BSEP-96-0287, Forwards Response to RAI Re Request for License Amends on Power Uprate

Forwards Response to RAI Re Request for License Amends on Power Uprate

ML20116E348
Person / Time
Site:	Brunswick
Issue date:	07/30/1996
From:	Campbell W CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BSEP-96-0287, BSEP-96-287, NUDOCS 9608050218
Download: ML20116E348 (10)

Text

m CP&L l Corolina Power & Light Company William R. Campbell PO Box 10429 j Vice President Southport NC 28461 0429 Brunswick Nuclear Plant l JUL 301996 SERIAL: BSEP 96-0287 i

10 CFR 50.90

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U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 l DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62

! SUPPLEMENTAL INFORMATION TO REQUEST FOR LICENSE AMENDMENTS -

POWER UPRATE l

i Gentlemen:

On July 9, Juiy 15, and July 18,1996, the NRC staff forwarded additional questions to Carolina Power & Light Company (CP&L) on the proposed amendment to the Technical Specifications for the Brunswick Steam Electric Plant (BSEP), Unit Nos.1 and 2 which was submitted on April 2,1996. These proposed amendments revise the BSEP Technical Specifications to allow uprate of the units to 105% of rated thermal power.

Enclosure 1 provides CP&L's formal response to the questions raised by the NRC staff.

Enclosure 2 provides a list of regulatory commitments.

Please refer any questions regarding this submittal to Mr. Tony Harris at (910) 457-3312.

Sincerely, I

William R. Campbell VUkah i Enclosures i

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Document Control Desk BSEP 96-0287 / Page 2 William R. Campbell, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power & Light Company.

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My commission expires: k /2,/79f pc: Mr. S. D. Ebneter, Regional Administrator, Region ll Mr. D. C. Trimble, Jr., NRR Project Manager - Brunswick Units 1 and 2 Mr. C. A. Patterson, Brunswick NRC Senior Resident inspector The Honorable H. Wells, Chairman - North Carolina Utilities Commission 1

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ENCLOSURE 1 l

BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2  !

NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS

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POWER UPRATE  !

SECOND NRC STAFF REQUEST F_QRADDITIONAL INFORMATION Question 1 In the RAI response dated July 2,1996, the last page contained a table of commitments. For the reliabihty monitoring of the HPCI and RCIC systems via the Maintenance Rule, the sheet is noted as "N/A" Why is this considered "Not Applicable"?

CP&L Response to Question 1 Enclosure 4 of CP&L's July 2,1996 response contains a list of commitments in the letter. The table has a column to denote the actual commitment, and a column to denote the date for implementation of the commitment. Since the commitment to monitor reliability of the HPCI and RCIC systems was being completed under the Maintenance Rule, and the commitment is an I ongoing commitment, the date is annotated as NA. CP&L remains committed to monitoring the reliability of both the HPCI and RCIC systems under the Maintenance Rule.

Question 2 Table 4-1 of the topical report NEDC-32466P provided the long-term suppression pool temperature response calculated for a postulated DBA-LOCA. In addition to this table, Section 4.2 on Page 4-6 of NEDC-32466P provided a different value for the long-term suppression pool temperature response. Provide an explanation for the different values provided in the report.

CP&L Response to Question 2 The Brunswick LFSAR contains analyses of four (4) cases of RHR cooling for the long-term  ;

suppression pool temperature response (Reference UFSAR Section 6.2.1.1.3.2). As stated in the UFSAR, Case C is the " design basis event" for peak bulk suppression pool temperature.

This case assumes a DBA-LOCA occurs with the following low pressure coolant systems operable: One (1) Core Spray loop, one (1) RHR loop containing one (1) RHR pump, one (1)

RHR heat exchanger, and two (2) Service Water pumps. The current UFSAR evaluation was performed at 2550 MWt, and the resultant peak suppression chamber temperature is 205*F.

General Electric evaluated Case C for power uprate conditions. In performing the evaluation, a more realistic decay heat model was used. In order to provide an appropriate comparison between the current power level (2436 MWt) and the uprated power level (2558 MWt), a calculation was performed for 102% of 2436 MWt using the more realistic decay heat model.

This calculation determined the current paak bulk suppression pool temperature to be 197'F.

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l The calculation was also performed for a power level of 102% of 2558 MWt, with a resultant peak suppression pool temperature of 201 *F. These values are provided in Table 4-1 of the topical report (NEDC-32466P).

In calculating the available Net Positive Suction Head (NPSH), the UFSAR states that Case B is the correct mode of operation to be used (Reference UFSAR Table 6.2.1-8). Assuming a power level of 102% of 2558 MWt, the peak bulk suppression pool Case B temperature is calculated to be 189"F. This value is provided on Page 4-6 of the topical report (NEDC-32466P).

Question 3 l

The License Amendment Request submitted a change in the SRV lift tolerance from 1% to i 3%. Was the overpressure analysis for the reactor coolant system performed using the 3%

l tolerance? If so, is the use of the 3% tolerance conservative?

l CP&L Response to Question 3 i j The BWROG submitted topical report NEDC-31753P to the NRC staff July 9,1990. The topical report provides the results of generic evaluations performed for the BWROG to support j changes to Technical Specification requirements for safety relief valves (SRVs). The NRC l provided a safety evaluation (SE) for this report on March 8,1993. The March 8,1993 NRC SE l provided six requirements for licensee analyses which would implement the proposed l specifications. CP&L evaluated the requested change in accordance with the guidelines )

established in the NRC Safety Evaluation for acceptance of General Electric topical report l NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report". Specifically, the six requirements from the NRC SE, and CP&L's evaluation of l these requirements, are provided belcw.

1. Transient analysis of all abnormal operational occurrences as described in NEDC-31753P, should be performed utilizing a 13% setpoint tolerance for the safety mode of the SRVs.

CP&L Evaluation: For power uprate, the analyses performed were identified in Table 9-1 of the topical report, with the initial conditions assumed. This table referenced the l analytical limits for the SRVs as shown in Table 5-1. These analytical limits included the l 3% lift tolerance.

2. Analysis of the design basis overpressure event using the 3% tolerance limit for the SRV setpoint is required to confirm that the vessel pressure does not exceed the ASME pressure vessel code upset limit.

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CP&L Evaluation: The results of the Brunswick specific analysis are discussed in Section

3.2 of the topical report. Section 3.1 of the report details the important assumptions for

{ this evaluation, including a SRV lift tolerance of13%.

j 3. The plant specific analyses described in Items 1 and 2 should assure that the number of j SRVs included in the analyses correspond to the number of valves required to be g operable in the technical specification.

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CP&L Evaluation: The License Amendment Request submitted on April 2,1996, included j a revision to the SRV Limiting Condition for Operation to specify the required number of SRVs. The plant-specific analysis described in Items 1 and 2 conservatively assumed one less SRV was operable than required by current Technical Specifications.

4. Reevaluation of the performance of high pressure systems (pump capacity, discharge pressure, etc.), motor-operated valves, and vessel instrumentation and associated piping must be completed, considering the 3% lift tolerance.

CP&L Evaluation: The BNP Power Uprate topical report (NEDC-32466P) details the results of the system evaluations for power uprate. RCIC system is addressed in Section 3.8 and the HPCI system is addressed in Section 4.2.1. The evaluations were performed considering a 3% SRV lift tolerance.

5. Evaluation of the 3% tolerance on any plant specific alternate operating modes (e.g.,

increased core flow, extended operating domain, etc.) should be completed.

CP&L Evaluation: The Power Uprate power / flow map includes the Maximum Extended Operating Domain (MEOD) which allows operation between 81% and 105% of rated core flow at the new rated power. This map is provided in the topical report as Figure 2-1.

The transient analyses performed for power uprate utilized this figure as the basis for the analyses, including the new SRV lift settings and the t3% tolerance. The results of the transient analyses indicate the overall capability of the design of the Brunswick units to meet all transient safety criteria for uprated operation. These results are reported in Section 9.1 of the topical report.

6. Evaluation of the effect of the 3% tolerance limit on the containment response during loss of coolant accidents and the hydrodynamic loads on the SRV discharge lines and containment should be completed.

CP&L Evaluation: The results of the Brunswick specific evaluation indicate that the Brunswick containment wi!I continue to perform its function and maintain the accident pressures and temperatures within the existing design limits assuming a 3% SRV lift tolerance. The results of the analysis are described in Section 4.1.2 of the topical report.

Question 4 In the application package, the licensee did not provide an analysis for single loop operation and any multipliers that may be used for limits such as MAPLHGR and MCPR. Please explain why no such discussion was included.

CP&L Response to Question 4 Extended single loop operation is not currently licensed for BNP and is not being requested under the power uprate license amendment request; therefore, no discussion was included in the package. However, the power uprate evaluation did review this operating mode and found it to be acceptable under power uprate conditions, as stated in Section 4.3 (Page 4-8) of the topical report. The SLO option is currently limited to 88% power. Under power uprate operation, the limiting power level would be limited to 83.8%.

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The current Brunswick practice is to submit any multipliers to the MCPR and MAPLHGR with the Core Operating Limits Report (COLR) for each cycle. The multipliers would be included in the COLR for the upcoming power uprate cycle for Brunswick Unit 1 (Cycle 11). Any future amendment request for single loop operation would ensure that the limitations imposed by the power uprate analyses are addressed.

Question 5 Regarding the licensee's answer to Question 2, explain why Case C and its associated peak bulk suppression pool temperature are an overly restrictive condition for calculating NPSH margin.

CP&L Response to Question 5 Cases A through D are defined in BNP UFSAR Section 6.2.1.1.3.2.1.b.2 (page 6.2.1-15).

Case A includes both RHR trains running, with no loop / equipment failures. Case B includes a sing'e RHR train running (other train assumed failed). Cases C and D have only a partial RHR train running (assumed failure of other loop a_nldr one RHR pump in operating train in Case C, and also an additional Service Water Pump failure in Case Di. Cases C and D include multiple unrelated failures and are therefore non-credible events an ayond the licensing basis of the plant. A note in UFSAR Table 6.2.1-8 (page 6.2.1-53) indicates that Cases A and B are the only credible scenarios.

Question 6 Regarding licensee's answer to Question 2, has the new GE decay model used for calculation of peak bulk suppression pool temperature been reviewed and approved by the NRC? If so, please provide references.

CP&L Response to Question 6 The "more realistic decay heat model" referred to in the response to Question 2 is the ANSI /ANS 5.1-1979 model. The generic power uprate licensing report (NEDC-31897P-A) described GE's methodology for containment evaluations, including use of the SHEX computer model and the ANSl/ANS 5.1-1979 decay heat model (see Appendix G and Reference 18 of that report).

NRC staff approval of NEDC-31897P-A provided conditional approval of GE's containment methodology. A December 1991 meeting was held with the NRC to clarify the NRC's position.

The outcome of the meeting was an understanding that the NRC accepted the use of SHEX and the ANS decay heat model, provided that, if the use of these models represented a change from the plant's original licensing basis, then the application would discuss any changes to input parameters and would provide benchmarking analyses (i.e., a containment evaluation at the original power level and an evaluation at the uprated power level).

GE wrote a letter to the NRC (MFN-026-93, dated March 12,1993) requesting that the NRC document their acceptance of the NEDC-31897P-A methodology. The NRC responded with a letter dated July 13,1993. The NRC has also reviewed and approved the use of SHEX and ANS decay he at model on other uprate applications.

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l The use of SHEX and the ANS decay heat model for Brunswick power uprate is described in NEDC-32466P, Section 4.1.1. The analysis performed included a benchmark at the current power level in addition to the analysis at the new uprated power level to ascertain the true affects due to power uprate vice the changes due to the decay heat model.

Question 7 Section 9.3.1 of NEDC-32466P, Supplement 1, Page 9-3, indicates that the "ATWS high pressure analytical limit" was increased by 50 psi. That value was then used as a key input for calculating peak vessel pressure during an ATWS event. What is the "ATWS high pressure analytical limit" referred to in this section. Is it the ATWS high reactor precsure trip setpoint?

Why was it necessary to increase this value by 50 psi for the Brunswick specific ATWS analysis at the uprated power?

CP&L Response to Question 7 l l

The ATWS high pressure analytical limit referred to in Section 9.3.1 of Supplement 1 to -

NEDC-32466P is the ATWS-RPT analytical limit. The ATWS-RPTinitiates a trip of the recirculation pumps in the event of an ATWS transient. This analytical limit is being changed from 1120 psig to 1170 psig, which is the 50 psig increase noted in our submittal. Although the analytical limit for the ATWS-RPT is being increased by 50 psi, the allowable value for this trip i increases from 1120 psig to 1143 psig (23 psi) and the trip setpoint increases from 1120 psig to 1137.8 psig (17.8 psi). Analysis of events u performed based on the analyticallimit increase of 50 psi.

As stated in the Basis for Change No. 24 on page E1-7 of the License Amendment Request  ;

(Letter BSEP 96-0123, dated April 2,1996), this value was increased in the power uprate ATWS evaluation to account for the increase in vessel operating pressure, SRV setpoints and tolerances, and to maintain operational margin. Raising the ATWS-RPT high pressure setpoint prevents unnecessary recirculation pump trips following pressurization transients with a reactor scram (e.g. turbine trip or load reject without bypass), while maintaining the requirements of 10 CFR 50.62. These recirculation pump trips lead to thermal stratification within the reactor vessel, as noted by the NRC in Infommtion Notice 93-62. Continued recirculation pump operation following a scram allows for better mixing of the reactor coolant and reduces vessel thermal stratification.

Question 8 Sec. 9.3.1 of NEDO-32466 Supplement 1 only discusses acceptability of ATWS analysis with respect to reactor pressure. It does not clearly discuss acceptability with respect to the other acceptance criteria associated with ATWS. Piease provide an appropriate discussion and conclusion for this.

CP&L Response to Question 8 in Supplement 1 to the topical report, the results of the ATWS analysis for a higher ATWS-RPT analyticallimit were provided. The change to this parameter only rasults in a significant effect on the peak vessel pressure. There is no effect on suppression pool temperature and peak fuel E1-5

conditions (maximum clad temperature and maximum local clad oxidation). Since the only significant impact was to the peak vessel pressure, this was the only discussion included in the topical report supplement.

i Question 9  !

Section 9.3.1 of NEDC-32466P Supplement 1 (Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant) describes a Brunswick specific analysis for Anticipated i Transients Without Scram (ATWS), which relies, in part, on a generic analysis described in l Sec. 3.7 of NEDC-31984P Supplement 1 (Generic Evaluations of General Electric Boiling Water Reactor Power Uprate). Sec. 9.3.1 (NEDC-32466 Supp.1) states that, per the generic ,

analysis (NEDC-31984P Supp.1), the inadvertent MSIV Closure (MSIVC) event is the limiting l ATWS event. In reviewing the generic ana!ysis, the NRC staff noted that a pressure regulator failure may be a more limiting ATWS ever.i with respect to some parameters analyzed. What is .

your basis for concluding that MSIVC is the most limiting ATWS event when all of these parameters are considered? Please address both the case where the ATWS trip setpoint was assumed to be increased by 20 psi and the case where the setpoint was increased by 50 psi.

Were both the MSIVC and pressure regulator failure events reanalyzed at the increased (50 psi increase) ATWS high pressure setpoint? If so, please provide the results at the +50 psi setpoint for each parameter analyzed and the associated acceptance criteria.

CP&L Response to Question 9 As indicated in the response to Question 7, increasing the ATWS-RPT setpoint would maintain l operational margin and reduce the probability of thermal stratification. Although the analytical limit for the ATWS-RPT is being increased by 50 psi, the allowable value for this trip increases from 1120 psig to 1143 psig (23 psi), and the trip setpoint increases from 1120 psig to 1137.8 psig (17.8 psi). Analysis of events was performed based on the analyticallimit increase of 50 psi.

As stated in response to Question 8, the only significant effect of an increase in the ATWS-'RPT analytical limit is an increase in the peak vessel pressure. The results in NEDC-31894P Supplement 1 (Section 3.7) indicate that for peak vessel pressure, the MSIV closure event is limiting. The analysis performed in NEDC-31894P, Supplement 1 was based on the original ATWS analysis performed for the BWR product lines detailed in NEDE-24222 " Assessment of BWR Mitigation of ATWS, Volume 11"(NUREG 0460 Alternate No. 3). In this analysis the Pressure Regulator Failure - Maximum Demand (PREGO) event was shown to follow the MSIVC event after the initial decrease in reactor power and pressure. Since the reactor pressure increases are limited by the action of the SRVs and the RPT, and the RPT high pressure setpoint was the only variable changed for the Brunswick specific analysis, the PREGO event was not reanalyzed for the increase in the ATWS-RPT setpoint. The MSIVC event remains the most limiting event for the Brunswick Plant considering the 50 psi increase in the ATWS-RPT analyticallimit.

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ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50-324 OPERATING LICENSE NOS. DPR-71 & DPR-62 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Carolina Power & Light Company in this document. Any other actions discussed in the submittal represent intended or planned actions by Carolina

- Power & Light Company. They are described to the NRC for the NRC's information and are not considered regulatory commitments. Please notify the Manager Regulatory Affairs at the Brunswick Nuclear Plant of any questions regarding this document or any associated regulatory commitments.

Commitment Committed Date No New Commitments Are Contained in This Letter N/A E1-7