ML19309D196
| ML19309D196 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/30/2019 |
| From: | Mark D. Sartain Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 19-225 | |
| Download: ML19309D196 (51) | |
Text
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 October 30, 2019 10 CFR 50.90 United States Nuclear Regulatory Commission Serial No.: 19-225 Attention: Document Control Desk NRA/GDM: RO Washington, D.C. 20555-0001 Docket Nos.: 50-280/281 License Nos.: DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST ADDITION OF ANALYTICAL METHODOLOGY TO THE CORE OPERATING LIMITS REPORT FOR A LARGE BREAK LOSS OF COOLANT ACCIDENT (LBLOCA)
Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Surry Power Station (Surry) Units 1 and 2. The proposed amendment requests the addition of the Westinghouse Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)" to the list of methodologies approved for reference in the Core Operating Limits Report (COLR) in TS 6.2.C. The proposed amendment also requests the deletion of TS Figure 3.12-2 "HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE".
Supporting information provided in the attachments to this letter is summarized below:
- Attachment 1 provides an evaluation of the proposed TS change.
- Attachment 2 provides marked-up TS pages.
- Attachment 3 provides proposed TS pages.
- Attachment 4 provides a technical evaluation of the proposed LAR Dominion Energy Virginia has evaluated the proposed amendment and determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in Attachment 1. Dominion Energy Virginia has also determined that operation with the* proposed change will not result in any significant increase in the amount of effluents that may be released offsite or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.
Serial No.19-225 Docket Nos. 50-280/281 Page 2 of 3 The proposed TS change has been reviewed and approved by the Surry Facility Safety Review Committee.
Dominion Energy Virginia requests approval of the proposed amendment by October 31, 2020, with a 90-day implementation period.
If there are any questions or if additional information is needed, please contact Mr. Gary Miller at (804)-273-2771.
Sincerely, Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this 3o'"""' day of Ge. +obec: , 2019.
My Commission Expires: ~ u.X: ~\ 1 '2..02 ~
Commitments made in this letter: None Attachments:
- 1. Discussion of Technical Specifications Change
- 2. Marked-up Technical Specifications Page
- 3. Proposed Technical Specifications Page
- 4. Westinghouse Technical Evaluation of the Proposed License Amendment Request.
Serial No.19-225 Docket Nos. 50-280/281 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector Surry Power Station Mr. Vaughn Thomas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 04 F-12 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738 State Health Commissioner Virginia Department of Health James Madison Building - 7th Floor 109 Governor Street Room 730 Richmond, Virginia 23219
Serial No.19-225 Docket Nos. 50-280/281 LAR - FSLOCA Methodology COLR Reference Attachment 1 DISCUSSION OF TECHNICAL SPECIFICATIONS CHANGE Virginia Electric and Power Company (Dominion Energy Virginia)
Surry Power Station Units 1 an_d 2
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1
- DISCUSSION OF TECHNICAL SPECIFICATIONS CHANGE SURRY POWER STRATION UNITS 1 AND 2 1.0
SUMMARY
DESCRIPTION Virginia Electric and Power Company (Dominion Energy Virginia) proposes changes to Surry Power Station (SPS) Units 1 and 2 Technical Specification (TS) 6.2.C to add Westinghouse Topical Report WCAP-16996-P-A, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," [Reference 1] to the list of methodologies approved for reference in the Core Operating Limits Report (COLR). The added reference identifies the analytical methods used to determine core operating limits for the Large Break Loss of Coolant Accident (LBLOCA) event described in the SPS Updated Final Safety Analysis Report (UFSAR), Section 14.5.1. In addition, .a legacy reference in TS 6.2.C that is no longer required to support SPS reload cores will be removed. Lastly, an administrative change to delete a TS figure that
. was previously relocated to the COLR is also proposed ..
A LBLOCA analysis has been completed for SPS Units 1 and 2 with the FULL SPECTRUM LOCA (FSLOCA) Methodology," [Reference 1]. The analysis was performed in compliance with the conditions and limitations included in the US Nuclear Regulatory Commission (NRG) Safety Evaluation (SE) in WCAP-16996-P-A, Revision 1. The analysis results confirm that SPS Units 1 and 2 continue to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.
The Westinghouse FSLOCA Evaluation Model (EM) has been generically approved by the NRG for Westinghouse 3-loop and 4-loop plants with cold leg Emergency Core Cooling System (ECCS) injection [Reference 1]. Since SPS Units 1 and 2 are Westinghouse designed 3-loop plants with cold leg ECCS injection, the approved method is applicable.
The proposed TS change has been reviewed, and it has been determined that no significant hazards consideration exists as defined in 10 CFR 50.92. In addition, it has been determined that the change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed TS change.
Page 1 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1 2.0 DETAILED DESCRIPTION 2.1. System Design and Operation The design requirements of . existing plant systems, structures, and components (SSCs) are used as inputs to the LBLOCA analysis, with appropriate technical conservatisms applied. Therefore, the LBLOCA analysis does not directly impact the existing design or configuration of any plant SSCs.
2.2. Current Technical Specification Requirement TS 6.2.C directs that core operating limits shall be established prior to each reload cycle and contains references to the approved analytical methods that are used to determine the core operating limits. The current method listed for LBLOCA is the Westinghouse Automated Statistical Treatment of Uncertainty Method (ASTRUM) EM identified as Reference 2 of TS 6.2.C.
2.3. Reason for the Proposed Change By letter dated April 28, 2017, Dominion Energy Virginia committed to submit a LBLOCA analysis for NRC review and approval that applies NRG-approved methods that include the effects of fuel pellet thermal conductivity degradation (TCD) [Reference 2]. The Westinghouse FSLOCA EM includes the effects of TCD, and this proposed change is being submitted to fulfill this commitment.
The commitment date provided in Reference 2 was later changed to October 31, 2019 by letter dated July 30, 2019 [Reference 6].
Reference 7 in TS 6.2.C (VEP-NE-3) is only applicable to a fuel type no longer used at SPS. Therefore, it is proposed to remove the legacy reference.
An administrative TS change is also being proposed to delete TS Figure 3.12-2, "Hot Channel Factor Normalized Operating Envelope." This figure was previously relocated to the COLR by SPS Units 1 and 2 TS Amendments 189/189 [Reference 3] but was not deleted from the TS at that time due to an administrative oversight.
2.4. Description of Proposed Changes The following mark-ups are included in Attachment 2 and the proposed TS pages are included in Attachment 3.
- The TS 6.2.C Core Operating Limits Report References list *currently includes ten documents that define the methods used to determine the Page 2 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1 core operating limits for SPS. The proposed revision to this list of references would replace the current Reference 7:
- 7. VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code" with WCAP-16996-P-A. The proposed change is as follows:
- 7. WCAP-16996-P-A, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," (Westinghouse Proprietary)
While it is our intent to use WCAP-16996-P-A to support future core reloads, existing TS 6.2.C Reference No. 2, WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," is being retained in the TS 6.2.C reference list. This will permit an orderly transition to FSLOCA in subsequent reload cycles for SPS Units 1 and 2.
- TS Figure 3.12-2, which displays the Hot Channel Factor K(z) curve, is being deleted. This curve should have been removed per TS Amendment Nos. 189/189 when the figure was relocated to the COLR.
As removal of this curve is fully bounded by NRG approval of Amendments 189/189, this change is considered administrative.
There are no references to TS Figure 3.12-2 within the SPS TS, and those TS that refer to the Hot Channel Factor K(z) curve cite the value as being specified in the COLR.
3.0 TECHNICAL EVALUATION
Attachment 4 provides the Westinghouse technical evaluation for the application of '
the FSLOCA EM to SPS. This analysis was performed in accordance with the NRG-approved FSLOCA EM in Westinghouse Topical Report WCAP-16996-P-A.
The application of the Topical Report to SPS only .involves the analysis for Region II (LBLOCA), as discussed in Section 1.0 of Attachment 4. This application is a replacement for the existing ASTRUM LBLOCA analysis, which addresses breaks from 1.0 ft2 to two times the area of the RCS cold leg piping.
The small break LOCA (SBLOCA) portion of the break spectrum is currently addressed for SPS Units 1 and 2 with the NOTRUMP EM (References 3 and 4 in the TS 6.2.C list). However, a separate license amendment request has been submitted for NRG review and approval to replace the NOTRUMP EM with Page 3 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1 Framatome Topical Report EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," as supplemented by the Surry-specific application report ANP-3676P [Reference 4].
The removal of the legacy reference and TS figure will have no technical impact on the ability to meet COLR limits.
4.0 REGULATORY EVALUATION
4.1. Applicable Regulatory Requirements and Criteria The FSLOCA EM in WCAP-16996-P-A satisfies the requirements of 10 CFR 50.46(b) paragraphs (1) through (4). The proposed change meets the current regulatory requirements and does not affect conformance with any General Design Criterion as described in the SPS Units 1 and 2 UFSAR.
4.2. Precedents The proposed change to TS 6.2.C adds Westinghouse Topical Report WCAP-16996-P-A to the list of approved methodologies for determining core operating limits at SPS. Numerous previous requests have been approved for addition of methodology reference changes to plant-specific TS COLR reference lists. The NRC has not previously approved the addition of WCAP-16996-P-A to the list of approved methodologies for determining core operating limits* at other plants. However, an LAR requesting a similar change has been submitted for Diablo Canyon Units 1 and 2 [Reference 5].
4.3. No Significant Hazards Consideration Virginja Electric and Power Company (Dominion Energy Virginia) proposes a change to Surry Power Station (SPS) Units 1 and 2 Technical Specification (TS) 6.2.C to add Westinghouse Topical Report WCAP-16996-P-A, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," to the list of methodologies approved for reference in the Core Operating Limits Report (COLR). The added reference identifies the analytical methods used to determine core operating limits for the Large Break Loss of Coolant Accident (LBLOCA) event described in the SPS Updated Final Safety Analysis Report (UFSAR),
Section 14.5.1. Administrative changes to delete a TS figure that was previously relocated to the COLR and to remove a legacy reference no longer utilized at SPS are also being requested.
Page 4 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1 Deletion of the TS Figure and removal of the legacy reference are considered administrative changes. Therefore, discussion of these items is categorically excluded from the Significant Hazards Considerations below.
Dominion Energy Virginia has evaluated whether a significant hazards consideration is involved with the proposed amendment, and a significant hazards evaluation was performed by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to TS 6.2.C permits the use of an NRG-approved methodology for analysis of the LBLOCA to determine if SPS Units 1 and 2 continue to meet the applicable design and safety analysis acceptance criteria. The proposed change to the list of NRG-approved methodologies in TS 6.2.C has no direct impact upon plant operation or configuration and does not impact either the probability of initiation of an accident or the mitigation of its consequences.
The results of the LBLOCA analysis demonstrate SPS continues to satisfy the 10 CFR 50.46(b)(1-4) ECCS performance acceptance criteria using an NRG-approved evaluation model.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not create the possibility of a new or different accident due to credible new failure mechanisms, malfunctions, or accident initiators not previously considered. There is no change to the parameters within which the plant is normally operated and no physical plant modifications are being made; thus, the possibility of a new or different type of accident is not created.
Page 5 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1 Therefore, the proposed change does not create the possibility of a new or different kind of accident or malfunction from those previously evaluated within the UFSAR.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
No design basis or safety limits are exceeded or altered by this change.
Approved methodologies will be used to ensure the plant continues to meet applicable design criteria and safety analysis acceptance criteria.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Implementation of the proposed license amendment is safe and will have no effect on plant operation. The proposed change does not result in any physical modifications to plant equipment or how the equipment is operated or maintained.
Based on the above information, Dominion Energy Virginia concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified~
4.4. Conclusion Based on the considerations presented above, there is reasonable assurance that: (1) the health and safety of the public will not be endangered by the demonstration that SPS continues to nieet applicable design criteria and safety analysis acceptance criteria, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 Environmental Considerations Dominion Energy Virginia has reviewed the proposed license amendment for environmental considerations in accordance with 10 CFR 51.22. The proposed license amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative Page 6 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 1 occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion- for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0 References
- 1. Westinghouse Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break. Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.
- 2. Letter from M. D. Sartain (Dominion) to USNRC (Serial No.17-158), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Emergency Core Cooling System (ECCS) Model Change Pursuant to ~he Requirements of 10 CFR 50.46, Submittal Schedule Commitment Change," April 28, 2017 (ADAMS Accession Number ML17125A256).
(Serial No.94-165), "Surry Units 1 and 2 - Issuance of Amendments Re: Core Operc1ting Limits Report (TAC Nos. M87004 AND M87005)," March 11, 1994 (ADAMS Accession Number ML012740164).
- 4. Letter from M. D. Sartain (Virginia Electric and Power Company) to USNRC (Serial No.18-249), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed License Amendment Request, Revision of Analytical Methodology Reference in Core Operating Limits Report for Small Break Loss of Coolant Accident," July 31, 2018 (ADAMS Accession Number ML18218A170).
- 5. Letter from Paula Gerten (PG&E) to USN RC (DCL No.18-100), "License Amendment Request 18-02, License Amendment Request to Revise Technical Specification 5.6.5b, 'Core Operating Limits Report (COLR)' for Full Spectrum Loss-of-Coolant Accident Methodology," Dec~mber 26, 2018 (ADAMS Accession Number ML19003A196).
- 6. Letter from M. D. Sartain (Virginia Electric and Power Company) to USNRC (Serial No.19-305), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Emergency* Core Cooling System (ECCS) Model Change, Pursuant to the Requirements of 10 CFR 50.46, Submittal Schedule Commitment Change," July 30, 2019 (ADAMS Accession Number ML19214A047).
Page 7 of 7
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGES Virginia Electric and Power Company (Dominion Energy Virginia)
Surry Power Station Units 1 and 2
TS FIG RE 3.12-2 Q2 Q1 9 1
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Amendment Nos. 18~ anEI I 8e
TS 6.2-2 08 12 14 The :Lflalytica'I methods used to determine t:hc core opcraling limits identified above shall be Lhos previously rn *icwcd and appro ed by th NRC. and identified below.
Th CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS rdcrcnced topical reports u cd to prepare Lhe CORE OPERATI~G LIMITS REPORT i.e.. report number. lillc. revision. date. and :Lfl}' upplcmcntst The co re operat ing limits shall be de termined so that ::ipplica bl li mits (e.g .. fuel thermal-me hanieal limiLS. core thermal-h, draulic limits. ECCS limit . nuclc-ar limits such as hutdown margin. and transient and a cidcnt anal sis limits ) of the afcty analysis arc m t. The CORE O:PERATI ' G LD,1IT REPORT. including any mid-c 'Cle revisions or upp'I rnents thereto. hall be prm*.ided for information for each reload c, clc to the J\RC Docume nt Con trol De k 1,vith eop ie Lo th R gional Administrator and Reside nt In poctor.
REFEREJ\CES I. VEP-FRD-41-A. bReload ~u lear Design I'tfrlhodolo~n*"
WCAP-16009-P-A. **Realistic ur~ Bn.'31.: LOCA faaluatioo Methodolog, Using the Automated Stati tical Treatmen of oceruiinty ~1 thod (ASTRUtvf)."
(\\'cstin~ouse Proprietary).
- 3. WCAP-10054-P-A. Westing use Smal l Break ECCS Evalu!ltion Model U ing t:h ~
NOTRUMP Cod :* (V. ProprictaJ)')
- 4. WCAP-1 79-P-A. .. NOTRUMP. A odal Transient Small Breai.: and General Network Code." rW Proprictar)')
- 5. WCA!P- I_610..P-A. **VANTAGE+ Fuel As.sembly Report." (W stinghousc Propric tary ,
- 6. VEP-NE---A, ..Statisticil DNBR faaluation Mel.hod loJ' 7'?1 VEP NE l A. Qti!lilteftl.ie11 ef lttt1 'NRB I Cl IF Cer,ela1ie11 i11 lhe Vir~i11is :Pewer COBRA Cg,11£..
i DOM-~AF-2-A...Reactor Core Thermal-Hydraulic Using the VIPRE-D Computer Code:* in luding Appcn<li B. "Qualification of Lhc \Vcslinghouse WRB-1CHF Correlation in the Dominion VIPRE-D Computer Code." and Appendix D.
QualifiC3lion of Lhe ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-DComput rCodc"
- 9. WCAP- 745-P-A. "De ign Bases for Th nnal O,*erpower Dclta-Tand Thermal 0Yen mperatiure Oe.lta-T Trip Function"
- 10. WCAP- 1_610..P-A andCENPD-404-P-A. Addend um I-A. "Oplimizcd ZIRLO."
CWestinghou se Proprietary)
WCAP-16996-P-A. "Realisric LOCA. Evaluation :to.le=") mcndm nt Nos. 283 888 283 Sp<<m1Ill of Breu Sizes (FULL SPECTR.l.J~,:! LOCA . fe!hodolo:y),~
(WesbnpoUSI! Proprietary)
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 3 PROPOSED TECHNICAL SPECIFICATIONS PAGES Virginia Electric and Power Company (Dominion Energy Virginia)
Surry Power Station Units 1 and 2
TS FIGURE 3.12-2 DELETED Amendment Nos.
TS 6.2-2 The analytical methods used to determine the core operating limits identified above shall be those previously reviewed and approved by the NRC, and identified below.
The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number, title, revision, date, and any supplements). The core operating limits shall be determined so that applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided for information for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
REFERENCES
.I. VEP-FRD-42-A, "Reload Nuclear Design Methodology"
.2. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"
(Westinghouse Proprietary).
- 3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," (W Proprietary)
- 4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," (W Proprietary)
- 5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Report," (Westinghouse Proprietary)
- 6. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology"
- 7. WCAP-16996-P-A, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)." (Westinghouse Proprietary)
- 8. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code"
- 9. WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function"
- 10. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO,"
(Westinghouse Proprietary)
Amendment Nos.
Serial No.19-225 Docket Nos. 50-280, 281 FSLOCA LAR Attachment 4 WESTINGHOUSE TECHNICAL EVALUATION OF THE APPLICATION OF THE FSLOCA EVALUATION MODEL TO SPS UNITS 1 AND 2 Virginia Electric and Power Company (Dominion Energy Virginia)
Surry Power Station Units 1 and 2
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO SURRY POWER STATION UNITS 1 AND 2
1.0 INTRODUCTION
An analysis with the FULL SPECTRUM' loss-of-coolant accident (FSLOCA') evaluation model (EM) has been completed for Surry Power Station Units 1 and 2. This license amendment request (LAR) for Surry Units 1 and 2 requests approval to apply the Westinghouse FSLOCA EM for the large-break loss-of-coolant accident analysis.
- The FSLOCA EM (Reference 1) was developed to address the full spectrum of loss-of-coolant accidents (LOCAs) which result from a postulated break in the reactor coolant system (RCS) of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine (DEG) rupture of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as Small and Large Break LOCAs.
The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as Small Break LOCAs (SBLOCAs ). Region II includes break sizes that are typically defined as Large Break LOCAs (LBLOCAs ).
Only the Region II (LBLOCA) analysis was performed for this application of the FSLOCA EM. A Region I (SBLOCA) analysis was not performed.
The FSLOCA EM explicitly considers the effects offuel pellet thermal conductivity degradation (TCD) and other bumup-related effects by initializing fuel rod conditions to fuel rod performance data input generated by the PADS code (Reference 2), which explicitly models TCD and is benchmarked to high bumup data in Reference 2. The fuel pellet thermal conductivity model in the WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet thermal conductivity degradation.
Three of the Title 10 of the Code of Federal Regulations (CFR) 50.46 criteria (peak cladding temperature (PCT), maximum local oxidation (MLO), and core-wide oxidation (CWO)) are considered directly in the FSLOCA EM. A high probability statement is developed for the PCT, MLO, and CWO that is needed to demonstrate compliance with 10 CFR 50.46 acceptance criteria (b)(l), (b)(2), and (b)(3) (Reference 3) when employing realistic methods to account for uncertainty. The MLO is defined as the sum of pre-transient corrosion and transient oxidation consistent with the position in Information Notice 98-29 (Reference 4). The coolable geometry acceptance criterion, 10 CFR 50.46 (b)(4), is met by compliance with acceptance criteria (b)(l), (b)(2), and (b)(3), and demonstrating that fuel assembly grid deformation due to combined seismic and LOCA loads does not extend to the in-board fuel assemblies, thus ensuring that a coolable geometry is maintained.
The FSLOCA EM has been generically approved by the Nuclear Regulatory Commission (NRC) for Westinghouse 3-loop and 4-loop plants with cold leg Emergency Core Cooling System (ECCS) injection (Reference 1). Since Surry Units 1 and 2 are Westinghouse designed 3-loop plants with cold leg ECCS injection, the approved method is applicable.
This report summarizes the application of the Westinghouse FSLOCA EM to Surry Units 1 and 2. The application of the FSLOCA EM to Surry Units 1 and 2 is consistent with the NRC-approved methodology (Reference 1), with exceptions identified under Limitation and Condition Number 2 in Section 2.3. A consistency review for the application of the FSLOCA EM to Surry Units 1 and with the conditions and FULL SPECTRUM and FSLOCA are trademarks of Westinghouse Electric Company LLC, its subsidiaries and/or affiliates in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be ~.ademarks of their respective owners.
Page 1 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 limitations as identified in the NRC's Safety Evaluation Report (SER) for Reference 1 is documented in Section-2.3.
Both Dominion Energy and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface processes, along with Westinghouse internal processes for assessing EM changes and errors, are used to identify the need for LOCA analysis impact assessments.
The major plant parameter and analysis assumptions used in the Surry Units 1 and 2 analysis with the FSLOCA EM are provided in Tables 1 through 4.
2.0 METHOD OF ANALYSIS 2.1 FULL SPECTRUM LOCA Evaluation Model Development In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 (Reference 3 and Reference 6) and Appendix K, "ECCS Evaluation Models," to permit the use of a realistic EM to analyze the performance of the ECCS during a hypothetical LOCA. Westinghouse's previously approved best-estimate LBLOCA EM is discussed in Reference 7. That EM is referred to as the Automated Statistical Treatment of Uncertainty Method (ASTRUM), and was developed following Regulatory Guide (RG) 1.157 (Reference 8).
When the FSLOCA EM was being developed, the NRC issued RG 1.203 (Reference 9) which expands on the principles of RG 1.157, while providing a more systematic approach to the development and assessment process of a PWR accident and safety analysis EM. Therefore, the development of the FSLOCA EM followed the Evaluation Model Development and Assessment Process (EMDAP), which is documented in RG 1.203. While RG 1.203 expands upon RG 1.157, there are certain aspects ofRG 1.157 which are more detailed than RG 1.203; therefore, both RGs were used for the development of the FSLOCAEM.
2.2 WCOBRA/TRAC-TF2 Computer Code The FSLOCA EM (Reference 1) uses the WCOBRA/TRAC-TF2 code to analyze the system thermal-hydraulic response for the full spectrum of break sizes. WCOBRA/TRAC-TF2 was created by combining
. a lD module (TRAC-P) with a 3D module (based on Westinghouse,modified COBRA-TF). The lD and 3D modules include an explicit non-condensable gas transport equation. The use ofTRAC-P allows for the extension of a two-fluid, six-equation formulation of the two-phase flow to the lD loop components.
This new code is WCOBRA/TRAC-TF2, where "TF2" is an identifier that reflects the use of a three-field (TF) formulation of the 3D module derived by COBRA-TF and a two-fluid (TF) formulation of the ID module based on TRAC-P.
This best-estimate computer code contains the following features:
- 1. Ability to model transient three-dimensional flows in different geometries inside the reactor vessel
- 2. Ability to model thermal and mechanical non-equilibrium between phases
- 3. Ability to mechanistically represent interfacial heat, mass, and momentum transfer in different flow regimes
- 4. Ability to represent important reactor components such as fuel rods, steam generators (SGs),
reactor coolant pumps (RCPs), etc.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 A detailed assessment of the computer code WCOBRA/TRAC-TF2 was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena for a LOCA. Modeling of a LOCA introduces additional uncertainties which are identified and quantified in the plant-specific analysis. The reactor vessel and loop noding scheme used in the FSLOCA EM is consistent with the noding scheme used for the experiment simulations that form the validation basis for the physical models in the code. Such noding choices have been justified by assessing the model against large and full scale experiments.
2.3 Compliance with FSLOCA EM Limitations and Conditions The NRC's SER for Reference 1 contains 15 limitations and conditions on the NRC-approved FSLOCA EM. A summary of each limitation and condition and how it was met is provided below.
Limitation and Condition Number 1 Summary
. The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.
Compliance The analysis for Surry Units 1 and 2 with the FSLOCA EM is only being used to demonstrate compliance with 10 CFR 50.46 (b)(l) through (b)(4).
Limitation and Condition Number 2 Summary The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.
Compliance Surry Units 1 and 2 are Westinghouse-designed 3-loop PWRs with cold-side injection, so they are within the NRC-approved methodology. The analysis for Surry Units 1 and 2 utilizes the NRC-approved FSLOCA methodology, except for the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in Reference 5, and with the exception of only including an analysis for Region IL After completion of the analysis for Surry Units 1 and 2, three errors were discovered in the WCOBRA/TRAC-TF2 code. The first error was regarding the calculation of radiation heat transfer to liquid, which could be incorrectly calculated under certain conditions. The second error was regarding vapor temperature resetting, where under certain conditions the vapor temperature could incorrectly be reset to the saturation temperature for heat transfer calculations. These first two errors were found to have a negligible impact on analysis results with the FSLOCA EM as described in Reference 12.
The third error impacted the gamma energy redistribution multiplier and was identified after completion of the analysis for Surry Units 1 and 2. The treatment for the uncertainty in the gamma energy redistribution is discussed on pages 29-75 and 29-76 of Reference 1, and the equation for the assumed increase in hot rod and assembly relative power is presented on page 29-76. The power increase in the hot rod and hot assembly due to energy redistribution in the application of the FSLOCA EM to Surry Units 1 and 2 was calculated incorrectly. This error resulted in a 0% to 5% deficiency in the modeled hot rod and hot assembly rod linear heat rates on a run-specific basis, depending on the as-sampled value for the Page 3 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 multiplier uncertainty. The effect of the error correction was evaluated against the application of the FSLOCA EM to Surry Units 1 and 2.
The error correction has only a limited impact on the power modeled for a single assembly in the core. As such, the error correction has a negligible impact on the system thermal-hydraulic response during the postulated LOCA. For the Region II analysis, parametric PWR sensitivity studies, derived from a subset of uncertainty analysis simulations covering various design features and fuel arrays, were examined to determine the sensitivity of the analysis results to the error correction. The PCT impact from the error correction was found to be different for the transient phases (i.e., blowdown versus reflood) based on the PWR sensitivity studies. The correction of the error is estimated to increase the Region II analysis PCT by 31 °F, leading to an analysis result of 1875 °F for the Region II analysis.
All of the analysis results, including the error correction, continue to demonstrate compliance with the 10 CFR 50.46 acceptance criteria.
Limitation and Condition Number 3 Summary For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are modeled). This includes utilizing a plant-specific initial containment temperature, and only talcing credit for containment coatings which are qualified and outside of the break zone-of-influence.
Compliance The containment pressure calculation for the Surry Units 1 and 2 analysis was performed consistent with the NRC-approved methodology. Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A minimum initial temperature associated with normal full-power operating conditions was modeled, and only containment coatings which are qualified and outside of the brealc zone-of-influence were credited.
Limitation and Condition Number 4 Summary The decay heat uncertainty multiplier will be sampled consistent with the NRC-approved methodology for the FSLOCA EM. The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.
Compliance The decay heat uncertainty multiplier for the Surry Units 1 and 2 Region II analysis was sampled consistent with the restrictions of the NRC-approved methodology for the FSLOCA EM. The analysis simulations were all executed for no longer than 10,000 seconds following reactor trip. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region II analysis results have been provided in units of sigma and approximate absolute units in Table 7.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Limitation and Condition Number 5 Summary The maximum assembly and rod length-average bumup must remain below the limits contained in the NRC-approved methodology for the FSLOCA EM.
Compliance The maximum analyzed assembly and rod length-average burnups for the Surry Units 1 and 2 analysis are less than or equal to the limits contained in the NRC-approved methodology for the FSLOCA EM.
Limitation and Condition Number 6 Summary The fuel performance data for analyses with the FSLOCA EM should be based on the PADS code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PADS methodology.
Compliance PADS fuel performance data is utilized in the Surry Units 1 and 2 analysis with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of Reference 2, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 2.
Limitation and Condition Number 7 Summary The YDRAG uncertainty parameter for Region I analyses should be treated as specified in the NRC-approved methodology for the FSLOCA EM.
Compliance A Region I uncertainty analysis was not performed in this application of the FSLOCA EM, so this Limitation and Condition is n:ot applicable.
Limitation and Condition Number 8 Summary Certain uncertainty contributors will be treated for Region I analyses as specified in the NRC-approved methodology for the FSLOCA EM.
Compliance A Region I uncertainty analysis was not performed in this application of the FSLOCA EM, so this Limitation and Condition is not applicable.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Limitation and Condition Number 9 Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm modeling selections for Region I analyses.
Compliance A Region I uncertainty analysis was not performed in this application of the FSLOCA EM and Surry Units 1 and 2 are Westinghouse 3-loop PWRs, so this Limitation and Condition is not applicable Limitation and Condition Number 10 Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to demonstrate that the applied break size boundary for Region I analyses serves the intended goal.
Additionally, the minimum sampled break area for the analysis of Region II should be 1 :ft2.
Compliance Surry Units 1 and 2 are Westinghouse-designed 3-loop PWRs, so this part of the Limitation and Condition is not applicable.
The minimum sampled break area for the Surry Units 1 and 2 Region iI analysis is 1 :ft2.
Limitation and Condition Number 11 Summary There are various aspects of this Limitation and Condition, which are summarized below:
- 1. Certain information regarding the Region I and Region II analyses must be declared and documented prior to performing the uncertainty analysis, and will not be changed throughout the remainder of the analysis once they have been declared and documented.
- 2. If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for PCT, MLO, and CWO which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.
- 3. Plant operating ranges which are sampled within the uncertainty analysis will be provided in the analysis submittal for both regions.
Compliance This Limitation and Condition was met for the Surry Units 1 and 2 analysis as follows:
- 1. The information specified in the NRC-approved methodology for the FSLOCA EM was declared and documented prior to analysis execution, and was not changed after it was declared and documented.
- 2. The analysis inputs were not changed once they were declared and documented.
- 3. The plant operating ranges which were sampled within the uncertainty analysis are provided for Surry Units 1 and 2 in Table 1.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Limitation and Condition Number 12 Summary The plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves must be adequately accounted for in analysis with the FSLOCA EM.
Compliance A bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Surry Units 1 and 2 analysis.
Limitation and Condition Number 13 Summary In plant-specific models for analysis with the FSLOCA EM, specific modeling considerations for the upper head spray nozzles should be followed as required by the NRC-approved methodology.
Compliance These specific modeling requirements for the upper head spray nozzles were adhered to for the Surry Units 1 and 2 Region II analysis.
Limitation and Condition Number 14 Summary For analyses with tJ?.e FSLOCA EM to demonstrate compliance against the current 10 CPR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.
- Compliance For the Surry Units 1 and 2 analysis, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then summed with the pre-existing corrosion for comparison against the 10 CPR 50.46 local oxidation acceptance criterion of 17 percent.
Limitation and Condition Number 15 Summary The Region II analysis will be executed twice; once assuming loss-of-offsite power (LOOP) and once assuming offsite power available (OPA). The results from both analysis executions should be shown to be in compliance with the 10 CPR 50.46 acceptance criteria.
The statistical analysis must adhere to the Limitation and Condition as specified in the NRC-approved methodology for the FSLOCA EM.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Compliance The Region II uncertainty analysis for Surry Units I and 2 was performed twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the 10 CFR 50.46 acceptance criteria (see Section 4.0).
The statistical analysis adhered to the Limitation and Condition as specified in the NRC-approved methodology for the FSLOCA EM.
3.0 REGION Il ANALYSIS 3.1 Description of Representative Transient A large-break LOCA transient can be divided into phases in which specific phenomena are occurring. A convenient way to divide the transient is in terms of the various heatup and cool down phases that the fuel assemblies undergo. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below.
Blowdown - Critical Heat Flux (CHF) Phase In this phase, the break flow is subcooled, the discharge rate of coolant from the break is high, the core flow reverses, the fuel rods go through departure from nucleate boiling (DNB), the cladding rapidly heats up, and the reactor is shut down due to the core voiding.
The regions of the RCS with the highest initial temperatures (upper core, upper plenum, and hot legs) begin to flash during this period. This phase is terminated when the water in the lower plenum and downcomer begins to flash. The mixture level swells and a saturated mixture is pushed into the core by the intact loop RCPs, still rotating in single-phase liquid. As the fluid in the cold leg reaches saturation conditions, the discharge flow rate at the break decreases significantly.
Blowdown - Upward Core Flow Phase Heat transfer is increased as the two-phase mixture is pushed into the core. The break discharge rate is reduced because the fluid becomes saturated at the break. This phase ends as the lower plenum mass is depleted, the fluid in the loops become two-phase, and the RCP head degrades.
Blowdown - Downward Core Flow Phase The break flow begins to dominate and pulls flow down through the core as the RCP head degrades due to increased voiding, while liquid and entrained liquid flows also provide core cooling. Heat transfer in this period may be enhanced by liquid flow from the upper head. Once the system has depressurized to less than the accumulator cover pressure, the accumulators begin to inject cold water into the cold legs.
During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is bypassed around the downcomer and sent out through the break. As the system pressure continues to decrease, the break flow and consequently the downward core flow are reduced. The system pressure approaches the containment pressure at the end of this last period of the blowdown phase.
During this phase, the core begins to heat up as the system approaches containment pressure, and the phase ends when the reactor vessel begins to refill with ECCS water.
Refill Phase The core continues to heat up as the lower plenum refills with ECCS water. This phase is characterized by a rapid increase in fuel cladding temperature at all elevations due to the lack of liquid and steam flow in the core region. The water completely refills the lower plenum and the refill phase ends. As ECCS water Page 8 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 enters the core at the end of the refill phase, the fuel rods in the lower core region begin to quench and liquid entrainment begins, resulting in increased fuel rod heat transfer.
Reflood Phase During the early reflood phase, the accumulators begin to empty and nitrogen is discharged into the RCS.
The nitrogen surge forces water into the core, which is then evaporated, causing system re-pressurization and a temporary reduction of pumped ECCS flow; this re-pressurization is illustrated by the temporary increase in RCS pressure. During this time, core cooling may increase due to vapor generation and liquid entrainment, but conversely the early reflood pressure increase results in loss of mass out through the broken cold leg. '
The pumped ECCS water aids in the filling of the downcomer throughout the reflood period. As the quench front progresses further into the core, the PCT elevation moves increasingly higher in the fuel assembly.
As the transient progresses, continued injection of pumped ECCS water refloods the core, effectively removes the reactor vessel metal mass stored energy and core decay heat, and leads to an increase in the reactor vessel fluid mass. Eventually the core fluid inventory increases enough that liquid entrainment is able to quench all the fuel assemblies in the core.
3.2 Analysis Results The Surry Units 1 and 2 Region II analysis was performed in accordance with the NRC-approved methodology in Reference 1 with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The analysis was performed assuming both LOOP and OPA, and the results of both of the LOOP and OPA analyses are compared to the IO CFR 50.46 acceptance criteria. The most limiting ECCS single failure of one ECCS train is assumed in the analysis as identified in Table 1. The results of the Surry Units 1 and 2 Region II LOOP and OPA uncertainty analyses are summarized in Table 5, and include the impact of the gamma energy redistribution error correction. The sampled decay heat multipliers for the Region II analysis cases are provided in Table 7.
Table 6 contains a sequence of events for the transient (OPA offsite power assumption) that produced the analysis PCT result. Figures 1 through 15 illustrate the key system thermal hydraulic parameters for this transient.
The containment pressure is calculated for each LOCA transient in the analysis using the COCO code (References 10 and 11). The COCO containment code is integrated into the WCOBRA/TRAC-TF2 thermal-hydraulic code. The transient-specific mass and energy releases calculated by the thermal-hydraulic code at the end of each timestep are transferred to COCO. COCO then calculates the containment pressure based on the containment model (the inputs are summarized in Tables 2 and 3) and the mass and energy releases, and transfers the pressure back to the thermal-hydraulic code as a boundary condition at the break, consistent with the methodology in Reference 1. The containment model for COCO calculates a conservatively low containment pressure, including the effects of all the installed pressure reducing systems and processes such as assuming that all trains of containment spray are available. The containment backpressure for the transient that produced the analysis PCT result is provided in Figure 9.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 4.0 COMPLIANCE WITH 10 CFR 50.46 It must be demonstrated that there is a high level of probability that the following criteria in 10 CFR 50.46 are met:
(b)(l) The analysis PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT is less than 2,200°F, the analysis with the FSLOCA EM confirms that 10 CFR 50.46 acceptance criterion (b)(l), i.e., "Peak Cladding Temperature less than 2,200°F," is demonstrated.
The results are shown in Table 5 for Surry Units 1 and 2.
(b)(2) The analysis MLO corresponds t'o a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO is less than 17 percent when converting the time-at-temperature to an equivalent cladding reacted using the Baker-Just correlation and adding the pre-transient corrosion, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Maximum Local Oxidation of the cladding less than 17 percent," is demonstrated.
The results are shown in Table 5 for Surry Units 1 and 2.
(b)(3) The analysis CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the resulting CWO is less than 1 percent, the analysis confirms that 10 CFR 50.46 acceptance criterion (b )(3), i.e., "Core-Wide Oxidation less than 1 percent," is demonstrated.
The results are shown in Table 5 for Surry Units 1 and 2.
(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains in a coolable geometry.
This criterion is met by demonstrating compliance with criteria (b)(l), (b)(2), and (b)(3), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. Criteria (b)(1 ), (b )(2), and (b )(3) have been met for Surry Units 1 and 2 as shown in Table 5.
It is discussed in Section 32.1 of the NRC-approved FSLOCAEM (Reference 1) that the effects ofLOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends to inboard assemblies beyond the core periphery (i.e.,
deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to combined LOCA and seismic loads is not calculated to occur for Surry Units 1 and 2.
(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS.
Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM (Reference 1).
Based on the analysis results for Region II presented in Table 5 for Surry Units 1 and 2, it is concluded that Surry Units 1 and 2 comply with the criteria in 10 CFR 50.46.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4
5.0 REFERENCES
- 1. "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," WCAP-16996-P-A, Revision 1, November 2016.
- 2. "Westinghouse Performance Analysis and Design Model (PADS)," WCAP-17642-P-A, Revision 1, November 2017.
- 3. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974.
- 4. "Information Notice 98-29: Predicted Increase in Fuel Rod Cladding Oxidation," USNRC, August 1998.
- 5. "U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017," LTR-NRC-18-30, July 2018.
- 6. "Emergency Core Cooling Systems: Revisions to Acceptance Criteria," Federal Register, V53, N180, pp. 35996-36005, September 1988.
- 7. "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005.
- 8. "Best Estimate Calculations of Emergency Core Cooling System Performance," Regulatory Guide 1.157, USNRC, May 1989.
- 9. "Transient and Accident Analysis Methods," Regulatory Guide 1.203, USNRC, December 2005.
- 10. "Westinghouse Emergency Core Cooling System Evaluation Model - Summary," WCAP-8339, June 1974.
- 11. "Containment Pressure Analysis Code (COCO)," WCAP-8327, June 1974.
- 12. "U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018," LTR-NRC-19-6, February 2019.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 1. Plant Operating Range Analyzed and Key Parameters for Surry Units 1 and 2 Parameter As-Analyzed Value or Range 1.0 Core Parameters a) Core power :S 2597 MWt +/- 0% Uncertainty b) Fuel type 15x15 Upgrade Fuel, Optimized ZIRLO' High Performance Cladding Material with IFMs c) Maximum total-core peaking factor (Fq), 2.50 including uncertainties d) Maximum hot channel enthalpy rise factor 1.70 (F Af!), including uncertainties e) Axial flux difference (AFD) band at 100% +/-11%
power t) Maximum transient operation fraction 0.5 2.0 Reactor Coolant System Parameters a) Thermal design flow (TDF) 88,500 gpm/loop b) Vessel average temperature (TAva) 564.4°F :S Tavg :S 581.6°F c) Pressurizer pressure (PRcs) 2190 psia :S PRcs :S 2310 psia d) Reactor coolant pump (RCP) model and Model 93A, 6000 hp power 3.0 Containment Parameters a) Containment modeling Containment response calculated for each transient using transient-specific mass and energy releases and the containment data in Tables 2 and 3 4.0 Steam Generator (SG) and Secondary Side Parameters a) Steam generator tube plugging level :S7%
b) Main feedwater temperature Nominal (421 °F) c) Auxiliary feedwater temperature Nominal (120°F) d) Auxiliary feedwater flow rate 233 gpm/SG Optimized ZIRLO is a trademark of Westinghouse Electric Company LLC, its subsidiaries and/or affiliates in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
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Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 1. Plant Operating Range Analyzed and Key Parameters for Surry Units 1 and 2 Parameter As-Analyzed Value or Range 5.0 Safety Injection (SI) Parameters a) Single failure configuration ECCS: Loss of one train of pumped ECCS Region II containment pressure: All SI trains are available to minimize containment pressure b) Safety injection temperature (Ts1) 37.5°P s Ts1 s 62.5°P c) Low pressurizer pressure safety injection 1715 psia safety analysis limit d) Initiation delay time from low pressurizer s 25 seconds (OPA) or 40 seconds pressure SI setpoint to full SI flow (LOOP) e) Safety injection flow Minimum flows in Table 4 6.0 Accumulator Parameters a) Accumulator temperature (TAce) 89°P S TACC S 113°P 3 3 b) Accumulator water volume (VAce) 965 ft 5 V ACC S 1035 ft c) Accumulator pressure CPAcc) 580 psia s P Aces 700 psia d) Accumulator boron concentration ~2250 ppm 7.0 Reactor Protection System Parameters a) Low pressurizer pressure reactor trip 1865 psia setpoint Page 13 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 2. Containment Data Used. for Region II Calculation of Containment Pressure for Surry Units 1 and 2 Parameter Value Maximum containment net free volume 1,819,000 ft3 Minimum initial containment temperature at full power operation 102°F RWST temperature for containment spray 37.5°F :'.S RWST Temp :'.S 62.5°F Minimum RWST temperature for broken loop spilling SI 37.5°F Minimum containment outside air/ ground temperature 9op Minimum initial containment pressure at normal full power operation 9.85 psia Minimum containment spray pump initiation delay from containment 2'.: 59 seconds (LOOP and OPA) high pressure signal time Maximum containment spray flow rate from all pumps 4625 gpm Maximum number of containment fan coolers in operation during 0 LOCA transient Maximum number of containment venting lines (including purge lines, 0 pressure relief lines or any others) which can be OPEN at onset of transient at full power operation Containment walls / heat sink properties Table 3 SI spilling flows 161.6 lbm/sec Page 14 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 3. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Surry Units 1 and 2 Wall Area (ft2) Thickness (ft) Material 1 8500 0.00025, 0.5 Paint, Concrete 2 62500 0.00025, 1.0 Paint, Concrete 3 47500 0.00025, 1.5 Paint, Concrete 4 8000 1.5 Concrete 5 12000 2.0 Concrete 6 9500 2.25 Concrete 7 4000 3.0 Concrete 8 44000 0.00025, 0.03125, 4.5 Paint, Carbon Steel, Concrete 9 2500 0.03125, 4.5 Carbon Steel, Concrete 10 26000 0.00025, 0.04167, 2.5 Paint, Carbon Steel, Concrete 11 12500 0.00025, 2.2 Paint, Concrete 12 12400 0.0005, 0.01967 Paint, Carbon Steel 13 3000 0.012 Carbon Steel 14 73260 0.0005, 0.03608 Paint, Carbon Steel 15 13740 0.03608 Carbon Steel 16 14615 0.0005, 0.07458 Paint, Carbon Steel 17 3885 0.07458 Carbon Steel 18 4000 0.0005, 0.14167 Paint, Carbon Steel 19 12500 0.0005, 0.24167 Paint, Carbon Steel 20 105000 0.005 Carbon Steel
\
21 48000 0.0097 Stainless Steel 22 17500 0.03567 Stainless Steel 23 2500 0.12783 Stainless Steel Page 15 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 4. Safety Injection Flows (Total in Intact Loops) Used for Region II Calculation for Surry Units 1 and 2 High Head Safety Injection Low Head Safety Injection Pressure (psia)
(HHSI) Flow (gpm) (LHSI) Flow (gpm) 14.7 253.2 2015.8 54.7 249.5 2015.8 64.7 248.6 2015.8 69.7 248.2 1843.8 89.7 246.4 1479.2 114.7 244.1 970.6 139.7 241.6 391.2 149.7 240.6 259.7 154.7 240.1 108.4 154.8 240.1 0.0 214.7 234.1 514.7 203.2 1014.7 144.0 1264.7 111.1 1414.7 89.5 1731.7 31.2 2014.7 0.0 Page 16 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 5. Surry Units 1 and 2 Analysis Results with the FSLOCA EM Outcome Region II Value (OPA) Region II Value (LOOP) 95/95 PCT* 1844°F + 31°F = 1875°P 1817°F + 31°F = 1848°F 95/95 l\.1LO 6.2% 6.5%
95/95 cwo 0.37% 0.43%
- The PCT values presented in the table shows the analysis-of-rec ord result, which is the sum of the uncertainty analysis result plus the impact of the energy redistrib ution error correction. The figures presenting the analysis results correspond to the uncertainty analysis results. The MLO and CWO were confirmed to demonstrate compliance with the 10 CPR 50.46 acceptance criteria with the error correction.
Table 6. Surry Units 1 and 2 Sequence of Events for Region II Analysis PCT Transient (OPA)
Event Time after Break (sec)
Start of Transient 0.0 Fuel Rod Burst Occurs -3.0 Safety Injection Signal 4.8 Accumulator Injection Begins 12.0 PCT Occurs 12.5 End ofBlowdown -20.0 Safety Injection Begins 29.8 Bottom of Core Recovery -30.0 Accumulator Empty -45.0 All Rods Quenched -350 End of Analysis Time 600 Page 17 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Table 7. Surry Units 1 a1,1d 2 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region II Analysis Cases Region Case DECAY_HT (units ofo') DECAY_HT (absolute units)*
PCT +0.893cr 4.37%
Region II MLO +0.150cr 0.72%
(LOOP) cwo +0.958cr 4.59%
PCT +0.224cr 1.08%
Region II MLO +0.052cr 0.26%
(OPA) cwo + l.039cr 4.90%
- Approximate uncertainty in total decay heat power at 1 second after shutdown as defined by the ANSI/ANS-5. l-1979 decay heat standard for 235 U, 23 9I>u, and 238 U assuming infinite operation.
Page 18 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 PCT 6 0 0 Dummy Rod 1 PC T 7 0 0 Dumm~ Rod 2 PCT 1 0 0 Ho t od PCT 2 0 0 Ho t Assemb l y PCT 3 0 0 Av e rage Rod 1 PCT 4 0 0 Average Rod 2 PCT 5 0 0 Low Power Rod 2000
1500 LL Q)
-+J CJ Q) a.
E
~ 1000 en C:
""O
""O CJ u .-, 1~
..::.t:.
I- \
~ .
CJ Q) : :
Q._
500 l ~'if"'(*;.-:-*-
- rJ_ .
,r : :1 :
~ rr-~--n( ~ ..........
o-+-......_...............-r-......_...............-r-......_...............-r-......_............__......_............__......_...............__......_............--l 0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 1: Surry Units 1 and 2 Peak Cladding Temperature for all Rods for the Region II Analysis PCT Case Assuming OPA Note: This figure presents the uncertainty analysis results without the PCT penalty for the gamma energy redistribution error correction.
Page 19 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 PC T-L DC 6 0 0 PEAK CLAD TEMP LDC.
12 10 8
C:
0 6 C
Q.>
w 4
2 0
0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 2: Surry Units 1 and 2 Peak Cladding Temperature Elevation Relative to the Bottom of the Active Fuel for the Region II Analysis PCT Case Assuming OPA Page 20 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 RMVM 16 6 0 Break Flow 50000 40000
- en E
..0
.;::::.. 30000 Q.)
C, 0:::
31::
0 ti: 20000 en en C
~
10000 .
0
-10000~...__._......__._............................._._.........................................--r--..........._.__'---r_.__......._.__...--,.......__.........__..
0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 3: Surry Units 1 and 2 Break Mass Flow Rate for the Region II Analysis PCT Case Assuming OPA Page 21 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 LO-LEVEL 0 0 COLLAPSED LIQ. LEVEL 8
2 o..................................._._. . .__.,. . . ._. _. . .__............_ ..........._ ............__.............._.._. .,. . . . . . . . . . . .~
0 100 200 300 400 500 600 700 Time After Break {sec)
Figure 4: Surry Units 1 and 2 Lower Plenum Collapsed Liquid Level Relative to the Inner Bottom of the Vessel for the Region II Analysis PCT Case Assuming OPA Page 22 of 33
Serial No.19-225 Westinghouse on-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 FGM 11 1 0 Bottom Cel FGM 11 12 OTopCel l I
/\
\
\ I\
\
I I \
~
\ . .
..... \ ...... ... .............. ....... ....... ... ...... .. ... .
500 I \ : : : : :
I
- en
\ : .....
E
\.J-' \
..0 0 .... ** , ** * * * ** ******* * **** A * * * ~ . -~~~~~
\ /..;..,- _..,
\
, rv
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\ I :
-500 .
-1000 -+-....._....__..........---..__..___.._....._....._,.................__..___.._-,-....._..______..---r......._._....._....._,........___.......................
0 5 10 15 20 25 30 Time After Break (sec)
Figure 5: Surry Units 1 and 2 Vapor Mass Flow Rate at the Top and Bottom Cell Faces of the Core Average Channel not Under Guide Tubes for the Region II Analysis PCT Case Assuming OPA Page 23 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Liquid Mass Flow Rate at the Top and Bottom of the Non-GT Avg Channel for the Analysis PCT Case F L~ 11 1 0 Bottom Cel I
- - - - - FU,1 11 12 0 Top Cel l 10000 5000 ~ . .. .. . ....... . ........ . ... . . . ...... . . . . . ...... .. .. . ... ... .
I Q) I C
a::::
31::
0 LL.
Cl) 0 Cl)
C
- iE
-5000
-1 0000-+-....__...................-,-......_................"""T"........___..............,......_..._........................._..._..................... . . . _ ~
0 5 10 15 2Y 25 JO Time After Break (sec)
Figure 6: Surry Units 1 and 2 Liquid Mass Flow Rate at the Top and Bottom Cell Faces of the Core Average Channel not Under Guide Tubes for the Region II Analysis PCT Case Assuming OPA Page 24 of 33
Serial No.19-225 Westinghouse on-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 PN 39 0 PRESS URE 80
- 0
- Cl) c...
60 (1.)
en en
~
a... ~
20 0
0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 7: Surry Units 1 and 2 RCS Pressure for the Region II Analysis PCT Case Assuming OPA Page 25 of 33
Serial No.19-225 Westinghouse on-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 MTH00 1 70 74 9 0 Component 74 MTH00172 84 9 0 Component 84 3000 2500 *********
2000
Cl)
E
...c
..._.. 1500 *******
(l}
C 0::::
3:
0 Li: 1000 Cl) en C
500 o~ ~~-' * .. .. .. ... . . .. . ......... - ~
- ~
- ,-.......~~~~~~~~~~~
-500 0 20 40 69 80 100 Time After Break lSec)
Figure 8: Surry Units 1 and 2 Accumulator Injection Flow per Loop for the Region II Analysis PCT Case Assuming OPA Page 26 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 PN 4 0 Cont a inm e nt Pr e ssur e 40-,------------------------------,
35 30
-~en Cl..
E 25 en en Cl) a..
20 15 10 0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 9: Surry Units 1 and 2 Containment Pressure for the Region II Analysis PCT Case Assuming OPA Page 27 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 VFMASS 0 0 0 VESS EL WA TE R MASS 160000 1-40000 120000
--E 100000
..c en en 0 80000
~
60000 20000 0...----....---............. . . . ._..._-,--......_._..........,._..._....._. . . . .__......................__.__.._......
0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 10: Surry Units 1 and 2 Vessel Fluid Mass for the Region II Analysis PCT Case Assuming OPA Page 28 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 LO- LEVE L 5 0 0 Hot Assemb l y Channe l LO-L EVEL 3 0 0 Non-G T Avg Channel LO-L EVE L 4 0 0 GT Avg Ch anne l LO-L EVE L 2 0 0 Low Power Channel 10 Sa cu>
Q)
_J
- -E C""' 6
- J
-0 Cl) en c...
0 u
0 4 2
0 0 100 200 300 ~ 500 600 700 Time After Break (sec)
Figure 11: Surry Units 1 and 2 Collapsed Liquid Level for Each Core Channel Relative to the Bottom of the Active Fuel for the Region II Analysis PCT Case Assuming OPA Page 29 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 MT H00 1 5 1 6 0 0 COLL A P S ED L I Q. LE V EL lO 25
? 20 Q)
Q)
_J
.:-E 5-
- .:J 15 "O
Q) en a.
..52 0
u 10 5
01 ......._.......______.,..........................._._....._......._._......_~-....._~__._-.-.......__.._'"'"'"-1 0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 12: Surry Units 1 and 2 Average Downcomer Collapsed Liquid Level Relative to the Bottom of the Upper Tie Plate for the Region II Analysis PCT Case Assuming OPA Page 30 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 MTH00164 25 6 0 Loop 2 SI MTH00166 35 6 0 Lo op 3 SI 140-
- 120-
- en
'e'
..c 100- *
~
cu C
c::: BO -
~
0 LJ... ._
en Cl)
C so-40 -
20- .
- I . I I I I I I I I I I I I I I I I I I I I I I I 0 I I I 0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 13: Surry Units 1 and 2 Total Safety Injection Flow Rate per Loop (not including Accumulator Flow) for the Region II Analysis PCT Case Assuming OPA Page 31 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 Hot Rod Hot Assemb l y Rod Average Rad 2.0
~ 1.5 0
CL
""'C Cl.)
C
,.... ------~-. , ,',-,',...,,*.-,,
..,, ,'I * '
, ... ,':"* .... *\
E 0 .,
,,' . " . .\, """*........, '.
- z: 1.0 ,, \
I
,' ' ~. ' ,
I '
II '\
I \
I \
I \
I \
I ',
o.s ~I I
I
.' \ , .
\
I \
\
().()
0 2 4 6 8 10 12 Elevation (ft)
Figure 14: Surry Units 1 and 2 Normalized Rod Axial Power Shapes for the Region II Analysis PCT Case Assuming OPA Note: The localized power decreases occur at grid elevations.
Page 32 of 33
Serial No.19-225 Westinghouse Non-Proprietary Class 3 Docket Nos. 50-280/281 Attachment 4 POWERF 0 0 0 RE LATIVE CORE POWER 1.2 1- . .
o.a-
-C:
0 z;
0 e o.s-L..
Q)
~
0 a..
_...___.1 1___._
I _ 1 1 1 _..._.l l ---,-_.I___...
I _.1_........,
I ___...
I __.I___._
I _.1
. __,._1 1 _..._.
1 1 --,-_.I_...___.
1 1___._
I ...........
1_ 1 1 _ ..__.
1 0
0 100 200 300 400 500 600 700 Time After Break (sec)
Figure 15: Surry Units 1 and 2 Relative Core Power for the Region II Analysis PCT Case Assuming OPA Page 33 of 33