ML13008A030

Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report. Cover Through Appendix a, Resumes and Qualifications A-37

ML13008A030
Person / Time
Site:	Beaver Valley
Issue date:	10/23/2012
From:	Beigi F, Guerra E M, Helffrich A L, Wakefield D J FirstEnergy Nuclear Operating Co, ABS Consulting
To:	Office of Nuclear Reactor Regulation
References
L-12-283
Download: ML13008A030 (354)

Text

Beaver Valley Power Station Unit 1 Near-Term Task Force Recommendation

2.3 Seismic

Walkdown Report October 23, 2012 Prepared by:.Donald J. Wakefield (ABS Consulting)

Farzin Beigi C n sulling)Eddie Guerra (ABS Csulling)Adam Helffrich (ABS.Consulting)

Reviewed by: Geog stbrook Jr. (FENOC)JoXReddington (FPNOO)R1. I'vmuelfer

('FENNOC)

,' .. ... .....Mohammed AMvi (FENOC)Approved by: -" Eugene Ebeck (FENOC)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 Page 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 ...........................................................................

4 4.0 SELECTIO N O F SSCs ...............................................................................................

5 4.1 Development of the SWEL 1 List (Related to Key Safety Functions)

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

5 4.2 Development of SWEL 2 for Spent Fuel Pool Related Items ...................................

9 5.0 SEISMIC WALKDOWN AND AREA WALK-BYS ..............................................

280 5.1 W alkdown Preparation

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

280 5.2 NTTF 2.3 W alkdowns ................................................................................................

281 5.3 Post W alkdown Activities

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

281 6.0

SUMMARY

OF THE WALKDOWN RESULTS ....................................................

281 6.1 W alk Down Item s and W alk-By Areas ......................................................................

281 6.2 W alk Down and Area W alk-By Findings ...................................................................

290 6.2.1 Seism ic W alkdown Findings ..........................................................................

290 6.2.2 Area W alk-By Findings ...................................................................................

291 6.3 Configuration Checks .................................................................................................

296 7.0 LICEN SIN G BA SIS EV A LU A TIO N ........................................................................

296 8.0 IPEEE V U LN ERA BILITIES .......................................................................................

299 9.0 PEER REV IEW ..............................................................................................................

300 10.0 REFEREN CES ...............................................................................................................

313 ii List of Tables Table 4-1 Base List I The Equipment Coming Out of Screen #3 and Entering Screen #4, for 5 S afety F un ction s ............................................................................................................................

11 Table 4-2 SWEL 1 Selected Equipment for 5 Safety Functions

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

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

274 Table 4-4 SWEL 2 (Selected Equipment for Spent Fuel Pool) .................................................

279 Table 6-1: Beaver Valley 1 NTTF 2.3 Walkdown Items (SWEL 1+2) .....................................

282 Table 6-2: Beaver Valley I NTTF 2.3 W alk-By Areas* ...........................................................

287 Table 6-3: Beaver Valley 1 NTTF 2.3 Inaccessible Items on SWELl+2* ...............................

288 Table 6-4: Beaver Valley 1 NTTF 2.3 Components Categorized by EPRI Classes ..................

289 Table 6-5: Potentially Adverse Seismic Conditions Identified from Seismic Walkdowns

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

292 List of Figures UFSAR Figure 2.5-1: Response Spectra DBE ..........................................................................

3 UFSAR Figure 2.5-2: Response Spectra OBE ..........................................................................

3 Figure 2-1: Beaver Valley Unit 1 Design SSE Spectra ..............................................................

4 Figure 6-1: Surface corrosion for valve 1AC-150 .....................................................................

294 Figure 6-2: Corroded valves with old deficiency tags ................................................................

295 List of Appendices APPENDIX A RESUMES AND QUALIFICATIONS APPENDIX B SEISMIC WALKDOWN CHECKLISTS (SWCs)APPENDIX C AREA WALK-BY CHECKLISTS (AWCs)APPENDIX D COMPONENT LIST FOR ANCHORAGE CONFIGURATION CHECK APPENDIX E MASONRY BLOCK WALLS VERIFIED UNDER IE BULLETIN 80-11 iii LIST OF ACRONYMS AWC Area Walk-by Checklist BV1 Beaver Valley Power Station Unit 1 DBE Design Basis Earthquake EPRI Electric Power Research Institute FENOC First Energy Nuclear Operating Company IPEEE Individual Plant Examination of External Events LERF Large Early Release Frequency LOCA Loss of Coolant Accident MCC Motor Control Center NPP Nuclear Power Plant NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor RAW Risk Achievement Worth SEL Seismic Equipment List SQUG Seismic Qualification Utility Group SSC Structures, Systems, and Components SWC Seismic Walkdown Checklist SWE Seismic Walkdown Engineer SWT Seismic Walkdown Team SWEL Seismic Walkdown Equipment List USI Unresolved Safety Issue iv

1.0 INTRODUCTION

This Report presents the results of the Seismic Walkdown conducted for the Beaver Valley Nuclear Power Plant Unit 1 (BVl) in support of FirstEnergy Nuclear Operating Company's (FENOC) response to NTTF Recommendation 2.3 in NRC 50.54(f) Letter, dated March 12, 2012. Consistent with the guidelines in Electric Power Research Institute (EPRI)Report 1025286, "Seismic Walkdown Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic" the walkdown implements the procedure described in Section 5.0 of this report.2.0 SEISMIC LICENSING BASIS The seismic licensing basis is contained in the Unit 1 Updated Final Safety Analysis Report (UFSAR).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 design basis earthquake (DBE). The OBE and DBE are considered equivalent to 1/2/2 Safe Shutdown Earthquake and the Safe Shutdown Earthquake (SSE), respectively.

The OBE is the earthquake which is of sufficient probability of occurrence to require its resulting ground accelerations at the site to be considered for operational loadings.

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 DBE/SSE is that earthquake giving rise to the maximum vibratory ground acceleration at a site which can be reasonably predicted from geologic and seismic evidence.The structures, systems and components designated Seismic Category I are designed to withstand, without loss of capability to protect the public, the most severe environmental phenomena ever experienced at the site with appropriate margins included in the design for uncertainties in historical data. All structures, systems and components including instruments and controls where failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of excessive amounts of radioactivity, and those structures and components vital to safe shutdown of the reactor are defined as Seismic Category I. Note that the 1 classification of Seismic Category I was previously designated as "Seismic Class I." These two terms are considered equivalent for which the above definition applies. 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 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.The design earthquakes, OBE and DBE, for the plant are specified by OBE and DBE design response spectra. These criteria are based on the plant site geologic investigations and seismologic recommendations as discussed in Section 2.5 and Appendix B of the Unit 1 UFSAR.The design is based on a DBE normalized to 0.125 g and for the Operational Basis Earthquake (OBE) normalized to 0.06 g. Analysis and design are based on response spectra as shown in UFSAR Figure 2.5-1 and 2.5-2 for the DBE and OBE, respectively.

Dynamic amplification factors used for these spectra are such as to give a maximum spectral acceleration of 0.44 g for two percent damping for the DBE with appropriate relative values for other amounts of damping.The spectra are flat from 2 to 5 Hz (0.2 to 0.5 second period) and reduce 'to an amplification ratio of unity for frequency exceeding 20 Hz.The response of plant structure are obtained through modal analysis of a multi-mass dynamic model which closely approximates the physical and response characteristics of the structure.

Masses are normally lumped at floor elevations and include the floor system, a portion of the walls above and the walls below the floor system, and major component and equipment loads. In addition, masses are located at elevations where any other response values are required.

Spring elements between masses represent building structural characteristics and are based on equivalent structural flexibilities.

These structural representations provide for the inclusion of torsional effects as a part of the dynamic output.2 FREDENCI1EMP~

FREGUENCY ICPS FIST1li 2.S-1 FI6URE 2.5-2 RESPOUSE SPECIRA ONE RESPO.NS SPECTRA ONE'PEATED .IAL SAFETY &,ALT$TT gUpsT UE0DATET VSAS SAFETt AFALYT ST OST UFSAR Figure 2.5-1: Response Spectra DBE UFSAR Figure 2.5-2: Response Spectra OBE The site of the station is underlain by approximately 100 ft of medium to dense sands and gravels laid in a high level terrace of the Ohio River. These are stable, relatively incompressible soils which provide a safe and adequate foundation for the power station. Settlements during construction were minor and settlements following operation will be negligible.

The surface soils of the terrace are slightly looser than the deeper lying soils and these near surface soils were removed beneath the structures and replaced with densely compacted granular fill. The surface of the terrace has been eroded within the limits of the turbine building to below desired foundation grade. Clay soils in this region were removed and replaced under the turbine building and the transformers with densely compacted granular fill to afford a safe and adequate foundation for these structures.

There is no hazard of liquefaction for the soils underlying the station under earthquake conditions.

Properties of the soil under dynamic loadings have been evaluated and proper cognizance taken of relative displacements between structures for piping design; the effects of earthquake loadings on lateral soil pressures on the containment structure and other earth retaining structures; and stability of slopes under earthquake and fluctuating water levels.For the Beaver Valley Nuclear Power Plant Unit I design SSC spectra refer to Figure 2-1.3

1.0 Beaver

Va ley Unit 1 SSE fS% Spectra 0.8 --.. ...... -r-" 0 0.6*Z.0.4 0.2 -0.0 0.1 1 10 100 Frequency

[Hz]Figure 2-1: The SSE response spectrum for Beaver Valley Unit 1 was digitized from BV1 FSAR Figure 2.5-1 3.0 PERSONNEL QUALIFICATIONS The following personnel worked together to formulate the list of selected equipment for the Beaver Valley Nuclear Power Plant Unit 1 NTTF Recommendation

2.3 Seismic

Walkdown: " J. Reddington

• R. Muller* D. Wakefield" F. Beigi The ABS Consulting Walkdown Team consisted of the following individuals: " F. Beigi" E. Guerra" A. Helffrich 4 Additionally, J. Reddington served as the reviewer of the Licensing Basis and of the Individual Plant Examination External Events (IPEEE). Mr. M. Alvi served as the lead peer reviewer for this effort.The seismic walkdown 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 these individuals (J. Reddington, M. Alvi, F. Beigi, E. Guerra and A. Helffrich) have also completed the EPRI Seismic Qualification Utility Group (SQUG) training.

Resumes and certificates of the walkdown 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 collectively represents previous Nuclear Power Plant walkdowns experience 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 also supported equipment selection, walkdown planning, equipment location determination, and selection of walk-by areas for the 2.3 Seismic Walkdown.4.0 SELECTION OF SSCS Consistent with the guidance in EPRI 1025286, "Seismic Walkdown Guidance," (Reference 1)dated May, 2012, the process of selecting the SSCs for inclusion of the Seismic Walkdown Equipment List (SWEL) 1 and SWEL 2 in support of the walkdown began with the creation of larger lists. The development of the list for SWEL 1 is presented first in Section 4.1 and it is followed by that for SWEL 2 in Section 4.2.4.1 DEVELOPMENT OF THE SWEL 1 LIST (RELATED TO KEY SAFETY FUNCTIONS)

The EPRI guidance document (Reference

1) says that using the previously developed IPEEE seismic equipment list 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 has assisted FENOC in developing a seismic equipment list (SEL) for use in a seismic probabilistic risk assessment (SPRA) for Beaver Valley Unit 1. An existing internal 5 PRA model is often 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 Beaver Valley Unit 1 PRA (i.e., working model BVlR5FLl, based on Reference 7)and the Beaver Valley Unit 1 IPEEE SEL list of 1920 SSCs (Reference 2). Duplicate SSCs, caused by (1) overlap between the two lists and (2) because the PRA contains multiple basic events for failure mode of a single component, were removed. Information about the original source of the remaining SSCs was retained.

In short the requirements in the EPRI walkdown guidance document in preparing the SSC SEL list were adequately satisfied.

However, during SSC sampling in preparation for the walkdown, selections were generally made preferentially from the IPEEE lists of SSCs. This is because the design packages were more likely to be available for these SSCs, so that advantage could be taken of the earlier design review work.SSCs from other sources were also chosen so that they were useful for seismic PRA purposes, but did not appear on either source list. For example, panels to be represented in the still evolving internal fire PRA and tanks represented in the PRA for internal floods were also reviewed for possible inclusion.

Again, duplicate SSCs were eliminated.

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

1) were also reviewed for completion.

Some SSCs were added as a result of this review.Nuclear steam supply system (NSSS) related SSCs were not required for this application and so were not added to the list. Also excluded were the supports for this equipment along with all the components mounted in or on this NSSS equipment.

Category 1 structures were also added in preparation for the seismic PRA, though they also are not required for the current walkdowns.

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 loss of coolant accident (LOCA) frequencies.

These SSCs were flagged as important to the containment safety function; i.e., they are involved in the computation of large early release frequency (LERF).Additionally, major new and replaced equipment, added to the plant since the performance of the IPEEE and the last Beaver Valley Unit 1 internal events PRA update are noted in a separate column of the developed lists titled "Screen 4d -Major New & Replacement Equip." These events were identified by consulting with long term plant operations staff that identified specific equipment items that had been replaced or overhauled, and by computerized searches of the 6 word "replace" in titles of existing engineering change packages (ECPs). Both lists were then evaluated to match equipment IDs appearing on Base List I with specific ECP numbers, that were judged to be of a major change.While there were no IPEEE vulnerabilities requiring plant changes identified for the Beaver Valley Unit 1 IPEEE, there were modifications performed in response to the assessment of A-46 outliers (References 3 and 4) in nearly the same time frame. The original list of A-46 outliers is documented in Reference 3 along with the initially proposed approaches to outlier resolution.

Some of these approaches involved changes to the plant equipment and were implemented right away. Where judged to be significant, these changes are noted in the column titled "Screen 4e -A46/1IPEEE Vulnerability" in Table 4-1. Additional effort was made to resolve other outliers, the majority of which were resolved by further analysis; i.e., without plant changes. Reference 4 then documented the final resolutions of the outliers that differed from the originally proposed approaches.

These changes were reviewed and those judged to involve significant plant changes are also documented in the column titled "Screen 4e -A46/ IPEEE Vulnerability" in Table 4-1.Once the initial list of SSCs was developed, it was first screened to retain only seismic category 1 quality, equipment.

Whether the SSC is regularly inspected, was also noted as this is justification for a second screen; e.g., for piping systems and containment penetrations.

Attributes of the retained SSCs were collected for the following information:

• Equipment ID" Brief SSC Description" SSC location -by building, elevation, and area description" The room environment where 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 Control, Reactor Coolant Inventory Control, Decay Heat Removal, and Containment Function) and several support system functions mentioned in the EPRI walkdown guidance.

Panels not previously evaluated for their associated safety functions (i.e., from the ongoing PRA for internal fires) were assigned the designator, "ESFAS", and retained for the selection process." Internal event PRA risk achievement worth (RAW) and Fussell-Vesely importance measures, if available.

7 The equipment ID and description fields were used to assign each retained SSC to one of the EPRI equipment categories (from Table A-1 of Reference

1) used for fragility analysis.

For some EPRI Categories (i.e., 0, 1, 2, 3 and 20), 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 the 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 walkdown because they are accounted for already in the selection of MCCs. The check valves and manual valves were assigned to Sub-Category Od, to avoid linking these numerous SSCs with SSCs also assigned to the EPRI other category.

A total of 5 SSCs were selected from the 0 and Od EPRI categories.

All of the EPRI categories were later employed as part of the SSC selection process. Except for EPRI Categories 11 (chillers), 12 (air compressors), and 13 (motor generators) at least one SSC was selected from the other EPRI categories.

Equipment in categories 11, 12, and 13 do appear on the combined list, however, at Beaver Valley Unit 1, none of these equipment are seismic Category 1 and therefore are screened from Base list 1.Base List 1, as defined in the EPRI walkdown guidance is attached as Table 4-1 for Beaver Valley Unit 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 SSCs, 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 of SSCs from which the SSC made its way onto the list. In some cases, SSCs appeared on both the original internal PRA and the IPEEE lists for Beaver Valley Unit 1. This is so indicated in the SSC source column.SWEL 1, as defined in the EPRI walkdown guidance (Reference

1) is attached as Table 4-2.The format is the same as that in the Base List 1, and the table is the same except that only the selected SSCs are shown. The equipment coming out of Screen #4 and entering the SWEL 1 bucket make up the SWEL 1 list. The selected SSCs have been chosen to account for a variety of systems, equipment types, room environments, and considering whether the SSCs involve new or replaced equipment since the completion of the IPEEE, or are subject to enhancements as a result of findings from the IPEEE and A-46 projects.SWEL 1 includes representative items from some of the variations within each of the above attributes.

A total of 113 SSCs were selected.

Beaver Valley Unit 1 plant operations staff was consulted in the SSC selection process. The selected list of SSCs is largely located in the service 8 building, safeguards building, and auxiliary building, but selections from the intake structure, diesel generator building, and containment are included.

Many of the selected SSCs are from support systems, but there are also SSCs selected from each frontline system. A total of 93 SSCs came from the original IPEEE or current internal events PRA model. SSCs are selected from each of the safety functions, including 9 related to the containment function.

There were 13 SSCs selected that are located in relatively high radiation areas and 22 that are often in damp areas. Most SSCs selected are in cool and dry areas. However, 10 are chosen from normally warm areas and 22 from relatively hot areas.The column in Table 4-2 labeled "Reason for Selection into SWEL I" 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. Also, SSCs subject to an enhancement resulting from the A-46 program or to an IPEEE enhancement are labeled as Screen 4e. 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 FVI>.005).

A representative set, but not all, of such risk significant SSCs were, therefore, included in the selected list. A number of selected SSCs are located inside the containment.

These SSCs were not accessible and therefore were not examined during the current walkdown effort and are scheduled to be inspected during the next refueling outage (IR22 in 2013).4.2 DEVELOPMENT OF SWEL 2 FOR SPENT FUEL POOL RELATED ITEMS For spent fuel pool related items, there was no starting list of SSCs with which to begin. Instead, the functions of the spent fuel pool systems were reviewed and equipment related to pool cooling and make up were included on a new list. Reference 5 details the operator actions to respond to a loss of spent fuel pool cooling or a loss of inventory.

The functions considered were normal spent fuel pool cooling, spent fuel pool makeup from demineralized water, spent fuel pool makeup using gravity feed from the refueling water storage tank (RWST), and spent fuel pool makeup from the fire protection system or from river water. The equipment identified for these functions in Reference 5 were included in the list along with the SSCs which make up the boundaries of the alternative makeup flow paths. The RWST and CVCS (i.e., from the blender)system were not included .in the spent fuel pool list of SSCs as those systems are included in Base List 1; i.e., see Section 4.1. Noticeably absent from the list are equipment related to spent fuel pool cleanup. The spent fuel pool cleanup equipment is not Seismic Category 1.9 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 walkdown guidance report (Reference

1) 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. At Beaver Valley Unit 1, the spent fuel pool cooling pumps and heat exchangers are Seismic Category 1 and therefore are included on Base List 2 Attributes of the retained SSCs were collected for the following information: " Equipment ID* Brief SSC Description

• SSC location -by building, elevation, and plant room 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 used for fragility analysis.

These EPRI categories were later employed as part of the SSC selection process.At Beaver Valley Unit 1, it is not possible to siphon the spent fuel pool level down to less than 9' 6.75" above the top of the spent fuel rack; i.e., failures resulting in a rapid drain-down cannot occur (Reference 6). Therefore, the rapid drain-down list of SSCs is empty for Beaver Valley Unit 1.SWEL 2, as defined in the EPRI walkdown guidance is attached as Table 4-4.There are no entries from rapid drain-down considerations; i.e., from Screen #4. The equipment coming out of Screen #3 and entering the SWEL 2 bucket in Figure 1-2 from the EPRI walkdown guidance report make up this second Seismic Walkdown Equipment List. The format is the same as that in the Base List 2, and the table entries are the same except that only the selected SSCs are shown. The selected SSCs have been chosen to account for a variety of equipment types and room environments.

Since Base List 2 is much shorter than that of Base List 1, and the number of applied screens smaller, the column labeled "Reason for Selection" simply contains the associated EPRI category and a text description of why each SSC was chosen. Since the types of Seismic Category 1 equipment related to the spent fuel pool are limited, so too is the variety of equipment types among the SSCs selected.10

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 (1) selection and execution of the configuration checks of selected anchorage, and (2) the verification of the seismic adequacy of block walls in the vicinity of equipment on the SWEL. 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.Prior to the walkdowns, the Seismic Walkdown Engineers (SWE) examined available plant documentation associated with (1) anchorage design, and (2) block wall capacity calculations, and correlated these 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 or block wall seismic adequacy.280 5.2 NTTF 2.3 WALKDOWNS The NTTF 2.3 walkdowns at Beaver Valley Unit 1 were performed over a duration of five days from September 10 to September 14, 2012. 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 A-46/IPEEE calculations.

Masonry walls in the vicinity of SWEL and non-SWEL items were recorded in the SWCs and AWCs. Subsequently, the SWEs verified the seismic adequacy of the block walls based on IE Bulletin 80-11 documentation.

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 A-46/IPEEE submittals, was 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 113 components, and SWEL2 included a total of 8 components.

From this total of 121 components, 108 components were walked down and 13 components were inaccessible and will require walkdown during the next plant's refueling outage. These thirteen items located inside the Containment Building will be walked down later during the next scheduled plant refueling outage. Notification No. 600788346 has been generated to have these walkdowns performed during the next refueling outage. Table 6-1 and Table 6-2 identify the walkdown items and walk-by areas, respectively, and Table 6-3 presents a list of items on the SWEL which were inaccessible while the plant is in operation.

These components will be walked down during the next plant's refueling outage. The areas walk-bys and the walkdown items are 281 cross correlated on the respective SWCs and AWCs. Table 6-4 provides the list of equipment that was walked down.Table 6-1: Beaver Valley 1 NTTF 2.3 Walkdown Items (SWEL 1÷2)Floor Equipment ID No Equip. Class Bldg El Area Description 1CC-E-IA 21. Tanks and Heat Exchangers AXLB 735 Primary Aux. Bldg 735'6" 1EE-EG-1 17. Engine Generators DGBX 735 DG Room Train A 1EE-EG-2 17. Engine Generators DGBX 735 DG Room Train B 1FC-E-1A 21. Tanks and Heat Exchangers FULB 735 Fuel Building 1FC-P-IA 5. Horizontal Pumps FULB 735 Beside H/X 1FW-PNL-I00B

20. Instrument and Control Panels SFGB 735 West Cable Vault ---- South East Comer IPC-145 0. Other -check/manual valve FULB 735 Fuel Building 1PC-1FC-102A

18. Instrument (on) Racks FULB 735 Fuel Building 1RW-189 0. Other -check/manual valve AXLB 735 Outlet of CCR HX 2. Low Voltage Switchgear and 480VUS-I-8-N SRVB 713 Emergency Switchgear Room Train A Breaker Panels 480VUS-1-9-P

2. Low Voltage Switchgear SRVB 713 Emergency Switchgear Room Train B 4KVS-1AE 3. Medium Voltage Switchgear SRVB 713 Emergency Switchgear Room Train A 4KVS-1DF 3. Medium Voltage Switchgear SRVB 713 Emergency Switchgear Room Train B BAT-I-1 15. Battery Racks SRVB 713, Battery Room #1 BAT-1-2 15. Battery Racks SRVB 713 Battery Room #2 16. Battery Chargers and BAT-CHG-1-1 SRVB 713 Emerg Swgear Rm Tr A Inverters 16. Battery Chargers and BAT-CHG I A SRVB 713 Emergency Switchgear Room Train B Inverters 16. Battery Chargers and BAT-CHG-1-3 SRVB 713 Emerg SW Rm Tr A Inverters 20. Instrument and Control BB-A1 SRVB 735 Control Room Panels 21. Tanks and Heat CC-TK- 1 AXLB 768 CCW Surge Tank Exchangers I _____1_752______Bldg_768 CH-BL-2 21. Tanks and Heat Exchangers AXLB 752 Aux Bldg 768 282 Table 6-1: Beaver Valley 1 NTTF 2.3 Walkdown Items (SWEL 1+2)Floor Equipment ID No Equip. Class Bldg Area Description El CH-P-1C 5. Horizontal Pumps AXLB 722 Charging Pump Cubicle 1C CH-P-2A 5. Horizontal Pumps AXLB 752 Boric Acid Pump Cubicle B CH-TK-lA 21. Tanks and Heat Exchangers AXLB 752 Boric Acid Tank IA 14. Distribution Panels and DC-SWBDI-1 SRVB 713 Emergency Switchgear Room Train A______________Automatic Transfer Switches 14. Distribution Panels and DC-SWBD1-2 SRVB 713 Emergency Switchgear Room Train B_______________Automatic Transfer Switches __________________

14. Distribution Panels and DC-SWBD 1-3 SRVB 713 Emergency Switchgear Room Train A_______________Automatic Transfer Switches _______14. Distribution Panels and DC-SWBDI-4 SRVB 713 Emergency Switchgear Room Train A_______________Automatic Transfer Switches _________________

.EE-P-IA 5. Horizontal Pumps DGBX 735 DG Room Train A EE-TK-2A 21. Tanks and Heat Exchangers DGBX 735 DG Room Train A LFCV-1CH-113A

7. Pneumatic-Operated Valves AXLB 722 Blender Room FCV-IFW-103B

7. Pneumatic-Operated Valves SFGB 735 QS/AFW Pump Room FW-59 0. Other -check/manual valve SFGB 735 QS/AFW Pump Room FW-P-2 5. Horizontal Pumps SFGB 735 QS/AFW Pump Room FW-P-3A 5. Horizontal Pumps SFGB 735 QS/AFW Pump Room FW-T-2 5. Horizontal Pumps SFGB 735 QS/AFW Pump Room HCV-ICH-186

7. Pneumatic-Operated Valves AXLB 722 Blender Room 16. Battery Chargers and INV-VITBUS 1-1 SRVB 713 Emergency Switchgear Room Train A Inverters 16. Battery Chargers and INV-VITBUS 1-3 SRVB 713 Emergency Switchgear Room Train A Inverters 16. Battery Chargers and INV-VITBUS 1-4 SRVB 713 Emergency Switchgear Room Train B Inverters LS-IEE-201-1

18. Instrument (on) Racks DGBX 735 DG Room Train A LT-IQS-100A

18. Instrument (on) Racks YARD 735 AT QS-TK-1 MCC-1-El 1. Motor Control Centers INTS 705 Intake Structure Pump Cubicle 1 (A)MCC-1-E12

1. Motor Control Centers SFGB 735 East Cable Vault 283 Table 6-1: Beaver Valley 1 NTTF 2.3 Walkdown Items (SWEL 1+2)Floor Equipment ID No Equip. Class Bldg El Area Description MCC-1-El3

1. Motor Control Centers SFGB 756 Motor Control Center Room MCC-l-E4 1. Motor Control Centers AXLB 735Primary Aux. Bldg 735'6"--West Wall Near CC-E- 1C MCC-1-E6 1. Motor Control Centers SFGB 735 East Cable Vault MCC-1-E7 1. Motor Control Centers DGBX 735 DG Room Train A MCC-1-E8 1. Motor Control Centers DGBX 735 DG Room Train B-- -East Wall MOV-1FW-151E 8A. Motor-Operated Valves SFGB 735 QS/AFW Pump Room MOV-IMS-105 8A. Motor-Operated Valves SFGB 751 Main Steam Valve Room MOV-IQS-101B 8A. Motor-Operated Valves SFGB 735Pipe Tunnel---W Area RR 745 -Shallow Pit MOV-IRW-102A2 8A. Motor-Operated Valves INTS 705 Intake Structure Pump Cubicle I (A)MOV-IRW-103A 8A. Motor-Operated Valves AXLB 722 Primary Aux. Bldg 722'6"--East Central MOV-IRW-103B 8A. Motor-Operated Valves AXLB 722 Primary Aux. Bldg 722'6"--East Central MOV-IRW-113B 8A. Motor-Operated Valves DGBX 735 DG Room Train A MOV-IRW-1 14B 8A. Motor-Operated Valves AXLB 722 Primary Aux. Bldg 722'6"--East Central MOV-1SI-860B 8A. Motor-Operated Valves SFGB 747 Valve Pit 688 MOV-I SI-862A 8A. Motor-Operated Valves SFGB 747 Valve Pit 689 MOV-RW- 117 8A. Motor-Operated Valves AXLB 722 Primary Aux. Bldg 722'6"--East Central PCV-IMS-lO1C 8B. Solenoid Valves SFGB 768 Main Steam Valve Room PI-IRW-102A1

18. Instrument (on) Racks AXLB 735 RW Inlet To CCR HTX'S PNL-IMS-lO0A

20. Instrument and Control Panels SFGB 751 Main Steam Valve Room PNL-IMS-101B

20. Instrument and Control Panels SFGB 751 Main Steam Valve Room PNL-AC1-BUS-iE

20. Instrument and Control Panels SRVB 713 Relay Room ---- West Wall PNL-AC 1-E1 20. Instrument and Control Panels SRVB 713 Emergency Switchgear Room Train A PNL-DC1-3

20. Instrument and Control Panels SRVB 735 Control Room PNL-DGEA-l

20. Instrument and Control Panels DGBX 735 DG Room Train A PNL-DIGEN-1

20. Instrument and Control Panels DGBX 735 DG Room Train A PNL-DIGEN-2

20. Instrument and Control Panels DGBX 735 DG Room Train B 14. Distribution Panels and PNL-VITBUSI-2 SRVB 735 Control Room Automatic Transfer Switches 284 Table 6-1: Beaver Valley 1 NTTF 2.3 Walkdown Items (SWEL 1+2)Floor Equipment ID No Equip. Class Bldg tEl Area Description

14. Distribution Panels and PNL-VITBUS 1-3 SRVB 735 Control Room Automatic Transfer Switches PT-IRW-l 13A 18. Instrument (on) Racks AXLB 735Primary Aux. Bldg 735'6--CCR HX Inlet HDR QS-P-1B 5. Horizontal Pumps SFGB 735 QS/AFW Pump Room REAC-TR-SWGR-

2. Low Voltage Switchgear and SRVB 713 Motor Generator Room IA Breaker Panels RK-1PRI-PROC-12 20. Instrument and Control Panels SRVB 713 Process Instrumentation Room RK-AUX-RPTST-A

20. Instrument and Control Panels SRVB 713 Process Instrumentation Room RK-NUC-INS-1

20. Instrument and Control Panels SRVB 735 Control Room RK-VV-REL-B

20. Instrument and Control Panels SRVB 713 Process Instrumentation Room RS-P-2B 6. Vertical Pumps SFGB 732 B RS Pump Cubicle RW-183 0. Other -check/manual valve AXLB 735 Primary Aux. Bldg 735'6"--East Central RW-57 0. Other -check/manual valve INTS 705 Intake Structure Pump Cubicle I (A)SAF-SW-65

20. Instrument and Control Panels SRVB 713 Emergency Switchgear Room Train A SI-P-lA 6. Vertical Pumps SFGB 751 LHSI Pump IA Primary Aux. Bldg 722'6"--At BIT SI-TK-2 21. Tanks and Heat Exchangers AXLB 722 Cubicle SV-MS-104B

7. Pneumatic-Operated Valves SFGB 768 Main Steam Valve Room TCV-ICH-144

7. Pneumatic-Operated Valves AXLB 722 Primary Aux. Bldg 722'6"--NE Coiner TI-1EE-301 19 Temperature sensors DGBX 735 DG Room #1 TRANS-1-8N

4. Transformers SRVB 713 Emergency Switchgear Room Train A TRANS-1-8N1

4. Transformers SRVB 713 Normal Switchgear Area TRANS-1-9P

4. Transformers SRVB 713 Emergency Switchgear Room Train B TRF-1015 4. Transformers SRVB 713 Emergency Switchgear Room Train A TRS-BIP-PNLI

20. Instrument and Control Panels SFGB 735 East Cable Vault TV-1BD-1OOA

7. Pneumatic-Operated Valves SFGB 722 Pipe Penetration C TV-1CV-102 8B. Solenoid Valves SFGB 722 Pipe Penetration C TV-1CV-150B 8B. Solenoid Valves SFGB 722 Pipe Penetration B TV-IMS-101C

7. Pneumatic-Operated Valves SFGB 768 Main Steam Valve Room 285 Table 6-1: Beaver Valley 1 NTTF 2.3 Walkdown Items (SWEL 1+2)Floor Equipment ID No Equip. Class Bldg Area Description

____ ____ ____El TV-1MS-105A

7. Pneumatic-Operated Valves SFGB 747 Main Steam Valve Room VB-A 20. Instrument and Control Panels SRVB 735 Control Room VS-AC-1A-10. Air handlers SRVB 713 AC Equip Rm BLOWER VS-D-22-2C

0. Other -check/manual valve DGBX 735 DG Room Train B VS-D-57A1

7. Pneumatic-Operated Valves INTS 705 Intake Structure Pump Cubicle 1 (A)VS-D-57B 1 7. Pneumatic-Operated Valves INTS 705 Intake Structure Pump Cubicle 2 (B)VS-D-57C1

7. Pneumatic-Operated Valves INTS 705 Intake Structure Pump Cubicle 3 (C)VS-F-55B 9. Fans SRVB 725 Service Building Floor 725 VS-F-57B 9. Fans INTS 705 Intake Structure Pump Cubicle 2 (B)WR-P-1A 6. Vertical Pumps INTS 705 Intake Structure Pump Cubicle 1 (A)WR-P-1B 6. Vertical Pumps INTS 705 Intake Structure Pump Cubicle 2 (B)Note: Equipment located in either SFG or MSCV are all designated to be in SFGB.286 Table 6-2: Beaver Valley 1 NTTF 2.3 Walk-By Areas*Area Bldg Floor El AC Equip RM SRVB 713 AT QS-TK-A YARD 735 Aux bldg768 AXLB 768 B RS Pump Cubicle SFGB 732 BATrg ROOM SRVB 713 Battery Room #2 SRVB 713 Blender Room AXLB 722 Boric Acid Pump Cubicle B AXLB 752 Boric Acid Tank 1 A AXLB 752 CCW Surge Tank AXLB 768 Charging Pump Cubicle I1A AXLB 722 Charging Pump Cubicle I1C AXLB 722 Control Room SRVB 735 DG Room #1 DGBX 735 DG Room Train A DGBX 735 DG Room Train B--East Wall DGBX 735 East Cable Vault SFGB 735 Emer Swgr Room Train A SRVB 713 Emer Swgr Room Train B SRVB 713 Fuel building FUBD 713 INTS Pump Cubicle 1 (A) INTS 705 INTS Pump Cubicle 2(B) INTS 705 INTS Pump Cubicle 3(C) INTS 705 LHSI Pump IB SFGB 751 Main Steam Valve Room SFGB 751 Motor Generator Room SRVB 713 Normal Switchgear Area SRVB 713 Pipe Penetration C SFGB 722 Prim Aux Bldg 735'6 HX Inlet AXLB 735 Prim Aux Bldg 735'6 W. Wall AXLB 735 Prim. Aux Bldg 735'6 E Central AXLB 735 Prim. Aux. BLDG 722'6 BIT Cub. AXLB 722 Prim. Aux. BLDG 722'6 E Cent. AXLB 722 Prim. Aux. BLDG 722'6 NE Comer AXLB 722 Prim. Aux. Bldg. 735'6 AXLB 735 287 Table 6-2: Beaver Valley 1 NTTF 2.3 Walk-By Areas*Area Bldg Floor El Prim. Aux. Bldg. 735'6 AXLB 735 QS-AFW Pump Room SFGB 735 Relay Room AXLB 713 Service Building Floor 725' SRVB 725 Valve Pit 688' SFGB 747 West Cable Vault -SE Comer SFGB 735* Does not include areas in Containment Building.Note: Equipment located in either SFG or MSCV are all designated to be in SFGB.Table 6-3: Beaver Valley 1 NTTF 2.3 Inaccessible Items on SWELl+2*Equip. ID Description Bldg [ El[ Area Description NITROGEN ACCUMULATOR REACTOR GN-TK-1B NIRGN RCBX 767 CONTAINMENT TANK GN-TK-1B BUILDING SGIA BIP NARROW RANGE OUTSIDE IA STEAM GEN LT-1FW-475 LEVEL TRANSMITTER LT-IFW- RCBX 718 475CUBICLE MOV1RC535PORV BLOCK VALVE MOV-RC-MOV-IRC-535 RCBX 767 PRESSURIZER CUBICLE PCV- 535 1 PCV-IRC- PZR PORV PCV-RC-455D RCBX 767 PRESSURIZER CUBICLE 455D CNMT AIR COMPR CHILLED REACTOR PCV-CC-101 CNTAR CUMPR CHLE RCBX 718 CONTAINMENT WATER SUP PRESS CONT BUILDING REACTOR COOLANT WIDE PT-lRC-403 RE , PT-RCDE3 RCBX 692 701 KEYWAY WALL RANGE PRESSURE, PT-RC-403 RH-P-IA RHRPUMP RH-P-IA RCBX 707 RHR PLATFORM RECIRC SPRAY HEAT RS-E-1D ECHANGER REAT RCBX 718 RECIRC SPRAY CLRS EXCHANGER RS-E-1 1D RS-P-1A INSIDE RECIR. PUMP RS-P-lA RCBX 692 AT CNMT SUMP LOOP 2 COLD LEG SI SUP CHECK SI2 VLE I24 RCBX 718 AT 1 B LOOP CUB VALVE SI-24 SOV-IRC- ON PZR CUBICLE 103B SOV-RC-103B RCBX 767 OUTSIDE WALL TV-ICC-107A RCP IA THERM BARR CCR OUT RCBX 718 -A RCP PP CUBICLE-ISOL,TV-CC-107A CNMT RECIRC AIR COOLERS REACTOR TI-1CC-131C CNMT REOR AI RCBX 692 CONTAINMENT INLET LOOP#3 BULDING (690)*These equipments will be walked down during the next scheduled refueling outage 288 Table 6-4: Beaver Valley 1 NTTF 2.3 Components Categorized by EPRI Classes EPRI Equipment Description Components Cat No. Walked Down 0 Other 7 1 Motor Control Centers and Wall-Mounted Contactors 7 2 Low Voltage Switchgear and Breaker Panels 3 3 Medium Voltage, Metal-Clad Switchgear 2 4 Transformers 4 5 Horizontal Pumps 10 6 Vertical Pumps 4 7 Pneumatic-Operated Valves 14 8 Motor-Operated and Solenoid-Operated Valves 16 9 Fans 2 10 Air Handlers 1 11 Chillers 0 12 Air Compressors 0 13 Motor Generators 0 14 Distribution Panels and Automatic Transfer Switches 6 15 Battery Racks 2 16 Battery Chargers and Inverters 6 17 Engine Generators 2 18 Instrument (on) Racks 7 19 Temperature Sensors 2 20 Instrumentation and Control Panels 17 21 Tanks and Heat Exchangers 9 Total 121 289 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 6 potentially adverse seismic conditions were identified during the Seismic Walkdowns.

Table 6-5 provides as summary of all 6 adverse finding conditions identified.

As shown in Table 6-5, only one condition report (CR 2012-14321) was 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 6-5: Potentially Adverse Seismic Conditions Identified from Seismic Walkdowns Description of Licensing Reference No I Equipment Class Adverse Seismic Evaluation for Condition Elt Justification Required No grout below SWC for IEE-EG-1 17. Engine Generators middle section of No EEEG-EDG's No grout below SWC for IEE-EG-2 17. Engine Generators middle section of No 1EEEG-2 EDG's 290 Table 6-5: Potentially Adverse Seismic Conditions Identified from Seismic Walkdowns Description of Licensing Reference Equipment ID Equipment Class Adverse Seismic Evaluation for NoCondition EvlainJustification Required Cable trays SWC for MCC-1-E6 1. Motor Control Center possible interaction No with MCCs MClE Cable trays SWC for MCC-1-E12

1. Motor Control Center possible interaction No with MCCs C-E1 Battery charger and CR 2012 DC-SWBDI-2

14. Distribution Panels switchboard not Yes 14321 bolted together Battery charger and CR 2012-BAT-CHGI-2-A

16. Battery Chargers and Inverters switchboard not Yes 14321 bolted together Masonry Block Walls Based on calculations presented in response to IE Bulletin 80-11, Masonry block walls identified in the vicinity of walked-down SWEL items, and in Area Walk-bys have adequate seismic capacity.

Appendix E presents the list of block walls associated with the nearby SWEL/Area Walk-by items, and as the referenced calculations used for verification of the block wall seismic capacity.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 10 potentially adverse seismic conditions were identified during the area walk-bys.

Table 6-6 provides a summary of all 10 potentially adverse seismic conditions identified.

As shown in Table 6-6, seven 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.291 Table 6-6: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys Licensing Floor Description of Adverse Basis Reference for Area Bldg El Seismic Condition Evaluation Justification Required INTS Pump Pump I WR-P- IC has 2 C u mp INTS 705 anchor bolts that are Yes CR 2012-13969 Cubicle 3(C) corroded.Black marks identified for AWC for DG DG Room Train A DGBX 735 EDG exhaust No Room Train A Security barrier near PT-Pipe Penetration C SFGB 722 1CV-101B Yes CR 2012-14020 Surface corrosion for valve Pipe Penetration C SFGB 722 IAC-150 Yes CR 2012-13971 Old deficiency tags CR 2012-13972 Various Various Various identified Yes CR21-37 identfiedCR 2012-13974 Main Steam Valve Pipe contact with knee AWC for Main Mntm SFGB 751 brace No Steam Valve Room Room Control Room ceiling tiles AWC for Control Room SRVB 735 No Control Room Potential for spray from AWC for Fuel Fuel Building FULB 735 space heater No Building Prim. Aux Bldg AXLB 735 Missing bolt near hanger Yes CR 2012-14047 735'6 E Central Corrosion on nuts on valve Pipe Penetration C SFGB 722 TV-ICC-110F2 Yes CR 2012-13970 As illustrated in Table 6-6, some of the condition reports issued will not require a Licensing Basis Evaluation since the findings were subsequently resolved.

These situations are described briefly below.292 o Security barrier near PT-iCV-IO1B During the area walk-bys, SWEs identified a security shield barrier near a pressure transmitter (PT-iCV-101B) at elevation 722' in the Pipe Penetration C Room located in the Safeguards Building.

This situation presented a potential seismic interaction concern due to the proximity of the barrier with respect to the pressure transmitter.

The Security Shift Supervisor was contacted and immediate action was taken to move the barrier away from the pressure transmitter and be relocated to a safer location.

No further action was required. (Ref. CR 2012-14020) 293

• Surface corrosion for valve JAC-150 While performing the area walk-by for the Pipe Penetration C Room, it was observed that Valve 1AC-150 presented mild corrosion on the valve body. Immediate action was taken and the engineering supervisor was notified about the condition observed.

Notification No. 600785632 was issued in order to clean and paint the corresponding valve. No further action was required since it was judged that the corrosive state will not affect the valve's intended design function.(Ref. CR 2012-13971)

Figure 6-1: Surface corrosion for valve 1 AC- 150 294

• Old deficiency tags identified During the area walk-bys, it was observed that deficiency tags have been in place for various components for an extended period of time apparently, without assessment of the condition.

Condition reports CR-2012-13972 and CR-2012-13974 address the specific components where these tags were indentified.

The SWT proceeded to initiate notification Nos. 600785633 and 600785694 in order to perform the work needed to assess the situation.

The SWE's concluded that the deficiencies identified will not affect the component's intended design functions.

Figure 6-2: Corroded valves with old deficiency tags.CR-2012-13972 issued to address these conditions 295

6.3 CONFIGURATION

CHECKS The SWELL 1+2 included 76 items, which were not in-line components such as valves. The process of verifying the anchorage configuration focused on 39 SWEL components arbitrarily selected prior to walkdown proceedings (this is 51% of the SWEL items with anchorage configurations).

Appendix D provides a list of the 39 components comprising the anchorage configuration list linked with the specific references used for verification purposes; i.e., A46/IPEEE Calculations, design drawings, etc.The anchorage configuration for each of the 39 SWEL components listed in Appendix D was verified based on A46/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 A46/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 38 instances.

7.0 LICENSING

BASIS EVALUATION Nine 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-2012-13969, CR-2012-13970, CR-2012-13971, CR-2012-13972, CR-2012-13974, CR-2012-14020, CR-2012-14047, CR-2012-14321, and CR-2012-14347.

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

CR-2012-13969 It was observed that two out of eight nuts for the bolts anchoring the Pump WR-P-1C to the floor are corroded.

This pump is located inside the intake structure at elevation 705' Cubicle C.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 296 judgment.

The two nuts are located side by side. No calculations or drawings are affected since the nuts are able to perform the intended function.Initiated Notification No. 600785650 to replace the nuts under work order 200530167.

CR-2012-13970 It was observed that two out of twelve nuts for the bolts for Component TV-ICC-1 10F2 are corroded.

This component is located inside Safeguard Building at elevation 722' in Pipe Penetration C Room.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 corroded nuts are located next to each other. No calculations or drawings are affected since they are able to perform the intended function.Initiated Notification No. 600785631 to replace the nuts under work order 200530160.

CR-2012-13971 It was observed that Valve 1AC-150 has mild corrosion deposited on the valve body. This component is located inside Safeguard Building at elevation 722' in Pipe Penetration C Room.Observation concluded that there is only surface corrosion on the valve body and it is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve only has surface corrosion and it is able to perform the design basis function.Initiated Notification 600785632 to clean the corroded areas under work order 200530161.

CR-2012-13972 It was observed that at elevation 735' in Safeguard Building Quench Spray Room, deficient tags have been in place on various components for extended duration which have not been addressed in a timely fashion. Specifically, following anomalies have been noted: 1. Component IAC-310: Inadequate thread engagement on packing studs and rust on the valve (Ref Tag ID 60734). 1/2 threads lacking engagement in one stud.2. Component TV-1CC-123-2:

Mild corrosion on the component 297

3. Component IAC-305: Mild corrosion on the component Observation concluded that the anomalies noted above are such that the components are still capable to perform their intended design basis function based on engineering judgment.

Mild corrosion does not affect calculations or drawings, and thread is still engaged on the other component so design basis is not affected.Initiated Notification Nos. 600785633, 600786401 and 600786402 to perform the work under work orders 200530162, 200530661, and 200530662.

CR-2012-13974 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that at elevation 747' in Safeguard Building Valve Garden, a deficient tag has been in place on the following component since 2003. Ref: Tag ID 45508 Component MOV-1SI-864B (Local Valve Indication Pin sheared off). The local valve indication pin is not required by design basis, and therefore does not violate the design basis documents.

Initiated Notification No. 600785694 to replace the indication pin and close the deficiency tag under work order 200015714.

CR-2012-14020 It was observed that a Security Shield Barrier was located too close to SR Component PT-1CV-1OIB at El. 722' Unit-1 Safeguard Building in Pipe Penetration C Room. This posed a seismic interaction concern that during a seismic event it had a potential to hit the pressure transmitter and potentially make the component not performing its intended design function.The security shift supervisor has been contacted and the shield was moved to a safer location.There is no infringement on the design basis.CR-2012-14047 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that Pipe Support H-70 has a missing anchor bolt. One of the six anchor bolts is missing.298 A review of calculation for Pipe Support H-70 was performed (Ref: Calculation 13387.02-S (B)-24CC2-B), and it concluded that the base plate was qualified with a five bolt configuration.

The missing bolt was noted. Therefore this is only a configuration issue and not a plant operability concern as the support design is in full compliance with the design basis requirements.

Initiated Notification No. 600785698 to update drawings and install a placard on the support noting that it has been analyzed.

Work to be done under work order 200530170.

CR-2012-14321 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that there is a potential of seismic interaction between the Switch Board DC-SWBD1-2 and Battery Charger BAT-CHG 1-2 as they are not connected to each other. These components are located side by side with zero gaps in between them. During a seismic event, there is a potential for these components to bang against each other and cause chatter of relays inside the battery charger.These components are located inside DF Switchgear Room at El 713' in U-I Service Building.This does not violate the design basis documents.

CR-2012-14347 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-I Plant during the week of September 10, 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. Additionally, an administrative CR (CR-2012-14210) was written which is not reflected in the matrix as it is not associated with any hardware concerns.Because there are no new issues identified in this CR, there is no affect on the design basis.8.0 IPEEE VULNERABILITIES 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 299 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 A46 enhancements as well as 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- 1:* 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 Valley Power Station, Unit-1 acted as Operations representative during the selection of the SWEL.300 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 I 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 the paragraph above.SWEL 1 for the Beaver Valley Power Station, Unit-1 included 113 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-I included 8 components.

==

Conclusion:==

No major concerns were identified by the peer review team in the selection process for SWEL 1 or SWEL 2.301 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 1 selection?

Y ON[-]See Attached Comments 2. Does SWEL I include an appropriate representation of items having the following sample selection attributes?

a. Various types of systems? Y ONE]See Attached Comments b. Major new and replacement equipment?

Y EN-See Attached Comments c. Various types of equipment?

Y ONE]See Attached Comments d. Various environments?

See Attached Comments e. Equipment enhanced based on the findings of the IPEEE (or equivalent) program?YZENE Y END See Attached Comments f. Were risk insights considered in the development of SWEL 1?Y END See Attached Comments 302 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 ONI-Y ONI]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 Revieww -ý Date.Y ONE]-(0 -k 93 -1 ZI Peer Reviewer #2: 110 Daft: [303 Peer Review Checklist for SWEL Comments on Question 1: A peer review of the SWEL selected for the Beaver Valley Power Station, Unit-I 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. 4.16KV AC Power C. 125V DC Power D. High Head Safety Injection E. Emergency Boration System F. Auxiliary Feedwater System G. Pressurizer H. Reactor Coolant Pumps I. 120V AC Power J. Main Feed Water K. River Water System L. 480V AC Power M. Quench Spray System N. Low Head Safety Injection Pumps and Suction 0. Instrument Air System P. Steam Generator Steam Relief System Q. Residual Heat Removal R. Recirculation Spray System S. High Pressure Makeup Supporting Systems T. Containment Isolation System 304 U. Heating Ventilating and Air Conditioning System V. Emergency Diesel generators W. Engineered Safety Features Actuation System X. Reactor Protection System Comments on Question 2b: The selected components represent many new and replacement equipment based on the following modifications:

A. ECP 11-0157-001:

Fused Test Jack Installation B. ECP 11-0157-002:

Fused Test Jack Installation C. ECP 08-0033-001:

Replacement of Inverter and Static Switch for BV-1 Vital Bus D. DCP 2422: Mounting Change E. ECP 02-0063: Inverter Replaced F. ECP 02-0283, ECP 06-005 and ECP 03-0428: Replaced Circuit Breakers G. ECP 12-0243-001:

Replace Thermal Overload H. ECP 10-0353-001:

Valve Replacement I. ECP 07-0203: Strainer Replaced J. ECP 02-0258-001:

Coupling Retrofit Modification K. ECP 02-0258-002:

Piping Modification L. ECP 08-0401-004:

Replaced, Reactor Coolant Gas Vent System Valve Upgrade M. ECP 08-0134-001 and 002: Reclosing Reactor Trip Breaker N. ECP 03-0575 and ECP 04-0403: Replaced Modules 0. ECP 02-0076: Replaced Station Battery Charger P. DCP-1741 and DCP-2416:

Replaced Level Transmitter Q. DCP-2424:

Replaced River Plant Water Pump 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 and Breaker Panels C. Medium Voltage Metal Clad Switchgear D. Battery Racks E. Battery Chargers and Inverters 305 F. Horizontal Pumps G. Distribution Panels and Automatic Transfer Switches 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 Comments on Question 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. Dry/Wet G. Warm/Cool Comments on Question 2e: Based on the review, the selected components represent equipment enhanced based on findings of the IPEEE.Comments on Question 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: 306 Spent Fuel Pool related items were considered and are adequately represented in SWEL 2.Comments on Question 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,"Seismic Walkdown Guidance." Also, an ex-SRO from the Beaver Valley Power Station, Unit-I 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 307 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.

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 eight of them resulted in generating condition reports as follows: CR-2012-13969, CR-2012-13970, CR-2012-13971, CR-2012-13972, CR-2012-13974, CR-2012-14020, CR-2012-14047, and CR-2012-14321.

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

However, a ninth CR was written to capture all the issues identified above in one roll up condition report (CR-2012-14347).

1. CR-2012-13969 It was observed that two out of eight nuts for the bolts anchoring the Pump WR-P-1C to the floor are corroded.

This pump is located inside the intake structure at elevation 705' Cubicle C.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.

The two nuts are located side by side. No calculations or drawings are affected since the nuts are able to perform the intended function.308 Initiated Notification No. 600785650 to replace the nuts under work order 200530167.

2. CR-2012-13970 It was observed that two out of twelve nuts for the bolts for Component TV-ICC-110F2 are corroded.

This component is located inside Safeguard Building at elevation 722' in Pipe Penetration C Room.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 corroded nuts are located next to each other. No calculations or drawings are affected since they are able to perform the intended function.Initiated Notification No. 600785631 to replace the nuts under work order 200530160.

3. CR-2012-13971 It was observed that Valve IAC-150 has mild corrosion deposited on the valve body. This component is located inside Safeguard Building at elevation 722' in Pipe Penetration C Room.Observation concluded that there is only surface corrosion on the valve body and it is capable to perform its intended design function based on engineering judgment.

No calculations or drawings are affected since the valve only has surface corrosion and it is able to perform the design basis function.Initiated Notification 600785632 to clean the corroded areas under work order 200530161.

4. CR-2012-13972 It was observed that at elevation 735' in Safeguard Building Quench Spray Room, deficient tags have been in place on various components for extended duration which have not been addressed in a timely fashion. Specifically, following anomalies have been noted: A. Component lAC-310: Inadequate thread engagement on packing studs and rust on the valve (Ref Tag ID 60734). 1/2 threads lacking engagement in one stud.B. Component TV-1CC-123-2:

Mild corrosion on the component C. Component IAC-305: Mild corrosion on the component 309 Observation concluded that the anomalies noted above are such that the components are still capable to perform their intended design basis function based on engineering judgment.

Mild corrosion does not affect calculations or drawings, and existing thread engagement was acceptable such that design basis is not affected.Initiated Notification Nos. 600785633, 600786401 and 600786402 to perform the work under work orders 200530162, 200530661, and 200530662.

5. CR-2012-13974 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that at elevation 747' in Safeguard Building Valve Garden, a deficient tag has been in place on the following component since 2003. Ref: Tag ID 45508 Component MOV-1SI-864B (Local Valve Indication Pin sheared off). The local valve indication pin is not required by design basis, and therefore does not violate the design basis requirements.

Initiated Notification No. 600785694 to replace the indication pin and close the deficiency tag under work order 200015714.

6. CR-2012-14020 It was observed that a Security Shield Barrier was located too close to SR Component PT-ICV-lOIB at El. 722' Unit-I -Safeguard Building in Pipe Penetration C Room. This posed a seismic interaction concern that during a seismic event it had a potential to hit the pressure transmitter and potentially make the component not performing its intended design function.The security shift supervisor has been contacted and the shield was moved to a safer location.There is no infringement on the design basis.7. CR-2012-14047 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that Pipe Support H-70 has a missing anchor bolt. One of the six anchor bolts is missing.A review of calculation for Pipe Support H-70 was performed (Ref: Calculation 13387.02-S (B)-24CC2-B), and it concluded that the base plate was qualified with five bolts configuration and the missing bolt was noted. Therefore this is only a configuration issue and not a plant 310 operability concern as the support design is in full compliance with the design basis requirements.

Initiated Notification No. 600785698 to update drawings and install a placard on the support noting that it has been analyzed.

Work to be done under work order 200530170.

8. CR-2012-14321 While performing seismic walkdowns per NRC Letter 50.54f Section 2.3, it was observed that there is a potential of seismic interaction between the Switch Board DC-SWBD1-2 and Battery Charger BAT-CHG1-2 as they are not connected to each other. These components are located side by side with zero gaps in between them. During a seismic event, there is a potential for these components to bang against each other and cause chatter of relays inside the battery charger.These components are located inside DF Switchgear Room at El 713' in U-1 Service Building.This does not violate the design basis documents.

==

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.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 6-5 and 6-6 in accordance with EPRI Document 1025286 and titled as "Potentially Adverse Seismic Conditions Identified from Component and Area Walk-Bys." Table 6-5 identified six potentially adverse seismic conditions, which resulted in generating one condition report. Adequate justification is documented in the checklists that provide the basis as why the remaining issues had insignificant impact on the design of the components and that the components are still capable of performing their intended design function while still meeting the design basis requirements.

Table 6-6 identified ten potentially adverse seismic conditions.

Seven of these conditions were entered in the corrective action program (CAP). Again, adequate justification is documented in 311 the checklists that provide the basis as to why the remaining issues had insignificant impact on the design of the surrounding components and that the components are still capable of performing their 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

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 appropriately 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.312

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 Unit 1 Probabilistic Risk Assessment, Individual Plants Examination of External Events", Submitted June 30, 1995 in response to U.S. Nuclear Regulatory Commission Generic Letter 88-20 Supplement 4, Duquesne Light Company.4. "Beaver Valley Power Station, Unit No. 1 and No. 2, BV-1 Docket No. 50-334, License No. DPR-66, BV-2 Docket No. 50-412, License No. NPF-73, Response to NRC request for Additional Information Regarding Unresolved safety Issue A-46 (TAC No. M69428), letter from Sushil C. Jain to U.S. Nuclear Regulatory Commission dated May 16, 1997.5. "Completion Report for Generic Letter 87-02, USI A-46 Program", Letter from T. S.Cosgrove and M. P. Pearson to L.W. Myers dated July 20, 2000, ND1NSM:9113.

First Energy Operating Company, Beaver Valley Power Station, Plant Services Department, Licensing Section, Unit 1.6. "Beaver Valley Power Station Unit 1, Spent Fuel Pool Cooling Trouble", Abnormal Operating procedure lOM-53C.4.1.20.1, Revision 1, November 16, 2011.7. "Beaver Valley Power Station Unit 1, Updated Final Safety Analysis Report", Revision 24, Section 9.5.3.1.8. "Beaver Valley Power Station Unit 1: Probabilistic Risk Assessment Update Report", Issue 5, December 31, 2010, First Energy Nuclear Operating Company.9. RG 1.29, Rev. 3, "Seismic Design Classification." 10. RG 1.60, Rev. 1, "Design Response Spectra for Seismic Design of Nuclear Power Plants." 11. RG 1.61, "Damping Values for Seismic Design of Nuclear Power Plants." 12. IEEE 344-1971, "IEEE Guide for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations." 13. ASME Boiler and Pressure Vessel Code Section III 1974 including Winter Addenda 1975.313 APPENDIX A RESUMES AND QUALIFICATIONS A-i 279 Dorchester Rd, Akron Ohio Phone 234-678-8262 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 80 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.

1980 -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.1979-1980 Wagner Martin Mechanical Contractors Richmond, IN Engineer/Project Manager Responsible for sprinkler 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 DONALD J. WAKEFIELD DONALD J. WAKEFIELD PROFESSIONAL HISTORY ABSG Consulting Inc., Irvine, California Senior Consultant, Operational Risk and Performance Consulting, 2000-Present EQE International, Senior Consultant, 1997-2000 PLG, Inc., Irvine, California, Senior Consultant, 1983-1997 Cygna Energy Services, Associate, 1981-1983 General Atomic Company, Engineer, 1974-1981 PROFESSIONAL

SUMMARY

Mr. Donald J. Wakefield has more than 30 years experience in all phases of the risk analysis of nuclear power plants and other complex facilities, including human reliability analysis.

He has served as principal investigator and project manager for the risk assessment of several nuclear plants in the United States and the Far East. He served as a key risk analyst on assessments of a floating, production, offloading and storage facility (FPSO), an oil tanker, and for the handling of abandoned chemical weapons in China. Mr. Wakefield is also Project Manager for the development of ABS Consulting's RISKMAN software for risk assessment applications.

He is now serving as the Chairman of the Low Power and Shutdown PRA Standard Writing Group (ANS 58.22) and serves on the ASME's Committee on Nuclear Risk Management (CNRM) and ANS's RISC Committee.

PROFESSIONAL EXPERIENCE In late 2006, Mr. Wakefield became the writing group chairman for the ANS PRA standard for Low Power and Shutdown Events (ANS-58.22).

This standard is still in development.

Mr.Wakefield has also been active recently in the modeling of shutdown events. He recently performed a review of the Seabrook Station, all power modes PRA model. He recently performed a Level 2 analysis for shutdown events of the KKG plant in Switzerland.

These efforts are in addition to his past Level I shutdown studies for HIFAR in Australia, Takahama-3/4, and for other plants in Japan.Mr. Wakefield recently served as the principle investigator for a fire risk analysis of the Watts Bar unit 2 plant to satisfy its FIVE licensing requirement.

This study was performed using CAFTA.A-5 ADS Consulting AN ABS GROUP COMPANY DONALD J. WAKEFIELD Mr. Wakefield has also performed human reliability analysis for nuclear plants. He served as task leader for the human factors analysis of the Three Mile Island (TMI) Unit 1 PSA.Performed the original human factors analysis for the PSA and then, nearly 20 years later, worked with the plant safety staff to update the analysis using the EPRI HRA Calculator.

More recently, Mr. Wakefield served as an independent reviewer for the South Texas Project upgrade to the latest EPRI HRA Calculator, and for a similar review effort for PG&E. Mr. Wakefield was co-author of the Electric Power Research Institute (EPRI) report on the SHARP-1 approach to HRA analyses for PSAs.Mr. Wakefield served as principal investigator for the Beaver Valley Units 1 and 2 PSA performed to satisfy U.S. Nuclear Regulatory Commission (USNRC) IPE and IPE for external event (IPEEE) requirements.

Mr. Wakefield also provided expertise in developing and analyzing the Sequoyah and Watts Bar PSA plant models to satisfy the individual plant examination (IPE).Mr. Wakefield served as project manager for the Salem PSA update and as technical consultant for a PSA of the new production (i.e., weapons materials) modular gas-cooled reactor.Mr. Wakefield was a key contributor to accident sequence modeling, including human factors analysis, and seismic analysis for the Diablo Canyon PSA.Mr. Wakefield served as principal investigator in charge of extending a fault tree linking PSA plant model for a pressurized water reactor in the Far East to accommodate the assessment of plant internal fires and seismic events.Mr. Wakefield served as consultant specializing in accident sequence modeling and plant systems analysis for probabilistic safety assessments (PSA). Recently, he served as technical advisor and sequence model architect for a risk assessment model for the excavation and disposal of abandoned chemical weapons in China. The study considered weapon handling errors, plant fires and weapon explosions there from. This assessment looked at all initiating events and the sequence development extended to payouts resulting from worker and population exposures, building and equipment losses and from environmental cleanup costs.Mr. Wakefield served as the technical lead and coordinated inputs from the Knoxville, San Antonio, and Irvine offices for use by the ABS Tokyo office.Mr. Wakefield served as senior analyst for the development of a QRA model for a Floating, Production, offloading and Storage (FPSO) facility hypothetically located in the Gulf of Mexico.This model, funded internally by ABS, looked at risk to the workers from pool fires and jet fires and environmental damage from potential oil spills. Also, in 1995, he performed risk assessment portion of an explosion analysis for the Agbami FPSO owned by Star Deep Water Petroleum Limited, and one for the GX Platform owned by Exxon Mobil for Mustang Engineering.

He also served as advisor for the PSA of a new, double-hulled oil tanker.Mr. Wakefield developed the CAFTA-based accident sequence model for a seismic margins assessment for the ACR-700 design for AECL.A-6 ABS Consulting AN ABS GROUP COMPANY DONALD J. WAKEFIELD Mr. Wakefield served as instructor for numerous PSA courses and provided extensive utility training sessions both in the U.S. and abroad. He served as course instructor to the US Nuclear Regulatory Commission for the risk assessment of external events and to describe the large event tree approach to sequence modeling.Mr. Wakefield provides technical direction and project management for the development of ABS Consulting's RISKMAN PSA software and administers the RISKMAN Technology Group (a utility users' group). This user's group, now in its eighteenth year, funds the maintenance and development of RISKMAN upgrades.

Mr. Wakefield provides the interface between the user's group members, and the RISKMAN development team.Mr. Wakefield was a substantial contributor to a 5-year high temperature gas-cooled reactor (HTGR) risk assessment study. He developed numerous improvements to severe accident consequence computer programs for the HTGR. Quantified uncertainties in severe accident source terms and dose assessment for the HTGR, the first such assessment ever accomplished for any reactor type. Developed a procedure for prioritizing HTGR safety research programs using PSA and formulated an initial set of research recommendations.

Prepared test specifications to implement research recommendations.

Mr. Wakefield has authored numerous scientific papers on the subject of probabilistic risk assessment methods including such topics as importance measures, comparison between event tree and fault tree linking, and human reliability analysis techniques.

EDUCATION M.S., Nuclear Engineering, University of California, Berkeley, 1974 B.S., Engineering Mathematics, University of California, Berkeley, 1973, with highest honors MEMBERSHIPS, LICENSES, AND HONORS American Nuclear Society Phi Beta Kappa, National Scholastic Honor Society Tau Beta Pi, National Engineering Honor Society Regents Fellowship, University of California, 1974 Department of Engineering Certificate Award, 1973 SELECTED PUBLICATIONS Wakefield, D.J., and Y. Xiong, "Importance Measures Computed in RISKMAN for Windows," PSAM 5, 5th International Conference on Probabilistic Safety Assessment and Management, November 2000.Johnson, D. H., D. J. Wakefield, and R. Cameron, "Use of PSA in Risk Management at a Research Reactor," presented at the American Nuclear Society, International Topical Meeting on Probabilistic Safety Assessmnent (PSA '99), Washington, D.C., August 22-25, 1999.Quilici, M., W. T. Loh, and D. J. Wakefield, "IPEEE Reports Survey," prepared for Computer Software Development Co., Ltd., Tokyo, Japan, PLG-1194, March 1998.A-7 ADS Consulting AN ABS GROUP COMPANY DONALD J. WAKEFIELD Wakefield, D. J., "PSA and RISKMAN Software Training Course," presented to Tennessee Valley Authority, Newport Beach, California, PLG-1195, February 2-6, 1998.Wakefield, D. J., and D. H. Johnson, "A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor," prepared for Department of Industry, Science and Tourism, Canberra, Australia, PLG-1200, January 1998.Wakefield, D. J., and D. H. Johnson, "Summary Report -A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor," prepared for Department of Industry, Science and Tourism, Canberra, Australia, PLG-1201, January 1998.Wakefield, D. J., and D. H. Johnson, "Technical Summary Report -A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor," prepared for Department of Industry, Science and Tourism, Canberra, Australia, PLG-1202, January 1998.Wakefield, D. J., M. A. Emerson, K. N. Fleming, and S. A. Epstein, "RISKMAN A System for PSA," Proceedings, Probabilistic Safety Assessment International Topical Meeting, Clearwater, Florida, pp. 722-729, January 1993.Wakefield, D. J., R. K. Deremer, and K. N. Fleming, "Accident Management Insights Obtained During the Beaver Valley Unit 2 Individual Plant Examination Process," Proceedings, Probabilistic Safety Assessment International Topical Meeting, Clearwater, Florida, pp. 1049-1053, January 1993.Contributing Author to: "Sequoyah Nuclear Plant Unit 1 Probabilistic Risk Assessment Individual Plant Examination," PLG, Inc., prepared for Tennessee Valley Authority, 1992."Watts Bar Nuclear Plant Unit 1 Probabilistic Risk Assessment Individual Plant Examination," PLG, Inc., prepared for Tennessee Valley Authority, 1992.Wakefield, D.J. and S.A. Nass, "Application of RISKMAN 2.0 to the Beaver Valley Power Station IPE," Probabilistic Safety Assessment and Management Conference, Beverly Hills, California, February 1991.Read, J.W., and D.J. Wakefield, "Diesel Generator Technical Specification Study for Indian Point 3," PLG, Inc., prepared for New York Power Authority, PLG-0690, December 1989.Wakefield, D.J., K.N. Fleming, et al., "Beaver Valley Unit 2 Probabilistic Risk Assessment," PLG, Inc., prepared for Duquesne Light Company, December 1989.Wakefield, D.J., H.F. Perla, D.C. Bley, and B.D. Smith, "Enhanced Seismic Risk Assessment of the Diablo Canyon Power Plant," Transactions of the Tenth International Conference on Structural Mechanics in Reactor Technology, Los Angeles, August 1989.Wakefield, D.J., H.F. Perla, et al., "Seismic and Fire Probabilistic Risk Assessment for a Typical Japanese Plant," PLG, Inc., prepared for Mitsubishi Atomic Power Industries, Inc., February 1988.Wakefield, D.J., "Three Mile Island Unit 1 Probabilistic Risk Assessment," PLG, Inc., prepared for GPU Nuclear Corporation, November 1987.A-8 ADS Consulting AN ABS GROUP COMPANY DONALD J. WAKEFIELD Wakefield, D.J., and C.D. Adams, "Quantification of Dynamic Human Errors in the TMI-1 PRA," International Topical Conference on Probabilistic Safety Assessment and Risk Management, Zurich, Switzerland, September 1987.Fray, R.R., B.D. Smith, R.G. Berger, M.L. Miller, H.F. Perla, D.C. Bley, D.J. Wakefield, and J.C.Lin, "Probabilistic Risk Assessment for Pacific Gas and Electric Company's Diablo Canyon Power Plant," presented at the International Conference on Radiation Dosimetnr and Safety, Taipei, Taiwan, March 1987.Wakefield, D.J., A. Singh, et al., "Systematic Human Action Reliability Procedures (SHARP)Enhancement Project; SHARP1 Methodology Report," PLG, Inc., prepared for Electric Power Research Institute, 1987.Wakefield, D.J., "Salem Nuclear Generating Station Reliability and Safety Management Program: Baseline Safety Assessment," PLG, Inc., prepared for Public Service Electric and Gas Company, July 1986.Wakefield, D.J., "PRA Procedures for Dependent Events Analysis, Volume II, Systems Level Analysis," PLG, Inc., prepared for Electric Power Research Institute, December 1985.PLG, Inc., "Application of PRA Methods to the Systems Interaction Issue," prepared for Electric Power Research Institute, PLG-0284, April 1984.Wakefield, D.J., D.C. Iden, and G. Paras, "Oyster Creek Conceptual HPCI System Risk Reduction Study," prepared for GPU Nuclear Corporation, PLG, Inc., PLG-0308, December 1983.Wakefield, D.J., R.K. Deremer, et al., "Probabilistic Risk Assessment and Systems Interaction Analysis Reference Manual," Cygna Energy Services Report to Texas Utilities, October 1982.Wakefield, D.J., and D. Ligon, "Quantification of Uncertainties in Risk Assessment Using the STADIC Code," International American Nuclear Sociehy/European Nuclear Society Topical Meeting on Probabilistic Risk Assessmnent, Port Chester, New York, September 20-24, 1981.Fleming, K.N., D.J. Wakefield, et al., "HTGR Accident Initiation and Progression Analyses Phase II Risk Assessment," United States Department of Energy Report, GA-A15000, UC-77, April 1978 A-9 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 Gbsgen 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-10 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-Il 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-12 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.* Managed the design and construction of seismic retrofits for production equipment and storage tanks at Coca Cola Co. in Japan." Seismic evaluation and design of retrofit for equipment, piping and structures at the UDS AVON Refinery located in Richmond, CA.* 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.

• 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.

• 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-13 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-14 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-15 ABS Consulting AN ABS GROUP COMPANY EDDIE M. GUERRA, E.I.T.Wind: Analysis and Design Revision of Wind Turbine Tower -October 2010 -Februanr 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-16 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-17 I -

ABS Consulting AN ABS GROUP COMPANY ADAM HELFFRICH, E.I.T.PROFESSIONAL HISTORY ABSG Consulting Inc., Contractor, Presently Paul C. Rizzo Associates, Inc., Pittsburgh, PA, Assistant Project Engineer, 2009-Present Penn DOT, Clearfield, PA, Intern, May 2008-August 2008 TNS, Indiana, PA, Surveyor, April 2007-August 2007 Shaler Area School District, Glenshaw, PA, Maintenance, May 2005-August 2006 PROFESSIONAL

SUMMARY

Mr. Adam Helffrich joins Paul C. Rizzo Associates, Inc. (RIZZO) as a Project Engineering Associate.

He recently received his Bachelor of Science in Civil Engineering from the University of Pittsburgh.

Prior to graduating, Mr. Helffrich was an Engineering Intern with RIZZO.PROFESSIONAL EXPERIENCE UAE Site A (Alternate)

NPP Site Selection/Site Characterization/PSAR and EIA -ENEC/KEPCO E&C, United Arab Emirates: May 2009- August 2009 RIZZO prepared the site investigation and submittal of a PSAR and ER to the Regulatory Authority for the siting of Nuclear Power Plants (technology to be decided).

Mr. Helffrich developed and reviewed boring logs for both sites; constructed drawings of cross sections for a site; and performed several checks and modifications to figures and slides for presentation purposes.Calvert Cliffs NPP Unit 3 -UniStar, Calvert County, Maryland: May 2009 -August 2009 Mr. Helffrich was responsible for cutting several cross sections of the sub surface for analysis purposes.A-18 ABS Consulting AN ABS GROUP COMPANY ADAM HELFFRICH, E.I.T.PREVIOUS EXPERIENCE Penn DOT -Clearfield, Pennsylvania:

May 2008 -August 2008 Intern: Conducted STAMPP program for roadway safety;Worked independently and unsupervised through several counties;Studied technical diagrams of roadways and foundations; and Applied gathered knowledge in roadway safety reports.TNS -Indiana, Pennsylvania:

April 2007 -August 2007 Surveyor: Conducted Research surveys and polls for various clients Shaler Area School District -Glenshaw, Pennsylvania:

May 2005 -August 2006 Maintenance:

Light Construction/Building Maintenance Janitorial EDUCATION 3-2 Pre-Engineer Program, Indiana University of Pennsylvania, Indiana, PA, Graduated 2008 COMPUTER SKILLS C++, Mathematica, AutoCAD A-19 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* Completed the Boiling Water Reactor (BWR) Plant Certification Course for Nine Mile Point Unit- I 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-20 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-21 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-22 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-23 Richard P. Mueller Street Address 1116 Vine Street, East Livrpool, Oh, 43920 Phone Number 330-3854-S63 Email address mu*126*,icomcast.net Work Experience Duquesne Light Company and First Energy Corporation House and Yard Laborer Coal and ash handles Nuler Operator Beaver Valley Unit 1 Liamsed Reactor Operator Beaver Valley Unit 1 Licensed Senior Reactor Operator Beaver Valley Unit 1 and Unit 2 9/8/19 7 5t D7/31/2011 Education Degrees Penn State University Associate Degree in Nuclear Engineer~ing Technology Nudear Power Plant Related Skills and Experiences Operttions lead for the last three NRC Triennial Fir Protectio inspectons.

Developed the Operations timelnes and strategies for the latest Appendix R and Safe Shutdown procedures for Beaver Valley Unit I and Unit 2 1OCFRS0.59 and Independent Qualified Reviewer (IQK) qualifiecL Opexations lead for refueling outage scheduling and plannmg.Developed alternate strategies and efficiency improvements for Operations procedwues and tests-Assisted in the development of the latest 10 year ASME testing program for Beaver Valley Unit I and Unit 2.A-24 m rme IREARCH INSTII'UE Certificate of Completion John Reddington Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012 .04 *0dA44qjav1III Date ROW~i K Kassawam EPRI Manager.Struotural Relibility

& Integrity A-25 I-U* U Eu U DrcR! i ýl Certificate ofAchlievement This is to Certify that Jo E. Reddinto n has Compktedthe T'ialSQVGA46 Walkdown Screening andSeismic Evaluation Training Course Heldfifovember 20-25, 1987 Richdan G. Starck, NM Asso0it.m Inc. Robert P. Kasawara, EPRI Traming Program Manager I ,!A-26 A-27 ELECTRIC POWER RESEARCH INSTITUTE Certificate of Completion Farzin Beigi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 13, 2012 Date Robert K, Kassawara EPRI Manager, Structural Reliability

& IntV*A-28 I A-29 Certificate of Completion

~Eddie Querra Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns W'*Vk\1Oaoqv, July 6, 2012 Nish R Va4 VP Advanced Eni Projects A-30 I A-31 Certificate of Completion Adam Heffiic/h Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns VP Advanced Eng Projects July 6. 2012 A-32 l!IQ Presents this Certfica te of chievement To Certify That Moha ed F. Alvi, P.E._, i"hasc" te ( ()" n" dh,, d vw n Screenin" andSeismic Evaluation Training Course Hefd November 4 th -9th, 1992 Ned P. Smth, Consmonwealth Edason-4 p SQU(G David A. F-reed, MPR As-iociates SQUG Training Coordinator Robert P. Kassawara, EPRI SQU(G Proam Manager A-33 EMMEM E=-I~ r i I REEARCH INS1ITUTE Certificate of Completion Mohammed Alvi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012P.K ato Date Robrt K Kassewaer EPRI Manager.stuctural Rwliability

& Integdty A-34 A-35 A-36 SQUC Tf rcrtttf iratc (..orf Aiitinrnit U19is is to (Ueriiftj tkat*oadmmcb .Ahti t~iz mr1d.~b the (~~~!aQ4 u.tmxl 'erif.Q~0ursr Augustt 5-- 2-5, 1992.~cLL(y2611'4

~zAA 2 2!J~zizU7 Paul W. Hayes. MPR 4980ciate8 Richard G. Starck 11, MPR Associates A-37