ML17241A194

License Amendment Request - Revise Technical Specifications Section 3.4.3 (Srvs/Svs) for the Remainder of the Current Operating Cycle for Unit 2 - Supplement 1 Response to Request for Additional Information

ML17241A194
Person / Time
Site:	Peach Bottom
Issue date:	08/29/2017
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF9705
Download: ML17241A194 (11)

Text

Exelon Generation ~

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 August 29, 2017 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Unit 2 Renewed Facility Operating License No. DPR-44 NRC Docket No. 50-277

Subject:

License Amendment Request- Revise Technical Specifications Section 3.4.3 (SRVs/SVs) for the Remainder of the Current Operating Cycle for Unit 2 -

Supplement 1 Response to Request for Additional Information

References:

1. Exelon letter to the NRC, "License Amendment Request -

Revise Technical Specifications Section 3.4.3 (SRVs/SVs) for the Remainder of the Current Operating Cycle for Unit 2," dated May 19, 2017 (ADAMS Accession No. ML17139D357)

2. Email from R. Ennis (USNRC) to D. Neff (Exelon), "Peach Bottom Unit 2 - Request for Additional Information -

Amendment Request Regarding Safety Relief Valve and Safety Valve Operability for Cycle 22 (CAC MF9705)," dated August 10, 2017 (ADAMS Accession No. ML17222A096)

In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) requested an amendment to Renewed Facility Operating License No. DPR-44 for Peach Bottom Atomic Power Station (PBAPS) Unit 2 (Reference 1). Specifically, the proposed change would revise the Technical Specifications (TSs) to decrease the number of safety relief valves (SRVs) and safety valves (SVs), required to be operable, when operating at a power level less than or equal to 3358 megawatts thermal (MWt). This change would be in effect for the current PBAPS Unit 2 operating cycle (Cycle 22) that is scheduled to end in October 2018.

During their technical review of the application, the NRC Staff identified the need for additional information. Reference 2 provided the Requests for Additional Information (RAls) from the NRC Reactor Systems Branch (SRXB). The attachment to this letter provides the responses to the RAls.

Exelon has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the U.S. Nuclear Regulatory Commission in Reference 1. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Further, the additional information provided in this

SRVOOS LAR Supplement 1 Response to Request for Additional Information August 29, 2017 Page2 submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania and the State of Maryland of this response by transmitting a copy of this letter to the designated State Officials.

There are no regulatory commitments contained in this letter.

Should you have any questions concerning this letter, please contact Mr. David Neff at (610) 765-5631.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of August 2017.

Respectfully,

~~~

Jam~arstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Response to Requests for Information from NRC Review Branch SRXB cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Pennsylvania Bureau of Radiation Protection S. T. Gray, State of Maryland

Attachment Peach Bottom Atomic Power Station, Unit 2 NRC Docket No. 50-277 Response to Request for Additional Information from NRC Review Branch SRXB

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 1of8 Responses to NRC Staff's Request for Additional Information By application dated May 19, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17139D357), Exelon Generation Company, LLC (Exelon) submitted a license amendment request (LAR) for Peach Bottom Atomic Power Station (PBAPS), Unit 2.

The amendment would revise the Technical Specifications (TSs) to decrease the number of safety relief valves (SRVs) and safety valves (SVs}, required to be operable, when operating at a power level less than or equal to 3358 megawatts thermal (MWt). This change would be in effect for the current PBAPS Unit 2 operating cycle (Cycle 22) that is scheduled to end in October 2018.

In an email dated August 10, 2017, from the NRC (Rick Ennis) to Exelon (David Neff) (ADAMS Accession No. ML17222A096}, the NRC provided Requests for Additional Information (RAls) seeking clarification of certain issues. In a phone call on August 10, 2017, Mr. Neff stated that Exelon would provide a response to the RAI questions by August 31, 2017.

SRXB-RAl-1: ASME Overpressure Analysis for New Operating Condition Overpressure protection for the reactor coolant pressure boundary (RCPB) during power operation is provided by SRVs and SVs and the reactor protection system. The NRC's acceptance criteria are based on: (1) draft GDC-9 1 , insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime; and (2) final GDC-31, insofar as it requires that the RCPB be designed with sufficient margin to assure that it behaves in a non-brittle manner and that the probability of rapidly propagating fracture is minimized.

The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at reduced reactor power level (two SRVs out*of-service (2SRVOOS), reactor powers 3358 MWt). In order to facilitate the NRC staff in performing an independent confirmatory calculation (simulating the limiting ASME overpressure event of all main steam isolation valve closures with scram on high neutron flux (MSIVF)), please provide the following information based on the current PBAPS Unit 2 reload licensing analysis (Cycle

22) and plant operating history:

1. Justification that the limiting ASME overpressure event is MSIVF for the new operating condition;

2. The main steam isolation valves' flow area (fraction of full open flow area) as a function of time during MSIVF event (LAR Figure 1);

1 As discussed in Appendix H to the PBAPS Updated Final Safety Analysis Report (UFSAR), during the construction/licensing process, PBAPS, Units 2 and 3, were evaluated against the then-current Atomic Energy Commission (AEC) draft of the 27 General Design Criteria (GDC) issued in November 1965. On July 11, 1967, the AEC published for public comment, in the Federal Register (32 FR 10213), a revised and expanded set of 70 draft GDC (hereinafter referred to as the "draft GDC"). Appendix H of the PBAPS UFSAR contains an evaluation of the design basis of PBAPS, Units 2 and 3, against the draft GDC. The licensee concluded that PBAPS, Units 2 and 3, conforms to the intent of the draft GDC. The licensees for PBAPS, Units 2 and 3, have made changes to the facility over the life of the plant that may have invoked the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the UFSAR and in other plant-specific design and licensing basis documentation.

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 2 of 8

3. The control rod insertion as a function of time during MSIVF event (LAR Figure 1); and

4. The test results for the last two refueling cycles' on SRVs/SVs lift setpoint tolerance.

RESPONSE

1. Two potentially limiting overpressure protection events are typically analyzed for BWRs: (1)

Main Steam Isolation Valve Closure with Scram on High Flux (MSIVF), and (2) Turbine Trip with Bypass Failure and Scram on High Flux. However, based on both plant initial core analyses and subsequent power uprate evaluations, the MSIVF is more limiting than the turbine trip (TI) event with respect to reactor overpressure. Closure of the MS IVs, which are closer to the reactor vessel than the turbine stop valves, reduces the amount of steam volume available to accommodate the pressurization (i.e., less steam line volume) compared to closure of the turbine stop valves and results in a higher pressurization rate and magnitude. PBAPS Extended Power Uprate (EPU) evaluations show a 40 to 50 psi difference between these two events. Operation at a reduced power level at PBAPS Unit 2 does not change the relative behavior/results of these two events and the MSIVF event will remain limiting for the new operating condition.

2. The Main Steam Isolation Valve (MSIV) flow area (fraction of full open flow) as a function of time assumed for the analysis of the MSIVF event is provided in the following table.

Closure Flow Area Time (sec) (fraction) 0.0 1.0 0.6 1.0 1.7 0.01 3.0 0.0

>3.0 0.0

3. The control rod (CR) insertion time assumed for the MSIVF event analysis is specified in PBAPS TS Table 3.1.4-1 (as a function of control rod notch position) and is provided in the table below (as a function of% insertion).

Scram Time CR Insertion (sec) (%)

0.0 0 0.2 0 0.49 5 0.9 20 2.0 50 3.5 90 3.875 100

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 3 of 8

4. The test results for the last two refueling cycles' on SRVs/SVs lift setpoint tolerance are provided in the tables below. The SRV/SV lift setpoints and setpoint tolerances are defined in TS Section 3.4.3, "Safety Relief Valves (SRVs) and Safety Valves (SVs)."

Fall 2016 PBAPS Unit 2 Refueling Outage Valve S/N Pressure As-Found Pressure Pressure Lift Result Set point Lift (psig) Setpoint Setpoint (PASS I (psig) Tolerance Deviation (psig FAIL)

(psig I%) I%)

20 1135 1135 +/- 34.1/+/-3% 010 PASS 21 1145 1124 +/- 34.4 / +/- 3% -21/-1 .8 PASS 22 1155 1150 +/-34.7/+/-3% -5 / -0.4 PASS 25 1155 1140 +/-34.7/+/-3% -15/-1.3 PASS SV14005 1260 1255 +/- 37.8 / +/- 3% -5 / -0.4 PASS Fall 2014 PBAPS Unit 2 Refueling Outage Valve S/N Pressure As-Found Pressure Pressure Lift Result Set point Lift Pressure Setpoint Setpoint (PASS I (psig) (psig) Tolerance Deviation (psig FAIL)

(psig I%) I%)

193 1135 1122 +/- 34.1/+/-3% -13/-1.1 PASS 83 1135 1141 +/- 34.1/+/-3% +61 +0.5 PASS 73 1145 1123 +/- 34.4 / +/- 3% -22/-1.9 PASS 88 1155 1137 +/-34.7/+/-3% -18/-1.6 PASS BL1095 1260 1212 +/- 37.8 / +/- 3% -48 / -3.8 FAIL BL 1104 1260 1272 +/-37.8 / +/-3% +12 / +1 .0 PASS SRXB-RAl-2: ATWS Overpressure Analysis for New Operating Condition An anticipated transient without scram (ATWS) is defined as an AOO followed by the failure of the reactor portion of the protection system specified in draft GDCs 14 and 15. Requirements related to ATWS events are specified in 10 CFR 50.62. The NRC staff review includes confirming that the peak reactor vessel bottom pressure will be less than the ASME Service Level C limit of 1500 psig during an ATWS overpressure event as protected by the SRVs/SVs and systems required in accordance with 10 CFR 50.62 (e.g., alternate rod injection system, standby liquid control system) .

The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at reduced reactor power level. In order to facilitate the NRC staff in performing an independent confirmatory calculation (simulating the limiting ATWS overpressure event), please provide the following information:

1. What is the limiting ATWS overpressure event for the proposed new operating condition (2SRVOOS, reactor powers 3358 MWt)? Please provide justification.

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 4 of 8

2. Clarify the differences among the following calculated peak reactor (vessel bottom) pressures:

(a) 1430 psig value discussed on pages 5 and 6 of Attachment 1 of the LAR; (b) 1419 psig value discussed on page 4 of Attachment 1 of the LAR and in Section 9.3.1.1 of NEDC-33720P 2 ; and (c) Value specified in the first sentence of Note 2 in Table 9-3 of NEDC-33720P.

Please confirm which one of the above is the licensing basis for current plant operation (i.e., 1SRVOOS, reactor powers 3951 MWt)?

3. The main steam isolation valves' flow area (fraction of full open flow area) as a function of time during MSIVC.

4. The sequence of events for the current ATWS overpressure licensing analysis (1SRVOOS, reactor powers 3951 MWt).

RESPONSE

1. The ATWS overpressure analysis is performed for two events - Main Steam Isolation Valve Closure (MSIVC) and Pressure Regulator Failure Open (PRFO). The PRFO event results in an MSIV closure after an initial reactor depressurization and the results are similar to the MSIVC event. The MSIVC event results in a higher peak reactor pressure by a small amount (12 psi). The Turbine Trip with Bypass Failure and Failure to Scram is not analyzed because, as noted in the response to SRXB-RAl-1 above, closure of the MSIVs, which are closer to the reactor vessel than the turbine stop valves, reduces the amount of steam volume available to accommodate the pressurization (i.e., less steam line volume) compared to closure of the turbine stop valves and results in a higher pressurization rate and magnitude. Operation at a reduced power level does not change the relative behavior/results of these two events and the MSIVC event will remain limiting. Therefore, the limiting ATWS overpressure event for the proposed new operating condition will remain the same (MSIVC).

2. The peak reactor pressure of 1430 psig value discussed on pages 5 and 6 of Attachment 1 of the LAR is the calculated result for the PBAPS Measurement Uncertainty Recapture (MUR) uprated condition (Reference 1) currently under NRC review. The peak reactor 2

GEH report NEDC-33720P, Revision 0, "Safety Analysis Report for Peach Bottom Atomic Power Station Units 2 & 3 Maximum Extended Load Line Limit Analysis Plus." was submitted to the NRC as Attachment 4 to the licensee's MELLLA+ application dated September 4, 2014 (ADAMS Accession No. ML14247A503). Attachment 4 is a proprietary (i.e., non-public) version of the report. A non-proprietary (i.e., public) version of the report is contained in to the MELLLA+ application.

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 5 of 8 pressure of 1419 psig value discussed on page 4 of Attachment 1 of the LAR and in Section 9.3.1 .1 of NEDC-33720P is the current analysis of record (licensing basis) result for the PBAPS CLTP conditions (EPU/MELLLA+).

Note 2 in Table 9-3 of NEDC-33720P refers to previous ATWS analyses performed to support implementation of PBAPS EPU prior to MELLLA+. The EPU ATWS overpressure analysis was performed utilizing the NRC approved ODYN code resulting in a peak reactor vessel pressure of 1461 psig. The ATWS overpressure analysis reported in NEDC-33720P was performed utilizing the NRC approved TRACG code. Note 2 in Table 9-3 of NEDC-33720P provides additional PBAPS EPU ATWS overpressure results utilizing the TRACG code to allow a direct comparison to the MELLLA+ results.

3. The MSIV flow area (fraction of full open flow) as a function of time assumed for the analysis of the MSIVC A TWS event using nominal values is provided in the following table. Since the ATWS event is an Anticipated Operational Occurrence (AOO), nominal initial conditions are used consistent with the NRC approved TRACG methodology.

Closure Flow Area Time (sec) (fraction) 0.0 1.0 0.8 1.0 3.8 0.01 4.0 0.0

>4.0 0.0 4 . The sequence of events for the current ATWS overpressure licensing analysis (MSIVC) is identified in the table below.

Seq. No. Event Response Time (sec) 1 MSIV Isolation Initiated 0.0 2 MSIVs Fully Closed 4.0 3 Recirculation Pump Trip 4.3 4 Qpenini:i of First Relief Valve 4.5 5 Peak Vessel Pressure 11.9 SRXB-RAl-3: Thermal Limits Assessment for New Operating Condition Draft GDCs 6, 14, 15, and 29 provide requirements related to core design and protection systems in order to assure that acceptable fuel damage limits are not exceeded. The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at reduced reactor power level (2SRVOOS, reactor powers 3358 MWt). To facilitate the staff's review, provide the following information based on the current PBAPS Unit 2 reload licensing analysis (Cycle 22):

1. What is the limiting AOO transient to determine the operating limit minimum critical power ratio (MCPR) multiplier, Kp, and power dependent linear heat generation rate (LHGR) multiplier (LHGRFAC(P)) for powers 85% of rated (reference Tables 4-2 and 5-3 of the PBAPS Unit 2, Cycle 22, core operating limits report (ADAMS Accession No. ML16327A068))?

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 6of8

2. What is the timing of the minimum MCPR and maximum LHGR and are there any SRVs required to lift prior to minimum MCPR and maximum LHGR for the limiting AOO transient identified above?

RESPONSE

1. The operating limit minimum critical power ratio (MCPR) multiplier, Kp, and power dependent linear heat generation rate (LHGR) multiplier (LHGRFAC(P)) for powers 85% of rated are determined by evaluating the limiting AOO transients (e.g., Load Rejection No Bypass [LRNBP], Feedwater Controller Failure (FWCF)) for each Equipment Out-of-Service (EOOS) combination. No one particular AOO transient establishes the Kp or LHGRFAC(P) values for all combinations or for all power levels within a combination. For instance, in the Base combination of PB2C22 Core Operating Limits Report (COLR) Table 4-2, the LRNBP is limiting at some power levels, but the FWCF is limiting at other power levels. In the Turbine Bypass System Out-of-Service (TBSOOS) combination, the FWCF is typically limiting for most, but not all, power levels. For the COLR Table 5-3, the FWCF is typically, but not always, limiting for LHGRFAC(P). The Kp and LHGRFAC(P) multipliers are developed in a manner that insures all AOO transients are bounded since no one AOO transient event is limiting at all off-rated conditions.

2. The most limiting MCPR and maximum LHGR values during pressurization events are calculated to occur prior to the opening of any SRVs or SVs. The limiting MCPR occurs approximately 0. 7-0.8 seconds prior to SRV opening and the maximum LHGR occurs approximately 1.1 seconds prior to SRV opening. Therefore, the most limiting MCPR and maximum LHGR values are not impacted by SRVs or SVs that may be out-of-service.

SRXB-RAl-4: Assessments of ECCS LOCA Performance and High Pressure System Performance for New Operating Condition The NRC's acceptance criteria related to emergency core cooling system (ECCS) and loss-of-coolant accident (LOCA) performance is based, in part, on: (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) draft GDCs 40 and 42, insofar as they require that protection be provided for engineered safety features against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; and (3) draft GDCs 37, 41, and 44, insofar as they require that a system to provide abundant emergency core cooling be provided so that fuel and clad damage that would interfere with the emergency core cooling function will be prevented.

The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at a reduced reactor power level (2SRVOOS, reactor power s 3358 MWt). It is stated on page 8 of Attachment 1 to the LAR that the ECCS/LOCA performance and high pressure performance systems have been assessed for the proposed new operating condition. However, there are no further details provided regarding what the assessment involved and how the assessment had been performed. Please provide these details.

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 7 of 8

RESPONSE

ECCS/LOCA Performance The impact of SRV/SV setpoint tolerance relaxation and SRV/SV OOS on the ECCS/LOCA performance was addressed in Reference 2. Reference 2 included evaluations for Containment Pressure and Temperature for Design Basis Accident (DBA) Loss of Coolant Accident (LOCA),

Small Steam Line Breaks, Intermediate and Small Line Break Accidents, Suppression Pool Temperature, and DBA LOCA Hydrodynamic Loads. Those evaluations concluded that the ECCS/LOCA results were not affected by SRV/SV setpoint tolerance relaxation and/or an SRV/SV OOS. A review of the evaluation bases and conclusions in Reference 2 determined that the conclusions would not change for an additional SRV/SV OOS. In addition, as noted in Reference 2, PBAPS has been previously analyzed and licensed to operate with 2 SRV/SV OOS.

Reference 2 addressed the impact from operation with an SRVOOS on SRV dynamic loads.

The report states that having an SRVOOS does not have an effect on SRV dynamic loads.

SRV loads are driven by SRV opening pressure and the SRV discharge line water level at the time of the second SRV actuation. The SRV discharge line water level is in turn a function of the time between the closure of the valves at the end of the first actuation and the time of the second actuation. Having an SRVOOS will not affect the operable valves' setpoints or the time between the initial and second valve actuations. As such, having an SRVOOS has no effect on SRV dynamic loads. Safety Valves (SV) have no impact on suppression pool dynamic loads.

These conclusions do not change for an additional SRV/SV OOS.

Therefore, it is concluded that operation with an additional (up to 2) SRV/SV OOS would have no adverse effect on the ECCS/LOCA performance, including SRV dynamic loads.

High Pressure System Performance The high pressure systems include the High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC) and Standby Liquid Control (SLC) systems. The Reference 2 report states that the most potentially significant effect of an SRV/SV OOS and/or relaxing the SRV/SV setpoint tolerance on the HPCI and RCIC systems' operations is the maximum reactor pressure at which they are required to deliver water to the reactor. Both systems are required to provide injection into the reactor pressure vessel at the pressure corresponding to the lowest group of SRV lift setpoints (including drift) - PBAPS includes four SRVs in the lowest setpoint group.

The condition of an SRV/SV OOS does not change the SRV lift setpoints. This statement applies equally to either one or two SRV/SV OOS. Therefore, it is concluded that the maximum reactor pressure at which the HPCI and RCIC systems are required to deliver water to the reactor would not be adversely impacted by operation with an additional (up to two) SRV/SV OOS. In addition, as noted in Reference 2, PBAPS has been previously analyzed and licensed to operate with two SRV/SV OOS.

The limiting ATWS case for long term results, including peak reactor lower plenum pressure during SLC system injection, has been evaluated for PBAPS CLTP conditions and the more limiting MUR condition (Reference 1) currently under NRC review. For the MUR uprated condition, at the time of SLC system injection at 124 seconds for the limiting MSIVC case, the reactor power is in the range of approximately 25% power. This power level is well within the available steam relief system capacity, even with an additional (up to two) SRV/SV OOS. In

SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 8 of 8 addition, the MUR analysis determines that there is significant SLC system relief valve margin of 183 psi. This margin is significantly greater than the 70 psi minimum requirement. Therefore, it is concluded that peak reactor lower plenum pressure during SLC system injection would not increase significantly and the SLC system performance would not be adversely impacted by operation with an additional (up to two) SRV/SV OOS.

References

1. Exelon Letter to NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated February 17, 2017 (ADAMS Accession No. ML17048A444)

2. "PBAPS Units 2 and 3 Safety Valve Setpoint Tolerance Increase Safety Analysis Report,"

NEDC-33533P, Rev. 1, dated May 2013 (ADAMS Accession No. ML13162A619)