BSEP-96-0449, Responds to RAI Re License Amend of Power Uprate
| ML20133G285 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/23/1996 |
| From: | Campbell W CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20133G289 | List: |
| References | |
| BSEP-96-0449, BSEP-96-449, TAC-M90644, TAC-M90645, NUDOCS 9701150284 | |
| Download: ML20133G285 (20) | |
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g CP&L Corolino Power & Light Company William R. Campbell PO Box 10429 Vice President Southport NC 28461 0429 Brunswick Nuclear Plant December 23,1996 SERIAL BSEP 96-0449 U. S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 LICENSE AMENDMENT NOS.183 AND 214 - POWER UPRATE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (NRC TAC NOS. M90644/M90645)
Gentlemen:
On November 14,1996, Carolina Power & Light Company (CP&L) and the U.S. Nuclear Regulatory Commission (NRC) held a telephone conference to discuss the station blackout (SBO) analysis performed in support of the power uprate application for the Brunswick Steam Electric Plant, Units 1 and 2. The purpose of this letter is to document the information provided in that call and provide further information requested by the NRC staff during the telephone conference (Enclosure 1).
CP&L has completed an in-depth review of the power uprate processes and products. We have concluded that the products associated with this project are sound; however, we have identified areas for improvement. We have, therefore, reexamined these areas, as well as key evaluations and calculations, to provide assurance that operation at the uprated power level is acceptable. The scope of this review and disposition of the results, a discussion of the cause of the SBO impact analysis error, and the proposed resolution of the specific SBO and LOCA analysis issues are provided in this letter.
Based on the results of the recently completed reviews, and the quality of the initial product, CP&L has concluded that the power uprate products are sound and support operation at the
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uprated power level. However, as discussed in the November 14,1996 conference call, and indicated in our December 3,1996 letter, CP&L is maintaining Brunswick Unit 1 at or below the previously approved licensed power level (2436 MWt) pending resolution of this issue with the NRC staff.
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f& V QC 9701150284 961223 3
i PDR ADOCK 05000324 I
P PDR Tel 910 457 2496 Fox 910 457 2803
Document Control Desk BSEP 96-0449 / Page 2 Please refer any questions regarding this submittal to Mr. Mark Turkal at (910) 457-3066.
9 Sincerely, William R. Campbell KAH/kah Enclosures cc:
U. S. Nuclear Regulatory Commission ATTN.: Mr. Stewart D. Ebneter, Regional Administrator 101 Marietta Street, N.W., Suite 2900 Atlanta, GA 30323-0199 Mr. C. A. Patterson NRC Senior Resident inspector - Brunswick Units 1 and 2:
U.S. Nuclear Regulatory Commission ATTN.: Mr. David C. Trimble, Jr. (Mail Stop OWFN 14H22) 11555 Rockville Pike Rockville, MD 20852-2738 The Honorable H. Wells Chairman - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510
ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 l
LICENSE AMENDMENT NOS.183 AND 214 - POWER UPRATE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (NRC TAC NOS. M90644/M90645)
TABLE OF CONTENTS I
Executive Summarv 1.0 Discussion of Circumstances.....
... page E1-2 2.0 Root Cause........
............... page E1 -3 3.0 Whv Analysis Discrecancies Were Identified Late in the Process.......
page E1-4 3.1 Design (Basic Process) page E1-4 3.2 Implementation (Basic Process)
.. page E1-5 4.0 Review for Similar issues..
. page E1-6 4.1 Calculations...
.. page E1-6 4.2 UFSAR, Technical Specifications, and NRC Safety Evaluation page E1-6 4.3 Analysis Basis Document (ABD).
. page E1-6 4.4 Re-evaluation of ECCS, Containment, and Special Events........ page E1-6 1
4.5 Improved Technical Specifications........
..... page E1-7 4.6 Final Affirmation by Responsible Engineers.
..... page E1-7 5.0 Results from Review for Similar issues.
.. page E1-7 5.1 Licensing Topical Report Items....
......... page E1-7 5.2 Technical Specification items.
. page E1-9 5.3 Modification Document Update Form (DUF) Items........
. page E1-11 6.0 Confirmation of Sucoression Chamber Desian Basis
. page E1-11 7.0 Soecific Resolution of SBO Issue
. page E1-12 8.0 Soecific Resolution of LOCA issue.
page E1-13 9.0 Lona-Term Corrective Actions.
. page E1-14 10.0 Summarv.
page E1-14 11.0 References.......
page E1-14 Table 1 Figures 1 & 2 E1-1
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1 Executive Summary Following issuance of the License Amendment and NRC Safety Evaluation on power uprate for Brunswick Units 1 and 2, CP&L identified an error in the Station Blackout (SBO) impact analysis which resulted in the resultant peak suppression pool temperature exceeding the acceptance criteria stated in the NRC SBO Safety Evaluation. CP&L placed a restriction on power ascension to the pre-uprate value of 2436 megawatts-thermal (MWt), and began an intensive review and reanalysis effort of the power uprate modification products.
The error in the SBO impact analysis was caused GE personnel using incorrect design input for the SBO impact analysis, and CP&L failing to clearly establish inputs and analytical and licensing acceptance criteria for containment analyses. The lack of clearly defined inputs and acceptance criteria contributed to CP&L not identifying the analysis error sooner.
i Our investigations also indicate that the 200'F value in the Brunswick Plant Technical Specifications Design Features Section for the suppression chamber maximum teniperature j
should be 220'F. CP&L intends to correct this discrepancy during the implementation phase of the improved Technical Specification conversion project.
No additional discrepancies were identified during the reviews which impacted the conclusions of the Brunswick Plant Power Uprate Licensing Topical Report and the NRC Safety Evaluation.
The additional reviews did identify other discrepancies in documentation and clarifications of design inputs which were required. These additional discrepancies are described in this enclosure.
CP&L believes that the recent reviews, as well as the original reviews and power uprate analyses indicate that operation at the uprated power level is acceptable; however, CP&L will continue to maintain Unit 1 at or below the pre-uprate power level of 2436 MWt until this issue j
is resolved with the NRC staff.
1.0 Discussion of Circumstances On November 1,1996, the NRC issued License Amen'dment Nos.183 and 214 for the Brunswick Steam Electric Plant, Units 1 and 2, respectively (Reference 2). These license amendments approved an increase in the maximum power level for each of the Brunswick units from 2436 MWt to 2558 MWt. The NRC Safety Evaluation (SE) for power uprate, Section 3.4.2.2 (Station Blackout) states: ".. Suppression pool temperature increases about 4*F and the containment pressure about 2 psi, but the increases are small enough to not affect equipment availability.. "
During the process of updating internal Station Blackout calculations to reflect the analyses previously completed in support of the Brunswick power uprate Licensing Topical Report (LTR) submittal (Reference 1), CP&L identified an error in the SBO power uprate impact analysis that caused the resultant peak suppression pool temperature to exceed the 200'F suppression chamber temperature limit identified in Section 5 (Design Features) of the Technical Specifications. The acceptance criteria for suppression pool peak temperature identified in the NRC staff's SBO SE (Reference 3) was based on the Technical Specification limit.
CP&L determined that the 200*F suppression pool Technical Specification Design Feature limitation should also be applied to the Loss-of-Coolant Accident (LOCA) peak suppression pool E1-2
temperature analysis. Since the LOCA peak suppression pool temperature calculated for i
power uprate was 201 *F (Reference 1, Table 4-1), the LOCA analysis was reexamined.
On November 5,1996, CP&L established a 95% hold point on reactor power based on the results of this revised calculation, and began an investigation into the cause of the analysis discrepancy and a validation effort of the power uprate modification processes and products.
On November 14,1996, CP&L and the NRC staff held a conference call to discuss the SBO analysis performed in support of the power uprate application. As a result of that call, CP&L committed to the NRC staff to maintain reactor power on Brunswick Unit 1 to 5 2436 MWt, and i
submit a letter describing the circumstances of this issue and the proposed resolution, ahng with a discussion of how CP&L could have reasonable. assurance that no other issues of this type exist with the power uprate modification. CP&L reiterated the commitment to maintain reactor power 5 2436 MWt in a letter dated December 3,1996.
As a result of the SBO analysis error, CP&L has developed and completed an action plan with respect to the power uprate project that addresses the following:
The root cause of the error (both CP&L and General Electric)
Why an analysis error was identified late in the process Whether similar issues exist with the power uprate modification Confirmation of the design basis of the suppression chamber Specific resolution of the SBO issue Specific resolution of the LOCA issue Long-term corrective actions Details of the completed reviews, a discussion of the cause of the error and actions taken by CP&L, and the proposed resolution to the SBO and LOCA analysis issues follow.
2.0 Root Cautite A summary of the root cause is contained in this section. In parallel with the root cause evaluations, several investigations were initiated (process reviews, calculation reviews, etc.).
These investigations were augmented as the root cause analysis was formalized.
As a result of the investigations, CP&L identified two p'rimary causes of the SBO impact analysis error. One cause was determined to be an error made by General Electric (GE) personnel while performing the power uprate SBO impact analysis. CP&L had provided an existing SBO suppression pool heatup calculation to GE which was performed at 2436 MWt for their use in their analysis of the power uprate impact on post-SBO peak suppression pool temperature. The calculation provided to GE contained Revision 0 through Revision 3 information. GE personnel selected an incorrect peak suppression pool temperature as a starting point in determining the impact of power uprate on the Brunswick SBO event. The value of 188 *F used by GE in their power uprate impact analysis was subsequently reflected in the Analysis Basis Document rather than the correct value of 197.9 *F.
A second cause of the impact analysis error was not developing clearly defined inputs and analytical and licensing acceptance criteria for the various containment analyses at the onset of the power uprate project (e.g.,200 *F,220 *F, NPSH Margin, etc.). Not having this information clearly defined exacerbated GE personnel's lack of familiarity with the Brunswick SBO design E1-3
basis and contributed to the failure of CP&L reviews to identify the error in a design input used by GE in their analyses.
3.0 Whv Analysis Discrepancies Were Identified Late in the Process To answer this question, CP&L and GE performed investigations of the process used for the i
design and implementation of the power uprate modification. The following discussion provides a brief overview of the basic process, and points out where the barriers broke down that led to the item not being discovered until late in the process.
3.1 Design (Basic Process)
Provided inputs to GE Analyses:
CP&L provided design basis documents, design verified calculations, etc. for use in their power uprate impact analyses (CP&L has verified that GE was provided the correct SBO calculation). However, CP&L did not provide clearly defined analytical and licensing basis acceptance criteria for the various containment analyses.
General Electric Performs Analyses :
GE performed analyses and evaluations in accordance with the generic guidance provided in NEDC-31897P " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate" to assess the impact of power uprate for the Brunswick Plant.
As a result of these analyses and evaluations, GE generated several reports for CP&L review, including:
An Analysis Basis Document (ABD)
This document details the results of the engineering analysis for power uprate affects on Anticipated Transient Without Scram (ATWS) events, High Energy Line Break events, Containment events, Radiological Evaluation, Reactor Vessel Fracture Toughness, Reactor Vessel Internals, Main Steam Line Flow Elements, a Station Blackout event, and Fire Protection (Appendix R) events.
System Evaluation Reports (SERs)
These reports describe the affects of power uprate on system performance to determine whether modifications are required to existing systems and components to support the operational power increase.
Transient Analysis Report This report describes the results of the limiting transients evaluated to demonstrate the capability of BNP to operate at uprated power conditions. The transient analysis results are used to quantify the transient response trends associated with power uprate and are not intended to replace the reload transient analysis results.
Reactor Pressure Vessel (RPV) Stress Report This report documents the ASME Boiler and Pressure Code, Section Ill, analysis of limiting Reactor Pressure Vessel (RPV) components for BNP reactor vessels.
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4 Piping Evaluation Report This report presents a summary of the evaluation of the NSSS and BOP piping impacted by operation at power uprate conditions.
GE has confirmed that an error was made in performing the SBO analysis. Revision 3 of the SBO calculation contained information from previous calculation revisions. The GE engineer selected a suppression pool final temperature (188 *F) from information in the calculation prior to adding the approximately 4 F impact from power uprate, instead of the correct value of 197.9 'F. Contributing to the cause is a lack of familiarity of GE personnel with the Brunswick SBO Design Basis.
Draft documents containing the impact evaluations and analyses were reviewed by the 1
responsible engineers and comments sent to GE. Final documents received an owner's review and were accepted by CP&L.
Third party line-by-line reviews of the Brunswick Licensing Topical Report (LTR)
(Reference 1) were performed. Comparisons of LTR information were made to information contained in the ABD and SERs.
The ABD and LTR reviews by CP&L did not adequately verify the design inputs used by GE in the SBO impact analysis.
Several reviews were conducted by CP&L of the GE work, both at Brunswick and at GE's San Jose and Wilmington offices. However, CP&L did not review inputs when a CP&L design verified calculation was the starting point of GE evaluations. Therefore, the inappropriate selection of a design input by GE was not identified by CP&L during the review of the design analyses.
Design Deliverables:
Impact evaluations and calculations determined necessary to support modification development and submittal of the LTR were completed before the LTR was submitted on November 20,1995, except for setpoint calculations and the Electrical Environmental Qualification evaluations. The LTR noted that these documents (i.e., setpoint calculations and electrical EQ evaluations) were not completed. Setpoint calculations were completed prior to the License Amendment Request submitted on April 2,1996.
3.2 implementation (Basic Process) j Develop Modification Package:
The Analysis Basis Document, Transient Analysis Report, Piping Evaluation Report, Vessel Stress Reports, and System Evaluation Reports formed the basis of the analyses and conclusions presented in the LTR, which was submitted on November 20, 1995. Development of the modification package was initiated with the submittal of the LTR and was based on the completed impact analyses. Based on the completed power uprate impact analysis, the SBO calculation was being updated to reflect the results of tha analyses contained in the above reports. This update was required to be completed prior to startup from outage B111R1.
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On November 4,1996, as a part of the required calculation updates, the responsible j
engineer discovered the problems with the design verified SBO calculation used as a starting point for the power uprate project. The Brunswick Plant Nuclear Safety Committee (PNSC) established a hold Unit 1 at 2436 MWt and startup commenced on November 6,1996.
4.0 Review for Similar issues CP&L initiated and has completed reviews and affirmations to determine that issues similar to the SBO discrepancy do not exist. (Note : An " affirmation" documents a review and is a signed statement that documents the scope and depth of the review.) The following discussion provides the tasks performed and areas reviewed during this effort. The results of these reviews are contained in Section 5.0.
4.1 Calculations CP&L responsible engineers have completed calculation affirmations to certify that the inputs to the calculation updates contained in the modification package Document Update Form (DUF) accurately reflect power uprate conditions, Technical Specifications, and the UFSAR.
1 To ensure the DUF contained in the modification package was complete, CP&L has completed a review of the calculations of record for Unit 1 and those common to both units.
The responsible GE engineers have conducted reviews of calculations and Design Basis Documents used in their analyses to ensure no similar issues existed. A CP&L manager traveled to GE's offices in San Jose and confirmed proper use of CP&L calculations in GE's analyses.
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4.2 UFSAR, Technical Specifications, and NRC Safety Evaluation CP&L has reviewed each numerical value provided in the LTR against the Technical Specifications, UFSAR, and tha NRC staff's Power Uprate SE to validate that no further changes are required.
4.3 Affirmation of Analyses CP&L has completed affirmations on the Analysis Basis Document, Piping Evaluation Report, Reactor Pressure Vessel Stress Report, and Transient Analysis Report, which formed a principal part of the basis for the LTR. These affirmations were performed to ensure the inputs accurately reflect power uprate conditions, Technical Specifications, and the UFSAR.
l 4.4 Re-evaluation of ECCS, Containment, and Special Events Analyses CP&L has performed an additional review of ECCS, Containment, and Special Event analyses to ensure that the assumptions, inputs, and outputs are consistent with the UFSAR, Technical Specifications, and power uprate.
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4.5 Improved Technical Specifications 4
CP&L believes that the intrusiveness of the ITE review provides additional assurance that the Technical Specifications contained in tha Brunswick ITS submittal, the UFSAR, and the power uprate project are consistent. Iteras being relocated from the Brunswick Plant current Technical Specifications to other documents as a result of the conversion to ITS (e.g., design feature suppression pool water temperature) will be reviewed for i
discrepancies as a part of their incorporation into the relocation document. CP&L has performed a review of the list of relocated items against the UFSAR to ensure no further discrepancies exist.
4.6 Final Affirroation by Responsible Engineers A final affirmation by CP&L responsible engineers of systems and programs (e.g.,
Appendix R, SBO, etc.) affected by power uprate which have an assigned CP&L Design Basis Document has been completed, representing 101 systems and programs. This affirmation confirmed the consistency between these systems and programs with the Brunswick design and licensing basis. This affirrcation also confirmed the consistency between the Brunswick design and licensing documents. Discrepancies identified during this review have been resolved and appropriate design and licensing documents are being revised through normal CP&L processes.
5.0 Results from Review for Similar issues:
This section provides the results of the reviews discussed in Section 4.0.
5.1 Licensing Topical Report Items
- 1. Section 2.5.1.1 Control Rod Drive Mechanism The CRD code of record for Unit 1 is stated in the LTR as ".1968 Edition up to and including the Winter 1969 addenda." It should read ".1968 Edition up to and including the Winter 1970 addenda." Additionally, the allowable stress reported in the LTR should be 25,860 psi rather than 26,060 psi. The calculated stress of 20,790 psiis unaffected.
- 2. Section 6.4.2.1 Discharge Limits Table 6-3. The Condenser Temperature Rise in this table is not a North Carolina State National Pollutant Discharge Elimination System (NPDES) permit limit. The NPDES permit effluent temperature limit pertains to the ocean discharge mixing zone temperature, which was evaluated in the non-radiological environmental assessment (Reference 1). The Cooling Water Flow per Unit is correctly stated in the table.
- 3. Section 6.5 Standby Liquid Control System (SLCS)
The statement on Page 6-7 of the LTR, "..to deliver the ATWS required flow rate of 86 gpm.. " should read "..to deliver the ATWS required flow rate equivalent to 86 gpm.. " The 86 gpm flow rate provided in the ATWS Rule (10 CFR 50.62) is the flow required with a sodium pentaborate concentration of 13 E1-7
weight percent for a vessel size of 251 inches. The Brunswick units have 218-inch diameter vessels. The equivalent flow for a 218 inch vessel at 13 weight percent is 66 gpm. The Brunswick Plant will continue to meet the ATWS rule with power uprate by having the capability to inject 13 weight percent sodium pentaborate at 66 gpm. This value has been accepted by the NRC staff for the Brunswick Plant in previously approved license amendments (References 6 and 7).
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- 4. Section 9.2 Design Basis Accidents in the NRC power uprate SE, the staff reviewed the main steam line break outside of containment (MSLB) accident by increasing the control room dose previously reported in an NRC staff SER dated February 16,1989. The results provided in the NRC SE for power uprate calculate a dose of 20 Rem thyroid to the operators, which is below the GDC-19 criteria. CP&L did not submit an analysis for dose in the main control room for the MSLB event. An existing bounding analysis (i.e., no credit for Control Building Emergency Air Filtration i
until after 10 minutes) was updated as part of the ITS effort which bounds the values in the NRC power uprate SE and indicates that the resultant control room dose remained less than GDC-19 limits. A calculation of the control room dose from a MSLB accident was completed that compares favorably with the NRC staff MSLB accident results provided in the power uprate SE.
CP&L also reviewed a calculation that had been performed to support the elimination of the Main Steam Line Radiation Monitor isolation of the mechanical i
vacuum pumps as a part of this review effort to validate that power uprate would not impact the results. This calculation had been completed to determine control room dose following a control rod drop accident. The calculation assumed no automatic isolation of the mechanical vacuum pumps for 10 minutes, even though CP&L modified the isolation logic to provide an isolation of the mechanical vacuum pumps from the main stack radiation monitor. While the calculation was determined to not be affected by power uprate, a detailed review of the calculation assumptions and methodology found an error in a key assumption of the calculation. The calculation inappropriately assumed the dose in the control room ended at 10 minutes, once the mechanical vacuum pumps were isolated.
CP&L has recalculated the control room dose assuming isolation of the mechanical vacuum pumps from the main stack. The recalculation indicates that the control room dose from a control rod drop accident remains bounded by the MSLB accident and is below the GDC-19 limits. CP&L has initiated a condition report on this discrepancy. Based on the recognized significance of this trip function, as a short-term corrective action, CP&L has completed testing of the main stack isolation logic for the mechanical vacuum pumps to ensure the function operates as designed. In additon, the applicable maintenance surveillances tests will be revised to require the operability of the mecharhcal vacuum pump isolation function as acceptance criteria for a successful test.
CP&L is reviewing this issue through the Brunswick Plant Corrective Action Program process to determine what additional short and long-term measures (e.g., Technical Specification amendment) are needed.
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- 5. Table 9-3 The " Site Boundary Distance (m)" should be clarified to read " Site Boundary Distance (m), Low Population Zone (m)" to be consistent with the two values given in this table.
- 6. Section 10.1.1 Temperature, Pressure, and Humidity Profiles During the final affirmation reviews of the Reactor Building Environmental Report (RBER) by the responsible engineer, an increase in the RWCU Heat-Exchanger Room pressure over the previously evaluated results from a High Energy Line Break (HELB) was identified. The pressure increased from 18.4 psi to 19.1 psi.
An evaluation has been completed that concludes the RWCU Heat-Exchanger j
Room can withstand the HELB pressure increase to 19.1 psi. This conclusion is j
consistent with the conclusions in the LTR, Section 10.1.
CP&L believes that the changes identified herein do not affect the conclusions of the NRC staff power uprate SE for the Brunswick Plant.
5.2 Technical Specification items 1.
HPCI Bases 3/4.5.1 1120 psig to 1164 psig.
CP&L has identified one Technical Specification Bases page (page B 3/4 5-1) which should have been updated to reflect changes due to power uprate. The change involves the operating pressure range of the HPCI system. The bases indicates that the HPCI system is designed to provide core cooling at reactor pressures between 1120 and 150 psig. As noted in the LTR (Reference 1, p. 4-7), "the (HPCI) system was also found to have the capability to deliver its design rated flow at the increased reactor pressure resulting from the SRV safety function pressure setpoint increase of 25 psi and the allowable setpoint tolerance of 3%.' Therefore, the Technical Specification Bases value of 1120 psig should have been revised to 1164 psig.
This change in higher operating pressures for the HPCI System was noted in the NRC SE (See Reference 2, p. 8). The change to this Bases page does not impact any conclusions in the NRC SE. A mark-up of the Bases pages for both units change is included in this document as Enclosure 2.
2.
The reviews identified a discrepancy in the current Technical Specification Bases that is not related to power uprate, in that the Bases of Technical Specification 3/4.5.1 is not consistent with the Specification. Specifically, the specification i
allows 14 days for one ADS valve out of service, and no out of service time for 2 or more inoperable ADS valves. The Bases note that one ADS valve can be out of service indefinitely, and 2 ADS valves can be out of service for up to 7 days provided HPCI is operable. The ADS bases discrepancy has been corrected in the ITS submittal (Reference 4).
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3.
Torus Volume.
During the ITS project reviews, CP&L identified a discrepancy between the Torus volume identified in the Technical Specifications and the calculated volume. This discrepancy was discussed in LER 1-96-15. As indicated in Supplement 1 to LER 1-96-15, CP&L will be submitting a separate license amendment request that will revise the suppression chamber water volumes to be consistent with the current Technical Specification level values by January 15,1997. This amendment corrects the non-conservatism between the actual suppression chamber water volumes associated with the Technical Specification levels of
-31" and -27" and those volumes currently indicated in the Technical Specifications. The power uprate analyses assumptions are consistent with the volumes in the proposed license amendment request and bound the volumes currently indicated in the Brunswick Technical Specifications.
4.
Vessel Volume - Technical Specification 5.4.2 Technical Specification 5.4.2 indicates the vessel volume is "Approximately 18,670 cubic feet." This value matches the FSAR vessel values used for input to the original containment analyses, which used the CONTEMPT-PS model. For power uprate, GE used the SHEX long-term containment model, which requires more discrimination in water and steam volume input parameters. The input parameters used in the SHEX model were approximately 18,320 cubic feet for vessel volume plus an additional feedwater contribution of approximately 3,500 cubic feet. SHEX provides a more realistic model of containment response. The vessel volume value used by GE was derived from the " Primary System Weights and Volumes" design drawings for Brunswick Units 1 and 2. CP&L confirmed that these design drawings is on the master parts list for the Brunswick Plant at the GE San Jose offices. This vessel volume value is relocated from the Technical Specifications during the ITS conversion project, and controlled in the Bases by direct referral to the containment section of the UFSAR, which includes the containment analysis parameters.
5.
Technical Specification 5.2.2.b (Suppression Chamber Design Temperature)
Technical Specification 5.2.2.b indicates that the suppression chamber design temperature is 200 'F. Based on our reviews of the suppression chamber design and the history of the design features section of the Brunswick TS, CP&L has determined that the Technical Specification Design Features Section value for the suppression chamber should be 220*F. A discussion of containment design basis is provided in Section 6.0 of this enclosure. This issue will be i
resolved during the Improved Technical Specification (ITS) process. As a part of I
that effort, the containment design values are being relocated to the UFSAR.
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CP&L plans to correct the value, in accordance with 10 CFR 50.59, during implementation of the ITS at the Brunswick Plant.
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5.3 Modification Document Update Form (DUF) Items CP&L reviews identified additional updates and/or clarifications to the UFSAR, calculation updates, and changes to the Design Basis Documents that will be made in conjunction with power uprate closure. These items will be added to the DUF for disposition and are being further investigated using the Brunswick Plant Corrective Action Program.
Based on the extensive reviews originally completed for the power uprate project, and the i
additional reviews presented in this enclosure, CP&L has reasonable assurance that no technical issues beyond those identified in this letter exist with respect to the power uprate project.
6.0 Confirmation of Suporession Chamber Desian Basis i
As a result of the containment issues identified, CP&L initiated reviews to confirm the j
suppression chamber design basis, including assumptions and acceptance criteria. The following provides the results of these reviews:
BNP Technical Specification 5.2.2 states that the maximum internal suppression chamber temperature is 200 *F. The original BNP Custom Technical Specifications stated that the design parameters of the primary containment shall be given in Final Safety Analysis Report (FSAR) Table 5.2-1. This statement is consistent with the statement in the Technical Specifications of many older BWRs with custom Technical Specifications.
l The Preliminary Safety Analysis Report Table (Table V-2-1) lists a maximum internal suppression chamber temperature of 220 F.
The original suppression chamber fabrication drawings for Unit i and Unit 2 specify a design temperature of 220 F.
The original FSAR Table (Table 5.2-1), which was not submitted to the NRC (formerly the AEC), listed the design temperature of the suppression chamber as 200 *F.
l Amendment 16 to the FSAR, which was issued to the AEC in April 1972, listed a j
suppression chamber internal design temperature of 220 F (Table 5.2-1).
The NRC (formerly the AEC) was provided a suppression chamber design temperature j
of 220 *F per the response to AEC question 5.54 (FSAR Appendix M5.54.1)
The Updated FSAR Table 6.2.1-1 specifies a maximum internal suppression chamber temperature of 220 F.
In November 1977, BNP converted the original custom Technical Specifications to the Standard Technical Specifications (STS). The 200 F value for suppression chamber design temperature was incorporated into the Design Features section of the STS at that time. CP&L has not identified any basis for this value; however, we have determined that 220 F is the correct value for the suppression structural design temperature. The error may have been introduced during E1-11
the conversion to the STS through the inappropriate use of the original FSAR Table 5.2-1(i.e.,
prior to Amendment 16).
Based on the results of this review, CP&L has concluded:
1.
Power uprate does not challenge the drywell and suppression chamber structural l
designs.
l 2.
The suppression chamber design temperature in the Technical Specifications should be 220*F.
3.
Other analytical parameters (e.g., ECCS NPSH values) provide a more restrictive limit on suppression chamber water temperature than the containment structural design for various events (see Table 1).
4.
The SE issued by the NRC for power uprate is correct and is not impacted.
l 7.0 Specific Resolution of SBO Issue CP&L has completed a revision of the SBO calculation using the analytical assumptions and decay heat calculation methodology consistent with the original hand calculation reviewed by the NRC staff for SBO (See Reference 3, Section 6.). The differences between the revised SBO calculation and the original SBO calculation are as follows:
The decay heat assumption used in the revised SBO calculation was refined. The revised calculation uses American Nuclear Society 5.1 (ANS 5.1 standard) for modehng of decay heat generation, as indicated was preferred by the NRC staff in the SBO SE (see Reference 3, p. 23). The refinement of the curve is accomplished by adding additional data points to the decay heat curve. Incorporating these additional data points into the decay heat curve results in a smoother, more accurate curve. The resulting decay heat energy that would reach the suppression pool as the result of a SBO event would be lower than that calculated using the original curves (See Figure 1 to this enclosure).
No credit is taken in the revised SBO suppression pool peak temperature calculation for heat transfer from the suppression chamber. This assumption is conservative with respect to the NRC staff SBO SE (Reference 3, p. 24), which indicated that, although some credit is reasonable, CP&L had credited excessive natural convection heat transfer in our original SBO calculation.
The suppression chamber initial water temperature used in the original calculation was increased in the revised calculation (up from 90"F to 95'F). The original 90 "F input was consistent with the normal expected maximum torus water temperature. However, CP&L elected to revise the analysis to incorporate the more conservative value of 95*F, which is consistent with the current Technical Specification allowed limit (Reference Technical Specification 3.6.2.1.a.2).
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The suppression chamber water volume used in the revised calculation was reduced from 87,600 ft' to 86,545 ft. This revised value corrects a discrepancy in the Brunswick 2
Plant Technical Specification between the volume and level allowed by Technical Specification 3.6.2.1.a.1, and is consistent with the values discussed in LER 1-96-15.
i An error in the original SBO calculation associated with the mass of water that enters the suppression pool from the reactor was corrected. The original calculation had accounted for the heat transferred to the suopression pool from the fluid, but did not account for the mass addition to the suppression pool.
The peak suppression pool temperature calculated for a postulated SBO event, using the assumptions and methodology described above, is 198.8 *F. This value is below the 200*F value in Technical Specification 5.2, which is the value indicated as acceptable in the NRC SBO SE (Reference 3, p. 8).
Based on the revised SBO calculation, the peak suppression pool temperature following a SBO event is less than the 200*F limit used by the NRC staff in the SBO SE (see Reference 3), and is less than the 200*F suppression chamber design value in the Section 5 of the Technical Specifications.
8.0 Specific Resolution of the LOCA lssue The LOCA peak suppression pool temperature analysis was reperformed. The differences between the revised LOCA calculation and the original power uprate LOCA calculation are as follows:
The decay heat assumption used in the revised LOCA calculation was refined. The revised calculation uses American Nuclear Society 5.1 (ANS 5.1 standard) for modeling of decay heat generation. The refinement of the curve is accomplished by adding more data points to the decay heat curve. Incorporating these additional data points into the decay heat curve results in a smoother, more acct. rate curve (see Figure 2). The resulting decay heat energy that would reach the suppression pool as the result of a LOCA would be. lower than that calculated using the original curves.
The original power upre.e LOCA calculation assumed a initial drywell temperature of 135'F. CP&L is pursuing a change to increase the drywell average air temperature from 135'F to 150*F. This amendment is being coordinated with the improved Technical Specification conversion effort (Reference 4). To be consistent with this Technical Specification change, the 150*F temperature was used an initial drywell temperature in the latest LOCA analysis.
The suppression chamber initial water temperature of 95*F used in the original power uprate calculation was also used in the revised LOCA analysis. The 95*F assumption is consistent with the current Technical Specification allowed limit (Reference Technical Specification 3.6.2.1.a.2).
The suppression chamber water volume used in the revised calculation was reduced from 87,600 ft* to 86,545 ft'. This revised value corrects a discrepancy in the Brunswick E1-13
Plant Technical Specification between the volume and level allowed by Technical Specification 3.6.2.1.a.1, and is consistent with the values discussed in LER 1-96-15.
The peak suppression pool temperature calculated for a postulated LOCA, using the assumptions and methodology described above, is 199.5 *F. This value is below the 200*F value in Technical Specification 5.2, and well below the suppression chamber structural design temperature (See Section 6.0).
9.0 Lona-Term Corrective Actions CP&L has initiated long-term corrective actions in response to this issue and will factor the lessons-learned from power uprate into the evaluations being conducted for the NRC design basis information request issued under 10 CFR 50.54(f). The long term actions include:
Revising the UFSAR and DBDs to ensure that the containment analysis inputs and acceptance criteria are clearly defined.
Performing an evaluation of the verification and validation practices for outsourced engineering work.
10.0 Summary:
Based on the completed revision of the SBO and LOCA calculations, using the analytical methods reviewed and approved by the NRC staff in the associated SEs, CP&L believes that compliance with the current 200*F Technical Specification limit has been adequately demonstrated. CP&L also believes that it has demonstrated reasonable assurance that no further issues exist with the power uprate modification. Upon NRC concurrence with this position, CP&L plans to continue power ascension beyond the currently maintained 2436 MWt power limit.
11.0 References 1.
NEDC-32466P, dated September 1995, " Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant, Units 1 and 2," transmitted to NRC staff by "BSEP-95-0535, Forwards non-proprietary & Proprietary Versions of TR Both Entitled, Power Update SAR for BSEP Units 1 & 2,Sept 1995. Proprietary Version Withheld Per [[CFR" contains a listed "[" character as part of the property label and has therefore been classified as invalid..790|letter dated November 20,1995]] (BSEP 95-0535) i 2.
NRC Letter to CP&L, " Issuance of Amendment No.183 to Facility Operating License No. DPR-71 and Amendment No. 214 to Facility Operating License No. DPR-62 Regarding Power Uprate - Brunswick Steam Electric Plant, Units 1 and 2 (BSEP 96-0123) (TAC Nos. M90644 and M90645)," November 1,1996.
3.
NRC Letter to CP&L, " Station Blackout Evaluation - Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. 68520 and 68521)," October 4,1990.
4.
CP&L Letter to NRC, BSEP96-0414, " Conversion to improved Standard Technical Specifications," November 1,1996.
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5.
CP&L Letter to NRC, BSEP 96-0123," Power Uprate License Amendment Request,"
April 2,1996.
6.
Letter from NRC to CP&L, Issuance of Amendment 106 to Facility Operating License No. DPR-71 for Brunswick Steam Electric Plant Unit 1, April 10,1987.
7.
Letter from NRC to CP&L, Issuance of Amendment No.143 to Facility Operating License No. DPR Brunswick Steam Electric Plant, Unit 2, Regarding SLCS,
' January 28,1988.
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l t
o Table 1
! Analysis}
Limit. Basis ~
Non dprhtoi UpratC ___
NPSH?
2436 MWt 4 2568 MWtt LimitJ DBA LOCA-CS NPSH required for CS Pumps Not 162.5'F 175.9'F*
Short Term Performed (4.32 ft.")
DBA LOCA-LPCI NPSH required for RHR Pumps Not 161.8'F 165.9'F*
Short Term Performed (1.22 ft.")
DBA LOCA NPSH required for CS Pumps 182.4*F 189.4*F 201.3'F*
Case B (6.34 ft.")
Containment Cooling DBA LOCA NPSH required for RHR Pumps 182.4*F 189.4*F 201.7'F*
Case B (6.57 ft.")
Containment Cooling DBA LOCA Limit is Technical Specification 204.6*F 199.5'F 211.8* F Case C Value of 200 *F*"
(UFSAR)
(16.6 ft.")
Containment 197'F Cooling (LTR)
- F on the T/S value of 200'F*"
(1.48 ft.")
Appendix R NPSH required for RHR Pumps 185'F 186.4'F 202.2* F (7.69 ft.")
Local Pool Temp.
Localized 203*F temperature near 196*F 198.1 *F N/A NUREG 0783 T-Ouencher (for complete (NPSH N/A) condensation)
No credit is taken for Containment Pressurization per Safety Guide 1 Margin to NPSH Limit (i.e., Available NPSH - Required NPSH)
With a suppression chamber structural design limit of 220 *F, NPSH provides a more restrictive limit.
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Figure 1: SBO Calculation -Input Decay Heat Comparison 0.025 t
u a
l 0.020 0
1 o
n.
m o
U a:
3 0
t i 0.015 o
a 1
o i!
o E
o j 0.010 o
0-m t*
E i
I m Original SBO Calc. Decay Heat Valtes 0.005 E
o Decay Heat Additional Data Points as 0.000 0
2000 4000 6000 8000 10000 12000 14000 16000 Time (Seconds) i I
i i
Figure 2: LOCA Calculation -Input Decay Heat Comparison 0.016 m
0.014 i
E Ei 0.012 t;
3 a:
I 0
- 0.010 x
C a
2v a
-2 0.008 a
E 0.006 i
E c
3 0.004 o
(
3 5*
m Original LOCA Calc. Decay Heat Values 0.002 o Decay Heat Additional Data Points 0.000 0
5000 10000 15000 20000 25000 30000 35000 40000 Time (Seconds)
P
-