ML20101J198
| ML20101J198 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/25/1996 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9604010332 | |
| Download: ML20101J198 (9) | |
Text
l Southzrn Nucisar Opertting Company Post Office Box 1295 Birmingham, Alabama 35201
- Tsliphone (205) 868-5131 k
L Southern Nudear Operating Company Dave Morey re r ect Marr.h 25,1996 the southem electnc system Docket Nos.: 50-348 10 CFR 50.46 50-364 U. S. Nuclear Regulatory Commission ATTN: Document ControlDesk' Washington, D.C. 20555 Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Reoort for 1995 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 l
with this requirement, enclosed is the Southern Nuclear Operating Company's report for Joseph M. Farley Nuclear Plant Units 1 and 2 for the calendar year 1995.
The annual report provides information regarding the effects of the ECCS Evaluation Model J
modifications on the peak cladding temperature (PCT) results since the 1994 annual report.
Also, the attached annual report provides a summary of the plant changes performed under the l
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).
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 Evaluation Model changes and errors applicable to Farley Units 1 and 2.
If you have any questions, please advise.
Respectfully submitted,
/
9604010332 960325 DR ADOCK 0500 a
Dave Morey i l
REM / cit:50-46 pct. doc l
. Attachment j
cc:
Mr. S. D. Ebneter, Region II Administrator
[
Mr. B. L. Siegel, NRR Senior Project Manager 0
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Mr. T. M. Ross, FNP Sr. Resident Inspector l
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ATTACHMENT Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model 1995 AnnualReport i
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ATTACHMENT JOSEPH M.FARLEY NUCLEAR PLANT 10 CFR 50.46 ECCS EVALUATION MODEL 1995 ANNUAL REPORT L
BACKGROUND Provisiocs in 10 CFR 50.46 require applicants and holders of operatmg licenses or construction permits to notify the Nuclear Regulatory Comnussion (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models on an annual basis. 10 CFR 50.46 requires that significant errors or changes in the ECCS Eval =hna Model be reported to the NRC l
within 30 days with a proposed schedule for providmg 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 error or change as one which results in a calculated fuel peak claddmg temperature (PCT) different by more than 50*F from the t..r.4are calculated for the limiting transient using i
the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective a...rigure changes is greater than 50*F.
In Reference 1, information was submitted to the NRC regardmg modifications to the W@iaghase large-break and small-break Loss-of-Coolant Accident (LOCA) ECCS Evaluation Models as applicable to the Farley Nuclear Plant (FNP) analyses for the calendar year 1994.
The followmg presents an assessment of the effects of modifications to the Westmghouse ECCS Evaluation Models on the Farley LOCA analysis results since the 1994 annual report for the calendar year 1995. The 1995 annual report also reflects the recent reanalysis of the Unit 2 large-break LOCA implemented in 1995 (Reference 5). This annual report has been prepared in l
accordance with the Westmghouse Methodology for Implementation of 10 CFR 50.46 Reporting I
(WCAP-13451, Reference 2). 'Ihe results presented in the annual report as an analysis-of-record for the large-break LOCA and small-break LOCA PCTs reflect the use of VANTAGE-5 fuel in both units (Reference 3).
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 The Isrge-break LOCA analyses for Farley Units 1 and 2 were w & -4 to assess the effects of the changes and errors in the Weingbaa large-break LOCA ECCS Evaluation Model on PCT results.
The large-break LOCA asialysis-of-record results for Farley Units 1 and 2 were calculated using the 1981 version of the Westmghouse large-break LOCA ECCS Evaluation Model incorporating the BASH analysis technology (Reference 4). The large-break LOCA analysis for Unit 2 was i
i revised and implemented in 1995 through the Cycle 11 reload safety evaluation process (Reference i
- 5) to support increasing hot assembly average power, P-bar, from 1.42 to 1.514, increasing the f
nuclear enthalpy hot channel factor, F%, from 1.65 to 1.70 (licensed value remamed at 1.65 during 1995), and increasmg the accumulator water temperature from 90*F to 120*F. As discussed i
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A*=d-at 10CFR50.46 ECCS Evaluaten Model 1995 Annual Report in Reference 5, in order to gain additional PCT margm in the Unit 2 reanalysis, the steam generator tube pluggmg limit was reduced from 20% to 10% (10% average,10% peak), admmistratively, in recognition of the fact that the actual pluggmg level was not ava~+d o exceed 10% average or t
peak during Cycle 11 (see Table 1). As seen in Table 1, the reduction in the steam generator tube plugging limit was also adopted for Unit I through the Cycle 14 reload safety evaluation process (Reference 6).
'Ihe Unit I and Unit 2 analyses assumed the following information important to the large-break LOCA in the BASH analysis.
Umt 1 Umt 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 Fo = 2.45 for VANTAGE-5 Fuel Fo = 2.32 for LOPAR Fuel Fo = 2.32 for LOPAR Fuel FM = 1.70 for VANTAGE-5 Fuel FM = 1.70* for VANTAGE-5 Fuel FM = 1.55 for LOPAR Fuel FM = 1.55 for LOPAR Fuel SGTP" = 20%
SGTP** = 20%
Upflow Configuration Downflow Configuration
- 'Ihe licensed value r==iaad at 1.65 during 1995.
" SGTP = Steam generator tube pluggmg limit assumed in the LOCA analysis. The limit was reduced admmistratively to 10% in 1995 in order to gain PCT margin (see Table 1).
For Farley Units 1 and 2, the hnutmg size break analysis of-record is a double-nded guillotine rupture of the cold leg piping with a discharge coefficient of Co = 0.4. 'Ihe limiting PCTs determmed for the Unit I 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 isolation, 8T for increased Tavg temperature uncertamty, and 67 for combined safe shutdown earthquake (SSE) and LOCA events. 'Ihe effects of containment mini-purge auto isolation and combined SSE plus LOCA events have been explicitly included in the Unit 2 revised analysis (Reference 5); however, the 8'F penalty for Tavg twindre uncertamty remams It is noted that the 50T transition core penalty has been removed by the Unit 1 Cycle 14 reload safety evaluation (Reference 6) since there are no LOPAR fbcl assemblies loaded in the Unit 1 Cycle 14 core. However, the 50T transition core penalty has been retamed by the Unit 2 Cycle 11 reload safety evaluation (Refeience 5) since there are still 33 LOPAR fuel assemblics loaded in the Unit 2 Cycle 11 core. In addition, Unit 1 Cycle 14 contams 1.5X IFBAs with 100 psi backfill pressure, wluch has shown to introduce a 7T PCT penahy for Unit 1 (Rds.cc 6). Unit 2 does not contairs any 1.5X IFBA with 100 psi backfill pressure and, as such, is unaffected by the additiond penalty (Reference 5).
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An=rhment 4
10CFR50.46 ECCS Evaluation Model 1995 Annual Report i
ILB 199510 CFR 50.46 LOCA MODEL ASSESSMENTS
'Ihere are no changes and errors in the WMiaghase ECCS Evaluation Models found in 1995.
Below are the previously reported changes and errors in the Westmghouse ECCS Evaluation Models affectmg the BASH Evaluation Model large-break LOCA analysis-of-record results.
II.B.1 PnorReported Assessments The prior large-break LOCA PCT assessments given in Table I were submitted to the NRC in March 1995 as part of the 1994 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).
ILC 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS As mentioned earlier and as noted in Table 1, the accumulator water temperature was increased from 90 F to 120*F for Unit I through the Cycle 14 reload safety evaluation process (Reference 6).
For Unit 2, the accumulator temperature of 120'F was explicitly used in the reanalysis (Reference 5).
i ILD TOTAL RESULTANT LARGE-BREAK LOCA PCT As discussed above, the changes and errors to the WMiagbar large-break LOCA ECCS Evaluation Model could affect the large-break LOCA analysis results by altering 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 22000F.
ILE LARGE-BREAK LOCA CONCLUSIONS i
An evaluation of the effects of changes and errors in the WM agbar large-break BASH ECCS Evaluation Model was performed on the large-break LOCA applicable to the Farley reference 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 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.
IILA SMALL-BREAK LOCA ANALYSIS-OF-RECORD The small-break LOCA analyses for Farley Units 1 and 2 were also examined to assess the effects of the changes and errors to the Westmghouse small-break LOCA ECCS Evaluation Models on PCT results. The small-break LOCA ECCS analysis results were calculated using the NOTRUMP small-break LOCA ECCS Evaluaten Model (Reference 7).
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A +'=^=aat 10CFR50.46 ECCS Evaluation Model 1995 Annual Report The Unit I and Unit 2 analyses assumed the followmg information important to the small-break LOCA analyses:
Umt i 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 FAH = 1.70 FAH = 1.70 Upflow Configuration Downflow Configuration For Farley Units 1 and 2, the limiting size break analysis-of-record for the VANTAGE-5 fuel analysis is a 3-inch diameter break in the cold leg. 'Ihe linutmg PCTs determmed for the Unit I and Unit 217x17 VANTAGE-5 small-break are shown in Table 2. Both the Unit I and Unit 2 analysis of-record limiting PCT values include a 20*F penalty due to the increased Tavg temperature uncertainty.
III.B 199510 CFR 50.46 LOCA MODEL ASSESSMENTS
'Ihe following changes and errors in the Weingbm ECCS Evaluation Models would affect the NOTRUMP small-break LOCA analysis results obtamed for the Farley VANTAGE-5 fuel analysis.
III.B.1 Prior Reported A=-mante The prior small-break LOCA PCT assessments shown in Table 2 were submitted to the NRC in Reference 1.
III.B.2 NOTRUMP Specific Enthalpv Error A typographical error was found in a line of coding in the NOTRUMP code. Although the equation in the NOTRUMP topical report is correct, the codmg represented the last term as a partial derivative with respect to the fluid node mixture region total energy instead of the mixture region total mass. The genene effect resulted in an M== tad penalty of 20*F for both Unit I and i
Unit 2.
III.C 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS
'lhere have been no non-zero non-model PCT assessments under 10 CFR 50.59 made against the reference VANTAGE-5 LOCA analysis results to date. It should be noted that the effects of all of the vplicable previous evaluations for both Farley Units I and 2 were incorporated into the VANTAGE-5 analysis.
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At**ehmaat i
10CFR50.46 ECCS Evaluation Model 1995 Annual Report i
III.D TOTAL RF3tULTANT SMALL-BREAK LOCA PCT As discussed above, the changes and errors in the Weiaghat-small-break LOCA ECCS l
Evaluaten Model could affect the small-break LOCA analysis results by altering the PCT as l
shown in Table 2.
l IILE SMALL-BREAK LOCA CONCLUSIONS l
An evaluation of the effects of changes and errors to the Weingivmse 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 detemuned that compliance with the requirements of 10 CFR 50.46 would be maintained for both Units 1 and 2.
IV.
REFERENCES
- 1. I.etter from D. N. Morey to USNRC, " Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1994," March 20,1995.
- 2. WCAP-13451, "Wst-gi.wse Methodology for Implementation of 10 CFR 50.46 Reporting,"
dated October 1992.
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- 3. NRC Safety Evaluation Report, " Issuance of Aa* No. 92 to Facility Operatmg License No. NPF-2 and Amendment No. 85 to Facility Operatmg License No. NPF-8 Regardmg the Use of VANTAGE-5 Fuel in Both Units and Allowmg Removal and Replace =aa' of the Resistance Temperature Detector Bypass 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 Weinghause 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 11 Reload Safety Evaluation - Revision 2 (10 CFR 50.59 Evaluation), letter CAF-NF-1455 dated April 7,1995.
- 6. Joseph M. Farley Nuclear Plant Unit 1 Cycle 14 Reload Safety Evaluation (10 CFR 50.59 Evaluation), letter CAF-NF-1479 dated September 1,1995.
l 7.
"WM nghause Small-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-i l
10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et. al, August 1985.
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Attachment I
10CFR50.46 ECCS Evaluation Model 1995 Annual Report TABLE 1 JOSEPH M.FARLEYNUCLEAR PLANT TOTAL RESULTANT LARGE-BREAK LOCA PCT (OF)
A. ANALYSIS-OF-RECORD (VANTAGE-5)
Unit 1. *F Unit 2.*F
- 1. ECCS Analysis 1896*
2120 "
- 2. Contamment Mini-Purge Auto Isolation 3
0"
- 3. Tavg Temperature Uncertamty 8
8" 4.
Combined SSE and LOCA Events 6
0"
- 5. Transition Core Penalty ON N
50 "
- 6. SG Tube Pluggmg Margin of10%
-40M
-40" 7.
1.5X IFBA 7
0 Total Analysis-of-Record PCT =
1880*
2138 "
B. 199510 CFR 50.46 MODEL ASSESSMENTS
- 1. Prior Reported As-ments 6"*
0"*
C. 10 CFR 50.59 PLANT MODIFICATIONS
- 1. Increased AccumulatorWaterTemperature 48 0"
D. TOTAL RESULTANT LARGE-BREAK LOCA PCT 1922 2138 "
(a)
The Unit 1 Transition Core Penalty has been removed since the core contains all VANTAGE-5 fuel.
(b)
Unit 2 still contains LOPAR fuel assemblies due to the redesign at EOC-10.
(c)
To gain additional PCT margin, the steam generator tube pluggmg limit us reduced from 20% to 10%, similar to that of Unit 2 (Reference 5).
The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points.
The Unit 2 results corisycad to the revised LOCA analysis performed as part of the increased peaking factors and accumulator water temperature (Reference 5).
The Structural Metal Heat Modeling correction (-250F) and the LUCIFER error correction (-6*F) for Unit 2 and the LUCIFER error correction (-6*F) for Unit I were submitted to the NRC in March 1995 as part of the 1994 Annual Repart (Reference 1). However, in the recent reanalysis of Unit 2 to increase the peakmg factors and accumulator water temperature, the above corrections were explicitly Mc=a*M for in the large-break LOCA for Unit 2 (Reference 5).
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4 A*eh-t 10CFR50.46 ECCS Evaluation Model 1995 Annual Report 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 Temperature Uncertainty
__20
_ 20 Total Analysis-of-Record PCT =
1805 1783 B.
199510 CFR 50.46 MODEL ASSESSMENTS
- 1. PriorReported Assessments 171*
56*
- 2. Boiling Heat Transfer Correlation Error 20 20 3.
Change in Burst and Blockagefrime in Life 17 "
8" C.
10 CFR 50.59 PLANT MODIFICATIONS None 0
0 D.
TOTAL RESULTANT SMALL-BREAY. LOCA PCT 2013 1867 i
Reported to the NRC under 10 CFR 50.46 in Reference 1.
For Burst and Blockagefrime in Life, penalties of 67*F for Unit I and 15*F for Unit 2 were included in B.1 above as previously reported to the NRC in Reference 1. Item B.3 reflects changes to the reported values due to the specific enthalpy error in B.2 and since the Burst and Blockagefrime in Life penalty is a function of PCT. Thus, the total penalties for change in Burst and Blockagefrime in Life are 84*F for Unit I and 23*F for Unit 2.
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