ML13008A049

Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Cover Through Appendix a, Resumes and Qualifications, Page A-36

ML13008A049
Person / Time
Site:	Perry
Issue date:	09/28/2012
From:	Lucarelli B A, Beigi F, Reny D, Guerra E, Reddington J FirstEnergy Nuclear Operating Co, ABS Consulting
To:	Office of Nuclear Reactor Regulation
References
L-12-283
Download: ML13008A049 (284)

Text

Perry Nuclear Power Plant Near-Term Task Force Recommendation

2.3 Seismic

Walkdown Revort September 28, 2012 Prepared by: Daniel Reny (ABS Consultol_-

Fariin Beiqi toS Consultinq)

Eddie Guerra (ABS Consulting)

Brian Lucarelli (ABS Consulting)

Brian Lucarefli (ABS Consulting) 4ý" ýý ./16hn Redding'Kh (FENOC)Reviewed by: Approved by: Mohammed Alvi (FENOC)Rich Siembor (FENOC)R.. Macko~vk FNC Notes: 1. Sections 1, 3, 4, 5, 6, and 10 have been prepared by ABS Consulting.

Sections 2, 7, 8, and 9 have been prepared by FENOC.2. The review and approval of this document by FENOC personnel constitutes the owner acceptance of work performed by ABS Consulting.

FirstEnergy Nuclear Operating Company (FENOC)

Table of Contents List of A cronym s ..................................................................................................................

iv 1.0 IN TRO D U CTIO N .......................................................................................................

1 2.0 SEISM IC LICEN SIN G BA SIS ..................................................................................

1 3.0 PERSO N N EL Q U A LIFICA TIO N S ...........................................................................

3 4.0 SELECTIO N O F SSC'S .............................................................................................

4 4.1 Development of the SWEL 1 List (Related to Key Safety Functions)

.......................

4 4.2 Development of SWEL 2 for Spent Fuel Pool Related Items ...................................

8 5.0 SEISMIC WALK-DOWN AND AREA WALK-BYS .............................................

118 5.1 W alk-down Preparation

..............................................................................................

118 5.2 NTTF 2.3 W alk-downs

...............................................................................................

119 5.3 Post W alk-down Activities

.........................................................................................

119 6.0

SUMMARY

OF THE WALK-DOWN RESULTS ...................................................

119 6.1 W alk Down Item s and W alk-By Areas ......................................................................

119 6.2 W alk Down and Area W alk-By Findings ...................................................................

126 6.2.1 Seism ic W alk-down Findings .........................................................................

126 6.2.2 Area W alk-By Findings ..................................................................................

132 6.3 Configuration Checks .................................................................................................

138 7.0 LICEN SIN G BA SIS EV A LU A TIO N ........................................................................

138 8.0 IPEEE V U LN EA RBILITIES

.......................................................................................

140 9.0 PEER REV IEW ..............................................................................................................

141 10.0 REFEREN CES ...............................................................................................................

152 ii List of Tables Table 4-1 Base List 1 The Equipment Coming Out of Screen 3 and Entering Screen 4, for Five S afety F unctions ..............................................................................................................

.............

10 Table 4-2 SWEL 1 Selected Equipment for Five Safety Functions

........................................

102 Table 4-3 Base List 2 -List of All SSCs for Spent Fuel Pool .................................................

112 Table 4-4 SW EL 2 (Spent Fuel Pool) .......................................................................................

117 Table 6-1: Perry NTTF 2.3 Walk-down Items (SWEL 1+2) .....................................................

120 Table 6-2: Perry NTTF 2.3 W alk-By Areas ..............................................................................

123 Table 6-3: Perry NTTF 2.3 Inaccessible Items on SWEL 1+2.................................................

124 Table 6-4: Perry NTTF 2.3 Components Categorized by EPRI Classes ...................................

125 Table 6-5: Potentially Adverse Seismic Conditions Identified from SWC's ............................

126 Table 6-6: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys ...............

132 List of Figures Figure 2-1: Perry Nuclear Power Plant Deign SSE Spectra ......................................................

3 Figure 6-1: Missing shim plates on MCC embed ...............................

128 Figure 6-2: Centered trolley on top of switchgear

.....................................................................

129 Figure 6-3: Storage locker near bus EH 12 .................................................................................

130 Figure 6-4: Scaffold storage rack close to HVAC control panel ...............................................

131 Figure 6-5: Branch lines for fire protection piping located in CC-I at elevation 574'. .............

133 Figure 6-6: Unanchored instrument air dryers located in CC-I on CC 574'. ............................

134 Figure 6-7: Cast iron piping identified with "Seismic Violation" Tags ....................................

135 Figure 6-8: View of scaffolding structure for chiller unit .........................................................

136 Figure 6-8: Potential flooding source on AX 599'. ...................................................................

137 List of Appendices APPENDIX A: RESUMES AND QUALIFICATIONS APPENDIX B: SEISMIC WALK-DOWN CHECKLISTS (SWCs)APPENDIX C: AREA WALK-BY CHECKLISTS (AWCs)APPENDIX D: COMPONENT LIST FOR ANCHORAGE CONFIGURATION CHECK APPENDIX E: PERRY NUCLEAR POWER PLANT SEISMIC INFORMATION (EXCERPTS FROM UFSAR)APPENDIX F: PERRY IPEEE VULNERABILITIES iii List of Acronyms ALARA As Low As Reasonably Achievable AWC Area Walk-By Checklist AX Auxiliary Building BWR Boiling Water Reactor BWST Borated Water Storage Tank CC Control Complex CFR Code of Federal Regulations CIEL Containment Isolation Equipment List CO Containment COLA Combined Construction and Operating License Applications DG Diesel Generator Building DHR Decay Heat Removal DW Drywell ECCW Emergency Closed Cooling Water ECP Engineering Change Package EPRI Electric Power Research Institute EW Emergency Service Water FENOC FirstEnergy Nuclear Operating Company FH Fuel Handling Building FV Fussel-Vessley GIP Generic Implementation Procedure HPCS High Pressure Cove Spray IB Intermediate Building IPEEE Individual Plant Examination of External Events LERF Large Early Release Frequency LOCA Loss of Coolant Accident LRR Low Ruggedness Relay MCC Motor Control Center MOV Motor Operated Valve MWO Maintenance Work Order NPP Nuclear Power Plant iv NSSS Nuclear Steam Supply System NTTF Near-Term Task Force OA Operator Action OBE Operating Basis Earthquake PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor PY Perry Nuclear Power Plant RAW Risk Achievement Worth RG Regulatory Guide RP Radiation Protection SEL Seismic Equipment List SFP Spent Fuel Pool SQUG Seismic Qualification Utility Group SRO Senior Reactor Operator SSC Structures, Systems, and Components SSE Safe Shutdown Earthquake SWC Seismic Walk-down Checklist SWE Seismic Walk-down Engineer SWT Seismic Walk-down Team SWEL Seismic Walk-down Equipment List TB Turbine Building USAR Updated Safety Analysis Report v

1.0 INTRODUCTION

This Report presents the results of the Seismic Walk-down conducted for the Perry Nuclear Power Plant in support of FirstEnergy Nuclear Operating Company's (FENOC) response to the NTTF Recommendation 2.3 in NRC 50.54(f) Letter, dated March 12, 2012 (Reference 1).Consistent with the guidelines in the Electric Power Research Institute (EPRI) Report 1025286,"Seismic Walk-down Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic" (Reference 3), the walk-down implements the procedure described in Section 5.0 of this report.2.0 SEISMIC LICENSING BASIS The seismic licensing basis is contained in the Updated Safety Analysis Report (USAR) and implemented through installation standard specifications (ISS Series).Geologic and seismologic surveys of the site were conducted to establish two design earthquakes with different intensities of ground motion. These are the Operating Basis Earthquake (OBE) and the Safe Shutdown Sarthquake (SSE).The OBE is postulated to be an earthquake which could reasonably be expected to affect the plant site during its operating life. The OBE produces the vibratory ground motion for which the Seismic Category I structures, systems and components are designed to remain operational without undue risk to the health and safety of the public. The OBE is considered to be a modified Mercalli Intensity VI as measured at the site.The SSE represents the strongest vibratory ground motion earthquake for which these features (as mentioned for OBE) are, as a minimum, designed to remain functional.

The SSE is considered to be a modified Mercalli Intensity VII as measured at the site.These Seismic Category I structures, systems and components, and the seismically analyzed systems and components of the plant are necessary to assure: (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, and/or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposure of 10 CFR 100.1 The design earthquakes, OBE and SSE, for the plant are specified by OBE and SSE design response spectra. These criteria are based on the plant site geologic investigations and seismologic recommendations as discussed in Section 2.5 of the USAR. These spectra represent earthquake ground motions that are potentially damaging to structures.

While these spectra could be exceeded by ground motion "spikes" above 10 Hz, such as those caused by the earthquake of January 31, 1986, extensive investigations concerning the effects of these high-frequency motions, both from structure/equipment evaluations as well as seismological considerations, demonstrate the adequacy of the spectra used for design.Design response spectra for the SSE and OBE, as shown in Figure 3.7-1, Figure 3.7-2, Figure 3.7-3, and Figure 3.7-4 (Appendix E), comply with Regulatory Guide 1.60. As shown in these figures, the vertical component is 2/3 of the horizontal component in the frequency region lower than 2.5 cps, and the vertical and horizontal components are equal in the frequency regions higher than 3.5 cps.Safety class structures are founded on shale or on materials with equivalent seismic properties, and hence, no site dependent analysis is used. A 12-inch layer of porous concrete is used between the shale and foundation slabs of the safety class structures.

The porous concrete has a modulus of elasticity in excess of 1.2 x 106 psi and a minimum shear wave velocity of 4,400 fps, which is equivalent to that of the underlying shale.A list of Seismic Category I structures and the methodology used to evaluate them is documented in the attached Table 3.7-2. The Seismic Category I systems and components located in these structures have also been designed to withstand the effects of the design basis earthquakes.

A list of applicable Regulatory Guides, Codes & Standards can be found in section 10: References.

For the Perry Nuclear Power Plant design SSE spectra refer to figure 2-1.2 0.70 -Perry-SSE Spectra-0.60 -i(5%-Damping) r-0.50 M 0.40*0.30 u 0.20 -0.10 ._ _ _ ____*_ 1 0.00 i , , 0.1 1 10 100 Frequency (Hz)Figure 2-1: The SSE response spectrum for Perry was digitized from Perry FSAR Figure 3.7-11 3.0 PERSONNEL QUALIFICATIONS The following FENOC personnel worked together to formulate the list of selected equipment for the Perry Nuclear Power Plant NTTF Recommendation

2.3 Seismic

Walk-down:

• J. Reddington" A. Zelaski" D. Reny" F. Beigi" R. Siembor The ABS Consulting Walk-down Team consisted of the following individuals: " F. Beigi" E. Guerra" B. Lucarelli Additionally, J. Reddington served as the reviewer of the Licensing Basis and of the Individual Plant Examination External Events (IPEEE). M. Alvi served as the lead peer reviewer for the walk-down.

3 The seismic walk-down personnel, peer reviewer and lead peer reviewer possess technical degrees from accredited universities and have been trained in the application of seismic experience data for seismic verification of nuclear power plant (NPP) structures, systems, and components (SSC). In addition to completion of the NTTF 2.3 training provided by EPRI, four of these individuals (J. Reddington, M. Alvi, F. Beigi and E. Guerra) have also completed the EPRI Seismic Qualification Utility Group (SQUG) training.

Resumes and certifications of the walk-down team members are presented in Appendix A of this report.The above mentioned individuals have experience in earthquake engineering and seismic analysis.

Additionally, the team has previous experience with NPP walk-downs associated with the A-46 program, IPEEE, and recent Fukushima related stress tests for plants outside the United States.Based on their knowledge of plant documentation, associated SSCs, equipment classes, and the previous IPEEE evaluation, these individuals supported equipment selection, walk-down planning, equipment location determination, and selection of walk-by areas for the 2.3 Seismic Walk-down.

4.0 SELECTION

OF SSC'S Consistent with the guidance in EPRI #1025286, "Seismic Walk-down Guidance," dated May 2012 (Reference 3), the process of selecting the SSCs for inclusion in the SWEL 1 and SWEL 2 lists began with the compilation of currently existing large lists. The development of the list for SWEL I is presented in Section 4.1 and SWEL 2 in Section 4.2.4.1 DEVELOPMENT OF THE SWEL 1 LIST (RELATED TO KEY SAFETY FUNCTIONS)

The EPRI guidance document (Reference

3) states that using the previously developed IPEEE seismic equipment list (SEL) as a starting point for Category 1 SSCs is acceptable provided it covers all of the five safety functions requested, including the containment function.ABS Consulting is in the process of assisting FENOC in developing a seismic equipment list (SEL) for use in a seismic probabilistic risk assessment (PRA) for Perry. The use of an existing internal PRA model is a prerequisite to developing such a seismic PRA. For example, the PRA modeling logic for non-seismic events was used as a starting point for the seismic PRA plant response model. It was therefore decided, to combine the lists of SSCs from both the currently available Perry internal events PRA (i.e., model PRA-PYl-FP-ROa) and the IPEEE Safe 4 Shutdown Equipment List (IPEEE SSEL) of SSCs from the "Individual Plant Examination of External Events for Severe Accident Vulnerabilities" (Reference 4). Duplicate SSCs, caused by overlap between the two lists and because the PRA contains multiple basic events for different failure modes of a single component, were removed. Thus the requirements outlined in the EPRI walk-down guidance document for preparing the SSC SEL list were adequately satisfied.

Additionally, during SSC sampling in preparation for the walk-down, selections were generally made preferentially from the IPEEE SSEL list of SSCs. This was due to the fact that design packages were more likely to be available for these SSCs, and would allow the team to take advantage of the earlier design review work.SSCs from other sources, such as internal flooding PRA, were also chosen so that they were useful for PRA purposes.

These did not appear on either of the two original source lists.Components from the internal flood PRA, the Containment Isolation Equipment List (CIEL), and the Spent Fuel Pool (SFP) equipment List were also reviewed.

Again, duplicate SSCs were eliminated.

The list of SSCs in Tables B-i and B-3 of EPRI 1025286 (Reference

3) were also reviewed for completeness.

Some SSCs were added as a result of this review.Careful attention was paid to the SSCs in the internal events PRA that are included in the modeling of the containment isolation function and for the evaluation of interfacing LOCA frequencies.

These SSCs were flagged as important to the containment safety function; i.e., they are involved in the computation of LERF.Once the initial list of SSCs was developed, it was first screened to retain only Seismic Category 1 quality equipment.

Regular inspection of the SSC was also noted as this was justification for a second screen; e.g., for piping systems and containment penetrations.

The following attributes of the retained SSCs were collected: " Equipment ID" Brief SSC Description" SSC location -building, elevation, flood area, and the column/row letter/number

• The room environment in which the SSC is located; including radiation level, moisture level, room temperature, and whether the location is inside or outside of plant buildings" System ID; including both frontline and support systems" Key associated safety function from among the list of five safe shutdown and containment functions (i.e., Reactor Reactivity Control, Reactor Coolant Pressure 5 Control, Reactor Coolant Inventory Control, Decay Heat Removal, and Containment Function) and several support system functions mentioned in the EPRI walk-down guidance.

Panels not previously evaluated for their associated safety functions were assigned the designator, "operator," and retained for the selection process.Internal event PRA Risk Achievement Worth (RAW) and Fussell-Vessely (FV)importance measures, if available.

The equipment ID and description fields were used to assign each retained SSC to one of the EPRI equipment categories (from Table A-I of EPRI NP-6041, Revision 1, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin [Revision 1]," Electric Power Research Institute, August 1991) used for fragility analysis.

For some EPRI Categories (i.e., 0, 1, 2, and 3), a sub-category was defined and tracked separately from the original category.

For example, category la was assigned for 480V breakers that are found within a motor control center (MCC)cabinet (i.e., Category 1). None of the breaker SSCs (i.e., assigned to category la) were separately selected for the walk-down because they are already accounted for in the selection of the MCCs. Check valves and manual valves were assigned to sub-category Od to avoid categorizing these SSCs with SSCs assigned to the EPRI "other" category.

Some SSCs were selected from both the 0 and Od EPRI categories.

All of the EPRI categories were later utilized as part of the SSC selection process. Except for EPRI categories 12 (air compressors), 13 (motor generators) and 19 (temperature sensors), at least 1 SSC was selected from each EPRI category.Base List 1, as defined in the EPRI walk-down guidance, is attached as Table 4-1. The equipment coming out of Screen 3 and entering Screen 4, make up the Base List 1. All SSCs in this table are Seismic Category 1, are not regularly inspected, and are associated with one of the safety functions and supporting systems defined in the EPRI guidance.

They are therefore candidates for the SSC selection process. The column labeled "SSC source" identifies the original list from which the SSC was selected.

In some cases, SSCs appeared on several lists for Perry, but only one source was listed.A review for Screens 4d (Major New & Replacement Equipment) and 4e (IPEEE Vulnerabilities) was conducted.

There were no 1PEEE Vulnerabilities identified for seismic equipment, and therefore no entries were made for Screen 4e. A thorough review of plant records for major changes or replaced equipment was conducted and 11 equipment types identified.

A component previously undergoing change or replacement was identified with the Engineering Change Package (ECP) number.6 The equipment coming out of Screen 4d and entering the SWEL 1 "bucket" made up the SWEL 1 list.SWEL 1, as defined in the EPRI walk-down guidance is attached as Table 4-2. The format is the same as the Base List I and the table is the same except that only the selected SSCs are shown.The selected SSCs have been chosen to account for a variety of systems, equipment types, room environments, and involvement in engineering change packages since the completion of the IPEEE.A total of 109 SSCs were selected for SWEL 1. Perry plant operations staff was consulted in the SSC selection process. The selected list of SSCs was from seven different buildings and represented a variety of functions and environments throughout the plant. Many of the selected SSCs are from support systems, but there were also SSCs selected from frontline systems.Although most components originated from the IPEEE SSEL or current internal events PRA model, some components came from the containment isolation equipment list and the spent fuel pool equipment list. SSCs are selected from each of the safety functions and from a range of environmental conditions.

Some SSCs from the Emergency Service Water System could have been included on either the SWEL 1 or the SWEL 2 lists but were included only on the SWEL 1 list. Therefore, some high radiation area components are only on the SWEL 1 list but also represent high radiation environments for SWEL 2. There were 11 SSCs selected for the SWEL 2 list of which some represent high radiation areas for SWELs. Most of the SSCs selected were in cool and dry areas. However, eight are chosen from warm and damp areas, and three are from relatively hot and humid areas.The column in Table 4-2 labeled "Reason for Selection into SWEL 1" summarizes the basis for selecting the chosen SSCs. The screens referred to for each SSC are associated with the screen numbers listed across the top of the table. SSCs which are new or subject to a major replacement are assigned a screen of 4d. For a number of SSCs, the internal events PRA importance rankings (i.e., screen 4f) indicated that the SSC is risk significant (i.e., RAW>2 or FV>0.005).

A representative set, but not all, of such SSCs were therefore included in the selected list.7

4.2 DEVELOPMENT

OF SWEL 2 FOR SPENT FUEL POOL RELATED ITEMS The functions of the spent fuel pool related systems were reviewed and equipment related to fuel pool cooling and make up were included on a new list. The functions included normal spent fuel pool cooling and spent fuel pool make-up from several available sources. Spent fuel pool cleanup equipment is not Seismic Category 1 and therefore, not included on SWEL 2.Base List 2 is attached as Table 4-3. The equipment coming out of Screen 2 and entering Screen 3 in Figure 1-2 of the EPRI walk-down guidance report make up "Base List 2." All SSCs on this list are Seismic Category 1 and involve equipment and systems related to the spent fuel pool.Attributes of the retained SSCs were collected for the following information: " Equipment ID" Brief SSC description" SSC location -by building, elevation, flood area, and the column/row letter/number" The room environment in where the SSC is located; including radiation level, moisture level, room temperature, and whether the location is inside or outside of plant buildings.

The equipment ID and description fields were used to assign each retained SSC to one of the EPRI equipment Categories.

These EPRI categories were later employed as part of the SSC selection process.At Perry, there is only one path that could result in a rapid drain-down of the SFP and leave less than 10 feet of water above the top of the spent fuel. The one line is a 6 inch drain line to the Cask Pit section of the SFP. This 6 inch drain line is embedded in concrete up to a manual isolation valve that is normally closed. The manual isolation valve is included on the SWEL 2 list. There are no other lines or equipment connected to the SFP in such a way that failure could result in a rapid drain-down below a water level of 10 feet above the top of the spent fuel. The other lines connected to the SFP all have pipe embedded in concrete and have siphon breakers located more than 10 feet above the top of the spent fuel to prevent draining.SWEL 2, as defined in the EPRI walk-down guidance is attached as Table 4-4. The equipment coming out of Screen 3 and entering the SWEL 2 'bucket" in Figure 1-2 from the EPRI walk-down guidance report make up this second Seismic Walk-down Equipment List. The format is the same as that in Base List 2, and the table entries are the same except that only the selected SSCs are shown. The selected SSCs have once again been chosen to account for a variety of equipment types and room environments.

Since the types of Seismic Category 1 equipment 8 related to the spent fuel pool are limited, so too is the variety of equipment types among the SSCs selected.9

5.0 SEISMIC

WALKDOWN AND AREA WALK-BYS This section summarizes the activities prior to, during, and after performing the NTTF 2.3 seismic walkdown and area walk-bys.

It also presents the results and findings of the walkdown and documents the checklists utilized to record the walkdown data.It is concluded that the approach implemented to conduct the seismic walkdowns and area walk-bys satisfies the characteristics and recommendations outlined in EPRI Report 1025286.Therefore, by following these guidelines, the walkdown approach and format of the results documented herein fulfills the requests established in the NRC 50.54(f) letter, Enclosure 3, Recommendation 2.3: Seismic.5.1 WALKDOWN PREPARATION The overall procedure directly implements the EPRI guidelines.

However, due to their unique nature, the following description gives special attention to the selection and execution of the configuration checks of selected anchorage.

EPRI guidelines recommend that a minimum of 50 percent of the equipment considered in the walkdown be examined to document the existing anchorage configurations, and assess this configuration relative to the design basis. It also recommends that the block wall maps be retrieved to document previous evaluations in support of NTTF 2.3. However, with the exception of one block wall, Beaver Valley Power Station Unit 2 does not have any safety related masonry block walls associated with Seismic Category 1 components, thus the process to verify block wall adequacy per IE 80-11 has been omitted for this walkdown.

It is noted-that the seismic adequacy of one block wall mentioned above was confirmed from existing design calculations.

Prior to the walkdowns, the Seismic Walkdown Engineers (SWE) examined available plant documentation associated with anchorage design and correlated this to relevant SWEL components and the respective Seismic Walkdown Checklists (SWC) and Area Walk-By Checklists (AWC). This pre-walkdown activity contributed to gaining familiarity and critical insights regarding the components and areas to be walked down. The relevant design documentation, drawings and calculations were uploaded to each of the SWEs electronic tablets used during the walkdown with the intention of verifying, if required, any anchorage configuration.

205 5.2 NTTF 2.3 WALKDOWNS The NTTF 2.3 walkdowns at Beaver Valley Unit 2 were performed over a duration of five days from September 17 to September 21, 2012 and a one day walkdown on October 5, 2012 for equipment that were inside the Containment building.

During the walkdowns, the SWEs completed the walkdown checklists as SWEL components were inspected.

Selected anchorage configurations were verified for 50% of the floor or wall mounted components on the SWEL with respect to design documentation, including anchorage design drawings and IPEEE calculations.

5.3 POST WALKDOWN ACTIVITIES The primary activity after the walkdown involved compiling the SWCs and the AWCs.Additional documentation, such as design calculations and/or IPEEE submittals, were also reviewed to support configuration checks. Photographs taken during the walkdown were linked to the respective checklists.

Some of the findings of the walkdown that could not readily be dispositioned during the walkdowns were evaluated further through additional calculation/modification package reviews for proper disposition.

The post walkdown activity also developed this walkdown report.6.0

SUMMARY

OF THE WALKDOWN RESULTS 6.1 WALK DOWN ITEMS AND WALK-BY AREAS The SWEL 1 included a total of 109 components, and SWEL2 included a total of 10 components.

From this total of 119 components, 105 components were successfully walked down during the week of September 17 to September 21, 2012. The SWT returned to the site on Friday, October 5 and walked down the remaining 14 components located inside the Reactor Building.

Table 6-1 and Table 6-2 identify the walkdown items and walk-by areas, respectively.

The areas walk-by and the walkdown items are cross correlated on the respective SWCs and AWCs. Table 6-3 provides the list of equipment that was walked down.206 Table 6-1: Beaver Valley 2 NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment ID No Equipment Class BldgJ El Area Description 2CCP-27A Od. Other-Check or Manual Valve AXLB 735 N-EAST 2 AXLB 735 2CCP-4 Od. Other-Check or Manual Valve AXLB 735 N-EAST 2 AXLB 735 2CCP-AOV107A

7. Pneumatic-Operated Valves RCBX 71.8 RCBX 721 -A RCP Pump Cub 2CCP-E21A

21. Tanks and Heat Exchangers AXLB 710 AXLB 710 HX 2CCP-FT107A

18. Instrument (on) Racks RCBX 718 RCBX 718 2CCP-MOV 112A 8A. Motor-Operated Valves RCBX 718 RCBX 718-Annulus 2CCP-MOV 119 8A. Motor-Operated Valves MSCV 773 IAC Room 2CCP-MOV 150-1 8A. Motor-Operated Valves MSCV 722 MSCV 718 2CCP-P21A

5. Horizontal Pumps AXLB 735 N/EAST 2 AXLB 735 2CCP-PTI07A

18. Instrument (on) Racks RCBX 718 RCBX 718 2CCP-TK21A

21. Tanks and Heat Exchangers AXLB 773 AXLB 773 Cool Surge Tank 2CHS-FCV1 14A 7. Pneumatic-Operated Valves AXLB 710 AXLB 710 2CHS-HCV 186 7. Pneumatic-Operated Valves AXLB 718 AXLB 718 2CHS-LCVlI 15B 8A. Motor-Operated Valves AXLB 718 AXLB 718 2CHS-MOV310 8A. Motor-Operated Valves RCBX 692 RCBX 692-Near Inner Stairs 2CHS-MOV8132A 8A. Motor-Operated Valves AXLB 718 AXLB 718 2CHS-P21A

5. Horizontal Pumps AXLB 735 AXLB-CP-735 2CHS-SOV206 8B. Solenoid Valves AXLB 755 AXLB 755 Boric Acid TK Rm 2CHS-TK21A

21. Tanks and Heat Exchangers AXLB 755 AXLB 755 Boric Acid TK Rm 2CVS-SOV102 8B. Solenoid Valves MSCV 718 MSCV 718 2CVS-SOV151A 8B. Solenoid Valves MSCV 718 MSCV 718 2DAS-AOV100A

7. Pneumatic-Operated Valves RCBX 718 RCBX 718-PEN-724 COL 9 2EGF-LIS203A

18. Instrument (on) Racks DGBX 732 EDG 2-1 2EGF-P21A

6. Vertical Pumps DGBX 732 EDG 2-1-2EGF-TK22A

21. Tanks and Heat Exchangers DGBX 732 EDG 2-1 2EGS-EG2-1

17. Engine Generators DGBX 732 EDG 2-1 2FNC-108 Od. Other-Check or Manual Valve FULB 729 FULB 729 PMP Room 2FNC-E21A

21. Tanks and Heat Exchangers FULB 740 FULB 741 HX Room 2FNC-EJM230A Od. Other-Check or Manual Valve FULB 729 FULB 729 PMP Room 2FNC-P21A

5. Horizontal Pumps FULB 729 FULB 729 PMP Room 2FNC-RV101 Od. Other-Check or Manual Valve FULB 740 FULB 741 HX Room 2FNC-TII01A

19. Temperature Sensors FULB 740 FULB 741 HX Room 2FNC"TI102A

19. Temperature Sensors FULB 740 FULB 741 HX Room 2FWE-FE101A

18. Instrument (on) Racks SFGB 741 SFGB 741 Cubicle A 207 Table 6-1: Beaver Valley 2 NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment ID No Equipment Class Bldg El Area Description 2FWE-HCV100D

7. Pneumatic-Operated Valves SFGB 741 SFGB 741 Cubicle C 2FWE-P22 5. Horizontal Pumps SFGB 718 SFGD 718 2FWE-P23A

5. Horizontal Pumps SFGB 718 SFGD 718 2FWS-FCV478

7. Pneumatic-Operated Valves SRVB 780 SRVB 780 2FWS-FCV479

7. Pneumatic-Operated Valves SRVB 780 SRVB 780 2FWS-HYVI57A

7. Pneumatic-Operated Valves MSCV 773 Main Steam Room El 778 2FWS-LT477F

18. Instrument (on) Racks RCBX 718 RCBX 718-Annulus 2HVC-ACU201A

10. Air Handlers CNTB 735 CNTB 735-AC Room 2HVD-DMP201A

7. Pneumatic-Operated Valves DGBX 759 EDG 2-1 Upstairs 2HVD-DMP22A

7. Pneumatic-Operated Valves DGBX 759 EDG 2-1 Upstairs 2HVD-FN270A

9. Fans DGBX 759 EDG 2-1 Upstairs 2HVP-CLC265A

10. Air Handlers AXLB 755 AXLB 755-MCC Room 2HVR-ACU207A

10. Air Handlers SFGB 741 SFGD 741-PLAT 2HVR-TI228

19. Temperature Sensors CNTB 735 Control Room 2HVR-TI228-1

19. Temperature Sensors CNTB 735 Control Room 2HVW-FN257A

9. Fans INTS 705 Intake Cubicle C 2HVW-MOD21A

7. Pneumatic-Operated Valves INTS 705 Intake Cubicle D 2HVZ-DMP215A

7. Pneumatic-Operated Valves MSCV 773 SWGR Vent Room 773 2HVZ-FN261A

9. Fans MSCV 773 SWGR Vent Room 773 2HVZ-FN261B

9. Fans MSCV 773 SWGR Vent Room 773 2MSS-AOV101A

7. Pneumatic-Operated Valves MSCV 773 Main Steam Room El 789 2MSS-SOV105A 8B. Solenoid Valves MSCV 773 Main Steam Room El 789 2MSS-SV101A Od. Other-Check or Manual Valve MSCV 773 Main Steam Rm Upper Plat.2QSS-297 Od. Other-Check or Manual Valve YARD 730 Yard 2QSS-MOV100A 8A. Motor-Operated Valves SFGB 718 SFGD 718 UP 2QSS-MOV101A 8A. Motor-Operated Valves SFGB 718 RSS Cubicle 2RCS-AOV101

7. Pneumatic-Operated Valves MSCV 718 MSCV 718 2RCS-PT440

18. Instrument (on) Racks MSCV 740 MSCV East 735 2RHS-E21B

21. Tanks and Heat Exchangers RCBX 707 RCBX 707 2RHS-HCV758A

7. Pneumatic-Operated Valves RCBX 692 RCBX 707 2RHS-MOV702A 8A. Motor-Operated Valves RCBX 718 RCBX 718-A RCP Pump 2RHS-MOV720A 8A. Motor-Operated Valves RCBX 718 RCBX 718-B RCP Pump 2RHS-P21A

6. Vertical Pumps RCBX 707 RCBX 707 2RHS-RV721A Od. Other-Check Manual Valve RCBX 692 RCBX 707 2RSS-TI150A

19. Temperature Sensors CNTB 735 Control Room 208 Table 6-1: Beaver Valley 2 NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment ID No Equipment Class Bldg El Area Description 2SIS-1 0d. Other-Check or Manual Valve SFGB 718 SFGD 718 2SIS-67 Gd. Other-Check or Manual Valve RCBX 718 RCBX 718-Annulus Col 4 2SIS-MOV863A 8A. Motor-Operated Valves SFGB 718 SFGD 718 West 2SIS-MOV867A 8A. Motor-Operated Valves AXLB 710 AXLB 710 Boron Tank 2SIS-MOV88 11 A 8A. Motor-Operated Valves SFGB 718 SFGD 718 West 2SIS-P21A

5. Horizontal Pumps -SFGB 718 SFGD 718 2SVS-HCV104

7. Pneumatic-Operated Valves MSCV 773 Main Steam Room El 778 2SVS-PCV101A Gd. Other-Check or Manual Valve MSCV 773 Main Steam Room Upper Plat.2SWS-57 Gd. Other-Check or Manual Valve INTS 705 Intake Cubicle D 2SWS-EJM221A

0. Other INTS 705 Intake Cubicle C 2SWS-MOV104A 8A. Motor-Operated Valves SFGB 718 SFGD 718 West 2SWS-MOV 106A 8A. Motor-Operated Valves VLVP 718 Valve Pit A 2SWS-MOVI 13A 8A. Motor-Operated Valves DGBX 732 EDG 2-1 2SWS-P21A

6. Vertical Pumps INTS 705 Intake Cubicle D 2SWS-PCV 118 7. Pneumatic-Operated Valves INTS 705 Intake Cubicle C 2SWS-PT-1 13A 18. Instrument (on) Racks VLVP 718 Valve Pit A 2SWS-PT-1 17A 18. Instrument (on) Racks INTS 705 Intake Cubicle D 480VUS-2-8

2. Low Voltage Switchgear SRVB 730 Emerg SWGR AE 480VUS-2-9

2. Low Voltage Switchgear SRVB 730 Emerg SWGR DF 4KVS-2AE 3. Medium Voltage Switchgear SRVB 730 Emerg SWGR AE 4KVS-2DF 3. Medium Voltage Switchgear SRVB 730 Emerg SWGR DF 52-BYA 2. Low Voltage Switchgear MSCV 755 CV&RC Area-Reac-2T-SWGR 52-RTA 2. Low Voltage Switchgear MSCV 755 CV&RC Area-Reac-2T-SWGR BAT-2-1 15. Battery Racks SRVB 730 Battery Room 2-1 BAT-CHG2-1

16. Battery Chargers and Inverters SRVB 730 Emerg SWGR AE BAT-CHG2-3

16. Battery Chargers and Inverters SRVB 730 Emerg SWGR AE DC-SWBD2-1

2. Low Voltage Switchgear SRVB 730 Emerg SWGR AE MCC-2-E01

1. Motor Control Centers INTS 705 Intake Cubicle D MCC-2-E03

1. Motor Control Centers AXLB 755 AXLB 755-MCC Room MCC-2-E05

1. Motor Control Centers MSCV 735 MSCV West MCC-2-E07

1. Motor Control Centers DGBX 732 EDG 2-1 MCC-2-E09

1. Motor Control Centers CNTB 707 Control BLDG MCC MCC-2-E 11 1. Motor Control Centers SFGB 737 SFGD 737 PNL 2DIGEN-1 20. Instrument and Control Panels DGBX 732 EDG 2-1 PNL DC2-07 14. Distribution Panels SRVB 730 Emerg SWGR AE 209 Table 6-1: Beaver Valley 2 NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment ID No Equipment Class Bldg El Area Description PNL DC2-19 14. Distribution Panels SRVB 730 Emerg SWGR AE PNL-2BLG-SER

20. Instrument and Control Panels CNTB 735 Control Room PNL-2RPU-A

20. Instrument and Control Panels CNTB 707 CNTB 707 PNL-SEQ-244

20. Instrument and Control Panels SRVB 730 Emerg SWGR AE PNL-VITBS2-1A 14..Distribution Panels CNTB 707 CNTB 707 SW Comer PNL-VITBS2-2C

14. Distribution Panels CNTB 735 Control Room 2QSS-LT 104A 18. Instrument (on) Racks YARD 730 Yard RK-2AUX-REL-C

20. Instrument and Control Panels CNTB 707 CNTB 707 RK-2PRI-PROC-1

20. Instrument and Control Panels CNTB 707 CNTB 707 RK-2PRI-PROC-2

20. Instrument and Control Panels CNTB 707 CNTB 707 RK-2SEC-PROC-A

20. Instrument and Control Panels CNTB 707 CNTB 707 RK-2NUC-INS

20. Instrument and Control Panels CNTB 735 Control Room TRF-2-8N 4. Transformers SRVB 730 Emerg SWGR AE UPS-VITBS2-1

16. Battery Chargers and Inverters SRVB 730 Emerg SWGRAE UPS-VITBS2-1-REG

4. Transformers SRVB 730 Emerg SWGR AE 210 Table 6-2: Beaver Valley 2 NTTF 2.3 Walk-By Areas Area Bldg Floor El AXLB 710 AXLB 710 AXLB 710 Boron Tank AXLB 710 AXLB 710 HX AXLB 710 AXLB 718 AXLB 718 AXLB 755 Boric Acid TK Room AXLB 755 AXLB 755-MCC Room AXLB 755 AXLB 773 Cool Surge Tank AXLB 773 AXLB-CP-735 AXLB 735 Battery Room 2-1 SRVB 730 CNTB 707 CNTB 707 CNTB 707 SW Corner CNTB 707 CNTB 735-AC Room CNTB 735 Control BLDG MCC CNTB 707 Control Room CNTB 735 CV&RC Area-Reac-2T-SWGR MSCV 755 EDG 2-1 DGBX 732 EDG 2-1 Upstairs DGBX 759 Emerg SWGR AE SRVB 730 Emerg SWGR DF SRVB 730 FULB 729 PMP Room FULB 729 FULB 741 HX Room FULB 741 IAC Room MSCV 773 Intake Cubicle C INTS 705 Intake Cubicle D INTS 705 Main Steam Room El 778 MSCV 778 Main Steam Room El 789 MSCV 789 Main Steam Room Upper Plat. MSCV 773 MSCV 718 MSCV 718 MSCV East 735 MSCV 735 MSCV West MSCV 735 N-EAST 2 AXLB 735 AXLB 735 RCBX 692-Near Inner Stairs RCBX 692 RCBX 707 RCBX 707 RCBX 718 RCBX 718 RCBX 718-A RCP Pump RCBX 718 RCBX 718-Annulus RCBX 718 211 Table 6-2: Beaver Valley 2 NTTF 2.3 Walk-By Areas Area Bldg Floor El RCBX 718-Annulus Col 4 RCBX 718 RCBX 718-B RCP Pump RCBX 718 RCBX 718-PEN-724 COL 9 RCBX 718 RCBX 721-A RCP Pump Cubicle RCBX 721 RSS Cubicle SFGB 718 SFGB 741 Cubicle A SFGB 741 SFGB 741 Cubicle C SFGB 741 SFGD 718 SFGB 718 SFGD 718 UP SFGB 718 SFGD 718 West SFGB 718 SFGD 737 SFGB 737 SFGD 741 -PLAT SFGB 741 SRVB 780 SRVB 780 SWGR Vent Room 773 MSCV 773 Valve Pit A VLVP 718 Yard YARD 730 212 Table 6-3: Beaver Valley 2 NTTF 2.3 Components Categorized by EPRI Classes EPRI Equipment De ti Components Cat No. eWalked Down 0 Other 13 1 Motor Control Centers and Wall-Mounted Contactors 6.2 Low Voltage Switchgear and Breaker Panels 5 3 Medium Voltage, Metal-Clad Switchgear 2 4 Transformers 2 5 Horizontal Pumps 6 6 Vertical Pumps 3" 7 Pneumatic-Operated Valves 17 8 Motor-Operated and Solenoid-Operated Valves. 20 9 Fans 4 10 Air Handlers 3 I i Chillers 0 12 Air Compressors 0 13 Motor Generators 0-14 Distribution Panels and Automatic Transfer Switches 4 15 Battery Racks 1 16 Battery Chargers and Inverters 3 17 Engine Generators I 18 Instrument (on) Racks 9 19 Temperature Sensors _ 5 20 Instrumentation and Control Panels 9 21 -Tanks and Heat Exchangers 6 Total 119 213 6.2 WALK DOWN AND AREA WALK-BY FINDINGS The examination of walkdown items and observations in area walk-bys confirms the general seismic robustness of the design and installation.

The plant is well maintained and no major issues related to potentially adverse conditions were uncovered.

In general, based on the number of minor potentially adverse seismic conditions identified during the walkdown, it can be concluded that most components and areas were found to be in good condition and that no major degraded or design non-conformances were identified.

Generally, the nature of the potentially adverse conditions is related to mild corrosive conditions, responsiveness for old deficiency tags and minor discrepancies between existing and as-designed conditions.

Several relatively minor findings are reported here. Observations in this respect are organized on the basis of potentially adverse seismic conditions identified during both Seismic walkdowns and area walk-bys.6.2.1 Seismic Walkdown Findings The following section presents potentially adverse seismic conditions and findings identified during the Seismic walkdowns.

A total of 7 potentially adverse seismic conditions were identified during the Seismic walkdowns.

Table 6-4 provides a summary of all 7 adverse finding conditions identified.

As shown in Table 6-4, six condition reports were issued, which required Licensing Basis Evaluation.

Justifications for findings for which a Licensing Evaluation is not required are provided in the Component's respective SWC provided in Appendix B.Table 6-4: Potentially Adverse Seismic Conditions Identified from Seismic Walkdowns Licensing quipment E quipment Class Description of Adverse Seismic Basis Reference for ID No Condition Evaluation Justification Required Corroded bolts for Vertical Pump CR-2012-2SWS-P21A

6. Vertical Pumps 2SWS-P21A.

Yes 14408 2SWS- 8A. Motor-Operated Deficiency tag for leakage in Yes CR-2012-MOV 113A Valves packing of MOV 113A 14409 MCC-2-E 11 1. Motor Control Unrestrained 55 gallon drum Yes CR-2012-Centers located near MCC 14420 214 Table 6-4: Potentially Adverse Seismic Conditions Identified from Seismic Walkdowns Licensing quipment EDescription of Adverse Seismic Basis Reference for ID No Equipment Class Condition Evaluation Justification Required PNL-SEQ- 20. Instrument and Interaction potential between Yes CR-2012-244 Control Panels lighting fixture and PNL-SEQ-244 14463 Corrosion identified on enclosure CR2012.2QSS- 18. Instrument (on) and anchorage of component-Yes 14744 2QSS-LT104A 2QSS-297 0. Other -Check or Corrosion identified on yoke of Yes CR-2012-Manual Valve manual valve 2QSS-297 14749 Substantial unsupported span SWC for 2FNC-P21A

5. Horizontal Pumps between discharge nozzle for 2FNC-P21A 2FNC-E21A

21. Tanks and Heat Pump 2FNC-P21A and inlet No & 2FNC-Exchangers nozzle for Heat Exchanger 2FNC- E21A AE21A.215 The following section provides additional insight into generally found scenarios and subsequently resolved conditions.

• Outage maintenance equipment located inside Reactor Building Several carts, storage boxes and general housekeeping equipment were identified in areas walked by inside the containment building.

It was noted that most of the carts and storage boxes in the area were properly restrained while others, such as ladders and maintenance tools, were not. It was confirmed by plant personnel that the nature of any unrestrained equipment identified in different areas was due to the works being performed as part of the ongoing plant outage in conformance with the plant house keeping requirements.

Figure 6-1: Maintenance equipment located inside Reactor Containment Building 216

• Corrosion identified on inspected Yard Components A corrosive state was identified for components inspected in the Yard area. A corrosion condition was identified for the anchorage of the level transmitter 2QSS-LT104A.

A corrosive state was also identified on the yoke of the manual valve 2QSS-297.

Both of these components are located outside in the yard area near the RWST 2QSS-TK21 at elevation 734'. Notification No. 600788283 was initiated in order to restore the condition of these components (see also condition reports CR-2012-14744 and CR-2012-14749).

Figure 6-2: Corrosive condition found for Yard components 217

• Substantial unsupportedpipe length between pump 2FNC-P21A and HX 2FNC-E21A While performing the walkdowns in the Spent Fuel Building, the SWT identified a long span of piping between the Pump 2FNC-P21A discharge nozzle and the Heat Exchanger 2FNC-E21A inlet nozzle that did not have any lateral restraints.

After further discussion regarding this finding, the SWT was informed that this portion of the Spent Fuel piping system is not seismic category 1. Nevertheless, pipe stress calculations for this piping were identified, which showed seismic adequacy of the piping and nozzles.Figure 6-3: View of piping system between Heat Exchangers and Pumps in the SFP area.218

• Interaction potential between lighting fixture and Panel PNL-SEQ-244 The SWT identified a potential for interaction between a chain hung lighting fixture and control panel PNL-SEQ-244 located in the Emergency Switchgear AE at elevation 730'. Lighting fixtures are located near the top of the panel which, during a seismic event, could potentially swing and hit the top section of the panel. Condition report CR-2012-14463 was issued to correct this impact concern.Figure 6-4: Lighting fixtures near Panel PNL-SEQ-244 219 6.2.2 Area Walk-By Findings The following section presents potentially adverse seismic conditions and findings identified during the area walk-bys.

A total of 7 potentially adverse seismic conditions were identified during the area walk-bys.

Table 6-5 provides a summary of all 7 potentially adverse seismic conditions identified.

As shown in Table 6-5, six condition reports were issued, which required Licensing Basis Evaluation.

Justifications for findings for which a Licensing Evaluation is not required are provided in the Area's respective AWCs provided in Appendix C.Table 6-5: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys Licensing Description of Adverse Basis Reference for Area Bldg Floor El Seismic Condition Evaluation Justification I jRequired Valve Pit A VLVP 718 Deficiency tag for component Yes CR-2012-14412 2SWM-MOV562.

MSCV 718 MSCV 718 Deficiency tag for component Yes CR-2012-14449 2SWS-MOVI160.

MSCV 718 MSCV 718 Deficiency tag for component Yes CR-2012-14450 2SWS-MOV 162.MSCV 718 MSCV 718 Deficiency tag for component Yes CR-2012-14452 2SWS-MOV152-1.

Corroded bolts on valve MSCV 718 MSCV 718 Corrod lts Yes CR-2012-14455 2SWS-MOV161.

Corroded flange nuts for AXLB-710 HX AXLB 710 nozzle outlet of component Yes CR-2012-14459 2CCP-E21 C.EDG Resistor Bank supported AWC for EDG 2-1 EDG 2-1 Upstairs DGBX 7 on 4 porcelain feet. No Upstairs 220 o Emergency Diesel Generator Ground Resistor 2EGS-GR2-1 While performing the area walk-by for the Diesel Generator Building 2-1, SWEs identified Resistor 2EGS-GR2-1 to be mounted on 4 porcelain feet and not top braced. The SWT discussed the possibility of an electrical safety concern if there were current running through the resistor bank at the time of an earthquake.

Since this equipment is only used for periodic testing and is not Category 1 equipment, SWE's judged the configuration to be adequate and not to present a seismic concern.Figure 6-5: Emergency Diesel Generator Ground Resistor 2EGS-GR2-1 221

6.3 CONFIGURATION

CHECKS The SWELL 1+2 included 70 items, which were not in-line components such as valves. The process of verifying the anchorage configuration focused on 35 SWEL components arbitrarily selected prior to walkdown proceedings (this is 50% of the SWEL items with anchorage configurations).

Appendix D provides a list of the 35 components comprising the anchorage configuration list linked with the specific references used for verification purposes; i.e., IPEEE Calculations, design drawings, etc.The anchorage configuration for each of the 35 SWEL components listed in Appendix D was verified based on IPEEE Calculations and Plant Design documentation.

SWEs referred to design drawings as the main reference for anchorage verification whenever it was possible to have a complete field inspection of the anchorage.

The design drawings were uploaded onto electronic tablets for quick accessibility during the walkdowns and verification of the as-installed configuration against the design drawings.

In cases where design basis drawings were not readily identifiable, SWEs referred to previous IPEEE Calculations to ensure that the configuration was assessed during the IPEEE program and no design concerns were identified.

These configuration checks verified consistency of as-installed conditions to that of the design drawings/calculations in all 35 instances.

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7.0 LICENSING

BASIS EVALUATION Thirteen condition reports (CR) were generated as a result of these walkdowns.

The following is a list of the condition reports written as a result of the walkdowns:

CR-2012-14408, CR-20 12-14409, CR-2012-14412, CR-2012-14420, CR-2012-14449, CR-2012-14450, CR-2012-14452, CR-2012-14455, CR-2012-14459, CR-2012-14463, CR-2012-14744, CR-2012-14749, and CR-2012-14758.

The following summarizes the condition and resolution to the condition reports written as a result of the walkdowns.

CR-2012-14408 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that two out of twelve nuts for the bolts anchoring the Service Water Pump 2SWS*P21A base to the anchor plate are corroded.

This pump is located inside the Intake Structure at elevation 705'Cubicle D.Observation concluded that there is only surface rust on the bolts and the nuts. Even though they are corroded, they are capable to perform their intended design function based on engineering judgment.

The two nuts are located near each other are readily visible as mildly corroded.

No calculations or drawings are affected since the nuts are able to perform the intended function.To avoid further degradation of the nuts due to presence of moist environment, the nuts should be replaced in the next system window and painted to mitigate corrosion.

Initiated Notification No. 600786691 to perform the work to be done under Work Order 200530757.

CR-2012-14409 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that a Deficient Tag (Tag ID 117472) for Component 2SWS-MOV1 13A has been in place with no date on it and no physical work performed to correct the deficiency.

This motor operated valve is located inside the Diesel Building (DGBX) at elevation 732'. The deficient tag says "Leakage in Packing." This leakage has also resulted in the corrosion of packing nuts.Observation concluded that there is only surface rust on the bolts and the nuts even though they are corroded are capable to perform their intended design function based on engineering judgment.

No calculations or drawings are affected since the nuts are able to perform the intended function.223 To avoid futher degradation of the nuts, the nuts should be replaced in the next system window and painted to mitigate corrosion.

Initiated Notification No. 600786692 to perform the work under Work Order 200481347.

CR-2012-14412 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that a Deficient Tag (Tag ID 63005) for Component 2SWM-MOV562 has been in place since 2007, and no physical work performed to correct the deficiency.

This motor operated valve is located inside the Valve Pit Room at elevation 718'. The deficient tag says "Valve has rust and needs to be cleaned/painted." Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve is able to perform the intended function.To avoid further degradation of the valve dueto corrosion, the corrosion should be cleaned/painted during the next system window. Initiated Notification 600786690 to perform the work under Work Order 200416834.-CR-2012-14420 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that a 55 gallon drum was located too close to MCC-2-E1 1 and was unrestrained.

The component is located at El 737' in U-2 Safeguard Building.

This posed a seismic interaction concern that during a seismic event it had the potential to hit the MCC and potentially impact the design function of inside components.

This condition has been corrected and this CR was generated to document an existing condition that was identified during the seismic walkdowns.

Currently no anomaly exists as such there are no operability concerns, and thus no violations of the design basis requirements.

CR-2012-14449 It was observed that Component 2SWS-MOV160 had a Deficient Tag (Tag ID 47014) and no physical work performed to correct the deficiency.

This motor operated valve is located inside the MSCV Room at elevation 718'. The deficient tag says "Surface rust to be cleaned and preserved." 224 Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since surface rust doesn't affect the function or the valve.Implement field work per Initiated Notification No. 600786757 to avoid further degradation of valve. Work to be done under Work Order 200531116.

CR-2012-14450 It was observed that Component 2SWS-MOV 162 has corrosion on the valve from yoke to the bonnet. This motor operated valve is located inside the MSCV Room at elevation 718'.Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve is able to perform the intended function.Clean the rust/corrosion per Initiated Notification No. 600786758 to perform the work. Work to be done under Work Order 200531117.

CR-2012-14452 It was observed that Component 2SWS-MOV152-1 had a Deficient Tag (Tag ID 47016) and no physical work performed to correct the deficiency.

Corrosion extends from yoke to bonnet. This motor operated valve is located inside the MSCV Room at elevation 718'. .The deficient tag says"Corrosion (Surface Rust) on the valve body at packing gland." Observation concluded that there is only surface rust and the valve, is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve is able to perform the intended function.Implement field work per Initiated Notification No. 600786761 to perform the work. Work to be done under Work Order 200531158.

CR-2012-14455 It was observed that the four bolts and nuts connecting the yoke to the bonnet for Component 2SWS-MOV161 are mildly corroded.

This motor operated valve is located inside the MSCV Room at elevation 718'.225 Observation concluded that there is only surface rust on the bolts as well as the nuts and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since they are able to perform the intended function.Recommend during the next available Maintenance Window, to replace the four nuts/bolts connecting the yoke to bonnet per Notification No. 600786762 to perform the work. Work to be done under Work Order 200531159.

CR-2012-14459 It was observed that the flange nuts at outlet nozzle for Component 2CCP-E21C are mildly corroded.

This heat exchanger is located at elevation 710' in Aux Building.Observation concluded that there is only surface rust on the nuts and the component is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since they are able to perform the intended function.Recommend during the next available Maintenance Window, to replace the flange nuts at outlet nozzle of heat exchanger per Notification No. 600786796 to perform the work. Work to be done under Work Order 200531161.

CR-2012-14463 There is a potential seismic interaction concern. associated with Panel PNL-SEQ-244 located at El. 730' in Service Building, Emergency Switchgear AE. Specifically, right near the top of this panel, there are lighting fixtures supported from unistruts hung from chains above. During a seismic event, there is a potential for these lighting fixtures to rattle and hit the top of safety related panel.Observation concluded that the impact force from these lighting fixtures on the panel, considering the weight of lighting fixtures and unistrut, will have insignificant impact on the components inside-the panel and they will perform their intended design function, by engineering judgment.

However, this CR is being generated as an enhancement to consider upgrading the condition and restraining the unistrut supporting the lighting fixture such that it would not hit the panel.226 No calculations or drawings are affected since the components are able to perform the intended function safely.CR-2012-14744 The enclosure and the anchorage for Component 2QSS-LT104A is corroded.

This component is located outside in the yard near RWST 2QSS-TK21 at elevation 734'. The corrosion does not affect the structural integrity of the component.

The design basis requirements are still being met.Replace the enclosure and the anchorage per Notification No. 600788282 to replace the enclosure.

Work to be done under Work Order 200532069.

CR-2012-14749 The yoke for the manual valve 2QSS-297 is corroded.

This component is. located outside in the yard near RWST 2QSS-TK21 at elevation 734'. The component is still capable to perform its intended design function.Replace the yoke per Notification No. 600788283 to replace the yoke, under Work Order 200532070.

CR-2012-14758 This CR has been generated to capture all the issues in one condition report (roll-up CR) that have been identified during.NRC 50.54f Letter Section 2.3 Seismic walkdowns performed at Beaver Valley Unit-2 Plant during the week of September 17, 2012.There are no new anomalies identified in this CR as individual CRs have already been generated as required and as identified in the attached matrix, as such there are no operability concerns associated with this CR.8.0 IPEEE VULNEARBILITIES There were no seismic vulnerabilities identified in the IPEEE submittal for Beaver Valley units 1 or 2. This was recognized by the NRC in NUREG'1437 Supplement 36 "Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 36, Regarding Beaver Valley Power Station Units 1 and 2." Page G20 and 21 states "The NRC staff also notes that the use of the integrated PSA to facilitate identification of SAMAs for external events, the prior 227 implementation of plant modifications for seismic and fire events, and the absence of external event vulnerabilities ensure that the search for external event SAMAs was reasonably comprehensive." Several submittals to the NRC covered IPEEE enhancements, but none would be classified as vulnerabilities.

Tables identifying IPEEE vulnerabilities are essentially based on these enhancements and the enhancements were incorporated into the walkdown component -selection to the extent possible.9.0 PEER REVIEW A peer review of the Submittal Report for the Near Term Task Force NTTF Recommendation 2.3 "Seismic Walkdowns" was performed using the guidance provided in Section 6 of EPRI Document 1025286, "Seismic Walkdown Guidance." Following are the peer reviewers for the Beaver Valley Power Station Unit-2:* Mohammed Alvi (Team Leader)" John Reddington The peer review process included the following activities: " Review the selection of the SSCs included on the SWEL* Review a sample of the checklists prepared for the seismic walkdowns and area walk-bys" Review the Licensing Basis Evaluations" Review the decisions for entering the potentially adverse conditions into the Corrective Action Program (CAP)." Review the submittal report" Summarize the results of the peer review process in the submittal report A. Review the Selection of the SSCs Included on the SWEL: The peer review concluded that the selection of Seismic Walkdown Equipment List (SWEL) was performed in accordance with guidance provided in Section 3 of EPRI Document 1025286"Seismic Walkdown Guidance." The peer reviewers used the checklist provided in Appendix F of this document which is enclosed.

Also, an ex-Senior Reactor Operator (SRO) from the Beaver 228 Valley Power Station, Unit-2 acted as Operations representative during the selection of the SWEL.Appropriate figures 1-1, 1-2 and 1-3 of the EPRI Document 1025286 were used and the final SWEL 1 and SWEL 2 were developed.

The peer review confirmed that the following EPRI screens were used in the selection of SWEL 1: Screen 1: Seismic Category I Screen 2: Equipment or System Screen 3: Support for the five safety functions Screen 4: Sample Considerations The station did use the existing documentation that resulted from IPEEE program in identifying the components.

A matrix/spreadsheet was prepared that identifies all the selected components on SWEL 1 and SWEL 2. It was confirmed that these two lists did include a variety of type of systems, major new and replacement equipment, a variety of equipment types, a variety of environments in which the components are located, and the equipment enhanced due to vulnerabilities identified during the IPEEE program.It was confirmed that the size of the sample was sufficiently large to include a variety of items that collectively included variations within all the attributes stated in theparagraph above.SWEL 1 for the Beaver Valley Power Station, Unit-2 included 109 components.

The peer review also confirmed that the station used the following EPRI screens in the development of SWEL 2: Screen 1: Seismic Category I Screen 2: Equipment or System Screen 3: Sample Considerations Screen 4: Rapid Drain-Down Similar process was used in the development of SWEL 2 as for SWEL 1. SWEL 2 for the Beaver Valley Power Station, Unit-2 included 10 components.

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==

Conclusion:==

No major concerns were identified by the peer review team in the selection process for SWEL 1 or SWEL 2.230 Peer Review Checklist for SWEL Instructions for Completing Checklist This peer review checklist may be used to document the review of the Seismic Walkdown Equipment List (SWEL)in accordance with Section 6: Peer Review. The space below each question in this checklist should be used to describe any findings identified during the peer review process and how the SWEL may have changed to address those findings.

Additional space is provided at the end of this checklist for documenting other comments.1. Were the five safety functions adequately represented in the SWEL I selection?

Y [N[-]See Attached Comments 2. Does SWEL 1 include an appropriate representation of items having the following sample selection attributes:

a. Various types of systems? Y [NI-See Attached Comments b. Major new and replacement equiipment?

Y ONE]See Attached Comments c. Various types of equipment?

See Attached Comments d. Various environments?

See Attached Comments e. Equipment enhanced based on the findings of the IPEEE (or equivalent) program?See Attached Comments f. Were risk insights considered in the development of SWEL 1?Y NF--" ONLI" ONF Y ENE See Attached Comments 231 Peer Review Checklist for SWEL 3. For SWEL 2: a. Were spent fuel pool related items considered, and if applicable included in SWEL 2?See Attached Comments b. Was an appropriate justification documented for spent fuel pool related items not included in SWEL 2?Y N[]Y OND-See Attached Comments 4. Provide any other comments related to the peer review of the SWELs.See Attached Comments 5. Have all peer review comments been adequately addressed in the final SWEL?Peer Revieu .Date.Y ONE](C)Peer Reviewer #2: Date: 1 0 [ % / z 232 Peer Review Checklist for SWEL Comments on Question 1: A peer review of the SWEL selected for the Beaver Valley Power Station, Unit-2 was performed to confirm that the selected components met the criteria set forth in Section 3 of EPRI Guidance Document 1025286. Specifically, Screen 3 calls out for assuring that the selected components represent are well associated with the five safety functions that are as follows: A. Reactor Reactivity Control B. Reactor Coolant Pressure Control C. Reactor Coolant Inventory Control D. Decay Heat Removal E. Containment Function The selected components represent the five safety functions stated above. A spreadsheet (Table 4-1) was prepared that documents this information.

Comments on Question 2a: The selected components represent various types of systems in the plant as indicated below: A. Primary Plant Component Cooling Water B. Chemical and Volume Control System C. 125V DC Power D. Containment Vacuum and Leakage Monitoring System E. Reactor Plant Vents and Drains F. 4 KV Station Service System G. Steam Generator Feedwater System H. Air Ventilation System Misc I. 120V AC Power J. Main Steam System K. Plant Process Control System L. 480V AC Power M. Containment Depressurization System N. Reactor Coolant System 0. Safety Injection System P. Service Water System Q. Residual Heat Removal R. Reactor Control and Protection System 233 S. In Core Instrumentation System Comments on Question 2b: The selected components represent many new and replacement equipment based-on the following modifications:

A. ECP 12-0242-001:

Replace Heat Exchanger B. ECP 11-0165-001:

Air Tube Replacement C. ECP 08-0504-025:

Replace Stem/Spline Key D. ECP 07-0259-003:

Motor Adapter Plate Modification E. ECP 05-009-001:

Replace Battery Charger F. ECP 05-009-003:

Replace Battery Charger G. ECP 02-0902: Replace Feedwater Control Valve Comments on Question 2c: The peer review concluded that the selected components represent various type of equipment installed in the plant. The various equipment types are indicated as follows: A. Tanks and Heat Exchangers B. Low Voltage Switchgear C. Medium Voltage Switchgear D. Battery Racks E. Battery Chargers and Inverters F. Horizontal Pumps G. Distribution Panels H. Engine Generators I. Pneumatic Operated Valves J. Check and Manual Valves K. Instrument on Racks L. Motor Control Centers M. Motor Operated Valves N. Solenoid Valves 0. Vertical Pumps P. Instrument and Control Panels Q. Transformers R. Fans S. Temperature Sensors T. Air Handlers 234 Comments on Ouestion 2d: The selected components are located in various types of environments found in the plant. The various plant environment types are as follows: A. Warm B. Damp C. Hot.D. Cool E. Dry F. Humid/Dry G. High Radiation Comments on Question 2e: Based on the review, the selected components represent equipment enhanced based on findings of the IPEEE.Comments on Ouestion 2f: The risk insights were considered in the development of SWEL 1. Specifically, Risk Achievement Worth (RAW) and Fussel-Vessley (FV) were considered.

Comments on Ouestion 3a: Spent Fuel Pool related items were considered and are adequately represented in SWEL 2 Comments on Ouestion 3b: Spent Fuel Pool components were considered.

Comments on Question 4: The peer review concluded that the selection of Seismic Walkdown Equipment List (SWEL) was performed in accordance with guidance provided in Section 3 of EPRI Document 1025286, 235 "Seismic Walkdown Guidance." Also, an ex-SRO from the Beaver Valley Power Station, Unit-2 acted as Operations representative during the selection of the SWEL.B. Review of a sample of the checklists prepared for the Seismic Walkdowns and Area Walk-Bys EPRI Document 1025286 on Seismic Walkdown Guidance required a review of the sample of the checklists prepared for the seismic walkdowns and area walk-bys by the peer reviewers.

The sample review should be between 10 percent and 25 percent.The following comments were identified during the early stages of peer review and were successfully resolved: A. In some cases, statements regarding minor anomalies (not resulting in a condition report)identified during the walkdowns did not have adequate justification for acceptability in meeting the design basis requirements.

B. In some cases, missing documentation/referenceS/checkmarks.

C. In some cases, minor anomaly stated but no justification provided.D. Editorial and typographical errors The above comments were discussed with the Seismic Walkdown Engineers (SWEs) and were successfully resolved in the final signed version of the checklists.

In addition, the peer reviewers also participated in a sample of walkdowns and observed the work performed by the SWEs during the inspections.

It was noted that the walkdown/inspection was intrusive, walkdown team members discussed, issues amongst themselves, and used engineering judgment in making decisions about whether there is any concern that should be noted. In some cases, the lead peer reviewer requested additional photographs.

The lead peer reviewer interviewed the SWEs to verify they followed the guidance in Section 4 of the EPRI Document "Seismic Walkdowns and Area Walk-Bys." The interview concluded that they did follow the said guidance and were knowledgeable about the walkdown requirements.

Questions asked were successfully answered during the interview as well as during the walkdowns.

236 Four SWEs participated in the walkdowns.

See their resumes for experience and background training.Conclusion:

The seismic walkdown and area walk-by checklists were completed in accordance with the guidance of EPRI Document 1025286 and no major issues were identified.

All comments were successfully resolved.

Adequate documentation has been provided in the checklists for the components that were walked down.C. Review of the Licensing Basis Evaluations The walkdowns identified several minor anomalies, however 12 of them resulted in generating condition reports as follows: CR-2012-14408, CR-2012-14409, CR-2012-14412, CR-2012-14420, CR-2012-14449, CR-2012-14450, CR-2012-14452, CR-2012-14455, CR-2012-14459, CR-2012-14463, CR-2012-14744, and CR-2012-14749.

Additionally, a thirteenth condition report was written to capture all the issues identified above in one condition report (CR-2012-14758).

1. CR-2012-14408 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that two out of twelve nuts for the bolts anchoring the Service Water Pump 2SWS *P21A base to the anchor plate are corroded.

This pump is located inside the Intake Structure at elevation 705'Cubicle D.Observation concluded that there is only surface rust on the bolts and the nuts. Even though they are corroded, they are capable to perform their intended design function based on engineering judgment.

The two nuts are located near each other are readily visible as mildly corroded.

No calculations or drawings are affected since the nuts are able to perform the intended function.To avoid further degradation of the nuts due to presence of moist environment, the nuts will be replaced in the next system window and painted to mitigate corrosion.

Initiated Notification No.600786691 to perform the work to be done under Work Order 200530757.

237

2. CR-2012-14409 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that a Deficient Tag (Tag ID 117472) for Component 2SWS-MOV1 13A has been in place with no date on it and no physical work performed to correct the deficiency.

This motor operated valve is located inside the Diesel Building (DGBX) at elevation 732'. The deficient tag says "Leakage in Packing." This leakage has also resulted in the corrosion of packing nuts.Observation concluded that there is only surface rust on the bolts and the nuts. Even though they are corroded, they are capable to perform their intended design function based on engineering judgment.

No calculations or drawings are affected since the nuts are able to perform the intended function.To avoid further degradation of the nuts, the nuts will be replaced in the next system window and painted to mitigate corrosion.

Initiated Notification No. 600786692 to perform the work to be done under Work Order 200481347.

3. CR-2012-14412 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that a Deficient Tag (Tag ID 63005) for Component 2SWM-MOV562 has been in place since 2007 and no physical work performed to correct the deficiency.

This motor operated valve is located inside the Valve Pit Room at elevation 718'. The deficient tag says "Valve has rust and needs to be cleaned/painted." Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve is able to perform the intended function.To avoid further degradation of the valve due to corrosion, the corrosion will be cleaned/painted during the next system window. Initiated Notification 600786690 to perform the work to be done under Work Order 200416834.

4. CR-2012-14420 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that a 55 gallon drum was located too close to MCC-2-E 11 and was unrestrained.

The component is located at El 737' in U-2 Safeguard Building.

This posed a seismic interaction concern that 238 during a seismic event it had the potential to hit the MCC and potentially impact the design function of inside components.

This condition has been corrected and this CR was generated to document an existing condition that was identified during the seismic walkdowns.

Currently no anomaly exists as such there are no operability concerns, and thus no violations of the design basis requirements.

5. CR-2012-14449 It was observed that Component 2SWS-MOV160 had a Deficient Tag (Tag ID 47014) and no physical work performed to correct the deficiency.

This motor operated valve is located inside the MSCV Room at elevation 718'. The deficient tag says "Surface rust to be cleaned and preserved." Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since surface rust doesn't affect the function of the valve.Initiated field work per Notification No. 600786757 to avoid further degradation of valve. Work will be done under Work Order 200531116.

6. CR-2012-14450 It was observed that Component 2SWS-MOV162 has corrosion on the valve from yoke to the bonnet. This motor operated valve is located inside the MSCV Room at elevation 718'.Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve is able to perform the intended function.Initiated Notification No. 600786758.

Work will be done under Work Order 200531117.

7. CR-2012-14452 It was observed that Component 2SWS-MOV152-1 had a Deficient Tag (Tag ID 47016) and no physical work performed to correct the deficiency.

Corrosion extends from yoke to bonnet. This motor operated valve is located inside the MSCV Room at elevation 718'. The deficient tag says"Corrosion (Surface Rust) on the valve body at packing gland." 239 Observation concluded that there is only surface rust and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve is able to perform the intended function.Initiated Notification No. 600786761 to perform the work. Work will be done under Work Order 200531158.

8. CR-2012-14455 It was observed that the four bolts and nuts connecting the yoke to the bonnet for Component 2SWS-MOV161 are mildly corroded.

This motor operated valve is located inside the MSCV Room at elevation 718'.Observation concluded that there is only surface rust on the bolts as well as the nuts and the valve is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since they are able to perform the intended function.Initiated Notification No. 600786762 to perform the work. Work will be done under Work Order 200531159.

9. CR-2012-14459 It was observed that the flange nuts at outlet nozzle for Component 2CCP-E21IC are mildly corroded.

This heat exchanger is located at elevation 710' in Aux Building.Observation concluded that there is only surface rust on the nuts and the component is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since they are able to perform the intended function.Initiated Notification No. 600786796 to perform the work. Work Will be done under Work Order 200531161.

10. CR-2012-14463 There is a potential seismic interaction concern associated with Panel PNL-SEQ-244 located at El. 730' in Service Building, Emergency Switchgear AE. Specifically, right near the top of this panel, there are lighting fixtures supported from unistruts hung from chains above. During a seismic event, there is a potential for these lighting fixtures to rattle and hit the top of safety related panel.240 Observation concluded that the impact force from these lighting fixtures on the panel, considering the weight of lighting fixtures and unistrut, will have insignificant impact on the components inside the panel and they will perform their intended design function, by engineering judgment.

However, this CR is being generated as an enhancement to consider upgrading the condition and restraining the unistrut supporting the lighting fixture such that it would not hit the panel.No calculations or drawings are affected since the components are able to perform the intended function safely.11. CR-2012-14744 The enclosure and the anchorage for Component 2QSS-LT104A is corroded.

This component is located outside in the yard near RWST 2QSS-TK21 at elevation 734'. The corrosion does not affect the structural integrity so drawings and calculations still apply. The design basis documents have not been violated.Initiated Notification No. 600788282 to perform the work. Work will be done under Work Order 200532069.

12. CR-2012-14749 The yoke for the manual valve 2QSS-297 is corroded.

This component is located outside in the yard near RWST 2QSS-TK21 at elevation 734'.Initiated Notification No. 600788283 to perform the work, under Work Order 200532070.

==

Conclusion:==

The licensing basis evaluations as documented in Section 7 of this report were reviewed.

In summary, they have been adequately evaluated against the design basis.requirements, the corrective actions taken are adequate, and no further action is required.241 D. Review of the decisions for entering the potentially adverse conditions into the CAP Process Section 6 of this report discusses the summary of walkdown results. Specifically, Section 6.2.1 discusses seismic walkdown findings associated with SWEL 1, and Section 6.2.2 discusses seismic walkdown findings associated with area walk-bys.

The potentially adverse conditions were documented in Tables 6-4 and 6-5 in accordance with EPRI Document 1025286 and titled as "Potentially Adverse Seismic Conditions Identified from Component and Area Walk-Bys." Table 6-4 identified seven potentially adverse seismic conditions, which resulted in generating six condition reports. Adequate justification is documented in the checklist that provides the basis as why the remaining issue had insignificant impact on the design of the components and that the component is still capable of performing its intended design function while still meeting the design basis requirements.

Table 6-5 identified seven potentially-adverse seismic conditions.

Six of these conditions were entered in the corrective action program (CAP). Again, adequate-justification is documented in the checklists that provide the basis as why the remaining issue had insignificant impact on the design of the surrounding components and that the component is still capable of performing its intended design function while still meeting the design basis requirements.

A review of the basis documented in the checklists for not entering these issues in the CAP concluded the decisions taken were appropriate.

==

Conclusion:==

The peer reviewers agree with the decisions taken for entering or not entering the identified potentially seismic walkdown findings in the corrective action program.E. Review of the Submittal Report

Conclusion:

A team of reviewers performed a review of this submittal report. Comments were successfully resolved.

Refer to the signature page for a listing of reviewers.

F. Summary of results of peer review process 242

==

Conclusion:==

The selected samples (SWEL 1 and SWEL 2) adequately represent and meet the criteria set forth in the selection process outlined in EPRI Document 1025286. An Operations person also participated in the sample selection process and the walkdowns.

The peer reviewers participated in sample walkdowns, observed the conduct of walkdown team members, and discussed issues while remaining independent.

The Seismic Walkdown Checklists (SWCs) and Area Walk-by Checklists (AWCs) were adequately prepared and the basis for justifications appropriatly documented.

The decisions taken to enter the findings or not to enter the findings into the CAP were appropriate.

Also, the resolution of the issues (License Basis Evaluations) identified in the condition reports was adequate.243

10.0 REFERENCES

1. NRC letter 50.54(f), March 17, 2012.2. EPRI 1025286, "Seismic Walkdown Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic," Final, June 2012.3. "Beaver Valley Power Station Unit 2: Probabilistic Risk Assessment Update Report", Issue 5A, August 31, 2012, First Energy Nuclear Operating Company.4. "Beaver Valley Unit 2 Probabilistic Risk Assessment, Individual Plants Examination of External Events", Submitted September, 1997 in response to U.S. Nuclear Regulatory Commission Generic Letter 88-20 Supplement 4, Duquesne Light Company.5. "Beaver Valley Power Station Unit 2, Spent Fuel Pool Cooling Trouble", Abnormal Operating procedure 20M-53C.4.2.20.1, Revision 1, November 16, 2011.6. "Beaver Valley Power Station Unit 2, Updated Final Safety Analysis Report", Revision 19, Section 9.1.3.2.7. RG 1.29, Rev. 3, "Seismic Design Classification." 8. RG 1.60, Rev. 1, "Design Response Spectra for Seismic Design of Nuclear Power Plants." 9. RG 1.61, "Damping Values for Seismic Design of Nuclear Power Plants." 10. RG 1.100, "Seismic Qualification of Electrical Equipment for Nuclear Power Plants." 11. IEEE 344-1975, Rev. 1, "IEEE Guide for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations." 12. ASME Boiler and Pressure Vessel Code Section III 1974 including Winter Addenda 1975.244 APPENDIX A RESUMES AND QUALIFICATIONS A-1 279 Dorchester Rd, Phone 234-678-8262 Akron Ohio 44313 E-mail jreddman@aol.com JOHN E. REDDINGTON Work experience January 2007 to present: Principal Consultant, Probabilistic Risk Analysis:

Lead fire PRA for the Davis-Besse fire PRA, including contractor oversight and coordination; specialization in HRA, including operations interface, model integration, dependency analysis and PWROG HRA Subcommittee; fire PRA peer reviews; currently technical lead for seismic PRA for FENOC fleet; mentor to junior and co-op engineers.

August 2004- January 2007: Principal Programs Engineer, Fleet office Akron, OH: responsible for the fire protection program for the FENOC fleet August 2003 to August 2004: Davis-Besse Nuclear Station Oak Harbor, OH Training Manager: Responsible for direction and implementation of site's accredited training programs.

Heavily involved with high intensity training required to get Davis-Besse back on line following a two year outage replacing the reactor head.January 2001 to August 2003 : Davis-Besse Nuclear Station Oak Harbor, OH Supervisor Quality Assurance Oversight for Maintenance:

Responsible for value added assessments based on performance as well as compliance.

Ensure industry best practices are used as standards for performance in-maintenance, outage planning, and scheduling.

1996 to January 2001, Superintendent Mechanical Maintenance Manage the short and long term direction of the Mechanical and Services Maintenance Departments.

Responsible for 8o to 90 person department with a budget between 7 and 15 million dollars a year. Direct the planning, engineering, and field maintenance activities.

Direct oversight of outage preparations and implementation.

One year assignment working with Technical Skills Training preparing for accreditation.

A-2 1993 -1996 Shift Manager Act as the on-shift representative of the Plant Manager. Responsible for providing continuous -management support for all Station activities to ensure safe and efficient plant operation.

Establish short term objectives for plant control and provide recommendations to the Shift Supervisor.

Monitor core reactivity and thermal hydraulic performance, containment isolation capability, and plant radiological conditions during transients and advise the operating crew on the actions required to maintain adequate shutdown margin, core cooling capability, and minimize radiological releases.1991 -1993 Senior System and Maintenance Engineer Provide Operations with system specific technical expertise.

Responsible for maintaining and optimizing the extraction steam and feedwater heaters, the fuel handling equipment and all station cranes.Acted as Fuel Handling Director during refueling outages.Responsibilities Included maintaining the safe and analyzed core configuration, directing operation personnel on fuel moves, directing maintenance personnel on equipment repair and preventative maintenance.

1986 -1991 Senior Design Engineer and Senior Reactor Operator student Activities included modification design work and plant representative on the Seismic Qualification Utilities Group and the Seismic Issues subcommittee.

Licensed as a Senior Reactor Operator following extensive classroom, simulator, shift training, and Nuclear Regulatory Commission examination.

1984 -1986 Sargent & Lundy Engineers Chicago, IL Senior Structural Engineer Responsible for a design team of engineers for the steel design and layout to support the addition of three baghouses on a coal fired plant in Texas.Investigated and prepared both remedial and long term solutions to structural problems associated with a hot side precipitator.

198o-1984 Structural Engineer Responsible for steel and concrete design and analysis for LaSalle and Fermi Nuclear Power plants. Performed vibrational load and stability analysis for numerous piping systems. Member of the on-site team of engineers responsible for timely in-place modifications to the plant structure at LaSalle.1979 -198o Wagner Martin Mechanical Contractors Richmond, IN Engineer/Project Manager Responsible for sprinlder system design through approval by appropriate underwriter.

Estimator and Project Manager on numerous mechanical projects up to 1.8 million dollars.A-3 Education 1975 -1979 Purdue University Bachelor of Science in Civil Engineering 1990- 1995 University of Cincinnati Master of Science in Nuclear Engineering West Lafayette, IN Cincinnati, OH Professional memberships Professional Engineer, State of Illinois, 1984 Professional Engineer, State of Ohio, 1986 Senior Reactor Operator, Davis-Besse Nuclear Power Plant, 1990 Qualified Lead Auditor, 2003 SQUG qualified 1987 Committee Chairman, Young Life Toledo Southside, Lake Erie West Region Sunday School Teacher- College age young people.Other A-4 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E.PROFESSIONAL HISTORY ABSG Consulting Inc., Oakland, California Senior Consultant, 2004-Present Technical Manager, 2001-2004 EQE International, Principal Engineer, 1990-2001 TENERA L.P., Berkeley, California, Project Manager, 1982-1990 PROFESSIONAL EXPERIENCE Mr. Beigi has more than 29 years of professional structural and civil engineering experience.

As a Senior Consultant for ABS Consulting, Mr. Beigi provides project management and structural engineering services, primarily for seismic evaluation projects.

He has extensive experience in the areas of seismic evaluation of structures, equipment, piping, seismic criteria development, and structural analysis and design. Selected project accomplishments include the following:

• Most recently, Mr. Beigi has been involved in performing seismic fragility analysis of equipment and structures at G6sgen Nuclear Power Plant in Switzerland, Lungmen Nuclear Power Plant in Taiwan, Oconee Nuclear station in U.S., Point Lepreau Nuclear Plant in Canada, Beznau Nuclear Power Plant in Switzerland, Olkiluoto Nuclear Power Plant in Finland, and Neckarwestheim Nuclear Power Station in Germany.* Provided new MOV seismic qualification (weak link) reports, for North Anna, Surry and Kewaunee nuclear plants to maximize the valve structural thrust capacity by eliminating conservatisms found in existing qualification reports and previously used criteria." At Salem Nuclear Power Plant Mr. Beigi developed design verification criteria for seismic adequacy of HVAC duct systems. He also performed field verification of as-installed HVAC systems and provided engineering evaluations documenting seismic adequacy of these systems, which included dynamic analyses of selected worst-case bounding samples.* Mr. Beigi has participated in several piping adequacy verification programs for nuclear power plants. At Watts Bar and Bellefonte Nuclear Plants, he was involved in the development of walkdown and evaluation criteria for seismic evaluation of small bore piping and participated in plant walkdowns and performed piping stress analyses.

At Oconee Nuclear Station, Mr. Beigi was involved in developing screening and evaluation criteria for seismic adequacy verification of service water piping system and performed walkdown evaluations, as well as, piping stress analyses.

At Browns Ferry Nuclear Plant, Mr. Beigi was involved in the assessment of seismic interaction evaluation program for large and small bore piping systems.A-5 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E.* Mr. Beigi performed a study for the structural adequacy of bridge cranes at DOE's Paducah Gaseous Diffusion Plant utilizing Drain-2DX non-linear structural program. The study focused on the vulnerabilities of these cranes as demonstrated in the past earthquakes.

• Mr. Beigi has generated simplified models of structures for facilities at Los Alamos National Lab and Cooper Nuclear Station for use in development of building response spectra considering the effects of soil-structure-interactions.

• Mr. Beigi has participated as a Seismic Capability Engineer in resolution of the US NRC's Unresolved Safety Issue A-46 (i.e., Seismic Qualification of Equipment) and has performed Seismic Margin Assessment at the Browns Ferry Nuclear Power Plant (TVA), Oconee Nuclear Plant (Duke Power Co.), Duane Arnold Energy Center (Iowa Electric Company), Calvert Cliffs Nuclear Power Plant (Baltimore Gas and Electric), Robinson Nuclear Power Plant (Carolina Power & Light), and Bruce Power Plant (British Energy -Ontario, Canada).He has performed extensive fragility studies of the equipment and components in the switchyard at the Oconee Nuclear Power Plant.* Mr. Beigi has developed standards for design of distributive systems to be utilized in the new generation of Light Water Reactor (LWR) power plants. These standards are based on the seismic experience database, testing results, and analytical methods.* Mr. Beigi managed EQE's on-site office at the Tennessee Valley Authority Watts Bar Nuclear Power Plant. His responsibilities included staff supervision and technical oversight for closure of seismic systems interaction issues in support of the Watts Bar start-up schedule.

Interaction issues that related to qualification for Category I piping systems and other plant features included seismic and thermal proximity issues, structural failure and falling of non-seismic Category I commodities, flexibility of piping systems crossing between adjacent building structures, and seismic-induced spray and flooding concerns.Mr. Beigi utilized seismic experience data coupled with analytical methods to address these seismic issues.* As a principal engineer, Mr. Beigi conducted the seismic qualification of electrical raceway supports at the Watts Bar Plant. The qualification method involved in-plant walkdown screening evaluations and bounding analysis of critical case samples. The acceptance criteria for the bounding analyses utilized ductility-based criteria to ensure consistent design margins. Mr. Beigi also provided conceptual design modifications and assisted in the assessment of the constructability of these modifications.

Mr. Beigi utilized similar methods for qualification of HVAC ducts and supports at Watts Bar, and assisted criteria and procedures development for HVAC ducting, cable trays, conduit and supports at the TVA Bellefonte nuclear power plant.Mr. Beigi also has extensive experience utilizing finite element computer codes in performing design and analysis of heavy industrial structures, systems, and components.

At the Texas Utility Comanche Peak Nuclear Power Plant, Mr. Beigi administered and scheduled individuals to execute design reviews of cable tray supports; evaluated generic design criteria for the design and construction of nuclear power plant systems and components and authored engineering evaluations documenting these reviews.A-6 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E.Mr. Beigi has also been involved in a number of seismic risk assessment and equipment strengthening programs for high tech industry, biotech industry, petrochemical plants and refineries, and industrial facilities.

Selected project accomplishments include: Most recently performed Seismic Qualification of Critical Equipment for the Standby Diesel Power Plants Serving Fort Greely, and Clear Air Force Station, Alaska. Projects also included design of seismic restraints for the equipment and design of seismic supports for conduit, cable tray, duct, and piping systems. Both facilities are designated by the Department of Defense as a Seismic User Group Four (SUG-IV) facility.

Seismic qualification of equipment and interconnections (conduit, duct and piping) involved a combination of stress computations, compilation of shake table data and the application of experience data from past earthquakes.

Substantial cost savings were achieved by maximum application of the experience data procedures for seismic qualification." Assessment of earthquake risk for Genentech, Inc., in South San Francisco, CA. The risk assessments included damage to building structures and their contents, damage to regional utilities required for Genentech operation, and estimates of the period of business interruption following a major earthquake.

Provided recommendations for building or equipment upgrades or emergency procedures, with comparisons of the cost benefit of the risk reduction versus the cost of implementing the upgrade. Project included identification of equipment and piping systems that were vulnerable under seismic loading and design of retrofit for those components, as well as, providing construction management for installation phase of the project.* Fault-tree model and analysis of critical utility systems serving Space Systems / Loral, a satellite production facility, in Palo Alto, CA.* Seismic evaluation and design of retrofits for equipment, tools and process piping, as well as, clean room ceilings and raised floors at UMC FABs in Taiwan.* For LDS Church headquartered in Utah, performed seismic vulnerability assessment and ranked over 1,200 buildings of miscellaneous construction types for the purpose of retrofit prioritization." Seismic evaluation and design of retrofits for clean room ceilings at Intel facilities in Hillsborough, Oregon.* Assessment of programmable logic controls as part of year 2000 (Y2K) turn over evaluation at an automatic canning facility in Stanislaus, ca.* Seismic evaluation and design of retrofits for equipment and steel storage tanks at the Colgate-Palmolive plant in Cali, Colombia.* Design of seismic anchorage for equipment and fiberglass tanks at the AMP facilities in Shizouka, Japan." Evaluation and design of seismic retrofits for heavy equipment, and piping systems at Raychem facilities in Redwood City and Menlo Park, CA.* Assessment of the seismic adequacy of equipment, structures and storage tanks at the Borden Chemical Plant in Fremont, CA.A-7 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E.* Design of seismic bracing for fire protection and chilled water piping systems at the Goldman Sachs facilities in Tokyo, Japan.* Design of seismic retrofits for low rise concrete and steel buildings and design of equipment strengthening schemes at AVON Products Co. in Japan.0 Managed the design and construction of seismic retrofits for production equipment and storage tanks at Coca Cola Co. in Japan.0 Seismic evaluation and design of retrofit for equipment, piping and structures at the UDS AVON Refinery located in Richmond, CA.0 Seismic assessment and peer review of the IBM Plaza Building, a 31 story high rise building located in the Philippines.

• Seismic evaluation and conceptual retrofit design for the headquarters building of the San Francisco Fire Department.

0 Equipment strengthening and detailed retrofit design for the Bank of America Building in San Francisco.

• Equipment strengthening and detailed retrofit design for Sutro Tower in San Francisco.

0 Equipment strengthening and detailed retrofit design for Pacific Gas & Electric (PG&E)substations in the San Francisco area.* Seismic evaluations and loss estimates (damage and business interruption) for numerous facilities in Japan, including Baxter Pharmaceuticals, NCR Japan Ltd., and Somar Corporation.

Seismic evaluation of concrete and steel buildings at St. Joseph Hospital in Stockton, Ca, in accordance with the guidelines provided in FEMA 178.EDUCATION B.S., Civil Engineering, San Francisco State University, San Francisco, CA, 1982 REGISTRATION Professional Engineer:

California Seismic Qualification Utilities Group Certified Seismic Capability Engineer Training on Near Term Task Force Recommendation 2.3 -Plant Seismic Walkdowns AFFILIATIONS American Society of Civil Engineers, Professional Member SELECTED PUBLICATIONS M. Richner, Sener Tinic, M. Ravindra, R. Campbell, F. Beigi, and A. Asfura, "Insights Gained from the Beznau Seismic PSA Including Level 2 Considerations," American Nuclear Society PSA 2008, Knoxville, Tennessee.

A-8 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E.U. Klapp, F.R. Beigi, W. Tong, A. Strohm, and W. Schwarz, ,Seismic PSA of Neckarwestheim 1 Nuclear Power Plant," 19th International Conference on Structural Mechanics in Reactor Technology (SMIRT 19), Toranto, Canada, August 12-17, 2007.A. P. Asfura, F.R. Beigi and B. N. Sumodobila.

2003. "Dynamic Analysis of Large Steel Tanks." 17th International Conference on Structural Mechanics in Reactor Technology (SMIRT 17), Prague, Czech Republic, August 17-22, 2003."Seismic Evaluation Guidelines for HVAC Duct and Damper Systems," April 2003. EPRI Technical Report 1007896. Published by the Electric Power Research Institute.

Arros, J, and Beigi, F., "Seismic Design of HVAC Ducts based on Experienced Data." Current Issues Related to Nuclear Plant Structures, Equipment and Piping, proc. Of the 6th Symposium, Florida, December 1996. Publ. by North Carolina State University, 1996.F.R. Beigi and J. 0. Dizon. 1995. "Application of Seismic Experience Based Criteria for Safety Related HVAC Duct System Evaluation." Fifth DOE Natural Phenomenon Hazards Mitigation Symposium.

Denver, Colorado, November 13-14, 1995.F.R. Beigi and Don R. Denton. 1995. "Evaluation of Bridge Cranes Using Earthquake Experience Data." Presented at Fifth DOE Natural Phenomenon Hazards Mitigation Symposium.

Denver, Colorado, November 13-14, 1995.A-9 ABS Consulting AN ABS GROUP COMPANY EDDIE M. GUERRA, E.I.T.PROFESSIONAL HISTORY ABSG Consulting Inc., Contractor, Presently Paul C. Rizzo Associates, Inc., Pittsburgh, PA, Assistant Project Engineering Associate, Presently Thornton Tomasetti, Inc., Philadelphia, PA, Structural Engineer Intern, January 2009-June 2009 Skanska USA, Inc., San Juan, Puerto Rico, Civil Engineering Intern, May 2008-July 2008 Network for Earthquake Engineering Simulation, Bethlehem, PA, Research Assistant, May 2007-July 2007 PROFESSIONAL

SUMMARY

Mr. Eddie M. Guerra, E.I.T. is an Assistant Project Engineering Associate with Paul C. Rizzo Associates, Inc. (RIZZO). Mr. Guerra has been involved primarily in the structural design and analysis of power generation structures in both nuclear and wind energy sectors. Mr. Guerra specializes in structural dynamics, Performance Based Seismic Design methodologies and elastic and inelastic behavior of concrete and steel structures.

He is fluent in both English and Spanish.PROFESSIONAL EXPERIENCE Nuclear: AP1000 HVAC Duct System Seismic Qualification

-October 2010 -Present SSM/ Westinghouse Electric Company, Pittsburgh, Pennsylvania:

Engineer for the seismic qualification of AP1000 HVAC Duct System.Structural dynamic analysis of all mayor steel platforms inside steel containment vessel.Investigation on the interaction of steel vessel and HVAC system displacements due to normal operational and severe thermal effects.Finite element modeling of HVAC access doors under static equivalent seismic loads.Followed AISC, ASCE and SMACNA standards for the qualification of steel duct supports.A-10 ABS Consulting AN ABS GROUP COMPANY EDDIE M. GUERRA, E.I.T.Wind: Analysis and Design Revision of Wind Turbine Tower -October 2010 -Februanj 2011 Siemens, Santa Isabel, Puerto Rico: Engineer for the analysis and design revision of a wind turbine tower to be constructed in Santa Isabel, Puerto Rico.Developed design criteria based on local building code requirements and the International Electrotechnical Commission (IEC) provisions for wind turbine design.Dynamic analysis of wind turbine.Design revision of turbine tower shell, bolted flange connections and global stability of the tower.EDUCATION M. Eng., Structural Engineering, Lehigh University, Bethlehem, PA -May 2010 B.S., Civil Engineering, University of Puerto Rico, Mayaguez, PR -Dec. 2008 SKILL AREAS Structural Analysis Seismic Design Reinforced Concrete Design Structural Steel Design Wind Aerodynamics Wind Turbine Design Plastic Steel Design Foundation Design COMPUTER SKILLS STAAD, ANSYS, AutoCAD, ADAPT, SAP2000, RAM, MATHCAD, PCA Column, MS Office REGISTRATIONS Engineer-In-Training:

Puerto Rico -2009 MEMBERSHIPS American Society of Civil Engineers (ASCE)American Concrete Institute (ACI)Network for Earthquake and Engineering Simulation (NEES)U.S. Dept. of Labor (OSHA)Society of Hispanic Professional Engineers (SHPE)A-1I ABS Consulting AN ABS GROUP COMPANY EDDIE M. GUERRA, E.I.T.HONORS AND AWARDS 2010 Recipient of the Thornton Tomasetti Foundation Scholarship Golden Key International Honor Society Tau Beta Pi Engineering Honor Society University of Puerto Rico at Mayaguez Dean's List PUBLICATIONS Guerra, Eddie M., "Impact Analysis of a Self-Centered Steel Concentrically Braced Frame," NEES Consortium, May -July 2007.A-12 ABS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.PROFESSIONAL HISTORY ABSG Consulting Inc., Contractor, Presently Paul C. Rizzo Associates Inc., Pittsburg, PA, Engineering Associate II, 2010- Present Engineers without Borders, Aquaculture Development, Makili, Mali, Africa September 2007 -December 2009, Southwestern Pennsylvania Commission, Pittsburgh, Pennsylvania, Transportation Intern, May 2008 -August 2008 PROFESSIONAL

SUMMARY

Mr. Lucarelli has experience providing engineering support for a number of domestic and international nuclear power plants. He has also completed RIZZO's in-house training course on NTTF 2.3 Seismic Walkdowns.

This course was delivered by RIZZO's senior staff that had completed the two day course.PROFESSIONAL EXPERIENCE February 2012 -July 2012 Vogtle NPP Units 3 and 4 -Westinghouse Electric Company, Burke County, Georgia: RIZZO conducted a settlement analysis to predict the total and differential settlements expected during construction of the Vogtle Units 3 and 4. Mr. Lucarelli was responsible for reviewing on-site heave and settlement data and the excavation sequence to calibrate the material properties in the settlement model. He was also responsible for creating a settlement model that implemented the expected AP1000 construction sequence and presenting the results in a report.January 2010 -June 2012 Levy County NPP Foundation Considerations

-Sargent & Lundy/Progress Energy, Crystal River, Florida: Mr. Lucarelli was extensively involved in the design and specification of the Roller Compacted Concrete Bridging Mat that will support the Nuclear Island foundation.

He has authored numerous calculations and reports related to the work conducted for this project, including responding to requests for additional information from the NRC. His analyses for this project included finite element analyses of the stresses within the Bridging Mat under static and dynamic loading and the determination of long-term settlement at the site.A-13 ABS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.Mr. Lucarelli also authored the Work Plan and served as on-site quality control during laboratory testing of RCC block samples in direct tension and biaxial direct shear. His responsibilities included inspection of the testing being performed and control of documentation related to testing activities.

September 2011 -March 2012 Akkuyu NPP Site Investigation

-WorleyParsons/Akkuyu Project Company, Mersin Province, Turkey: RIZZO conducted a geotechnical and hydrogeological investigation of the proposed site for four VVER-1200 reactors.

This investigation entailed geotechnical and hydrogeological drilling and sampling, geophysical testing, and geologic mapping. Mr. Lucarelli served as on-site quality control for this project. His responsibilities included controlling all records generated on site, interfacing with TAEK (Turkish Regulatory Agency) auditors, and tracking nonconformances observed during the field investigation.

Mr. Lucarelli also assisted in the preparation of the report summarizing the findings of the field investigation.

May 2010 -November 2010; July 2011 -January 2012 Calvert Cliffs NPP Unit 3 -Unistar, Calvert County, Maryland: RIZZO completed a COLA-level design of the Ultimate Heat Sink Makeup Water Intake Structure at the Calvert Cliffs site. Mr. Lucarelli authored and checked a number of calculations to determine the design loads to be used in a Finite Element model of the structure.

Mr.Lucarelli was also responsible for ensuring that the design met the requirements of the Design Control Document.Mr. Lucarelli has also performed a settlement analysis for the Makeup Water Intake Structure.

February 2010 -March 2010 C.W. Bill Young Regional Reservoir Forensic Investigation

-Confidential Client, Tampa, Florida: RIZZO conducted a forensic investigation into the cause of soil-cement cracking on the reservoir's upstream slope. This investigation involved a thorough review of construction testing results and documentation to determine inputs for seepage and slope stability analyses.Mr. Lucarelli reviewed construction documentation and conducted quality control checks on the data used for the analyses.

Mr. Lucarelli also prepared a number of drawings and figures that presented the results of the forensic investigation.

Previous Experience:

September 2007 -December 2009 Aquaculture Development

-Makili, Mali, Africa: The University of Pittsburgh Chapter of Engineers Without Borders designed and constructed an aquaculture pond in rural Mali, Africa with a capacity of 3.6 million gallons. This pond is designed to maintain enough water through a prolonged dry season to allow for year-round cultivation of tilapia. As the project technical lead, Mr. Lucarelli was involved in developing A- 14 ABS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.conceptual design alternatives and planning two site assessment trips. These scope of these site assessment trips included topographic surveying, the installation of climate monitoring instrumentation, soil sampling and characterization, and laboratory soils testing.As the project coordinator, his primary responsibilities included maintaining a project schedule, developing a budget for project implementation, and coordinating technical reviews of project documentation with a Technical Advisory Committee.

May 2008 -August 2008 Southwestern Pennsylvania Commission

-Pittsburgh, Pennsylvania:

As a transportation intern, Mr. Lucarelli analyzed data in support of various studies dealing with traffic forecasting, transit use, and highway use. He also completed fieldwork to assess the utilization of regional park-and-ride facilities.

EDUCATION B.S., Civil Engineering, University of Pittsburgh, Pittsburgh, PA, 2009 B.S., Mathematics, Waynesburg University, Waynesburg, PA, 2009 CONTINUING EDUCATION Short Course on Computational Geotechnics and Dynamics, August 2011 ASDSO Estimating Permeability Webinar, December 2010 COMPUTER SKILLS SAP2000, PLAXIS, SEEP/W, SLOPE/W, THERM, AutoCAD, ArcGIS, Phase2, Slide, MathCAD REGISTRATIONS Pennsylvania:

Engineer-in-Training

1. ET013562 MEMBERSHIPS American Concrete Institute (ACI)-ACI Committee 207 (Mass Concrete)

-Associate Member American Society of Civil Engineers (ASCE)Engineers Without Borders (EWB)A-15 Resume of Mohammed F. Alvi, P.E.

SUMMARY

" Thirty-three years of experience as an engineering professional (27 years in nuclear)" Professional Engineer, registered in the State of New York, USA 0 Completed the Boiling Water Reactor (BWR) Plant Certification Course for Nine Mile Point Unit-1 Nuclear Station" Experience as a Structural Design Engineer, Engineering Supervisor for Structural/Mechanical Design and Plant Support Engineering, Manager Mechanical/Structural Design and Project Manager" Innovative and resourceful engineer with problem solving skills" Excellent leadership skills with proven record" Excellent analytical, design, decision making, communication, organizational, and interpersonal skills" Proficient in computer skills EXPERIENCE:

June 2012 -First Energy Nuclear Operating Company Present Senior Consulting Engineer Project Manager for Seismic Probabilistic Risk Assessment (SPRA) Project. Responsibilities include vendor oversight for 50.54(f) Letter Seismic 2.1 and 2.3 as well as technical overview of the SPRA project.March 2008 -Entergy Nuclear Operations May 2012 James A. Fitzpatrick Nuclear Power Plant Oswego, New York Supervisor, Mechanical/Civil Design Engineering Responsible for supervising a group of 10 mechanical/civil/structural engineers at the James A. Fitzpatrick Nuclear Plant. Responsibilities included issuing plant modifications, evaluations, engineering changes, equivalency changes, supporting refueling and forced outages, acted as engineering duty manager, identified training needs, participated in the daily fleet telephone calls, resolved operability issues related to degraded conditions, assisted in resolving plant emergent issues, responded to US Nuclear Regulatory Commission (NRC) Resident questions, supported emergency response organization duties, etc. Oversight of construction activities, owner acceptance of A/E Consulting Firm design. Performed duties of acting design engineering manager, trained staff on technical/administrative skills, etc.February 2007 -Public Service Electricity

& Gas (PSEG) Nuclear A-16 February 2008 Hope Creek Nuclear Generating Station Branch Manager, Mechanical/Structural Design Responsible for managing a staff of 8 Mechanical/Structural engineers at Hope Creek Nuclear Generating Station. Responsibilities included analysis, design of Structures, Systems, Components, resolving operability issues, preparing design change packages, evaluating non-conforming conditions, addressing short and long term issues for the station, supporting outages, address training needs of the group, participate in Plant Health Committee, interface with resident NRC inspectors, etc.I was also responsible for performing the duties as the site reviewer of all Structural/Mechanical related license renewal documents being prepared by the License Renewal Group. I was implementing the Hope Creek primary containment (Drywell and Torus) ageing management program to support the license renewal process. I was also assisting in the implementation of FatiguePro software at Hope Creek.1988 -Oct. 2006 Nine Mile Point Nuclear Station (Constellation Nuclear)Oswego, New York Engineering Supervisor/Principal Engineer Responsible for analysis, design and maintenance of various nuclear power plant structures at Nine Mile Point Nuclear Station Units 1 & 2. Analysis includes design of reactor building superstructure, turbine building superstructure, yard structures, masonry wall design, piping analysis and supports for safety related systems, cable tray supports and various electrical and mechanical components supports, etc.Supervised a group of 10 engineers/designers, coordinated projects with site engineering consultants, performed engineering evaluations and cost benefit studies for various projects for an economical design.As one of the leaders of the engineering organization, I directed and supervised individuals technically and administratively to make sure the job is done correctly the first time and per schedule.

I had the decision making authority for all structural engineering issues at the station.License Renewal: I was also the Manager for Fatigue Monitoring Program for Nine Mile Point Nuclear Station, Units 1 & 2. I was involved in setting up the software "FatiguePro" at the station for a cost of $500K. This was in commitment to the Nuclear Regulatory Commission as part of License Renewal program for NMP station. This program included identifying the various transients that the plants were originally designed for, historical count of transients, identifying cumulative usage factors at critical locations, identifying what locations CUFs will be exceeded for a 60 year plant life and what actions were needed to resolve the same. Also addressed the environmental fatigue issues.A-17 I was also responsible for managing all structural aspects of license renewal program at the station. This included preparation of program basis documents (e.g., masonry walls, bolting, monitoring of structures, etc.), scoping documents, ageing management program documents, time limiting ageing analysis (TLAAs), performed walkdowns for defining boundaries.

I was also part of the design team that gave a presentation to NRC license renewal team at Rockville, MD regarding the primary containment ageing management program for torus and drywell shell thickness at Nine Mile Point Unit-1.Note: I was also the Nine Mile Point Nuclear Station Lead for the NRC Component Design Bases Inspection (CDBI) that was conducted in September/October 2006. I successfully lead the NMP team, supported the inspection with no major violations for the station. This project started in May 2006 which included self assessment (mock inspection), taking appropriate corrective actions prior to the actual inspection for a successful outcome.Acting Manager, Engineering Unit 1 Nine Mile Point Nuclear Station Performed the duties of an engineering manager, attended the daily leadership meetings, resolved the plant issues, prioritized and coordinated the work activities of various disciplines in Engineering, conducted branch staff and safety meetings, successfully resolved all engineering issues during this period for safe operation of the plant.Supervisor, Civil/Structural Engineering, Unit 1 Nine Mile Point Nuclear Station Responsible for all structural engineering issues at Nine Mile Point Unit.Major accomplishments as Structural Supervisor included implementation of Structural Maintenance Rule Program, development of various engineering specifications and drawings for the older vintage plant.Attended various structural seminars on Seismic Qualification Utility Group (SQUG), concrete and masonry walls, structural maintenance program, completed various training on leadership skills, supervisory skills, performance appraisals, effective communication, Labor training, Leadership Academy and completed two weeks of training at Institute of Nuclear Power Operations (INPO)-Atlanta for Engineering Supervisors Professional Development Seminar.1983 -1988 Sargent & Lundy Engineers Chicago, Illinois Lead Structural Engineer Responsible for analysis and design of various nuclear power plant structures using ACI and AISC codes, was responsible for designing pipe supports, conduit supports, pipe whip restraints, masonry walls, steel frames, used various in-house computer programs for analysis A-18 design, performed walk-downs, performed structural calculations, resolved non-conformance reports, performed seismic qualification calculations, etc.1978 -1983 Klein & Hoffman, Inc Consulting Engineers, Chicago, Illinois Structural Engineer Structural engineer responsible for analysis and design of schools, parking garages, industrial buildings, high rise buildings, sewage treatment plant structures, etc. Extensively used AISC and ACI codes and various in'house computer programs for analysis and design.EDUCATION:

• Master of Science (Structural Engineering), University of Illinois, Chicago (1977)* Bachelor of Engineering (Civil), Bhopal University, India (1976)PROFESSIONAL LICENSES/CERTIFICATIONS:

• Registered Professional Engineer, State of New York* Boiling Water Reactor (BWR) Plant Certification Course for Nine Mile Point Unit-I Nuclear Station PROFESSIONAL SOCIETY MEMBERSHIP:

• Member, American Society of Civil Engineers (ASCE)

REFERENCES:

Provided upon request CITIZENSHIP:

Citizen of the United States of America A-19 ABS Consulting DANIEL A. RENY PROIFISSION AL 141STORY ABSG Consulting Inc, Irvine, CA, Senior Consultant, 200-Present ARES Corporation, Los Angeles, CA, Senior Consultant, 2002-2008 SCIENTECI-K Inc., Kent, WA, Senior Consultant, 1997-2002 Jason Associates Corporation, Idaho Falls, ID, Senior Consultant, 1993-1997 TENERA, L. P., Idaho Falls, ID, Senior Consultant, 1991-1993 EG&G Idaho, Inc., Idaho Falls, ID, Senior Engineering Specialist, 1987-1991 Rockwdl International, Anaheim, CA, Senior Engineer, 1985-1987 PLC Inc., Irvine, CA, Engineering Analyst, 1981-1985 PRO IF SSI ON AL SUM MARY Senior Nuclear Power Plant PRA, Risk and Reliability Consultant and Project Manager with over 31 years professional experience providing services to Nuclear Power Operatois in USA and former Soviet Union countries, USNRC, US DOE, and NASA.Mr. Reny has over 31 years experience as technical lead and project manager conducting probabilistic risk analyses (PRA), [PEs, IPEEEs seismic and fire PRAs, human reliability analyses (HRA), safety and risk informed management applications for domestic and international commercial nuclear power utilities and DOE nuclear facilities.

Developed PRA tools such as SAPIBRE, and have performed PRAs using SAPHIRE, CAFTA, WINNUPRA, and RISKMAN.PROFESSIONAL EXPEWRIENCE ABS Consulting, Technical Manager, trvine, CA (May 2008 present)Mr. Reny served as Technical Manager performing PRA analyses for South Texas Project, First Energy Corporation, Diablo Canyon, Entergy and Swiss NOK nuclear plants. He performed PRA analyses for shutdown risk, internal flooding, seismic, internal fires, external flooding, other external events, data updates, and PRA program planning and procedures for PRA model updates and other risk-informed applications.

A-20 ASS Consulting DANILL A. RENY ARES Corporation, Senior Consultant, Los Angeles, CA (2002 -May 2008)Mr. Reny served as Project Manager and Senior Consultant performing PRA, external events including fire PRA, and risk-informed analyses for Diablo Canyon Nuclear Power Plant. He performed PRA analyses in support of upgrades and improvements of plant PRA model for tire PRA uses and other risk-informed applications.

Mr. Reny was technical lead on Northrop Crumman/Boeing teams performing PRA and reliability assessments of NASA Orbital Space Plane (OSP) and Crew Exploration Vehicle (CEV). Performed PRA studies and conducted reliability programs on system architectures, vehicle designs, lunar and earth orbit misions.Performed risk assessments, conducted risk management programs and prepared risks for proposals for Jupiter Icy Moons Orbiter nuclear vehide, NASA Constellation architecture studies, NASA human and robotic research initiatives for Boeing, Strategic Border Initiative (SBINET) for Boeing, GOES satellite for Lockheed Martin, and Future Combat System for Raytheon.Mr. Reny performed hazard analyses and preparation of safety analysis report for Los Alamos Neutron Beam (LANSCE) facility; including review and walkdown of facility systenms and operations and analysis of procedures.

SCIENTECH, Inc., Senior Consultant, Kent, WA (1997- 2002)Member of USNRC sponsored international review panel providing safety review and systems upgrade of the Lithuanian RBMK plant. Participated m safety reviews of design and operations, preparation and review and approval of l&C Systenms design, installation, and PRA analyses.

He also conducted Human Reliability Assessment (1RA) for Ukraine nuclear power plants.Performed PRA updates, fault tree/event tree modeling, data analysis, Human Reliability Analysis, IPEEE, seismic and fire PRAs, quantification, and consequence analysis of nuclear power plant PRAs at Indian Point 21 Wolf Creek, Columbia, Diablo Canyon, San Onofre and other nuclear facilities.

He performed plant system walkdowns, operational data gathering, and operator interviews to update PRAs. He participated in development of models for real-time plant Safety Monitor applications.

He performed RAM analyses for nuclear plant emergency operations center, and loss of offsite power risk study for San Onofre Nuclear plant.Jason Associates Corporation, Office Manager -Senior Consultant, Idaho Falls, ID (1993-1997)

Office Manager, Project Management, Administration Management, Program Development for DOE Facilities.

Mr. Reny served as Idaho Falls Office Manager responsible for 35 employees and $6 to $10 million worth of contracts with DOE A-21 AS Consufting DANIEL A. RENY Project Safety engineer responsible for development of SARs for Pit 9 buried waste remediation facilitv at INEL Radioactive Waste Management Complex.Program Manager for DOE technical support contract (-$5M annually) responsible for all tasks supporting DOE-Il) office in preparation and review of environmental and safety prqects.Mr. Reny was the project safety analyst for Spent Fuel Project at Hanford responsible for safety design requirements and SAR preparation for Cold Vacuum Drying Facility.TENERA, L. P., Senior Consultant, Idaho Falls, ID (1991-1993)

Senior Consultant providing PRAs, IPE/IPEEE, Human Factors, Project Management, and DOE Nudear Facilities Safety Analysis Reports Mr. Reny performed various analyses and modeling tasks in development of PRAs,[PEs, and IPEEEs for Dresden, Quad Cities, Wolf Creek and other nuclear power plants.Tasks included data analysis, systems analyses, event tree and fault trees, seismic and fire PRAs, HRAs, Level 2 and Level 3 modeling and assessments.

He also prepared PSAR for Pit 9 Project safety design requirements for bid and proposal preparation of project EG&G Idaho, Inc, Idaho National Engineering Laboratory, Sr. Eng Specialist, Idaho Falls, ID (1987-1991)

Senior Engineering Specialist providing Task Management, PRA, Risk Management/Decision Analysis, Reliability/Availability/Maintainabiliht Assessment, Accident Analysis, and Software Development for commercial nuclear power plants and DOE facilities.

Mr. Reny applied PRA models in risk management/decision analyses, publication of NUREG report and testimony in support of NRC Generic Safety Issues resolution Principal Investigator with project responsibility for Air Force RAM analyses on over 2 Air Force bases. Authored and implemented RAM analysis guidelines to be used for Air Force facilities emergency power and utilities systems.Mr. Reny also served as safety engineer for PREPP experimental TRU waste incinerator at INEL responsible for preparation of facility design requirements and SAR.Rockwell International, Senior Engineer, Anaheim, CA (1985-1987)

Senior Reliability Engineer for Naval Shipboard Weapons Systems RAM Engineering, Reliability Testing, and MIL Standard Programs for the development of naval shipboard A-22 MS Consulting DANIEL A. RENY systems, guidance control systems for missile programs, and advance avionics for aircraft.Program Reliability Engineer responsible for development and implementation of reliability programs in conjunction with development and procurement of naval electronics equipment.

Responsibilities included program management, planning, methodology, and direction.

Analytical responsibilities included performing reliability modeling, predictions, reliability testing and growth analysis, failure reporting, failure analvsis, and correctie action, and EEE parts program.PLG Inc., Engineering Analyst, Irvine, CA (1981-1985)

Mr. Reny served as Engineering Analyst performing Probabilistic Risk Assessment/

Probabilistic Safety Assessment, and Reliability/

Availability/

Maintainability Assessment.

He also performed various level I and 2 internal event, seismic and fire PRA analyses and modeling.Mr. Reny performed PRA tasks, systems analysis, fault tree and event sequence modeling for TMI Unit 1 and Sequoyah nuclear power plants. He performed plant visits, data collection and systems walkdowns.

He developed and demonstrated risk assessment methods and tools for EPRI and NRC NUREG applications.

Mr. Reny performed RAM analyses and design decision input for procurement and design of a coal fired power plant, geothermal power plant, and various i&C systems designs.EDUCATION B.S., Applied Physics, University of California, Irvine 1982 A-23 Richard P. Mueller Streer Address 1116 Vine Street, East Liverpool, Oh, 43920 Phone Number 330-3854-5633 Email address mue12b6(comcastnet Work Experience Duquesne Light Company and First Energy 9/8/19 7 3 to 7/31/2011 Corporation Honse and Yard Laborer Coal and ash handler Nuclear Operator Beaver Valley Unit 1 Licensed Reactor Operator Beaver Valley Unit I Licensed Senior Reactor Operator Beaver Valley Unt I and Unit 2 Education Degrees Penn Stare University Associate Degree in Nncear Engineering Technology Nuclear Power Plant Related Skills and Experiences Operations lead for the last three NRC Trennial Fire protection ispecnons-Developed the Operations timelines and strategies for the latest Appendix R and Safe Shutiown proced7res for Beaver Valley Unit 1 and Unit 2 10CFRS0.59 and Independent Qualified Reviewer (IQR) qualified.

Operations lead for refueling outage scheduling and planning.Developed alternate strategies and improvements for Operations procedumes and tests.Assisted in the development of the latest 10 year ASME testing program for Beaver Valley Unit I and Unit 2.A-24 C:I2 RmSARco tINsTITU7E Certificate of Completion John Reddington Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012 Dae RObert K Kassawwa EPRI Struotural Reliablity

& Integft A-25 I mm UEE ire-, Certificate o fA chievement This is to Certify that 6,ý, f f fl.John:E.:Reddington has Compfeted the TriafSQUGA46 Walkdown Screening andSeismic Evaluation Training Course[feldNovember 20-25, 1987 Richard G. Slarck"., NIPP, A 5500 am, TW,.Trainirn C~ocdhinor Rob~ertP.

KassawamEPRF Progmmr~ Managci I A-26 4 SQUC Terrtif iratr of Ardtirurmrnd

~K1'I W14is is to aertifgj tha!irYarzin isri qi i f[ii'4as jormplrtrb tl~r §(QUG Watkbown§rrerning l: anb -.erismir Evaluation

(!0 oursr Trlb Maij 3-7, I19,3 ,1 i &.- 6ýuýA ý -, _.4 -David A. Freed. MPR Associates SQUG Training Coordinator Neil P. Smith, Commonwealth Edison SQUG Chairman Robert P. Kassawara.

EPRI SQUG Program Manager lilJ 4 A-27 eF~w IRESEARCH INSTITUTE Certificate of Completion Farzin Beigi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 13, 2012 4 44 4, Date Robert K, Kassawara EPRI Manager, StructuraJ Reliability

& Integrity A-28 A-29 Certificate of Completion Eddie Guerra Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns Nish R V a VP Advanced E Projects July 6, 2012 A-30 I Certificate of Completion Brian Lucarell Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns VP Advanced Eng Projects July 6, 2012 A-31 U I Presents this Certificate of) chi'evement To Certify That 7i M RA oha ed F. AIVI', P*Ee LIN----------------

fins rComi.pftedf the C)' )G 144,lfkdA a n Screening andSeismic Evaluation Training Course Hefd November 4 t -9 th, 1992 Neil P. Smith. C ornimmwaith Edismn David A. Freed. MPR Associates SQUG Tramiing Coordinator Robert P. Kassawara.

EPRI SQUG Program Manager A-32 ca~rl IRESEARCH

~NS1TtUTE Certificate of Completion Mohammed Alvi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012 Rob~ert K. Kassawwar EPRI Mianager, Structuwat Reliability

& Integrity Date A-33 pV ii A-34 W.-SQUG I ($crttfiatzt

'of Acijiu cu-tm QTl 1 is is to (grr1if~ II~at* n1~txnnu~b Aturi 1i~is c!J~ompi~h'~b tip TAMIr IEttnlrndirni tiiiiu~ QhTursc!I 4ei-j BeUack. MPR ASsociates A-35 A-36