ML17222A096

Request for Additional Information - Amendment Request Regarding Safety Relief Valve and Safety Valce Operability for Cycle 22

ML17222A096
Person / Time
Site:	Peach Bottom
Issue date:	08/10/2017
From:	Richard Ennis Plant Licensing Branch 1
To:	Deborah Neff Exelon Generation Co
Ennis R
References
CAC MF9705
Download: ML17222A096 (4)

Text

From: Ennis, Rick Sent: Thursday, August 10, 2017 10:42 AM To: David Neff Cc: David Helker; Richard.Gropp@exeloncorp.com

Subject:

Peach Bottom Unit 2 - Request for Additional Information -

Amendment Request Regarding Safety Relief Valve and Safety Valve Operability for Cycle 22 (CAC MF9705)

Attachments: final01 rai m9705.doc By application dated May 19, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17139D357), Exelon Generation Company, LLC (Exelon, the licensee) 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 Cycle 22 that is scheduled to end in October 2018.

The Nuclear Regulatory Commissions (NRC) staff is reviewing your submittal and has determined that additional information is needed to complete its review. The specific request for additional information (RAI) questions are enclosed. The RAI questions were provided in draft form to Mr. David Neff of the Exelon staff via e-mail on August 3, 2017. The draft questions were sent to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed.

In a phone call on August 10, 2017, Mr. Neff said a clarification call was not needed to discuss the draft RAI questions. Mr. Neff stated that Exelon would provide a response to the RAI questions by August 31, 2017.

If you have any questions, please contact me at (301) 415-1420.

Richard B. Ennis, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-277

REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT SAFETY RELIEF VALVE AND SAFETY VALVE OPERABILITY FOR CYCLE 22 EXELON GENERATION COMPANY, LLC PEACH BOTTOM ATOMIC POWER STATION - UNIT 2 DOCKET NO. 50-277 By application dated May 19, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17139D357), Exelon Generation Company, LLC (Exelon, the licensee) 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 Cycle 22 that is scheduled to end in October 2018.

The Nuclear Regulatory Commissions (NRC) staff is reviewing your submittal and has determined that additional information is needed to complete its review. The specific request for additional information (RAI) questions are shown below. The RAI questions were provided in draft form to Mr. David Neff of the Exelon staff via e-mail on August 3, 2017. The draft questions were sent to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed.

In a phone call on August 10, 2017, Mr. Neff said a clarification call was not needed to discuss the draft RAI questions. Mr. Neff stated that Exelon would provide a response to the RAI questions by August 31, 2017.

Reactor Systems Branch (SRXB)

Reviewer: Shie-Jeng Peng SRXB-RAI-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 NRCs acceptance criteria are based on: (1) draft GDC-91, 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.

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.

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 power 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);

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.

SRXB-RAI-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 power 3358 MWt)? Please provide justification.

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-33720P2; and 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 licensees 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.

(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 power 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 power 3951 MWt).

SRXB-RAI-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 power 3358 MWt). To facilitate the staffs 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 power 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))?

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?

SRXB-RAI-4: Assessments of ECCS LOCA Performance and High Pressure System Performance for New Operating Condition The NRCs 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 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.