Fo = 2.50 for VANTAGE + Fuel-FAH = 1.70 forVANTAGE+ Fuel SGTP = 20%

For Farley Units 1 and 2, the limiting size break analysis-of-record is a split break of the cold leg piping with a discharge coefficient of Co = 1.0. 'Ihe limiting PCT values determbaxi for the Unit I and Unit 2 large-break LOCAs are shown in Table 1.

II.B 199810 CFR 50.46 LOCA MODEL ASSESSMENTS The following changes and errors in the WMi=@=m ECCS Evaluation Model would affect the BE-LOCA Model.

II.B.1 Prior Repo:talAssesunsta A power uprate for both Farley Unit I and Unit 2 was implemented since the last reponing period. With the power uprate, a new ECCS model (BE-LOCA) was developed; therefore, there are no prior assessments noted.

II.B.2 1998 PCT Assessments Vessel Channel DX Error (impacts vessel channel flow modeline In the gap flow wall friction and interfacial drag coefficient calculation the incorrect cell height was used.

Rather than using cell specific heights (DX) at each level only one DX value was used. (Reference 9).

Increased Containment Sorav Mow Assessment An evaluation of RWST Uncertainties and Increase in Contamment Spray Flow Effects was performed.

In this evaluation, it was determined that there was an effect on PCT. (Reference 10).

Note that this error along with the impact of containment spray flow caused the large-break LOCA resuhs to be limiting at the first reflood rather than the second The sum of the absolute magnitude of the above changes in the first reflood PCT is 5'F.

1 a

r Enclosure Page 3 '

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De Vessel Channel DX error was ceirw&4 in the Steam Generator Replacement submittal to the NRC (R L s 11).

ILC 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS He adchtion of p...m :,1orage boxes in containment was evaluated and found not to cause a change to PCT (RLs 7).

ILD TOTAL RESULTANT LARGE-BREAK LOCA PCT As discussed above, the changes and errors to the W=h.g.ause large-break LOCA ECCS Evaluc tion Model could affect the large-break LOCA analysis results by alterung the PCT. As shown in Table 1, the large-break LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200*F.

ILE LARGE-BREAK LOCA CONCLUSIONS An evaluatma of the effects of changes and errors in the WMing6an large-break BE-LOCA ECCS Evaluation Model was pufvi.. 4 on the large-break LOCA applicable to the Farley.J..cc analysis.

When the effects of the large-break ECCS Evaluation Model changes and errors were combined with those of plant changes and the large-break LOCA analysis-of-record results, it was determined that Farley Units 1 and 2 are in compliance with the requirements of 10 CFR 50.46.

This ananal report is also serving as a 30 day Significant Error Report for large-break LOCA. His is due to an error in the Vessel DX calenlation as well as an increase in u=tainment spray flow assessment, which resulted in a total PCT impact greater than 50*F, as reported by Weiag6m to SNC on March 5,1999. %c large-break LOCA has been re-analyzed for Steam Generator Replacement (SGR). De re-analysis corrects this error and has already been submitted to the NRC.

IIL SMALL-BREAK LOCA Table 2 shows the small-break LOCA PCT rack-ups for both Unit I and Unit 2.

ILLA SMALL-BREAK LOCA ANALVMS-OF-RECORD De small-break LOCA analyses for Farley Units 1 and 2 were also examined to assess the effects of the changes and errors to the W~tiag6m small-break LOCA ECCS Evaluation Models on PCT results.

De small-break LOCA ECCS analysis results were calculated using the NOTRUMP small-break LOCA i

ECCS Evaluation Model (Reference 4).

De Unit I and Unit 2 analyses assumed the followmg ' formation important to the small-break LOCA m

analyses:

ENCLOSURE Page 4 Umt 1 Umt 2 Core Power = 1.02 X 2775 MWT Core Power = 1.02 x 2775 MWT 17x17 VANTAGE + Fuel Assembly 17x17 VANTAGE + Fuel Assembly

Fq = 2.50 FQ = 2.50 FMI = 1.70 FAH = 1.70 Upflow Configuration Downflow Omfiguration For Farley Units I and 2, the limiting size break analysis of-record for the VANTAGE + fuel analysis is a 3-inch diameter break in the cold leg. The limiting PCT values deterr.iined for the Unit I and Unit 2 17x17 VANTAGE + small-break are shown in Table 2. An analysir, for each Unit was performed due to Farley Power Uprate on Unit I and Unit 2 (Reference 3). Followbg the Significant Error Report (Reference 12) the limiting case became the high Tavg analysis.

III.B 199810 CFR 50.46 LOCA MODEL ASSESSME'.4TS The following changes and errors were identi6ed:

III.B.1 Prior Reported Asssi..ciits Since the SB-LOCA was reanalyzed for power uprate (Reference 3) and the Significant Error Report (Reference 12), there are no prior assessments noted.

III.B.2 1998 PCT Assessments There were no 1998 assessments other than those previously reported in reference 12.

III.C 10 CFR 50,59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS The use of Annular Pellets on Farley Unit 1 (previously reported for Farley 2 in Reference 12) has increased 10 CFR 50.59 Non-Model Impacts to PCT. Along with the Annular Pellet increase to PCT, coupled with this is an increase in the Burst and Blockage /fime in Life PCT. Both of these impacts were noted for Unit 2 in the last Signific. ant Error Report (Reference 12).

III.D TOTAL RESULTANT SMALL-BREAK LOCA PCT l

As discussed above, the changes and errors in the Westmginuse small-break LOCA ECCS Evaluation Model could affect the small-break LOCA analysis results by altering the PCT. As shown in Table 2, the small-break LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200*F.

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Enclosure Page 5 III.E SMALL-BREAK LOCA CONCLUSIONS An evaluation of the effects of changes and errors to the Westinghouse ECCS Evaluation Model was performed for the small-break LOCA analysis results, When the effects of the small-break ECCS Evaluation Model changes and errors were combined with those of plant changes and the small-break LOCA analysis-of-record results, it was deternuned that compliance with the requirements of 10 CFR 50.46 is maintamed for both Units 1 and 2.

. REFERENCES

1. Letter from D. N. Morey to USNRC, " Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1997," March 17,1998.

2. WCAP-13451, "Westmghouse Methodology for Implementation of 10 CFR 50.46 Reporting," dated October 1992.

3. Imer from D. N 'w to USNRC, " Joseph M. Farley Nuclear Plant Facility Operatmg Licenses and Technical Speutications Change Request for Power Upratmg," June 20,1997.

4. "Weia@m Small-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et. al, August 1985.

5. Joseph M. Farley Nuclear Unit 1 Cycle 16 Reload Safety Evaluation (10 CFR 50.59 Evaluation).

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6. Joseph M. Farley Nuclear Unit 2 Cycle 13 Reload Safety Evaluation (10 CFR 50.59 Evaluation).

7. SECL-97-062. Rev. ), October 20,1997.

8. ALA-98-176, "RCS Flow Report and Revised PCT Summary Sheets - SBLOCA Uprate Analysis with Original Steam Generators," August 13,1998.

9. ALA-99-041, "10 CFR 50.46 Annual Noti 6 cation and Reporting for 1998," March 5,1999.

10. ALA-98-333, " Evaluation of RWST Uncertainties and Increase in Containment Spray Flow Effects,"

December 16,1998 I1. Letter from D. N. Morey to USNRC, " Joseph M. Farley Nuclear Plant Steam Generator Replacement Related Technical Specifications Cimnge Request," December 1,1998.

12. Letter from D. N. Morey to USNRC. " Joseph M. Farley Nuclear Plant 10 CFR 50.46 Significant Error Report for SBLOCA," September 10,1998.

Enclosure Page 6 TABLEI JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT LARGE-BREAK LOCA PCT (OF)

A. ANALYSIS-OF-RECORD (VANTAGE-5)

Unit 1. *F Unit 2. 'F

1. ECCS Analysis ~

2004*

2004*

Total Analysis-of-Record PCT =

2004*

2004*

B. 199810 CFR 50.46 MODEL ASSESSMENTS i

1. Prior Reposted A-Tets 0

0

2. Vessel Channel DX 56" -

56 "

I

3. Evaluation ofImpact ofIncreased Contamment Spray Flow 9***

9*"

i C. 10 CFR50.59 PLANT MODIFICATIONS

~ 1. Addition of Permanent Storage Boxes in Containment 0""

0""

D. TOTAL RESULTANT LARGE-BREAK LOCA PCT 2069 2069 De PCT values are rounded up to the next highest integer number to avoid reporting in decimal points. These values were calculated using the BE-LOCA M6talagy as submitted by SNC and approved by the NRC for the uprate analysis. Herefore they do not directly compare with the results submitted with the 199710 CFR 50.46 Annual Report.

A Vessel Channel DX Error (Reference 9) was found. He code was only using one cell height rather than each individual cell height.

A WMiagbm evaluation ofIncrease of C=*=ia-t Spray Flow resulted in an increase in PCT (Reference 10).

"" The addition of Permanent Storage Boxes in Containment did not increase PCT (Reference 7).

o Enclosure Page 7 l

TABLE 2 I

t JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (DF)

I A.

ANALYSIS-OF-RECORD (VANTAOE-5)

Unit 1. *F_

Unit 2. *F

1. ECCS Analysis 1923*

1891*

2. Annular Pellets (Farley 2 Only) 0 10 "

3. Burst and Blockagefrime in Life 79 61 Total Analysis-of-Record PCT =

2002 1952 1

i

.B.

- 199710 CFR 50.46 MODEL ASSESSMENTS

1. Prior Reported A= = =ts 0

0 t

I C.

10 CFR 50.59 PLANT MODIFICATIONS

1. Additional Permanent Storage Boxes in Contamment 0

0

2. Change in Burst and Blockagefrime in Life 9*"

0

3. AnnularFuelPellets 10"*

0 D.

TOTAL RESULTANT SMALL-E.'J.AK LOCA PCT 2021 1962 ECCS Analysis following Farley Poo er Uprate and Significant Error Report (Reference 12)

He 10 CFR 50.59 Plant Modification for Annular Pellets was previously reported to the NRC in l

the PCT Error Report sent September 10,1998.

l De addition of Annual Pellets on Farley Unit 1 Cycle 16 caused in increase to PCT as well as a change to the Burst and Blockagefrime in Life values. (References 5 and 6)