ML20217D480
| ML20217D480 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/24/1998 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9803300031 | |
| Download: ML20217D480 (9) | |
Text
,
q Dave Morey Sotahera Nuclear Vice President.
Op:ta!ng Company Farley Project P.O. Box 1295 -
Birmingham Alabama 35201 Tel 205.992.5131 SOUTHERN h March 24,1998
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Docket Nos. 50-348 50-364
~U.S. Nuclear Regulatory Commission ATrN: Document Control Desk
' Washington, IX' 20555 Joseph M. Farley Nuclear Plant - Units I and 2
^
10 CFR 50.46 Annual ECCS Evaluation Model Chances Report for 1997 Ladies and Gentlemen:
Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to annually notify the Nuclear Regulatory Commission (NRC) of changes and errors in the Emergency Core Cooling System (ECCS) Evaluation Models. In compliance with this requirement, enclosed is the Southern Nuclear Operating Company's report for Joseph M. Farley Nuclear Plant Units I and 2 for the calendar year 1997.
The annual report provides information regarding the effects of the ECCS Evaluation Model I
modifications on the peak cladding temperature (PCT) results since the 1996 annual report. 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. The report is in accordance with the Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting (WCAP-13451).
I It has been determined that compliance with the requirements of 10 CFR 50.46 is maintained when the effects of plant design changes are combined with the effects of the ECCS Evaluation Model changes and errors applicable to Farley Units I and 2.
If there are any questions, please advise.
Respectfully submitted, f
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.ao D. N. Morey REM:maf 97PCTRI'T. doc
-Attachment cc:
Mr. L. A. Reyes, Region II Administrator-Mr. J. I. Zimmerman, NRR Project Manager Mr. T.' M. Ross, FNP Sr. Resident Inspector (b
l 9903300031 980324 "
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PDR ADOCK 05000348 Y'
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l ATTACHMENT Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Eval==* lam Model 1997 Annual Resort l
s e
JOSEPH M. FARLEY NUCLEAR PLANT 10 CFR 50.46 ECCS EVALUATION MODEL 1997 ANNUAL REPORT L
BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to notify the Nuclear Paa"1=*~y Commission (NRC) of errors and changes in the Esgy Core Cooling System (ECCS) Evaluation Models on an annual basis. 10 CFR 50.46 requires that sigmficant errors or changes in the ECCS Evaluation Model be soported to the NRC within 30 days with a proposed schedule for providag a reanalysis or taking other action as may be needed to show complimace with 10 CFR 50.46 requirements 10 CFR 50.46 defines a sigmficant error or change as one which results in a c=Im1**d fuel peak claddmg temperature (PCT) difforcat by more than 50*F from the 6p4are calculated for the limiting transient 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 regarding modifications to the Westinghouse large-break and small-break las-of-Coolant Accident (LOCA) ECCS Evaluation Models as applicable to the Farley Nuclear Plant (FNP) analyses for the calendar year 1996.
The followmg presents an assessment of the effects of anaddicmhans to the W~*iag6m ECCS Evaluation Models on the Farley LOCA analysis results (for the calcad=r year 1997) since the 1996 annual report (Reference 1). The 1997 annual report also reflects the recent reanalysis of the Unit 2 large-break LOCA impla==aa'ad in 1996 (Reference 5). This annual report has been prepared in accordance with the Wa*Iag6m Methodology for Impla=aa+=*iaa of 10 CFR 50.46 Reportmg (WCAP-13451, Reference 2). The results presentad in the snual repost as an analysisef-record for the large-break LOCA and small-break LOCA PCTs reflect the use of VANTAGE-5 fuel in both units (Reference 3).
11.
LARGE-BREAK LOCA Table I shows the large-break LOCA PCT rack-ups for both Unit I and Unit 2.
l II.A LARGE-BREAK LOCA ANALYSIS-OF-RECORD The large-break LOCA analyses for Farley Units 1 and 2 were aamined to assess the effects of the changes and errors in the PCT results.
The large-break LOCA analysis of record results for Farley Units I and 2 were calculated using the 1981 version of the Westing 6m large-break LOCA ECCS Evaluation Model incorporating the BASH analysis technology (Reference 4). 'Ihe large-break LOCA analysis for Unit 2 was revised and implemented in 1996 through the Cycle 12 reload safety evaluation process (Reference 5) to support the g&==+ +iaa of ZlRLO" claddiaa As discussed in Reference 5, to gain additional PCT margin in the Unit 2 reanalysis, the steam generator tube plugging limit was administratively reduced from 20% to 13%
(13% average,13% peak) in recognition of the fact that the actual plugging level was not expected to exceed 13% average or peak during Cycle 12 (see Table 1). Also, as da====ad in the Unit 1 Cycle 15 RSE (Reference 6), the steam generator tube plugging limit was raised from 10% to 13% to scrannt for a potential increase in the number of tubes that needed to be plugged during the refueling outage (see Table 1).
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ATTACHMENT Page 2 :
' %e Unit 1 and Unit 2 analyses===amad bs fa W inrn,==tian important to tbs large break LOCA in t
the BASH analysis. The ah*= of==tal==aat mini puras auto isolation and==hi =i safe shutdown earthquake (SSE) plus LOCA events have been explicidy included in the Unit 2 revised analysis (Reference 5).
Unit 1 Unit 2 Core Power = 1.02 x 2652 MWT Core Power = 1.02 x 2652 MWT 17x17 VANTAGE-5 Fuel Assembly 17x17 VANTAGE-5 Fuel Assembly Fq = 2.45 for VANTAGE-5 Fuel Fq = 2.45 for VANTAGE-5 Fuel Fq = 2.32 for LOPAR Fuci Fq = 2.32 for LOPAR Fuci FAH = 1,70 for VANTAGE-5 Fuel FAH = 1.70* for VANTAGE-5 Fuel FAH = 1.55 for LOPAR Fuel FAH = 1.55" for LOPAR Fuel SGTP*" = 20%
SGTP*" = 20%
Upflow Configuration Downflow C=rumrataan
- De trannad value r
- a~i at 1.65 during 1996.
" The licensad value was reduced to 1.30 during 1996.
"*SGTP = Steam generator tube plugging limit===umad in the LOCA analysis. He limit was administratively revised to 13% in 1997 (Unit 1) and 13% in 1996 (Unit 2); see Table 1.
For Farley Units 1 and 2, the limiting size break analysis-of-record is a double-ended guillotme rupture of the cold leg piping with a discharge coefficient Co f 0.4. %e limiting PCTs determined for the Unit I o
and Unit 2 large-break are shown in Table 1. The Unit I analysis of-record limiting PCT value includes 3T for containment mini-purge automatic i@daa. 87 for increased Tavg temperature uncertainty, and 6T for combined SSE and LOCA events. It is nouxi that the 50T transition core penalty has been removed by the Unit 1 Cycle 15 reload safety evaluatson (Reference 6) since there are no LOPAR fuct l
assemblics loaded in that core. Althoudi there are still 28 LOPAR fuel assemblics in the Unit 2 core, the l
transition core penalty was also removed for Unit 2 by taking credit for reduced powers in the l
VANTAGE-5 assemblics that are adjacent to LOPAR assemblies (see Reference 5). In addition, both l
units contain 1.5X IFBAs with 100 psi hackfill pressure, which introduces a 77 PCT penalty for Unit 1 l
(Reference 6) and a IST PCT penalty for Unit 2 (Reference 5).
II.B 199710 CFR 50.46 LOCA MODEL ASSESSMENTS
%e following changes and errors in tbs Westingimuse ECCS Evaluation Model ah the 1981 Evaluation Model with BASH results obtained for tbs Farley analysis.
H.B.1 Prior Raponed Anacasmants
%e prior large-bensk LOCA PCT===ansaients given in Table I were submittad to the NRC on April 1, 1997, as part of the 1996 Annual Report (Reference 1). It is noted in Table 1 that the previous changes and errors were corrected in the recent reanalysis of the large-break LOCA for Unit 2 (Reference 5).
.11.B.2 1997 PCT Assessments
ATTACHMENT Page 3 ILC ' 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS As reported in Reference 1 and as noted in Table 1, the accumulator water temperature was increased from 90*F to 120*F for ' Unit 1. For Unit 2, the ---* temperature of 120*F was explicdly used in the reanalysis (IL', a 5).
J II.D TOTAL RESULTANT EARGE-BREAK LOCA PCT As A=ca=W above, the changes and errors to the Westeghouse lars> break LOCA ECCS Evaluation Model could affect the large4 reek I4CA analysis results by ahoring the PCT. As shown in Table 1, the largo 4reak LOCA analysis PCT results for both units am below tha 10 CFR 50.46 limit of 22000F.
]
II.E LARGE-BREAK LOCA CONCLUSIONS An evaluation of the effects of changes and errors in the Westmghnuse large-break BASH ECCS Evaluation Model was performed on the large-break LOCA applicable to the Farley referece analysis.
When the effects of the large-break ECCS Evaluation Model changes and errors were cMiH with those of plant changes and the large-break LOCA analysis-of-record results, it was determmed that Farley Units 1 and 2 were in compliance with the requirements of 10 CFR 50.46.
IIL SMALL-BREAK LOCA Table 2 shows the small break LOCA PCT rack-ups for both Unit I and Unit 2.
III.A SMALL-BREAK LOCA ANALYSIS-OF-RECORD h small-break LOCA analyses for Farley Units 1 and 2 were evamiaM to assess the effects of the changes and errors to the WwiaM > small-break LOCA ECCS Evnin=tian Models on PCT results
'Ihe small-break LOCA ECCS analysis resuks were c=Ma*M using the NOTRUMP small-break LOCA ECCS Evaluation Model(ILLw 7).
The Unit I and Unit 2 analyses assumed the following informatina important to the small-break LOCA analyses:
Umt 1 Unit 2 Core Power = 1.02 x 2775 MWT Core Power = 1.02 x 2775 MWT 17x17 VANTAGE-5 Fuel Assembly 17x17 VANTAGE-5 Fuel Assembly Fq = 2.50 Fq = 2.50 FMI=1.70 FAH = 1.70 Upflow Configuration Downflow Configuration For Farley Units 1 and 2, the innitag siac break analysis-of-record for the VANTAGE-5 fuel analysis is a 3-inch diameter break in the cold leg. h hmiting PCTs determuuvi for the Unit 1 and Unit 217x17
. VANTAGE-5 small-break are shown in Table 2. Both the Unit 1 and Unit 2 analysis-of-record limiting PCT values include a 20*F pensky due to the increased Tavs Qeure uncertainty.
=
ATTACHMENT Page 4 III.B 199710 CFR 50.46 LOCA MODEL ASSESSMENTS The following changes and errors were identirmi:
III.B.1 Prior Reoorted Amee<ments The prior small-break LOCA PCT assessments shown in Table 2 were submitted to the NRC in Reference 1.
III.B.2 SBLOCTA Clad Burst Strain Error As reported in Reference 8, an error was discovered in the SBLOCTA code related to improper calculation of clad post burst strain. Specifically, the error occurs hacane although the burst nade is predicted to strain out upon the occurrence of burst, incorrect codmg logic enued this burst strain to be negbread in subsequent time steps. The main effect for small break transient Whtban is that for high peak clad temperature cases in excess of 1800*F, which are also limited by the rapid zirc-water reaction accompanying incipient clad burst, the smaller clad diameters miuce the burst temperature spike and upon correction of the codmg, a net PCT penalty may result.
For Farley Units I and 2, the effects have been determined to add no PCT penalty for each unit since a prior version of SBLOCTA without the error was used.
III.C 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS There have been no non-zero non-model PCT assessments under 10 CFR 50.59 made against the reference VANTAGE-5 LOCA analysis results. It should be noted that the effects of all of the applicable previous evaluations for Farley Units 1 and 2 were incorporated into the VANTAGE-5 analysis.
III.D TOTAL RESULTANT SMALL-BREAK LOCA PCT As dimMabove, the changes and errors in the Westmghouse small-break LOCA ECCS Evaluation Model could affect the small-break LOCA analysis results by altering the PCT as shown in Table 2.
III.E SMALL-BREAK L.OCA CONCLUSIONS l
An evaluation of the effects of changes and errors to the Westinghouse ECCS Evaluation Model was l
performed for the small-break LOCA analysis results. Wien the effects of tim 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 determined that compliance with of 10 CFR 50.46 was maintained for Units 1 and 2.
ATTACHMENT Page 5 IV, ' REFERENCES
- 1. Letter from D. N. Morey to USNRC, " Joseph M. Faricy Nucleu Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1996," April 1,1997,
- 2. WCAP-13451, "Westmghouse Methodology for Implemenwinn of 10 CFR 50.46 Peporting," dated October 1992.
- 3. NRC Safety Evaluation Report, "h="== of Ammiment No. 92 to Facility Operating f inensa No.
NPF-2 and An** No. 85 to Facility Opera 6ag I icema No. NPF-8 Regarding the Use of VANTAGE-5 Fuel in Both Units and Allowung Removal and Replacement of the Resistance Temperature Detector Bypa.is Manifold System in Unit 2 - Joseph M. Farley Nuclear Plant, Units 1 and 2 (TAC Nos. M81025 and M81026)," March 11,1992.
4.
"The 1981 Version of the Westmghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, Rev. 2 (Proprietary), Young, M. Y., et. al, March 1987.
- 5. Joseph M. Farley Nuclear Plant Unit 2 Cycle 12 Reload Safety Evaluaten (10 CFR 50.59 Evaluation), letter CAF-NF-1564 dated October 2,1996.
- 6. Joseph M. Farley Nuclear Plant Unit ! Cycle 15 Reload Safety Evaluation (10 CFR 50.59 Evaluation), letter CAF-NF-1597 dated February 28,1997.
- 7. " Westinghouse 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.
- 8. Ixtter from T. W. Wallace to D. N. Morey, "10 CFR 50.46 Annual Notification and Reporting for 1997."
l
TABLE 1 JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT LARGE-BREAK LOCA PCT (DF)
A. ANALYSIS-OF-RECORD (VANTAGE-5)
Unit 1. T Unit 2_T
- 1. SCCS Analysis 1896* '
2042 "
- 2. rnataia===t Mini-Purge Auto taalatian 3
0**
- 3. Tavg Temperature Uncertainty 8
8"
0"
- 5. Transition Core Penalty 0(*)
0*)"
l 6.
SO Tube Pluggmg Margin
-28 )
-28*
- 6 7.
1.5 x IFBA 7
__U.
Total Analysis-of-Record PCT =
1892*
2037 "
I B. 199710 CFR 50.46 MODEL ASSESSMENTS
- 1. Prior Reported A-. sits 9
15 2.
1997 PCT Assessaunts 0
0 C. 10 CFR 50.59 PLANT MODIFICATIONS
- 1. Increased Ace _-w~ Water Temperature 48 0"
l D. TOTALRESULTANTLARGE-BREAKLOCAPCT 1949 2052 (a)
De Unit 1 Transition Core Penalty has beni removed since the core contains all VANTAGE-5 fuel.
(b)
The Unit 2 Transition Core Penalty has been removed by taking credit for reduced powers in the VANTAGE-5 assemblies that are a4acent to the LOPAR assemblies (see Refermcc 5).
(c)
To account for a potential increase in steam generator tube plugging, the Unit I limit was raised from 10% to 13% administrative limit (Reference 6).
(d)
To gain additional PCT margin, the steam generator tabe plugging limit was reduced from a 20% to a 13% administrative limit (Reference 5).
He PCT values are rounded up to the next highest integer number to avoid reporting in decimal Pomts He Unit 2 results correspond to the revised LOCA analysis performed to support the use of ZlRLO and a revised floodmg rate (Reference 5).
g TABLE 2 JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (OF)
A.
ANALYSIS-OF-RECORD (VANTAGE-5)
Unit 1. *F Unit 2. *F
- 1. ECCS Analysis 1785*
1763*
- 2. Tavg Tanperature Uncertamty
_2Q
_2Q Total Analysis-of-Record PCT =
1805 1783 B.
199710 CFR 50.46 MODEL ASSESSMENTS
- 1. Prior Reported Assessments 232*
104*
- 2. SBLOCTA Clad Burst Strain Error 0"
0"
- 3. Change in Burst and Blockagefrine in Life 0"*
0*"
C.
10 CFR 50.59 PLANT MODIFICATIONS None 0
0 D.
TOTAL RESULTANT SMALL-BREAK LOCA PCT 2037 1887 19% Annual Report to the NRC under 10 CFR 50.46 in Reference 1.
See Referenoc 8.
According to Reference 1, the total r-h for change in Burst and Blockagefrime in Life are 96*F for Unit I and 31*F for Unit 2.