ML20137L289
| ML20137L289 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/31/1997 |
| From: | Mccoy C SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-0998, LCV-998, NUDOCS 9704070192 | |
| Download: ML20137L289 (11) | |
Text
' C. K. McCoy Seatern Nuclear -
Vce President
. 0perating Company,Inc.~
i
' Vogtle Project.
40 inverness Center Parkway -
. P.O. Box 1295 LMarch'31,"l'997 BkminpamAahma 35201 Tel 205 992.7122 Fax 205.932.0403 Docket Nos. 50-424 50-425 Energy to Serve YourWorld" U. S.' Nuclear Regulatory Commission -
LCV-0998 ATTN: Document Control Desk Washington, D. C. 20555 :
Gentlemen VOGTLE ELECTRIC GENERATING PLANT
- 10CFR50.46 ECCS EVALUATION MODELS 1996 ANNUAL REPORT AND SIGNIFICANT ERROR REPORT
~ Attached is Southern Nuclear Operating Company, Inc.'s (Southern Nuclear) 10CFR50.46 Emergency Core Cooling System (ECCS) Evaluation Models 1996 Annual Report based on WCAP-13451 and in compliance with the reporting requirements of 10CFR50.46(a)(3)(ii). It is based on information provided by Westinghouse of errors and changes assessed against the L
Vogtle Electric Generating Plant (VEGP) ECCS Evaluation Models since the 1995 Annual Report.
The attached report summarizes the effects of changes and errors in the ECCS Evaluation Models on peak clad temperature (PCT). Also, the report provides a summary of the plant change safety evaluations performed under the provisions of 10CFR50.59 that also affect PCT.
The report results will be incorporated into a future Final Safety Analysis Report (FSAR) update.
Upon review of the information provided by Westinghouse for this 1996 Annual Report, it was discovered that a new assessment against the VEGP LBLOCA EM resulted in an assessed sum of the absolute magnitude of the assessments to be slightly greater than 50*F PCT for the BASH l
LBLOCA Model results. Therefore, this annual report also serves as a Significant Error Report. However, the net effect of the assessment is a minor reduction in the LBLOCA PCT
/
for VEGP. The items which resulted in an absolute sum of 56 F are 1) steam generator flow
/p area application, 2) structural metal heat model, 3) LUCIFER error correction, and 4) translation of fluid conditions from SATAN to LOCTA. The SBLOCA PCT results are not affected by these items. Southern Nuclear proposes not to reanalyze the BASH part of the LBLOCA ECCS anlaysis at this time for the following reason:
)
.(1)
The current magnitude of the PCT for the LBLOCA eDent maintains considerable margin to the 2200*F acceptance criteria. In fact, the cumulative assessments result in a net benefit (6*F in PCT) to the previously reported analysis-of-record PCT.
F 9704070192 970331 PDR ADOCK 05000424 p
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07002t a
i l
l U. S. Nuclear Regulitory Commission Page 2 j
l (2) The overall conservatism in the LBLOCA Evaluation Model has been approved by the NRC.
(3) The revision of the BASH code is expected to result in a net PCT benefit upon future reanalysis.
(4)
The PCT analysis-of-record value of 191l'F was based on recent model changes involving partial reanalysis of the LBLOCA event using the LOCBART code.
(5)
The assessments performed by Westinghouse to address new issues and phenomena since the previous analyses were appropriately conservative.
Southern Nuclear will incorporate the new BASH model at the next licensing action requiring reanalysis of the LBLOCA and for which the BASH code would be used within the Evaluation Model.
Based on the attached 1996 Annual Repod, it has been determined that compliance with the requirements of 10CFR50.46 continues to be maintained when the effects of plant design changes are combined with the effects of the ECCS Evaluation Models assessments applicable to VEGP Units 1 and 2.
Ifyou have any questions regarding this report, please contact this office.
Sincerely, C. K. McCoy i
CKM/ HWM/gmb Attachment cc: Southern Nuclear Operating Company Mr. J. B. Beasley Mr. M. Sheibani NORMS U. S Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. L. L. Wheeler, Licensing Project Manager, NRR Mr. C. R. Ogle, Senior Resident Inspector, Vogtle LCWO327-B l
ATTACHMENT VOGTLE ELECTRIC GENERATING PLANT 10 CFR'50.46 ECCS EVALUATION MODELS 1996 ANNUAL REPORT AND SIGNIFICANT ERROR REPORT BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models on an annual basis when the errors and changes are not significant, and within 30 days of discovery when the errors and changes are significant. A significant error or change, as defined by 10 CFR 50.46, is one which results in a calculated fuel peak cladding temperature (PCT) different by more than 50 F from the temperature calculated for the limiting transient using the last acceptable model, or a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50*F.
The following presents a summary of the effects of errors and changes to the Westinghouse ECCS Evaluation Models on the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 loss of coolant accident (LOCA) analyses since the 1995 annual report (Reference 1). This annual repon has been prepared in accordance with the methodology presented in WCAP-13451 (Reference 2). This annual report also serves as a significant change report since the large-break LOCA assessments exceed an absolute sum of 50 F PCT. The LBLOCA and SBLOCA 4
analyses, Evaluation Model assessments, and safety evaluation results reponed herein will be included in a future VEGP Final Safety Analysis Report (FSAR) update.
LARGE-BREAK LOCA ECCS Evaluation Modej Since the previous report (Reference 1), one new assessment against the VEGP LBLOCA analysis has been identified. When this assessment is combined with the earlier assessments listed in Reference 1, the absolute sum of the assessment values slightly exceeds 50*F PCT.
The LBLOCA analysis results are based on the Westinghouse BASH large-break ECCS Evaluation Model (Reference 3), as approved by the NRC for VEGP-specific application (References 4 and 5) and the latest acceptable LOCBART model. The limiting size break analysis continues to assume the following information important to the LBLOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly 1
o Core Power = 1.02
- 3565 MWT o
Vessel Average Temperature = 571.9*F o
Steam Generator Plugging Level = 10%
o Fq = 2.50 o
Fall = 1.65
ATTACHMENT Page 2
]
~
I For VEGP Units 1 and 2, the limiting size break continues to be the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of Co = 0.6. The LBLOCA
~ LOCBART analysis-of-record calculated PCT value was reanalyzed as a result of a LOCBART clad creep and burst error.
j The error was discovered in the LOCBART code related to improper modeling of fuel rod cladding creep and burst. The high temperature creep model did not properly account for pellet / clad contact which can often occur during creepdown in the first several seconds of a.
large break LOCA when there is still a compressive stress exerted on the clad. The incorrect clad strain which was calculated during this time period then contributed some residual i
cumulative effect on the stress and strain calculations of clad ballooning and burst during the l
later phase of the transient. For the burst model, logic previously used in the code only j
examined the highest temperature axial node at any given time step to determine if the temperature and hoop stress met burst conditions. However, since for the ZIRLO clad model there is an additional requirement that a preburst strain of 10% must have occurred in order to -
i
. burst, it is possible on occasion that a node which is not the highest temperature one for a given time step simultaneously meets all of the burst criteria first. Due to the small magnitude of effects and the interaction between these two items, they are being evaluated as a single, closely related effect. A LOCBART reanalysis for VEGP with the corrected model resulted in a new analysis-of-record PCT value of 1911 F. This new PCT value of 1911 F replaces the older value of1915 F.
The containment purge, T uncertainty, and transition core penalty items continue to be listed m
separately per the format of WCAP-13451. The items are listed separately because these items are not explicitly modeled. The PCT assessment values on these items remain 10,11, and 50*F, respectively.
As reported in Reference 1, VEGP had included a 41*F PCT penalty assessment for fuel burnups less than 150 MWD /MTU for core designs using the 1.5X integral fuel burnable absorber (IFBA) fuel rod design. The fuel in both VEGP Units 1 and 2 has accumulated a greater than 150 MWD /MTU burnup. Therefore, this previous penalty assessment of 41 F is no longer applicable.
VEGP has begun using ZIRLO clad fuel rods in the Vogtle-2 Cycle 6 core design. ZIRLO clad fuelis also planned for Vogtle 1 Cycle 8 scheduled for startup in October 1997. The use of ZIRLO clad fuel rods results in a penalty of 5 F PCT as calculated by the latest acceptable LOCBART model.
VEGP is using the 1.5X IFBA, ZlRLO clad fuel rod with a backfill pressure of 100 psig in the
' Vogtle-2 Cycle 6 core design. A 1.25X IFBA, ZIRLO clad fuel rod is planned for use in the Vogtle-1 Cycle 8 core design scheduled for startup in Ocmier 1997. The 1.5X IFBA, ZIRLO clad rod results in a penalty of 21*F PCT as calculated by the htest acceptable LOCBART model.
I
1
, ATTACHMENT Page 3 Because there is a mix of Zircaloy and ZIRLO clad fuel rods and IFBA and non-IFBA rods in VEGP core designs, VEGP will continue to show an analysis-of-record LOCBART calculated PCT value based on non-IFBA, Zircaloy fuel rods (191l'F) and will apply PCT penalties when the VEGP cores contain the ZIRLO clad fuel rod.
Prior BASH Large-Break ECCS Evaluation Model Assessments The steam generator flow area application, structural metal b:st modeling, and LUCIFER error correction assessments continue to be listed separately per the format of WCAP-13451.
These items are prior BASII large-break ECCS Evaluation Model assessments. The PCT assessment values on these items are 10, -25, a:4 -or, respectively (Reference 7).
1996 BASH Large-Break ECCS Evaluation Model Assessments Since the 1995 annual report, one new assessment to the BASH large-break ECCS Evaluation Model that would afTect the VEGP LBLOCA PCT analysis has been identified.
The assessment is a translation error of fluid conditions from SATAN to LOCTA. The error was discovered in the coding related to the translation of fluid conditions between the SATAN blowdown hydraulics code and the LOCTA code used for subchannel analysis of the fuel rods In performing axial interpolations to translate the SATAN fluid conditions onto the mesh nodalization used by the LOCTA code, the length of the lower core channel fluid connection to the lower plenium node was incorrectly calculated. A PCT penalty of 15 F has been assessed to the VEGP LBLOCA BASH ECCS Model Analysis-of-Record.
LBLOCA ECCS Evaluation Model Assessment Summary The abseNte sum of the previous and the new PCT assessment now slightly exceeds 500F.
However, the net efrect of the assessment is a small reduction in theVEGP LBLOCA PCT results (e g., a 6*F benefit to the 2200*F acceptance criteria). Therefore, Southern Nuclear proposes not to reanalyze the BASH portsn of the LBLOCA ECCS analysis at this time for j
the following reasons:
(1)
The current magnitude of the PCT for the LBLOCA event maintains considerable margin to the 2200*F acceptance criteria. In fact, the cumulative assessments result in a net benefit (6*F in PCT) to the previously reported analysis-of-record PCT.
(2) The overall conservatism in the LBLOCA Evaluation Model has been approved by the NRC.
4 (3) The revision of the BASH code is expected to result in a net PCT benefit upon future reanalysis.
, A'TTACHMENT Page 4 (4)
The PCT analysis-of-record value of 191l'F was based on recent model changes involving partial reanalysis of the LBLOCA event using the LOCBART code.
(5)
The assessments performed by Westinghoac to address new issues and phenomena since the previous analyses were appropriateh conservative.
Southern Nuclear will incorporate the new BASH model at the next licensing action requiring reanalysis of the LBLOCA and for which the basil code would be used within the Evaluation Model.
10 CFR 50.59 Evaluation Assessments There are two plant modifications pursuant to 10 CFR 50.59 which affect the LBLOCA analysis results. Combining the PCT effects from the two evaluations concerning the permanent radiation shield and trisodium phosphate baskets still results in only a IT PCT assessment.
Licensing Basis LBLOCA PCT Based on the above discussions concerning the VEGP-specific application of the Westinghouse B ASH large-break ECCS Evaluation Model, the licensing basis LBLOCA PCT is as follows:
A.
1996 Annual Report LBLOCA BASli ECCS Model Analysis-of-Record
- 1. LOCBART Reanalysis Result 1911.0*F
- 2. Evaluation for Containment Purging
+ 10.0*F
- 3. Evaluation for +/- 6T Uncertainty Band
+ 11.0*F
- 4. Evaluation for Transition Cycle Penalty
+ 50.0 F
- 5. Core Designs Using the 1.5X IFBA ZIRLO Clad Fuel Rod With a Backfill Pressure of 100psig. (Unit 2 only)
+ 21.07
- 6. ZlRLO Clad Fuel Rods (Unit 2 only)
+
5.0 F B.
Prior BASH Large-Break ECCS Model Assessments
- 1. Steam Generator Flow Area Application
+ 10.0*F
- 2. Structural Metal Heat Modeling
- 25.0 F
- 3. LUCIFER Error Corrections 6.0T C.
1996 BASH Large-Break ECCS Model Assessment Translation of Fluid Conditions from SATAN to LOCA
+ 15.0 F NOTE: Absolute sum of B+C assessment = 56T.
Algebraic sum of B+C assessment = -6*F.
, A'TTACUMENT Page 5 t
D.
10 CFR 50.59 Evaluations
- 1. Permanent Radiation ShieldffSP Baskets
+ 1.07 1977.0T Licensing Basis LBLOCA PCT (Unit 1)
=
2003.0T (Unit 2)
=
Conclusion An evaluation of the effect of the assessment to the Westinghouse BASH large-break ECCS Evaluation Model was performed on the LBLOCA analysis results. When the effects of the B ASH ECCS Evaluation Model assessment and safety evaluations were combined with the VEGP LBLOCA analysis results, it was determined that the sum of the absolute magnitude of the asse'ssment slightly exceeded 50T PCT. However, the cumulative effect of the assessment is a reduction of 67 PCT in the LBLOCA PCT results. As discussed earlier, Southern Nuclear does not propose to reanalyze the BASH portion of the LBLOCA s.alysis at this time. Southern Nuclear will incorporate the new model at the next licensing action requiring reanalysis of the LBLOCA and for which the 1981 BASH model would be used.
[ITTACHMENT j
. Page 6 -
l
[
SMALL-BREAK LOCA l
4
' ECCS Evaluation Model-t 4
Since the last annual report (Reference 1), no new assessments were identified against the small-break LOCA (SBLOCA) analysis PCT for VEGP Units 1 and 2. The current SBLOCA
' analysis results are based on the earlier Westinghouse NOTRUMP small-break ECCS L
- Evaluation Model (Reference 6) as approved by the NRC for VEGP-specific application (References 4 and 5) and the latest acceptable SBLOCTA model.' The limiting size break analysis continues to assume the following information important to the SBLOCA analyses:
{
i i
L 17x17 VANTAGE-5 Fuel. Assembly-o Core Power = 1.02
- 3565 MWT i
o. Vessel Average Temperature = 571.9*F
-l o.
Steam Generator P!ugging Level = 10%
l o.
Fo = 2.48 at 9.5 ft o
FAH = 1.70 l
For VEGP Units I and 2, the limiting size small-break continues to be a three-inch equivalent diameter break in the cold leg. The SBLOCA analysis-of-record SBLOCTA calculated PCT i
value was reanalyzed as a result of a SBLOCTA fuel rod initialization error.
The error was discovered in the SBLOCTA code related to the adjustments which are made as i
part of the fuel rod initialization process which is used to obtain agreement between the SBLOCTA model and the fuel data supplied from the fuel thermal-hydraulic design y
calculations at full power, steady-state conditions. Specifically, an adjustment to the power which is made to compensate for adjustments to the assumed pellet diameter was incorrect.
Additionally, updates were made to the fuel rod clad creep and strain model to correct logic errors that could occur in certain transient conditions. These model revisions also had a small i
affect on the fuel rod initialization process, and can produce small affects during the transient.
l Due to the small magnitude of the affects and the interaction between the two items, they are being evaluated as a single, closely related affect. An SBLOCTA reanalysis for VEGP resulted in a new analysis-of-record PCT value of 1778 F. This new value of 1778 F replaces the older value of1770 F.
As a result of this slight increase in the new analysis-of-record value, the Burst and Blockageffime in Life calculation was reperformed and resulted in an additional 15 F PCT -
penalty from th'at combined and reported in Reference 7. Therefore, this additional 15 F PCT assessment of the Burst and Blockage /fime in Life recalculation is listed separately.
- The steam generator lower level tap relocation and T.v, uncertainty items continue to be listed separately per the format of WCAP-13451. The items are listed separately because these items are not explicitly modeled. The PCT assessment values on these items are 15*F and 4*F, respectively.
i f, ATTACHMENT Page 7 t
VEGP has begun using ZIRLO clad fuel rods in the Vogtle-2 Cycle 6 core design. ZIRLO clad fuel rods are planned for use in the Vogtle-1 Cycle 8 core design scheduled for startup in October 1997. The use of 2IRLO clad fuel rods results in a penalty of 3 F PCT as calculated in the latest acceptable SBLOCTA model. This penalty applies to both IFBA and non-IFBA j
rods.
j i
Prior NOTRUMP Small-Break ECCS Evaluation Model Assessments As reported in the last annual report (Reference 7), there were seven prior model assessments that were combined as shown in Table 2. They are: (1) Safety Injection (SI) Flow into the
. Broken RCS Loop, (2) Improved Steam Condensation Model, (3) Drift Flux Flow Regime i
Error, (4) LUCIFER Error Corrections, (5) Burst and Blockage / rime in Life item, (6) Boiling j
Heat Transfer Correlation Error, and (7) Steam Line Isolation Logic error. Because of the i
recalculated Burst and Blockagefrime in Life item, it was decided to remove the 15*F PCT penalty from this accumulated result and combine it with that listed in the analysis-of-record Section A.
1 i
The NOTRUMP specific enthalpy error continues to be listed separately in accordance with i
WCAP-13451 since it was not combined with the prior model assessments (see Reference 1).
1996 NOTRUMP Small-Break ECCS Evaluation Model Assessments No new assessments have been identified to the NOTRUMP SBLOCA ECCS Evaluation Model that would affect the VEGP analysis results.
SBLOCA ECCS Model Assessment Summary The absolute sum of the new SBLOCA PCT assessments is less than 50*F for the VEGP NOTRUMP SBLOCA ECCS model since that last reported in LCV-0579 (Reference 1).
10 CFR 50.59 Evaluation Assessments i
As reponed in Reference 1, there is only one plant modification which affects the SBLOCA analysis results. The evaluation concerning the loose pan in the VEGP Unit 1 RCS remains in effect.
Licensine Basis SBLOCA PCT Based on the above discussions concerning the VEGP-specific application of the Westinghouse NOTRUMP small-break ECCS Evaluation Model, the licensing basis SBLOCA PCT is as follows:
j
ATTACHMENT i
Page 8 j
3 A. 1996 Annual Report SBLOCA NOTRUMP ECCS Model Analysis-of-Record
- 1. SBLOCTA Reanalysis Result 1778.0 F l
- 2. Evaluation for Steam Generator Lower Level Tap Relocation
+ 15.0*F j
l
- 3. Evaluation for +/- 6*F Uncertainty Band
+
4.0*F j_
- 4. Burst and Blockage / Time in Life Recalculation as a
+ 30.0 F Result of SBLOCTA Reanalysis Results j
- 5. ZlRLO Clad Fuel Rods (Unit 2 only)
+
3.0 F B. Prior NOTRUMP Small-Break ECCS Model Assessments l
l
- 1. SIin Broken Loop / Improved Condensation Model 17.0*F l
(pending NRC approval of addendum to WCAP-10054-P-A),
i i
Drift Flux Flow Regime, LUCIFER Error Corrections, i
i Boiling Heat Transfer Correlation Error, Steam Line Isolation Logic Error, and Burst and Blockage / Time in Life j
as reported to the NRC in Reference 7.
- 2. NOTRUMP Specific Enthalpy Error
+ 20.0*F C. 1996 NOTRUMP Small-Break ECCS Model Assessments a
No new assessments were identified in 1996.
D.10 CFR 50.59 Evaluations
- 1. Loose Part (VEGP Unit 1 only)
+
2.0 F Licensing Basis SBLOCA PCT (Unit 1) =
1832 F (Unit 2) =
1833 F Conclusion An evaluation of the effect of assessments to the Westinghouse NOTRUMP small-break ECCS Evaluation Model was performed on the SBLOCA analysis results. When the effects of the NOTRUMP ECCS Evaluation Model assessments were combined with the 10 CFR 50.59 i
evaluation and the VEGP SBLOCA analysis results, it was determined that compliance with the requirements of 10 CFR 50.46 would be maintained for both Units 1 and 2.
I 1
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,' 'ITACHMENT Page 9 REFERENCES 1.
LCV-0780 "Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Evaluation Models 1995 Annual Report," letter from C. K. McCoy (GPC) to USNRC, dated March 25, 1996.
2.
WCAP-13451, " Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," dated October 1992.
l 3.
"The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, Rev. 2, (Proprietary), March 1987.
l 4.
ELV-02166, "Vogtle Electric Generating Plant, Request for Technical Specifications Changes VANTAGE-5 Fuel Design," letter from W. G. Hairston, III, to USNRC, dated l
November 29,1990.
5.
ELV-03375, "Vogtle Electric Generating Plant, Licensing Change Power Uprating,"
letter from C. K. McCoy (GPC) to the NRC, dated February 28,1992.
d 6.
"destinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"
WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985.
7.
LCV-0579, "Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Evaluation Models 1994 Annual Report," letter from C. K. McCoy (GPC) to USNRC, dated March 17, 1995.
j i
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