ML17328A163

Peach Bottom Atomic Power Station, Units 2 and 3 - Report for the Audit of Licensee Responses to Interim Staff Evaluations Open Items Related to NRC Order EA-13-109

ML17328A163
Person / Time
Site:	Peach Bottom
Issue date:	11/30/2017
From:	Auluck R C Beyond-Design-Basis Engineering Branch
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Lee B E
References
CAC MF4416, CAC MF4417, EA-13-109, EPID L-2014-JLD-0053
Download: ML17328A163 (31)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 30, 2017 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUBJECT: PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 -REPORT FOR THE AUDIT OF LICENSEE RESPONSES TO INTERIM STAFF EVALUATIONS OPEN ITEMS RELATED TO NRC ORDER EA-13-109 TO MODIFY LICENSES WITH REGARD TO RELIABLE HARDENED CONTAINMENT VENTS CAPABLE OF OPERATION UNDER SEVERE ACCIDENT CONDITIONS (CAC NOS. MF4416 AND MF4417; EPID L-2014-JLD-0053) Dear Mr. Hanson: On June 6, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13143A334), the U.S. Nuclear Regulatory Commission (NRC) issued Order EA-13-109, "Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Condition," to all Boiling Water Reactor licensees with Mark I and Mark II primary containments. The order requirements are provided in Attachment 2 to the order and are divided into two parts to allow for a phased approach to implementation. The order required licensees to submit for review overall integrated plans (OIPs) that describe how compliance with the requirements for both phases of Order EA-13-109 will be achieved. By letter dated June 30, 2014 (ADAMS Accession No. ML 14181A301), Exelon Generation Company, LLC. (the licensee) submitted its Phase 1 OIP for Peach Bottom Atomic Power Station, Unit 2 and Unit 3 (Peach Bottom). By letters dated December 19, 2014, June 30, 2015, December 15, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 30, 2016, December 15, 2016, and June 30, 2017 (ADAMS Accession Nos. ML 14353A125, ML 15181A018, ML 15364A015, ML 16182A012, ML 16350A265, and ML 17181A034, respectively), the licensee submitted its 6-month updates to the OIP. The NRC staff reviewed the information provided by the licensee and issued interim staff evaluations (ISEs) for Phase 1 and Phase 2 of Order EA-13-109 for Peach Bottom by letters dated February 12, 2015 (ADAMS Accession No. ML 15026A469), and August 2, 2016 (ADAMS Accession No. ML 16099A272), respectively. When developing the ISEs, the staff identified open items where the staff needed additional information to determine whether the licensee's plans would adequately meet the requirements of Order EA-13-109. The NRC staff is using the audit process described in letters dated May 27, 2014 (ADAMS Accession No. ML 14126A545), and August 10, 2017 (ADAMS Accession No. ML 17220A328), to gain a better understanding of licensee activities as they come into compliance with the order. As part of the audit process, the staff reviewed the licensee's closeout of the ISE open items.

B. Hanson -2 -The NRC staff conducted teleconferences with the licensee on May 18, 2017, and November 2, 2017, respectively. The enclosed audit report provides a summary of that aspect of the audit. If you have any questions, please contact me at 301-415-1025 or by electronic mail at Rajender.Auluck@nrc.gov. Docket Nos. 50-277 and 50-278 Enclosure: Audit report cc w/encl: Distribution via Listserv Sincerely, Rajender Auluck, Senior Project Manager Beyond-Design-Basis Engineering Branch Division of Licensing Projects Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AUDIT REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION AUDIT OF LICENSEE RESPONSES TO INTERIM STAFF EVALUATIONS OPEN ITEMS RELATED TO ORDER EA-13-109 MODIFYING LICENSES WITH REGARD TO RELIABLE HARDENED CONTAINMENT VENTS CAPABLE OF OPERATION UNDER SEVERE ACCIDENT CONDITIONS BACKGROUND EXELON GENERATION COMPANY, LLC. PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-277 AND 50-278 On June 6, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13143A334), the U.S. Nuclear Regulatory Commission (NRC) issued Order EA-13-109, "Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Condition," to all Boiling Water Reactor (BWR) licensees with Mark I and Mark II primary containments. The order requirements are divided into two parts to allow for a phased approach to implementation. Phase 1 of Order EA-13-109 requires license holders of BWRs with Mark I and Mark II primary containments to design and install a Hardened Containment Vent System (HCVS), using a vent path from the containment wetwell to remove decay heat, vent the containment atmosphere (including steam, hydrogen, carbon monoxide, non-condensable gases, aerosols, and fission products), and control containment pressure within acceptable limits. The HCVS shall be designed for those accident conditions (before and after core damage) for which containment venting is relied upon to reduce the probability of containment failure, including accident sequences that result in the loss of active containment heat removal capability or extended loss of alternating current power (ELAP). The order required all applicable licensees, by June 30, 2014, to submit to the Commission for review an overall integrated plan (OIP) that describes how compliance with the Phase 1 requirements described in Order EA-13-109 Attachment 2 will be achieved. Phase 2 of Order EA-13-109 requires license holders of BWRs with Mark I and Mark II primary containments to design and install a system that provides venting capability from the containment drywell under severe accident conditions, or, alternatively, to develop and implement a reliable containment venting strategy that makes it unlikely that a licensee would need to vent from the containment drywell during severe accident conditions. The order required all applicable licensees, by December 31, 2015, to submit to the Commission for Enclosure

-2 -review an OIP that describes how compliance with the Phase 2 requirements described in Order EA-13-109 Attachment 2 will be achieved. By letter dated June 30, 2014 (ADAMS Accession No. ML 14181A301), Exelon Generation Company, LLC. (Exelon, the licensee) submitted its Phase 1 OIP for Peach Bottom Atomic Power Station, Unit 2 and Unit 3 (Peach Bottom). By letters dated December 19, 2014, June 30, 2015, December 15, 2015 (which included the combined Phase 1 and Phase 2 OIP), June 30, 2016, December 15, 2016, and June 30, 2017 (ADAMS Accession Nos. ML 14353A125, ML 15181A018, ML 15364A015, ML 16182A012, ML 16350A265, and ML 17181A034, respectively), the licensee submitted its 6-month updates to the OIP, as required by the order. The staff reviewed the information provided by the licensee and issued interim staff evaluations (ISEs) for Phase 1 and Phase 2 for Peach Bottom by letters dated February 12, 2015 (ADAMS Accession No. ML 15026A469), and August 2, 2016 (ADAMS Accession No. ML 16099A272), respectively. When developing the ISEs, the staff identified open items where the staff needed additional information to determine whether the licensee's plans would adequately meet the requirements of Order EA-13-109. The NRC staff is using the audit process in accordance with the letters dated May 27, 2014 (ADAMS Accession No. ML 14126A545), and August 10, 2017 (ADAMS Accession No. ML 17220A328), to gain a better understanding of licensee activities as they come into compliance with the order. The staff reviews submitted information, licensee documents (via ePortals), and preliminary Overall Program Documents (OPDs)/OIPs, while identifying areas where additional information is needed. As part of this process, the staff reviewed the licensee closeout of the ISE open items. AUDIT SUMMARY As part of the audit, the NRC staff conducted teleconferences with the licensee on May 18, 2017 and Novemeber 2, 2017, respectively. The purpose of the audit teleconferences was to continue the audit review and provide the NRC staff the opportunity to engage with the licensee regarding the closure of open items from the ISEs. As part of the preparation for these audit calls, the staff reviewed the information and/or references noted in the OIP updates to ensure that closure of ISE open items and the HCVS design are consistent with the guidance provided in Nuclear Energy Institute (NEI) 13-02, Rev. 1 and related documents (e.g. white papers (ADAMS Accession Nos. ML 14126A374, ML 14358A040, ML 15040A038 and ML 15240A072, respectively) and frequently asked questions (FAQs), (ADAMS Accession No. ML 15271A148)) that were developed and reviewed as part of overall guidance development. The NRC staff audit members are listed in Table 1. Table 2 is a list of documents reviewed by the staff. Table 3 provides the status of the ISE open item closeout for Peach Bottom. The open items are taken from the Phase 1 and Phase 2 ISEs issued on February 12, 2015, and August 2, 2016, respectively. FOLLOW UP ACTIVITY The staff continues to audit the licensee's information as it becomes available. The staff will issue further audit reports for Peach Bottom, as appropriate. Following the licensee's declarations of order compliance, the licensee will provide a final integrated plan (FIP) that describes how the order requirements are met. The NRC staff will

-3 -evaluate the FIPs, the resulting site-specific OPDs, as appropriate, and other licensee documents, prior to making a safety determination regarding order compliance. CONCLUSION This audit report documents the staff's understanding of the licensee's closeout of the ISE open items, based on the documents discussed above. The staff notes that several of these documents are still preliminary, and all documents are subject to change in accordance with the licensee's design process. In summary, the staff has no further questions on how the licensee has addressed the ISE open items, based on the preliminary information. The status of the NRC staff's review of these open items may change if the licensee changes its plans as part of final implementation. Changes in the NRC staff review will be communicated in the ongoing audit process. Attachments: 1. Table 1 -NRC Staff Audit and Teleconference Participants 2. Table 2 -Audit Documents Reviewed 3. Table 3-ISE Open Item Status Table Table 1 -NRC Staff Audit and Teleconference Participants Title Team Member Organization Team Lead/Sr. Project Manager Rajender Auluck NRR/DLP Project Manager Support/Technical Support -Containment / Ventilation Brian Lee NRR/DLP Technical Support-Containment/ Ventilation Bruce Heida NRR/DLP Technical Support -Electrical Kerby Scales NRR/DLP Technical Support -Balance of Plant Kevin Roche NRR/DLP Technical Support -l&C Steve Wyman NRR/DLP Technical Support -Dose John Parillo NRR/DRA Attachment 1 Table 2 -Audit Documents Reviewed PM-0546, "Torus Hardened Vent-Flow Calculation" PB-MISC-025, "MAAP Analysis to Support HCVS Design: Containment Vent Cycling" ARA-002611002611 "Tornado Missile TORMIS Analysis" PM-1190, "HCVS Dose Assessment" PE-0301, "FLEX Electrical Loading and Voltage Drop" PE-0308, "HCVS Battery Sizing and Selection" PM-1207, "HCVS Phase II Dose Assessment" 2493544-28, "Temperature Profile for Areas During an ELAP" PM-1034, "Temperature Profile for Cable Spreading Room and Computer Room During an ELAP" PM-1174, "Temperature Profile for Refurl Floor Area During an ELAP" BWROG-TP-008, "Severe Accident Water Addition Timing" BWROG-TP-011, "Severe Accident Water Management Supporting Evaluations" PB-MISC-023, "MAAP Analysis to Support HCVS Design" Attachment 2 ISE Open Item Number Requested Action Phase 1 ISE 01 1 Make available for NRG staff audit guidelines and procedures for HCVS operation. Peach Bottom Atomic Power Station, Units 2 and 3 Vent Order Interim Staff Evaluation Open Items: Table 3 -ISE Open Item Status Table Licensee Response -Information NRG Staff Close-out notes provided in 6 month updates and on the ePortal The guidelines and procedures for HCVS The NRG staff reviewed the operation are complete and posted to the information provided in the 6-ePortal. month updates and on the ePortal. A list of the guidelines and procedures for HCVS operation is provided below: The guidelines and procedures for HCVS operation are complete -AO 56E.2-3, Rev O 13, E 134-T-B, E 134-and consistent with the guidance W-A, AND E-13A4-EC-A MOTOR in NEI 13-02. CONTROL CENTERS OR E134 AND E13A4 EMERGENCY No follow-up questions. -AO 56E.4-2, Rev 019, E324-R-B, E324-R-D, E324-D-A, E324-T-B OR E324-0-A MOTOR CONTROLCENTERS OR E324 EMERGENC -CC-PB-118, Rev 003, PEACH BOTTOM IMPLEMENTATION OF DIVERSE AND FLEXIBLE COPING STRATEGIES (FLEX) AND SP -FSG-010-2, Rev 001, ALIGNING FLEX GENERATOR TO PANEL 2AS1061 AND FOR FUEL OIL TRANSFER -FSG-010-3, Rev 001, ALIGNING FLEX GENERATOR TO PANEL 3AS1061 Safety Evaluation (SE) status Closed; Pending; Open (need additional information from licensee) Closed [Staff evaluation to be included in SE Section 5.1] Attachment 3

-2 --FSG-011-2, Rev 0000, ALIGNING FLEX GENERATOR TO PANEL 2BS1061 AND FOR FUEL OIL TRANSFER -FSG-011-3, Rev 001, ALIGNING FLEX GENERATOR TO PANEL 3BS1061 -FSG-013-2, Rev 0000, ELAP AC LOAD ALIGNMENT -FSG-013-3, Rev 001, ELAP AC LOAD ALIGNMENT -GP-2-2, Rev 008, NORMAL PLANT START-UP -OP-PB-108-103-2, Rev 007, LOCKED VALVE LIST-PBAPS UNIT 2 -PF-72H, Rev 007, RADWASTE BLDG RADWASTE COMPACTING AND STORAGE -ELEVATION 135'-0" -PF-75, Rev 006, RADWASTE BLDG RADWASTE CONTROL ROOM DEGON AND MEDICAL STATION ELEVATION 135'-0" -RT-M-O?B-950-2, Rev 003, TORUS HARDENED VENT RUPTURE DISC INTEGRITY TEST -RT-O-O?K-900-2, Rev 0000, CONTAINMENT EMERGENCY VENT SYSTEM BOTTLE PRESSURE VERIFICATION TEST -RT-0-57P-100-2, Rev 0000, CONTAINMENT EMERGENCY VENT

-3 -SYSTEM COMPONENT OPERATION TEST -RT-0-57P-745-2, Rev 0000, CONTAINMENT EMERGENCY VENT BATTERY MONTHLY INSPECTION -RT-0-57P-746-2, Rev 0000, CONTAINMENT EMERGENCY VENT BATTERY QUARTERLY INSPECTION RT-0-100-505-2, Rev 040, EMERGENCY OPERATING PROCEDURE TOOL INVENTORY -SE-4 BASES, Rev 031, FLOOD -BASES -SE-4 PROCEDURE, Rev 041, FLOOD -PROCEDURE -SE-11 BASES, Rev 022, LOSS OF OFF-SITE POWER -BASES -SO 7B.1 .A-2 COL, Rev 011, CONTAINMENT ATMOSPHERE INERTING -SO 7K.1.A-2 COL, Rev 0000, CONTAINMENT EMERGENCY VENT ARGON RUPTURE PURGE -SO 16C.1.A-2 COL, Rev 001, BACKUP INSTRUMENT NITROGEN TO HARDENED CONTAINMENT VENT SYSTEM -SO 57P.1 COL, Rev 0000, CONTAINMENT EMERGENCY VENT SYSTEM

-4 --SO 57P.1, Rev 001, HARDENED CONTAINMENT VENT SYSTEM BATTERY CHARGER STARTUP AND NORMAL OPERATIONS -SO 58C.1.A COL, Rev 047, NORMAL 120 VAC SYSTEM COMMON PLANT -ST/LLRT 20.078.11, Rev 003, TORUS 18 INCH VENT ISOLATION VALVES TO S.B.G.T. 0-RING SEALS OF A0-2-078-2511 -ST-M-16B-250-2, Rev 019, SAFETY GRADE INSTRUMENT GAS SYSTEM FUNCTIONAL AND INSERVICE TEST -T-200-2, Rev 014, PRIMARY CONTAINMENT VENTING -T-200J-2, Rev 004, CONTAINMENT VENTING VIA THE TORUS HARDENED VENT -TQ-PB-201-0113, Rev 008, PEACH BOTTOM TRAINING DEPARTMENT SIMULATOR EXAMINATION SECURITY Phase 1 ISE 01 2 The NRC staff reviewed Procedure CC-The NRC staff reviewed the Closed PB-118, which contains the HCVS out of information provided in the 6-Make available for NRC staff service requirements and is posted to the month updates and on the [Staff evaluation to be audit the site specific ePortal. ePortal. included in SE Section controlling document for HCVS 3.1.2.13] out of service and The out of service requirements are The site specific controlling compensatory measures. described in the licensee's 5th Six-Month document for HCVS out of service Status Report, dated December 15, 2016. and compensatory measures is consistent with the guidance in NEI 13-02. No follow-up questions.

-5 -Phase 1 ISE 01 3 The design of the HCVS includes a The NRC staff reviewed the Closed control switch in the MCR [main control information provided in the 6-Make available for NRC staff room] for transferring power to the month updates and on the [Staff evaluation to be audit a technical justification solenoid valve (SV) that does not include ePortal. included in SE Section for the use of jumpers in the a PCIS [primary containment isolation 3.1.2.6] HCVS strategy. system] isolation signal, eliminating the The licensee has changed its need for jumpers. strategy and the use of jumpers is no longer needed. Applicable Unit 2 drawings (M-1-S-23, Sheet 348 and E-403, Sheet 5) and Unit No follow-up questions. 2 ECR package 15-00148 have been uploaded to the ePortal. Phase 1 ISE 01 4 Calculation PM-0546, Torus Hardened The NRC staff reviewed the Closed Vent-Flow demonstrates that the HCVS information provided in the 6-Make available for NRC staff has the capacity to vent the steam/energy month updates and on the [Staff evaluation to be audit analyses demonstrating equivalent of one percent licensed/rated ePortal. included in SE Section that HCVS has the capacity to thermal power. 3.1.2.1] vent the steam/energy Calculation PM-0546 assumes equivalent of one percent of The primary containment design pressure the rated thermal power of both licensed/rated thermal power is 56 psig [per square inch gauge] Units 2 and 3 is 3951 MWt (unless a lower value is (UFSAR [updated final safety analysis [megawatt thermal]. The justified), and that the report] [Section] 5.2.3.1 ). The primary calculation shows a choked flow suppression pool and the containment pressure limit is 60 psig of 74,300 lbm/sec. The maximum HCVS together are able to (UFSAR [Section] 5.2.3.6). PM-0546 required flow for 1 % power is absorb and reject decay heat, shows that the HCVS capacity exceeds 40,700 lbm/sec. Calculation PM-such that following a reactor one percent of licensed/rated thermal 0546 results show that at shutdown from full power power at the lower of these values. containment design pressure of containment pressure is Calculation PM-0546 has been uploaded 56 psig, the vent capacity has a restored and then maintained to the ePortal. 26% margin for Unit 2 and a 30% below the primary containment margin for Unit 3. design pressure and the Additionally, MAAP [modular accident primary containment pressure analysis program] calculation PB-MISC-Calculation PB-MISC-025 limit. 025 was developed using MAAP 4.0.6 to confirmed that the size of the investigate the response of Peach Bottom wetwell portion of the HCVS Unit 2 and Unit 3 containment venting provides adequate capacity to using the HCVS vent parameters and the meet or exceed the order criteria. use of RPV [reactor pressure vessel] alternate injection with assumed No follow-up questions. immediate RCIC failure. The MAAP analysis demonstrates that the

-6 -suppression pool and the HCVS together are able to absorb and reject decay heat, such that following a reactor shutdown from full power containment pressure is restored and then maintained below the primary containment design pressure and the primary containment pressure limit. Calculation PBMISC-025 has been uploaded to the ePortal. Phase 1 ISE 01 5 Reference drawing M-395: The Reactor The NRC staff reviewed the Closed Building Exhaust System. information provided in the 6-Make available for NRC staff month updates and on the [Staff evaluation to be audit descriptions or diagrams The RB Exhaust System Fans, including ePortal. included in SE Section of reactor building ventilation the Refuel Floor Exhaust Fans, RB 3.1.2.3] including exhaust dampers Exhaust Fans, and RB Equipment No follow-up questions. failure modes to support the Exhaust Fans, have Fail-Close dampers licensee justification for the in exhaust ducts to prevent uncontrolled HVAC release point being or unmonitored release from the RB in the below and 150 feet from the event of loss of power to the solenoid reactor building ventilation valves associated with the dampers. Fail-release point. Close dampers will eliminate pathway into the RB in the event of use of the HCVS in an ELAP. Drawing M-395 has been uploaded to the ePortal. Phase 1 ISE 01 6 Exelon prepared a Tornado Missile The NRC staff reviewed the Closed TORMIS analysis (ARA-002611002611 which information provided in the 6-Make available for NRC staff calculated probabilities of damage to the month updates and on the [Staff evaluation to be audit details to justify the external piping that would crimp the pipe ePortal. included in SE Section deviation from the tornado to a point of not being able to perform as 3.2.2] protection standards provided expected under severe accident The licensee's "reasonable in NEI 13-02 or make available conditions following an ELAP event. protection evaluation," which a description of how the HCVS calculated the damage will comply with the tornado The conclusions of the analysis are probabilities to the HCVS external protection standards provided provided in a table in the 5th Six-Month piping to a point of not being able in NEI 13-02. Status Report. Frequencies for the to perform as expected under Hardened Vent targets for each unit are severe accident conditions lower than the threshold frequency of following an ELAP event are less 1.0E-06 per year. than the numerical criterion stated in the NRC staff established

-7 -In addition, compensatory measures are position in the (TGRMIS) SER available in the event the HCVS external dated May 7, 1983. piping does become crimped. This analysis has been reviewed by the NRC The licensee also has established and determined to be acceptable based a "reasonable qualitative on the probability values provided in the argument" to indicate defense-in-analysis and the qualitative reasonable depth by having compensatory assurance argument provided. TORMIS measures available in the event analysis ARA-002611002611has been posted to the HCVS does become crimped. the ePortal. No follow-up questions. Phase 1 ISE 01 7 FLEX modification ECR 15-00126 The NRC staff reviewed the Closed improved the PBAPS communication information provided in the 6-Make available for NRC staff system to be functional in the event of an month updates and on the [Staff evaluation to be audit documentation that ELAP. This modification provided ePortal. included in SE Section demonstrates adequate equipment and connections to form a 3.1.1.1] communication between the workable system to provide in-plant The communication methods are remote HCVS operation communications during and after an the same as accepted in Order locations and HCVS decision ELAP/BDBEE [beyond-design-basis EA-12-049. makers during ELAP and external event]. The existing 3-way severe accident conditions. antenna coupler located in the northeast No follow-up questions. corner of the reactor building at El. 135'-O", was replaced with a 4-way coupler to accommodate a new connection installed from a new cabinet to power the existing plant antenna network. The new cabinet is located in the same area and was vendor supplied with all equipment as required. The cabinet houses an 8-hour rectifier power system and associated batteries, radio repeater that duplicates the frequency of the operations channel 1 (ops 1 ), a radio frequency duplexer to control the signals to and from the portable radio antenna and associated cabling that will be deployed after the event. The cabinet is powered from a division 1 power source which will be powered from the FLEX portable diesel during an ELAP. The permanently

-8 -installed equipment in the reactor building along with all conduits for the cabling are seismically mounted. This system is deployed using procedure FSG-020. Operators in the remote HCVS operating location (Radwaste Building El. 135') will utilize this system to communicate with the Main Control Room staff. ECR 15-00126 and procedure FSG-020 have been uoloaded to the ePortal. Phase 1 ISE 01 8 Peach Bottom will utilize an Argon purge The NRC staff reviewed the Closed system to address combustible gases in information provided in the 6-Provide a description of the the HCVS piping. month updates and on the [Staff evaluation to be final design of the HCVS to ePortal. included in SE Section address hydrogen detonation A summary of the design features is 3.1.2.11] and deflagration. included in the December 2015 OIP. The licensee's design is consistent with Option 3 of the endorsed white paper HCVS-WP-03. The Argon purge system will have switches for the control valve in the MCR and will be operable from the Remote Operating Station (ROS), in case of a direct current power or control circuit failure. The Argon purge will only be utilized following severe accident conditions when hydrogen is being vented. The installed capacity for the Argon purge system will be sized for at least 8 purges within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the ELAP. This number of vent cycles is the same value used for sizing the primary containment isolation valve (PCIV) pneumatic supply. The design will allow for Arqon bottle

-9 -replacement for continued operation past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The replacement bottles will be stored at the same location as the ROS. The Argon purge system can also be used to breach the rupture disc if venting is required before reaching the rupture disc set-point. No follow-up questions. Phase 1 ISE 01 9 Temperature and radiological conditions The NRC staff reviewed the Closed have been evaluated. The evaluated information provided in the 6-Make available for NRC staff conditions support safe access and month updates and on the [Staff evaluation to be audit an evaluation of operation of control and supporting ePortal. included in SE Sections temperature and radiological equipment by operating personnel. 3.1.1.2 and 3.1.1.3] conditions to ensure that ECR 15-00148 indicates that the operating personnel can safely Temperature conditions are evaluated in ROS (Room #241) will be a access and operate controls Section 3.14 of ECR 15-00148 and maximum of 120°F and 90% and support equipment. radiological conditions are evaluated in relative humidity during an Calculation PM-1190. Unit2 ECR 15-accident. The ECR notes that 00148 and Calculation PM-1190 have during the first 4-hours after the been posted to the ePortal. start of a severe accident access to the room will be unaffected due to the mass of concrete between the containment and the ROS. Operators will access the room for system alignment during that period. If ROS re-entry is required after the area heats up, operators will have ice vests and limited stay times per plant procedures. Access to other areas has been addressed under Order EA-12-049.

-10 -Temperature and radiological conditions should not inhibit operator actions needed to initiate and operate the HCVS during an ELAP with severe accident conditions. No follow-up questions. Phase 1 ISE 01 10 Descriptions of all instrumentation and The NRC staff reviewed the Closed controls necessary to implement this information provided in the 6-Make available for NRC staff order including qualification methods are month updates and on the [Staff evaluation to be audit descriptions of all provided in ECR 15-00148, Section 3.2, ePortal. included in SE Section instrumentation and controls 3.5, 3.14, 3.15 & 3.39. ECR 15-00148 3.1.2.8] (existing and planned) has been posted to the ePortal. ECR 15-00148, Sections 3.2, 3.5, necessary to implement this 3.14, 3.15 & 3.39 discusses the order including qualification environmental conditions during methods. an accident at the locations containing l&C components. The staff's review indicated that the environmental qualification met the order requirements. No follow-up Questions. Phase 1 ISE 01 11 Final sizing evaluation for HCVS The NRC staff reviewed the Closed battery/battery charger is contained in information provided in the 6-Make available for NRC staff Calculation PE-0308. month updates and on the [Staff evaluation to be audit the final sizing evaluation ePortal. included in SE Section for HCVS batteries/battery Incorporation of HCVS loading to FLEX 3.1.2.6] charger including incorporation DG loading is contained in Calculation The licensee stated that all into FLEX DG loading PE-0301. electrical power required for calculation. operation of HCVS components is Calculations PE-0308 and PE-0301 have provided by a dedicated HCVS been posted to the ePortal. battery charger and batteries. The battery sizing calculation (PE-0308) confirmed that the HCVS batteries have a minimum capacity capable of providing power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without

-11 -recharging, and therefore is adequate. The licensee revised the FLEX Diesel Generator calculation (PE-0301) to add the loads for the HCVS. The NRC staff reviewed the revised calculation and confirmed that with the additional load added, it's still within the capacity and capability of the Peach Bottom FLEX diesel generator. No follow-up questions. Phase 1 ISE 01 12 Descriptions of local conditions and The NRC staff reviewed the Closed confirmation that components required for information provided in the 6-Make available for NRC staff HCVS venting are capable of performing month updates and on the [Staff evaluation to be audit the descriptions of local their functions during ELAP and severe ePortal. included in SE Section conditions (temperature, accident conditions are contained in ECR 3.1.1.4] radiation and humidity) 15-00148, Sections 3.14 & 3.15. Unit2 ECR 15-00148, Sections 3.14 & anticipated during ELAP and ECR 15-00148 has been posted to the 3.15 discuss the environmental severe accident for the ePortal. conditions during an accident at components (valves, the locations containing l&C instrumentation, sensors, components. The staffs review transmitters, indicators, indicated that the environmental electronics, control devices, qualification met the order and etc.) required for HCVS requirements. venting including confirmation that the components are No follow-up questions. capable of performing their functions during ELAP and severe accident conditions. Phase 1 ISE 01 13 Documentation that the existing The NRC staff reviewed the Closed containment isolation valves, which are information provided in the 6-Make available for NRC staff relied upon for HCVS, will open under the month updates and on the [Staff evaluation to be audit documentation of an maximum expected differential pressure ePortal. included in SE Section evaluation verifying the during BDBEE and severe accident 3.1.2.1] existing containment isolation wetwell venting (60 psig) is provided in Data sheet DS-FP Sheet 1161A valves, relied upon for the indicates a maximum operating

-12 -HCVS, will open under the specification M-00117 and valve data differential pressure of 60 psid maximum expected differential sheet DS-FP, Sheet 1161. [per square inch differential] in the pressure during BDBEE and normal direction. The staff severe accident wetwell Specification M-00117 requires that PCIV reviwed Specification M-00117, venting. A0-2(3)-078-2(3)511 be able to operate Addendum 1 for Primary with a 62 psig pressure differential and Containment Isolation Butterfly valve data sheet DSFP, Sheet 1161 Valves for the Peach Bottom states that the design pressure of PCIV Atomic Power Station Units 2 and A0-2(3)-0?b-8(9)0290 is 150 psig. 3, Calculation ME-0167, Specification M-00117 and valve data "Determine the Maximum sheet DS-FP, Sheet 1161 have been Pressure Gradient Allowable posted to the ePortal. Across the Containment Venting Valves (All Acceptable Vent Paths) While Maintaining Valve Integrity and Operability", and MOD 5236 PBAPS Torus Hard Vent Design Input Document. Licensee addressed the ability of the HCVS PCIVs to operate under maximum design differential pressure. No follow-up questions. Phase 1 ISE 01 14 As described in the OIP, the HCVS torus The NRC staff reviewed the Closed vent path in each unit, starting at and information provided in the 6-Provide a description of the including the downstream PCIV, will be a month updates and on the [Staff evaluation to be strategies for hydrogen control dedicated HCVS flow path. There are no ePortal. included in SE Section that minimizes the potential for interconnected systems downstream of 3.1.2.12] hydrogen gas migration and the downstream, dedicated HCVS PCIV. The HCVS wetwell pipe in each ingress into the reactor Interconnected systems are upstream of each unit provides a dedicated building or other buildings. the downstream HCVS PCIV and are HCVS flowpath from the wetwell isolated by normally shut, fail shut PCIVs penetration PCIVs to the outside which, if open, would shut on an ELAP. with no interconnected There is no shared HCVS piping between downstream piping. The staff's the two units. review of the proposed system indicates that the licensee's design appears to maintain hydrogen below flammability limits.

-13 -No follow-up questions. Phase 1 ISE 01 15 The Hardened Containment Vent is The NRC staff reviewed the Closed isolated from the SGTS by valve A0-2-information provided in the 6-Make available for NRC audit 078-2512 (Unit 2) and A0-3-078-3512 month updates and on the [Staff evaluation to be documentation confirming that (Unit 3). The boot seal of A0-2(3)-078-ePortal. included in SE Section HCVS will remain isolated 2(3)512 will be maintained inflated from 3.1.2.1] from standby gas treatment the dedicated HCVS nitrogen supply to No follow-up questions. system during ELAP and ensure that A0-2(3)-078-2(3)512 remains severe accident conditions. closed/isolated. Unit 2 ECR 15-00148 has been posted to the ePortal. Phase 2 ISE 01 1 Eguii:1ment and Controls The NRC staff reviewed the Closed information provided in the 6-Licensee to evaluate the Plant instrumentation for SAWM that is month updates and on the [Staff evaluation to be SAWA [severe accident water qualified to RG 1.97 or equivalent is ePortal. included in SE Sections addition] equipment and considered qualified for the sustained 4.1.1.4 and 4.2.1.4] controls, as well as ingress operating period without further EC 618957 along with calculation and egress paths for the evaluation. The following plant PM-1207 shows that temperature expected severe accident instruments are qualified to RG 1.97: and radiological conditions should Conditions (temperature, not inhibit operator actions or humidity, radiation) for the PR/TR-4805 (Unit 2) SAWA equipment and controls sustained operating period. PR/TR-5805 (Unit 3) needed to initiate and operate the Ll-8123A (Unit 2) HCVS during an ELAP with Ll-9123A (Unit 3) severe accident conditions. Not all components that are either in the No follow-up questions. PR/TR-4(5)805 instrument loops or that may affect the loops are RG 1.97 qualified. These components include PT-4805, PT-5805, 2DC834 and 3DC834, and are evaluated in EC 618957, Attachment 9A for the radiological, thermal and humidity conditions expected during a severe accident. These components are determined to be either acceptable or have no adverse effect on the PR/TR-4805 and PR/TR-5805 instrument loops if failure occurred. Passive components that do not need to change state after initially establishing

-14 -SAWA flow do not require evaluation beyond the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time they are expected to be installed and ready for use to support SAWA/SAWM. The following additional equipment performing an active SAWNSAWM function is considered: -SAWA/SAWM flow instrument -SAWA/SAWM pump (FLEX pump) . -SAWA/SAWM generator (FLEX generator) -Active valves in SAWA flow path Ingress and Egress A quantitative evaluation of expected. dose rates has been performed per HCVS-WP-02 in Calculation PM-1207 and found the dose rates at deployment locations including ingress/egress paths are acceptable. Calculation PM-1207 has been posted to the ePortal. Unit 3 EC 618957 will be posted to the ePortal upon approval of the EC. Phase 2 ISE 01 2 Egui1;1ment and Controls The NRC staff reviewed the Closed information provided in the 6-Licensee to demonstrate how Plant instrumentation for SAWM that is month updates and on the [Staff evaluation to be instrumentation and equipment qualified to RG 1.97 or equivalent is ePortal. included in SE Sections being used for SAWA and considered qualified for the sustained 4.1.1.4 and 4.2.1.4] supporting equipment is operating period without further Technical Evaluation 2493544-28 capable to perform for the evaluation. The following plant along with calculation PM-1207 sustained operating period instruments are qualified to RG 1.97: shows that temperature and under the expected radiological conditions should not temperature and radiological PR/TR-4805 (Unit 2) inhibit operator actions or SAWA conditions PR/TR-5805 (Unit 3) equipment and controls needed Ll-8123A (Unit 2) to initiate and operate the HCVS

-15 -Ll-9123A (Unit 3) during an ELAP with severe accident conditions. See response to Phase 2 ISE-1 for discussion on components that are No follow-up questions. associated with the PR/TR-4(5)805 instrument loops that are not RG 1.97 qualified. Passive components that do not need to change state after initially establishing SAWA flow do not require evaluation beyond the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time they are expected to be installed and ready for use to support SAWA/SAWM. The following additional equipment performing an active SAWA/SAWM function is considered for temperature and radiation effects: -SAWA/SAWM flow instrument -SAWA/SAWM pump (FLEX pump) -SAWA/SAWM generator (FLEX generator) -Active valves in SAWA flow path Temperature The location of SAWA equipment and controls are the same or similar as FLEX with the exceptions noted below. The areas/rooms of PBAPS that require operator actions or passage to implement the SAWA/SAWM strategy are evaluated. The evaluation concluded that the expected thermal conditions associated with a severe accident are considered to be acceptable to accomplish the required SAWA/SAWM actions for HCVS Phase 2.

-16 -Details of the evaluation are contained in EC 618957, Attachment 98. The SAWA/SAWM action time line is the same as the FLEX action time line with the exception of performing procedure FSG-043 "Defeating RCIC Interlocks" and FSG-032 "Establishing HPCI/RCIC/Sump Room Ventilation, Lighting and Water Removal". These procedures are not required to be performed as part of the SAWA/SAWM strategy since the actions performed in these procedures are to support RCIC/HPCI operation and RCIC and HPCI are assumed to fail for SAWA/SAWM. Technical Evaluation 2493544-28 established a temperature profile for areas of PBAPS during an ELAP that require operator actions or passage to implement the FLEX strategy. This evaluation considered loss of ventilation (until FLEX generators are deployed) combined with decay heat from the reactor and various subsystems and heat from plant equipment fed by station batteries such as relays, emergency lighting and other electrical components credited in the FLEX strategy. Active plant calculations involving loss of HVAC were evaluated to identify calculations where area temperatures would be bounding when compared to those temperatures anticipated during an ELAP. The following rooms were evaluated for FLEX in Technical Evaluation 2493544-28; however, they are not required to be accessed as part of the SAWA/SAWM strategy.

RCIC Pump Room and HPCI Pump Room -17 -Since RCIC and HPCI are assumed to fail immediately for SAWA/SAWM, there is no requirement to access the RCIC or HPCI pump rooms; however, operators will likely open the RCIC and HPCI pump room doors per procedure SE-11 and determine if the RCIC or HPCI pumps can be started. The temperatures listed for these rooms in Technical Evaluation 2493544-28 would remain bounding for the relatively short time operators would spend determining if the RCIC or HPCI pumps can be started since the heat load from the RCIC or HPCI pump operation would not contribute to the overall heat load of the room. Reactor Building Sump Room and Core Spray Pump Room The reactor building sump room and core spray pump room are required to be accessed as part of the FLEX strategy per procedure FSG-032; however, these rooms are not required to be accessed as part of the SAWA/SAWM strategy since FSG-032 is not required to be performed as stated above. South Isolation Valve Room (SIVR) Procedure FSG-044 aligns the backup instrument nitrogen bottles to the ADS valves as part of the FLEX and SAWA/SAWM strategy. This is achieved by bypassing the A and B backup instrument nitrogen solenoid valves via a

-18 -local bypass line. The B backup instrument nitrogen solenoid valve is located in the SIVA and cannot be bypassed using the local bypass line since the Unit 3 SIVA is uninhabitable during a Severe Accident. Therefore, the Engineering Change associated with SAWNSAWM provides a new bypass line which is routed outside of the SIVA into the Control Rod Drive (CRD) Equipment Area which is evaluated below. The remaining areas identified in Technical Evaluation 2493544-28 are required to be accessed as part of the SAWNSAWM strategy. The temperatures listed for the Stairwells, Reactor Building Closed Cooling Water (RBCCW) Rooms, CAD Equipment Areas, Corridors, and Operating Area are associated with a Loss of Coolant Accident (LOCA). As stated in Technical Evaluation 2493544-28, these area temperatures are greater than those anticipated in an ELAP. These LOCA temperatures are considered to be acceptable for application to SAWNSAWM actions under a severe accident since these temperatures are conservative in nature under an ELAP, the SAWA/SAWM actions will be completed within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> into the event where containment heat up is restricted to core damage prior to vessel breach, and the drywell shield wall provides restrictive heat transfer which will mitigate the propagation of the higher drywell temperature. The temperatures listed for

-19 -the Cable Spreading Room and Computer Room are from calculation PM-1034; these area temperatures are reasonable and would not be expected to have any notable increase due to location/proximity to primary containment. The Refuel Floor Area temperature is determined by calculation PM-1174, Rev. 0 and this temperature is also not expected to increase since the inputs and assumptions of the calculation remain the same. Per HCVS-FAQ-01, no evaluation is required for use of the MCR as the primary control station. Actions to open Reactor Building Railroad Doors and Refuel Floor Roof Hatch to establish natural circulation of the Secondary Containment atmosphere and establish MCR, Battery Room and Switchgear Room ventilation per FSG-030, FSG-031 and FSG-033 are maintained as part of the SAWNSAWM strategy. Technical Evaluation 2493544-28 established area/room temperatures during an ELAP for PBAPS and determined that these temperatures are acceptable to implement FLEX actions. Based on the determination that the temperatures listed in Technical Evaluation 2493544-28 can be applied to SAWNSAWM actions and that these temperatures are acceptable to accomplish the required FLEX actions, the expected thermal conditions associated with SAWNSAWM are considered to be acceptable to accomplish the required SAWNSAWM actions for HCVS Phase 2.

-20 -Radiation For equipment locations within the Reactor Building, a quantitative evaluation of expected normal operation and containment shine (vent line shine) total integrated dose has been performed per HCVS-FAQ-12 (HCVS-WP-02) in Calculation PM-1207 and found the total integrated dose at deployment locations is acceptable. For locations outside the Reactor Building between 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 7 days when SAWA is being utilized, a quantitative evaluation of expected dose rates has been performed per HCVS-WP-02 in Calculation PM-1207 and found the dose rates at deployment locations are acceptable. Calculation PM-1207 and Procedures FSG-030, FSG-031, FSG-032, FSG-033 and FSG-043 have been posted to the ePortal. The revised Procedure FSG-044 will be posted to the ePortal upon approval of the procedure revision. Unit 3 EC 618957 will be posted to the ePortal upon approval of the EC. Phase 2 ISE 01 3 The wetwell vent has been designed and The NRC staff reviewed the Closed installed to meet NEI 13-02, Rev. 1 information provided in the 6-Licensee to demonstrate that guidance which will ensure that it is month updates and on the [Staff evaluation to be containment failure as a result adequately sized to prevent containment ePortal. included in SE Sections of overpressure can be overpressure under severe accident 4.1 and 4.2] prevented without a drywell conditions. The SAWM strategy will BWROG-TP-15-008 vent during severe accident ensure that the wetwell vent remains demonstrates adding water to the conditions. functional for the period of sustained reactor vessel within 8-hours of operation. the onset of the event will limit the peak containment drywell temperature significantly reducinq

-21 -PBAPS will follow the guidance (flow rate the possibility of containment and timing) for SAWA/SAWM described in failure due to temperature. BWROG-TP-15-008 and BWROG-TP-15-Drywell pressure can be 011. These documents have been posted controlled by venting the to the ePortal for NRC staff review. The suppression chamber through the wetwell vent will be opened prior to suppression pool. exceeding the PCPL value of 60 psig. Therefore, containment over BWROG-TP-011 demonstrates pressurization is prevented without the that starting water addition at a need for a drywell vent. high rate of flow and throttling after approximately 4-hours will not increase the suppression pool level to that which could block the suppression chamber HCVS. The suppression pool will be provided with sufficient makeup to maintain a heat sink for reactor decay heat, which will maintain containment within design limits. The SAWM strategy will ensure that the wetwell vent remains functional for the period of sustained operation. No follow-up questions. Phase 2 ISE 01 4 Peach Bottom will implement the initial The NRC staff reviewed the Closed SAWA flow rate of 500 GPM [gallons per information provided in the 6-Licensee to demonstrate how minute] used in the reference plant month updates and on the [Staff evaluation to be the plant is bounded by the evaluation. A site specific MAAP ePortal. included in SE Section reference plant analysis that evaluation for determining an appropriate 4.2.1.1] shows the SAWM strategy is site specific initial SAWA flow rate is not Peach Bottom was used as the successful in making it unlikely required. Using Figure 2.1.C from the reference plant. The staff that a drywell vent is needed. combined Phases 1 and 2 OIP, compare concurs that it is unlikely the the reference plant parameters to the suppression chamber HCVS plant specific parameters. could become blocked leading to a successful SAWA/SAWM strategy. Therefore, it is unlikely a drywell vent would be required Reference Plant to maintain containment inteqrity.

-22 -Torus freeboard volume is 525,000 gallons, SAWA flow is 500 GPM at 8 No follow-up questions. hours followed by 100 GPM from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, Peach Bottom Torus freeboard volume is 525,000 gallons, SAWA flow is 500 GPM at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> followed by 100 GPM from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to168 hours. The above parameters for Peach Bottom compared to the reference plant that determine success of the SAWM strategy demonstrate that the reference plant values are bounding. Therefore, the SAWM strategy implemented at Peach Bottom makes it unlikely that a drywell vent is needed to prevent containment overpressure related failure. Phase 2 ISE 01 5 The SAWA/SAWM pump (FLEX pump) The NRC staff reviewed the Closed and SAWA manual valve location is information provided in the 6-Licensee to demonstrate that adjacent to the 3 Startup Switchgear month updates and on the [Staff evaluation to be there is adequate Building north of the Unit 3 Reactor ePortal. included in SE Section communication between the Building. This location is shown in EC 4.1] main control room (MCR) and 618957, Attachment 12. Unit 3 EC The communication methods are the operator at the FLEX 618957 will be posted to the ePortal upon the same as accepted in Order manual valve during severe approval of the EC. EA-12-049. accident conditions. PBAPS [Peach Bottom Atomic Power No follow-up questions. Static] utilizes radio communications to communicate between the MCR and the operator at the SAWA flow control location. This communication method is the same as accepted in Order EA-12-049. These items will be powered and remain powered using the same methods as evaluated under EA-12-049 for the period of sustained operation, which may be lonQer than identified for EA-12-049.

-23 -Phase 2 ISE 01 6 For locations outside the Reactor Building The NRC staff reviewed the Closed between 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 7 days when SAWA information provided in the 6-Licensee to demonstrate the is being utilized, a quantitative evaluation month updates and on the [Staff evaluation to be SAWM flow instrumentation of expected dose rates has been ePortal. included in SE Sections qualification for the expected performed per HCVS-WP-02 in 4.1.1.3 and 4.2.1.3] environmental conditions. Calculation PM-1207 and found the dose EC 618957 discusses the SAWM rates at deployment locations are flow instrumentation qualification. acceptable. The selected instrument is The NRC staff found through designed for the expected flow rate, engineering judgement that the temperature and pressure for SAWA over accuracy of the flow meter and the period of sustained operation. the environmental qualifications related to the performance of the SAWA Flow Instrument flow meter meet the intent of Order EA-13-109. 3.3 to 1100 GPM, -4 to 140 °F, 740 PSI No follow-up questions. Ex~ected SAWA Parameter Qualification Range 100 to 500 GPM, 10 to 95 °F, 300 PSI The SAWA flow instrument qualifications are contained in EC 618957, Attachment 4.

8. Hanson -3 -SUBJECT: PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 AND 3 -REPORT FOR THE AUDIT OF LICENSEE RESPONSES TO INTERIM STAFF EVALUATIONS OPEN ITEMS RELATED TO NRC ORDER EA-13-109 TO MODIFY LICENSES WITH REGARD TO RELIABLE HARDENED CONTAINMENT VENTS CAPABLE OF OPERATION UNDER SEVERE ACCIDENT CONDITIONS DATED November 30, 2017 DISTRIBUTION: PUBLIC PBEB R/F RidsNrrDorllpl1 Resource RidsNrrPMPeachBottom Resource RidsNrrLaSLent Resource ADAMS Accession No. ML 17328A163 RidsRgn1 MailCenter Resource TBrown, NRR RAuluck, NRR Blee, NRR RidsACRS_MailCTR Resource *via e-mail OFFICE NRR/DLP/PBEB/PM NRR/DLP/PBMB/LA NRR/DLP/PBEB/BC NRR/DLP/PBEB/PM NAME RAuluck SLent TBrown RAuluck DATE 11/28/2017 11/28/2017 11/30/2017 11/30/2017 OFFICIAL RECORD COPY