ML13008A041

Beaver Valley Power Station Unit 2, Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report. Cover Through A-40

ML13008A041
Person / Time
Site:	Beaver Valley
Issue date:	10/31/2012
From:	Lucarelli B A, Beigi F FirstEnergy Nuclear Generation Corp, ABS Consulting
To:	Office of Nuclear Reactor Regulation
References
L-12-283
Download: ML13008A041 (223)

Text

Beaver Valley Power Station Unit 2 Near-Term Task Force Recommendation

2.3 Seismic

Walkdown Report October 31, 2012 Prepared by.Reviewed by.F~aweln W~igr ( ConstiI1enig Eddie Grrerra (ABS Cwskiiting) nnLicweIfi (AaS Corsuts~lruqj)

Mohammled AMu J(FNOC)w- ~Approved by.muellwi( (FEwopr -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)

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

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

205 5.1 W alkdown Preparation

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205 5.2 NTTF 2.3 W alkdowns ................................................................................................

206 5.3 Post W alkdown Activities

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206 6.0

SUMMARY

OF THE WALKDOWN results ..........................................................

206 6.1 W alk Down Item s and W alk-By Areas ......................................................................

206 6.2 W alk Down and Area W alk-By Findings ...................................................................

214 6.2.1 Seism ic W alkdown Findings ..........................................................................

214 6.2.2 Area W alk-By Findings ..................................................................................

220 6.3 Configuration Checks .................................................................................................

222 7.0 LICEN SIN G BA SIS EV A LU A TIO N ........................................................................

223 8.0 IPEEE V U LN EA RBILITIES

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227 9.0 PEER REV IEW ..............................................................................................................

228 10.0 REFEREN CES ...............................................................................................................

244 ii List of Tables Table 4-1 Base List 1 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

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

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

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

207 Table 6-2: Beaver Valley 2 NTTF 2.3 W alk-By Areas .............................................................

211 Table 6-3: Beaver Valley 2 NTTF 2.3 Components Categorized by EPRI Classes ..................

213 Table 6-4: Potentially Adverse Seismic Conditions Identified from Seismic Walkdowns

....... 214 Table 6-5: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys ...............

220 List of Figures UFSAR Figure 3.7B-1: Response Spectra SSE ..........................................................................

2 UFSAR Figure 3.7B-2: Response Spectra 1/22 SSE ......................................................................

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

4 Figure 6-1: Maintenanceequipment located inside Reactor Containment Building .................

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

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217 Figure 6-3: View of piping system between Heat Exchangers and Pumps in the SFP area ...... 218 Figure 6-4: Lighting fixtures near Panel PNL-SEQ-244

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219 Figure 6-5: Emergency Diesel Generator Ground Resistor 2EGS-GR2-1

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221 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 iii List of Acronyms AWC Area Walk-by Checklist BV2 Beaver Valley Power Station Unit 2 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 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 Power Station Unit 2 (BV2) 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 2 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 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.Seismic Category I instrumentation and electrical equipment are designed to maintain the capability to: 1. Initiate a protective action during the safe shutdown earthquake (SSE), 2. Withstand seismic disturbances during post-accident operation without loss of safety function.I Instrumentation and electrical equipment are seismically qualified in accordance with general instructions for earthquake requirements (UFSAR Section 3.7B.3.1).

These requirements conform with, and exceed, those outlined in IEEE Standard 344-1971, and are in agreement with the acceptance criteria in SRP 3.10, Rev. 1, 11-75 (NUREG-75-087).

Although not required (due to Beaver Valley's docket date being before October 27, 1972), IEEE 344-1975 was employed for seismic qualification of Seismic Category I electrical equipment when feasible.Instrumentation and electrical equipment may be tested as individual components, as part of a simulated structural section, or as part of a completely assembled module or unit.1-1 -11 FIYEMUECI C151 FIGURE 3.?R- I DESIGN RESPONSE SPECTRA SAFE SHUTDOWN EARTHOUAKE B AVER VALLEY POWER STAITO-UNIT 2 FINAL SAFETY ANALYSIS REPORT FIGURE 378-2 DESIGN RESPONSE SPECTRA 1/2 SAFE SHUTDOWN EARTHQUAKE SEAVER VALLEY POWERSTATION-UNITZ FINAL SAFETY ANALYSIS REPORT UFSAR Figure 3.7B-1: Response Spectra SSE UFSAR Figure 3.7B-2: Response Spectra 1/2/ SSE The response of racks, panels, cabinets, and consoles is considered in assessing the seismic capability of instrumentation and electrical equipment.

As a minimum, mounted equipment is qualified to acceleration levels consistent with those transmitted by supporting structures.

A design objective is to minimize amplification of floor acceleration by supporting members to mounted equipment.

Determination of amplification and seismic adequacy of instrumentation 2

and electrical equipment are implemented by the analysis and testing methods outlined in UFSAR Section 3.7B.3.1.Supports for Seismic Category I electrical equipment, instrumentation, and control systems are seismically qualified by the analysis and testing procedures outlined in Section 3.7B.3. 1.Supports are designed to withstand the combined effects of normal operating loads acting simultaneously with horizontal and vertical components of earthquake loading and must retain their functional capability and structural integrity as applicable.

When qualified by analysis, stress levels permitted under applicable codes. If there are no applicable codes, the stress level under the combined loading for an operating basis earthquake (OBE) does not exceed 75 percent of the minimum yield strength of the material in accordance with the ASTM specification.

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 Sections 2.5 and 3.7 through 3.10 of the Unit 2 UFSAR. These spectra represent earthquake ground motions which are potentially damaging to structures.

While these spectra could be exceeded by ground motion "spikes" above 10 Hz, extensive investigations concerning the effects of these high-frequency motions, both from structure/equipment evaluations as well as seismological considerations, demonstrate the adequacy of the spectra used for design.The horizontal design response spectra used for seismic analysis are shown on UFSAR Figures 3.7B-1 and 3.7B-2. The spectra for the safe shutdown earthquake (SSE) correspond to a maximum ground surface acceleration of 0.125g, and the spectra for the 1/2 safe shutdown earthquake (1/2 SSE) correspond to a maximum ground acceleration of 0.06g. (The operating basis earthquake, which is referenced in Section 3.2, and Regulatory Guide 1.143, is equivalent to 1/2 of the SSE.) These spectra differ from the spectra in Regulatory Guide 1.60. The Beaver Valley Power Station -Unit 2 (BVPS-2) spectra are based in Appendices 2C and 2D of the BVPS-2 PSAR, and as revised in the response to USAEC Regulatory Position 3 of May 25, 1973 (Question 3.15, BVPS-2 PSAR, Amendment 7, July 9, 1973). The vertical design response spectra are taken to be two-thirds of the horizontal design response spectra.For the Beaver Valley Nuclear Power Plant Unit 2 design SSC spectra refer to Figure 2-1 3 1.0 2 4-j M 0.8 0.6 0.4 0.2 0.0 Beaver Valley Unit 2 SSE Spectra (5% damping)0.1 1 10 100 Frequency

[Hz]Figure 2-1: The SSE response spectrum for Beaver Valley Unit 2 was digitized from BV2 FSAR Figure 3.7B-2 3.0 PERSONNEL QUALIFICATIONS The following personnel worked together to formulate the list of selected equipment for the Beaver Valley Nuclear Power Station NTTF Recommendation

2.3 Seismic

Walkdown: " J. Reddington" R. Mueller" D. Reny* F. Beigi The ABS Consulting Walkdown Team consisted of the following individuals: " F. Beigi" E. Guerra* B. Lucarelli Additionally, J. Reddington served as the reviewer of the Licensing Basis and of the Individual Plant Examination External Events (IPEEE). 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 4 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 B. Lucarelli) 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 (PRA) for Beaver Valley Unit 2. An existing internal 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 2 internal events PRA (i.e., working model BV2REV5F based on Reference

2) and the Beaver Valley Unit 2 IPEEE SEL list of 1443 SSCs (Reference 3).Duplicate SSCs, caused by (1) overlap between the two lists and (2) because the PRA contains 5 duplicate basic events for multiple failure modes of an SSC, 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 2 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 word "replace" in titles of existing engineering change packages (ECPs). Both lists were then evaluated to match equipment IDs appearing on Base List 1 with specific ECP numbers, that were judged to be of a major change.6 There were several IPEEE vulnerabilities requiring plant changes identified for the Beaver Valley Unit 2 IPEEE. These vulnerabilities were associated with RCP Seal LOCA, Station Blackout, containment bypass/isolation failure, loss of switchgear HVAC and transients without scram. Modifications performed in response to these vulnerabilities included new operator action procedures for mitigation of loss of emergency switchgear ventilation, failure of 4,160 V fast bus transfer, and battery load shedding.

In addition some hardware modifications were implemented such as capability to crosstie Unit 1 and 2 diesel generators.

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 retained for the selection process." Internal event PRA risk achievement worth (RAW) and Fussell-Vesely importance measures, if available.

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

A total of 10 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 2, 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 2. 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 2. 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.SWEL 1 includes representative items from some of the variations within each of the above attributes.

A total of 109 SSCs were selected.

Beaver Valley Unit 2 plant operations staff was consulted in the SSC selection process. The selected list of SSCs is from most all of the major buildings including the containment.

Two components are from the valve pit and one (Refueling Water Storage Tank) is from the yard. Many of the selected SSCs are from support systems, but there are also SSCs selected from each frontline system. A total of 94 SSCs came from the original IPEEE or current internal events PRA model. Another 10 SSCs came from the list of panels reviewed for the Fire PRA (Fire Panels). SSCs are selected from each of the safety functions, including 7 related to the containment function.

There were 13 SSCs selected that are located in relatively high radiation areas and 11 that are often in damp or humid areas and 2 that 8 are in wet areas. Most SSCs selected are in cool and dry areas. However, 77 are chosen from normally warm areas and 10 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 IPEEE vulnerability 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 FV>.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 September walkdowns.

Those SSC's located in containment were walked down on October 5th 2012 during refueling outage 2R16.4.2 DEVELOPMENT OF SWEL 2 FOR SPENT FUEL POOL RELATED ITEMS For spent fuel pool repeated 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 4 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 4 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.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 2, the spent fuel pool cooling pumps and heat exchangers are Seismic Category 1 and therefore are included on Base List 2 9 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 2, it is not possible to siphon the spent fuel pool level down to less than 10' above the top of the spent fuel rack; i.e., failures resulting in a rapid drain-down cannot occur (Reference 5.). Therefore, the rapid drain-down list of SSCs is empty for Beaver Valley Unit 2.SWEL 2, as defined in the EPRI walkdown guidance is attached as Table 4-4. A total of 10 equipment items are included in SWEL 2.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.5.2 NTTF 2.3 WALKDOWNS The NTTF 2.3 walkdowns at Davis-Besse were performed over a duration of four days from July 11 to July 14, 2012. The overall task man-hour, including walkdowns and post walkdown 146 preparations, was 620 hours0.00718 days <br />0.172 hours <br />0.00103 weeks <br />2.3591e-4 months <br />. 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, A-46/JPEEE SEWS and A-46 calculations.

Anchorage configuration could not be verified for some SWEL components MCCs due to lack of accessibility inside the cabinet or the presence of recently added fire proofing material obscuring the anchorage for panels inside the Control Room. These situations were addressed by verifying that the A-46 anchorage calculations were consistent with the design drawings.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.147 6.0

SUMMARY

OF THE WALKDOWN RESULTS 6.1 WALKDOWN ITEMS AND WALK-BY AREAS The SWEL 1 included a total of 109 components, and SWEL 2 included a total of 7 components.

From this total of 116 components, 108 components were walked down and 8 components were inaccessible and will require walkdown during the next plant's refueling outage. These eight items located inside the Containment Building will be walked down later during the next scheduled plant refueling outage. W.O. # 200529380 has been generated to have these walkdowns performed during 18RFO. Cabinets and panels listed in Table 6-3a will have their walkdowns completed by opening the cabinet doors and inspecting the internals consistent with the FAQ distributed 9-17-12 titled "Opening Cabinets".

These components will be Completed no later then 18RFO. 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 refueling outage scheduled for 2014. The areas walk-bys and the walkdown items are cross correlated on the respective SWCs and AWCs. Table 6-4 provides the total number of walked down components arranged by their respective equipment classes.Table 6-1: Davis-Besse NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment Equip. Class Bldg Floor El Room ID No 2N 15. Battery Racks AUXB 603 428A 2P 15. Battery Racks AUXB 603 428A AF19 0. Other -check/manual valve AUXB 565 237 AF608 8A. Motor-Operated Valves AUXB 585 303 0. Other -check valve or manual BW10_valve AUXB 565 209 BW21 0. Other -check valve or manual AUXB 585 304 valve.Cl 3. Medium Voltage Switchgear AUXB 585 325 Cll-1 12. Air Compressors AUXB 585 318 C1-2 11. Chillers CTMT9 565 217 C21-1 9. Fans AUXB 643 603 C25-3 9. Fans AUXB 585 319 C31-4 9. Fans AUXB 545 105 C3615 20. Instrument and Control Panels AUXB 585 318 C3645 20. Instrument and Control Panels AUXB 585 325 148 Table 6-1: Davis-Besse NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment BdTRo EqID No Equip. Class Bldg Floor El Room C4606 2. Low Voltage Switchgear AUXB 603 428 C5702 20. Instrument and Control Panels AUXB 623 505 C5706 20. Instrument and Control Panels AUXB 623 505 C5712 20. Instrument and Control Panels AUXB 623 505 C5755 20. Instrument and Control Panels AUXB 623 502 C5792A 20A. Inst. in control panel/cabinet AUXB 623 502 LB2 C73-1 9. Fans AUXB 565 237 C78-2 9. Fans AUXB 603 428A CC 1469 7. Pneumatic-Operated Valves AUXB 545 113 CS 1530 8A. Motor-Operated Valves AUXB 585 303 CV159 20A. Inst. in control panel/cabinet CTMT9 565 217 CV-5005 0. Other -check/manual valve AUXB 643 600 CV5070 7. Pneumatic-Operated Valves ANULS 623 127 CV5080 0. Other -check/manual valve ANULS 623 127 D1 3. Medium Voltage Switchgear AUXB 585 323 D DED 1. Motor Control Centers AUXB 603 429 DIN 14. Distribution Panels AUXB 603 429A D2 ED 1. Motor Control Centers AUXB 603 428 D2N 14. Distribution Panels AUXB 603 428B D2P 14. Distribution Panels AUXB 603 428 DA-3783 8B. Solenoid Valves AUXB 585 318 DBC1PN 16. Battery Chargers and Inverters AUXB 603 429 DBC2P 16. Battery Chargers and Inverters AUXB 603 428 DBC2PN 16. Battery Chargers and Inverters AUXB 603 428 DH1O1 0. Other -check valve or manual AUXB 585 312 valve DH77 0. Other -check/manual valve CTMT9 565 214 DH9B 8A. Motor-Operated Valves AUXB 545 225 El 2. Low Voltage Switchgear AUXB 603 429 El lB 1. Motor Control Centers AUXB 585 304 E12B 1. Motor Control Centers AUXB 585 318 E12C 1. Motor Control Centers INTK 576 51 E22-1 21. Tanks and Heat Exchangers AUXB 585 328 E22-2 21. Tanks and Heat Exchangers AUXB 585 328 E27-1 21. Tanks and Heat Exchangers AUXB 545 113 E27-2 21. Tanks and Heat Exchangers AUXB 545 113 E37-1 10. Air Handlers CTMT9 585 317 149 Table 6-1: Davis-Besse NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment Equip. Class Bldg Floor El Room EF12C 1. Motor Control Centers INTK 576 52 Fl 2. Low Voltage Switchgear AUXB 603 428 F108-1 0. Other -sub-component AUXB 585 318 Fl IA 1. Motor Control Centers AUXB 603 427 F1-2 0. Other -EDG Intake Filter INTK 585 50 F12A 1. Motor Control Centers AUXB 603 428 F12D 1. Motor Control Centers INTK 576 52 FD1062 0. Other -Fire Damper AUXB 603 428 FIS 1612 20. Instrument and Control Panels AUXB 585 312 FTHP3C 18. Instrument (on) Racks AUXB 565 208 FV6451 8B. Solenoid Valves AUXB 565 238 FV6452 20A. Inst. in control panel/cabinet AUXB 565 237 HIS 5889A 20. Instrument and Control Panels AUXB 623 505 HIS 7528 20. Instrument and Control Panels AUXB 623 505 HP2B 8A. Motor-Operated Valves AUXB 565 236 HP2C 8A. Motor-Operated Valves AUXB 565 208 HV5314 0. Other AUXB 623 515 IA-636 7. Pneumatic-Operated Valves AUXB 565 208 ICS 11A 7. Pneumatic-Operated Valves AUXB 643 602 K5-1 17. Engine Generators AUXB 585 318 K5-2 17. Engine Generators AUXB 585 319 L311 20A. Inst. in control panel/cabinet AUXB .623 502 L511 20A. Inst. in control panel/cabinet AUXB 623 502 LI- 1525A 20. Instrument and Control Panels AUXB 623 502 LSHHSP9B6

18. Instrument (on) Racks AUXB 623 502 LT-1402 18. Instrument (on) Racks AUXB 623 501 LT-2787 18. Instrument (on) Racks AUXB 585 321A LTSP9A6. 18. Instrument (on) Racks CTMT9 565 220 MS 101 7. Pneumatic-Operated Valves AUXB 643 601 MS5889A 7. Pneumatic-Operated Valves AUXB 565 237 MU242 0. Other -check/manual valve CTMT9 565 214 P14-1 5. Horizontal Pumps AUXB 565 237 P14-2 5. Horizontal Pumps AUXB 565 238 P3-2 6. Vertical Pumps INTK 576 52 P372B 5. Horizontal Pumps AUXB 565 225 P4-1 0. Other -Screen Wash Pump INTK 585 50 P42-1 5. Horizontal Pumps AUXB 545 -105 P43-2 5. Horizontal Pumps AUXB 585 328 150 Table 6-1: Davis-Besse NTTF 2.3 Walkdown Items (SWEL 1+2)Equipment Equip. Class Bldg Floor El Room ID No P58-1 5. Horizontal Pumps AUXB 545 105 PS3689D 18. Instrument (on) Racks AUXB 623 501 PSL 106C 18. Instrument (on) Racks AUXB 565 237 PSL4928A 18. Instrument (on) Racks AUXB 565 237 RC3701 20. Instrument and Control Panels AUXB 585 314 SF11 0. Other -check valve or manual AUXB 585 304 valve SF16 16A 0. Other -check valve or manual AUXB 545 122 valve SF47 0. Other -check valve or manual AUXB 585 312 valve ___ _ 585_312 SP17A7 7. Pneumatic-Operated Valves AUXB 643 602 SW1399 8A. Motor-Operated Valves INTK 565 53 SW3963 7. Pneumatic-Operated Valves INTK 565 53 SW-5896 7. Pneumatic-Operated Valves AUXB 643 603 SW82 0. Other -check/manual valve INTK 565 251 T10 21. Tanks and Heat Exchangers AUXB 565 PT T12 21. Tanks and Heat Exchangers AUXB 623 501 T153-1 21. Tanks and Heat Exchangers YARD YARD YARD T46-1 21. Tanks and Heat Exchangers AUXB 585 321A TE-5329 19. Temperature Sensors AUXB 585 318 TS-5261 18. Instrument (on) Racks AUXB 638 603 XCEI-1 4. Transformers AUXB 603 429 XDF1-2 4. Transformers AUXB 603 428 Y105 20. Instrument.

and Control Panels AUXB 603 429 Y2 14. Distribution Panels AUXB 603 428 YE2B 4. Transformers AUXB 585 304 YF1 14. Distribution Panels AUXB 585 319 YRF2 16. Battery Chargers and Inverters AUXB 603 428 YV2 16. Battery Chargers and Inverters AUXB 603 428 YV4 16. Battery Chargers and Inverters AUXB 603 428 151 Table 6-2: Davis-Besse NTTF 2.3 Walk-By Areas*Room Bldg [ Floor El 50 INTK 585 51 INTK 576 52 INTK 576 53 INTK 565 105 AUXB 545 113 AUXB 545 122 AUXB 545 208 AUXB 565 209 AUXB 565 225 AUXB 565 236 AUXB 565 237 AUXB 565 238 AUXB 565.251 INTK 565 303 AUXB 585 304 AUXB 585 312 AUXB 585 314 AUXB 585 318 AUXB 585 319 AUXB 585 323 AUXB 585 325 AUXB 585 328 AUXB 585 427 AUXB 603 428 AUXB 603 429 AUXB 603 50s AUXB 623 502 AUXB 623 505 AUXB 623 515 AUXB 623 600 AUXB 643 601 AUXB 643 602 AUXB 643 603 AUXB 638 428A AUXB 603 429A AUXB 603 152 Table 6-2: Davis-Besse NTTF 2.3 Walk-By Areas*Room Bldg Floor El 429B AUXB 603 PT AUXB 565 INTK INTK 576* Does not include areas in Containment Building (i.e., Rooms 127, 214, 220, and 317)Table 6-3: Davis-Besse NTTF 2.3 Inaccessible Items on SWEL Equip. ID Description

[ Bldg El Rm #CV5070 VACUUM BREAKERS ANNU 623 127 CV5080 VACUUM BREAKERS ANNU 623 127 C1-2 CAC1-2 Chiller Air condition CTMT9 565 217 CV159 CAC 1-1 DROPOUT REGISTER CTMT9 565 217 DH77 STOP-CHECK VALVE DH 77 CTMT9 565 214 E37-1 CAC COIL 1-1 (SW SIDE) CTMT9 585 317 LTSP9A6 LEVEL TRANSMITTER LTSP9A6 CTMT9 565 220 MU242 STOP-CHECK VALVE MU 242 CTMT9 565 214 Table 6-3a: Davis-Besse NTTF 2.3 Cabinets to be Opened Equipment Equip. Class Bldg Floor El Room Description Closed ID No Cabinet?20. Instrument CONTROL PANEL C3645 and Control AUXB 585 325 (AUX Panels FEEDWATER)

YES 20. Instrument Control Room C5755 and Control AUXB 623 502 Panels cabinet room YES DIN 14. Distribution AUXB 603 429A PNL DIN Panels YES D2N 14. Distribution AUXB 603 428A PNL D2N Panels YES 14. Distribution D2P Panels AUXB 603 428 PNL D2P YES Y2 14. Distribution AUXB 603 428 PNL Y2 Panels YES 20. Instrument Relay Cabinet in RC3701 and Control AUXB 585 314 Mechanical Panels Penetration Room 4 YES 153 Table 6-4: Davis-Besse NTTF 2.3 Components Categorized by EPRI Classes EPRI Equipment Description Components Cat No. Walked Down 0 Other 10 1 Motor Control Centers and Wall-Mounted Contactors 10 2 Low Voltage Switchgear and Breaker Panels 3 3 Medium Voltage, Metal-Clad Switchgear 2 4 Transformers 3 5 Horizontal Pumps 6 6 Vertical Pumps 1 7 Pneumatic-Operated Valves 9 8 Motor-Operated and Solenoid-Operated Valves 8 9 Fans 5 10 Air Handlers 1 a 11 Chillers lb 12 Air Compressors 1 13 Motor Generators 0c 14 Distribution Panels and Automatic Transfer Switches 5 15 .Battery Racks 2 16 Battery Chargers and Inverters 6 17 Engine Generators 2 18 Instrument Racks 9 19 Temperature Sensors 1 20 Instrumentation and Control Panels 15 21 Tanks and Heat Exchangers 18' E37- 1, Located inside Containment b Cl-2, Located inside Containment No Category I Motor Generators at the plant Total 116 6.2 WALKDOWN 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 154 of 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 credible interaction effects and minor discrepancies between existing and as-designed conditions.

Several relatively minor findings are reported here. These are generally of the nature of seismic interactions.

Observations in this respect are organized on the basis of potentially adverse seismic conditions identified during both Seismic Walkdowns and Area Walk-Bys.While performing the area walk-by walkdowns some lack of thread engagement in some bolted connections was observed; however based on Davis-Besse procedure DB-MM-09266 these conditions were judged to be acceptable and required no further action.6.2.1 Seismic Walkdown Findings No potential adverse seismic conditions were identified during the Seismic Walkdowns.

All findings were resolved and judged not to present credible and/or significant seismic concerns based on sound engineering judgment and precedent design documentation.

Field notes and finding resolutions are presented in their respective SWCs included in Appendix B.* 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 have adequate seismic capacity.

Appendix E presents the list of block walls associated with the nearby SWEL item as well as the referenced calculations used for verification of the block wall seismic capacity.Other conditions which were noted but subsequently resolved are briefly described below.* Unprotected Fluorescent Light Tubes During the walkdowns hanging light fixtures without proper cover to prevent falling of dislodged fluorescent tubes were noted in virtually all rooms inspected.

It was subsequently verified that these fluorescent light fixtures are attached via socket & plunger type connection and not the twist-in type. It was further verified from previous SEWS from SQUG documentation 155 (Reference Calculation No. C-CSS-D1) that these fluorescent type fixtures were tested in order to assess the holding capacity of the fixtures.Thus it is concluded that fluorescent tubes do not represent a credible interaction source.Figure 6-1: Unprotected Fluorescent Light Tubes 156

• Wooden Scaffold in Battery Room 429B (Auxiliary Building)A wooden platform in the Battery Room 429B was noted during the walkdowns.

The overall stability of this platform was judged to present a potential interaction hazard for battery racks 2N and 2P. Based on plant's control process log, FENOC personnel confirmed that this scaffold is a temporary structure complying with the working period limits and will be removed after work completion.

That platform was braced, and no further action is required.Figure 6-2: Wooden Scaffold in Battery Room 429B 157 e Missing Nuts along Cooler Fan Housing One comer of the housing for the Cooler Fan 34-1 was missing all screw nuts along the vertical edge connection.

FENOC personnel were informed of the situation.

It was confirmed with maintenance that sheet metal screws are used and nuts are not required Figure 6-3: Missing Nuts along Cooler Fan Housing 158 9 Differential Displacement between Aux. Building and Containment During inspection of valve CS1530 and its attached piping, it was observed that the attached piping crosses from the Auxiliary Building to the Shield Building and that it may not have adequate flexibility to accommodate the differential displacements between these two buildings in a seismic event. The effect of differential building displacements (or Seismic Anchor Movements) on this piping was verified from Pipe Stress Calculation lB R/12 and found acceptable.

Figure 6-4: Support Conditions for Component CS1530 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 23 potentially adverse seismic conditions were identified during the area walk-bys.

Table 6-5 provides as summary of all 23 adverse finding conditions identified.

As shown in Table 6-5, only two condition reports were issued, which would require 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.159 Table 6-5: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys Licensing Reference Description of Adverse Seismic Basis for Room Bldg Floor El Condition Evaluation fon RequiredJustification

______Required Interaction hazards: Maintenance AWC Room 208 AUXB 565 equipment w/o adequate restraint and fire N 208 extinguisher w/o wall strap.Interaction hazards: Maintenance AWC Room 209 AUXB 565 equipment w/o adequate restraint and fire N 209 extinguisher w/o wall strap.225 AUXB 565 RP cart and dolly not restrained.

N AWC Room 225 236 AUXB 565 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. N 236 238 AUXB 565 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 238 AWC Room RC3703 has a 1 inch conduit at the top 314 314 AUXB 585 that is missing a nut on the bracket that Y(CR-2012-attaches the conduit to the unistrut.10920)318 AUXB 585 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 318 319 AUXB 585 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 319 323 AUXB 585 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap and supply cabinet left open. 323 325 AUXB 585 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 325 328 AUXB 585 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 328 427 AUXB 603 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 427 Interaction hazards: Maintenance 428 AUXB 603 equipment w/o adequate restraint and fire N A C o extinguisher w/o wall strap. 428 AWC Room Crack observed in Masonry Wall, 502 502 AUXB 623 Unrestrained trash can, light bulb storage Y(CR-2012 container, and I&C cart. 1012-10973)160 Table 6-5: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys Licensing Reference Room Bldg Floor El Description of Adverse Seismic Basis for Condition Evaluation fon Required Justification

_____ ______ _________________________

Required AWC Room 505 AUXB 623 Small podium not anchored N 505 515 AUXB 623 Dolly loosely tied to column adjacent to N AWC Room MCC. 515 Unrestrained storage containers observed AWC Room 601 AUXB 643 in area N 601 602 AUXB 643 Unrestrained storage containers observed N AWC Room in area 602 Fire extinguishers not restrained.

I&C N AWC Room 603 AUXB 643 Cart not restrained.

603 Anchor threads shown with substantial AWC Room length past nut 251 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 50 52 INTK 576 Interaction hazards: Fire extinguisher w/o N AWC Room wall strap. 52 53 INTK 566 Ladder in area is not restrained N AWC Room 53 As illustrated in Table 6-5, most of the outlined conditions correspond to potential interaction hazards, which were noted but subsequently resolved.

These are briefly described below.* Lateral Restraint of Fire Extinguishers Fire extinguishers were identified in various locations around the plant lacking the proper fixity to restrain against vertical movement during a seismic event. Common practice suggests the use of metal straps around the extinguisher bolted to the wall as well as to provide an adequate overall encasing.

However, based on previous calculations, the vertical peak spectral accelerations are generally less than 1.Og and it is judged that fire extinguishers will remain supported and therefore do not represent a credible seismic condition.

161 Figure 6-5: Typical Wall Mounting for Fire Extinguishers

• Unrestrained Housekeeping/Storage Boxes Various maintenance-related equipment such as ladders, storage boxes and dollies, were found to be loosely tied or without any restraint to prevent contact with surrounding items. For every case where any of these situations were encountered, the SWEs notified the associated owner about the finding, in addition to ensuring its temporary condition in the area. These conditions were judged to not pose a significant seismic interaction on both the operability and integrity of any surrounding component.

No condition report (CR) was written based on judgments made above.Figure 6-6: Unrestrained Maintenance Equipment 162

6.3 CONFIGURATION

CHECKS The SWELL 1+2 included 77 items, which were not in-line components such as valves. The process of verifying the anchorage configuration focused on 45 SWEL components arbitrarily selected prior to walkdown proceedings (this is about 62% of the SWEL items with anchorage configurations).

Appendix D provides a list of the 45 components comprising the anchorage configuration list linked with the specific references used for verification purposes; i.e., A-46/IPEEE SEWS, design drawings, etc.The anchorage configuration for each of the 45 SWEL components listed in Appendix D was verified based on A-46/IPEEE SEWS, A-46 calculation and/or 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 A-46/IPEEE SEWS or A-46 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 45 instances.

7.0 LICENSING

BASIS EVALUATION Two condition reports (CR) were generated as a result of this walkdown.

CR-2012-10920 identifies what appeared to be a missing nut on the strap that holds a conduit to its unistrut support. Upon further inspection with assistance from a qualified electrician, the conduit strap was an approved strap that is "nutless" and relies on thread engagement through the strap itself.The "as-found" strap is a Unistrut model P1112. Direction for the types of straps used is specified in Drawing E-302A, and it does allow this model number of strap or equivalent.

The Unistrut strap was replaced, however, with a type that accepts a nut and, the subject strap was replaced on 8/3/12. The second condition report CR-2012-10973 identified a crack in a concrete masonry unit wall in the cabinet room adjacent to the control room. The crack was evaluated to be a cosmetic, non-structural crack through a mortar joint and did not invalidate the calculation for the wall. These cracks were previously identified during maintenance rule walkdown of Room 503 and dispositioned as not a structural concern.163 Several other questionable items were noted during the walkdown.

These items were researched and validated to be within their design basis. They are noted in Table 6-5, and their resolution is provided in the walkdown checklists.

8.0 IPEEE

VULNERABILITIES A summary of the IPEEE Vulnerabilities is provided in Appendix G. Toledo Edison Serial Number 2316 (August 29, 1995) is the submittal for resolution of Generic Letter 87-02,"Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue A-46." A sampling of these vulnerabilities were verified to have been corrected during both the component and area walkdowns.

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 Davis-Besse Nuclear Power Plant: " Mohammed Alvi (Team Leader)* Tim Ridlon 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 164 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 Davis-Besse Nuclear Power Station acted as Operations representative during the selection of the SWEL.Appropriate figures 1-1, 1-2 and 1-3 of the EPRI Document 1025286 were used and the final SWEL 1 and SWEL 2 were developed.

The peer review confirmed that the following EPRI screens were used in the selection of SWEL 1: Screen 1: Seismic Category I Screen 2:.Equipment or System Screen 3: Support for the five safety functions Screen 4: Sample Considerations The station did use the existing documentation that resulted from IPEEE and USI A-46 programs in identifying the components.

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

The peer review also confirmed that the station used the following EPRI screens in the development of SWEL 2: Screen 1: Seismic Category I Screen 2: Equipment or System 165 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 Davis Besse Nuclear Power Plant included 7 components.

==

Conclusion:==

No major concerns were identified by the peer review team in the selection process for SWEL I or SWEL 2.166 Peer Review Checklist for SWEL Instructions for Completing Checklist This peer review checklist may be used to document the review of the Seismic Walkdown Equipment List (SWEL)in accordance with Section 6: Peer Review. The space below each question in this checklist should be used to describe any findings identified during the peer review process and how the SWEL may have changed to address those findings.

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

YEND See Attached Comments 2. Does SWEL 1 include an appropriate representation of items having the following sample selection attributes:

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

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

Y END See Attached Comments d. Various environments?

See Attached Comments e. Equipment enhanced based on the findings of the IPEEE (or equivalent) program?Y ONO Y END See Attached Comments f. Were risk insights considered in the development of SWEL 1?Y ONE]See Attached Comments 167 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 ONE Y ONEI 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?Y NL1 Peer ________________

C 4r Dale.~Dotw ~ZI2j 168 Peer Review Checklist for SWEL Comments on Question 1: A peer review of the SWEL selected for Davis-Besse Nuclear Power Station 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. 480V B. 4160V C. AC Power D. DC Power E. Auxiliary Feedwater (AFW)F. Borated Water Storage Tank (BWST)-G. Containment Air Cooling (CAC)H. Component Cooling Water (CCW)J. Containment Isolation (CI)J. Core Spray (CS)K: Service Water System (SW)L. Decay Heat Removal (DH)M. Emergency Diesel Generators (EDGs)N. High Pressure Injection (HPI)0. Heating, Ventilation and Air Conditioning (HVAC)P. Instrument Air (IA)Q. Main Steam (MSTM)R. Make Up (MU)169 S. Reactor Coolant System (RCS)T. Reactor Protection System (RCS)U. Safety Features Actuation System (SFAS)V. Steam Feedwater Rupture Control System (SFRCS)W. Vacuum Breakers Comments on Question 2b: The peer reviewers concluded that the selected components represent inclusion of major new and replacement equipment.

Containment Air Coolers were new and installed via ECP No. 03-0533-00.

Reactor Trip Breakers were replaced with Mod No. 00-0031 w/Framotome Reactor Trip Model.Comments on Question 2c: The peer reviewers concluded that the selected components represent various type of equipment installed in the plant. The various equipment types are indicated as follows:-A. Motor Control Centers B. Air Handlers C. Distribution Panels D. Battery Racks E. Battery Chargers and Inverters F. Engine Generators G. Instrument on Racks H. Low Voltage Switchgear I. Instrument and Control Panels J. Instrument in Control Panel Cabinets K. Tanks and Heat Exchangers L. Medium Voltage Switchgear M. Transformers N. Horizontal Pumps 0. Pneumatic Operated Valves P. Motor Operated Valves Q. Solenoid Valves R. Fans S. Temperature Sensors T. Vertical Pumps U. Air Compressors V. Chillers 170 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. High Radiation B. Dry C. Damp D. Cool E. Dry/Wet F. Warm G. Hot Comments on Question 2e: Based on the review, the selected components represent equipment enhanced based on findings of the IPEEE. Approximately 20 percent of the SWEL indicates this selection.

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 Question 3a: Spent Fuel Pool related items were considered and are adequately represented in SWEL 2.Comments on Question 3b: Spent Fuel Pool components were considered and approximately 10 percent were included as part of the sample.171 Comments on Ouestion 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 Davis-Besse Nuclear Power Station 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. However, the lead peer reviewer reviewed 100 percent of the checklists to ensure that all the work has been performed in compliance with the requirements.

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 E. In some cases, weakness in documenting 50 percent anchor check documentation.

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 lead peer reviewer also participated in the walkdowns and observed the work performed by the SWEs during the inspections.

It was noted that the walkdown/inspection was intrusive, walkdown team members discussed, issues amongst themselves, and used engineering 172 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.The lead peer reviewer also observed the configuration check in some cases and assured that the installed configuration did match the plant drawings/documentation.

He also reviewed several masonry wall calculations to make sure that they were seismically designed according to the design basis requirements.

==

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.173 C. Review of the Licensing Basis Evaluations The walkdowns identified several minor anomalies, however two of them resulted in generating condition reports as follows: A. CR-2012-10920:

RC3703 has a 1 inch conduit coming in the top that is missing a nut on the bracket that attaches the conduit to the unistrut.B. CR-2012-10973:

Crack in the concrete masonry wall near the control room.The station performed the licensing basis evaluations for the above two CRs which are documented in Section 7 of this report.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 results were documented in Table 6-5 in accordance with EPRI Document 1025286 and titled as "Potentially Adverse Seismic Conditions Identified from Area Walk-Bys." No potential adverse seismic conditions were identified in association with SWEL 1 list of components.

All findings with SWEL 1 were judged to be acceptable.

Adequate justification is documented in the checklists that provide the basis as why these issues have 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 6-5 identified 23 potentially adverse seismic conditions.

Two of these conditions were entered in the corrective action program (CAP). Again, adequate justification is documented in the checklists that provide the basis as why these 21 issues have insignificant impact on the 174 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.

Section 9.C above discusses the nature of anomalies documented in two condition reports.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 lead peer reviewer participated in the walkdowns, observed the conduct of walkdown team members, and discussed issues while remaining independent.

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

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

Also, the resolution of the issues (License Basis Evaluations) identified in the condition reports was adequate.175

10.0 REFERENCES

1. NRC letter 50.54(f), March 17, 2012.2. Nuclear Regulatory Commission letter, "Request for Information pursuant to Title 10 of the Code of Federal Regulations 50.54(f) regarding Recommendations 2.1, 2.3, and 9.3, of the Near Term Task Force Review of Insights from the Fukushima Dai- Ichi Accident," dated March 12, 2012.3. Nuclear Regulatory Commission letter, "Endorsement of Electric Power Research Institute (EPRI) Draft Report 1025286, "Seismic Walkdown Guidance", dated May 31, 2012.4. EPRI 1025286, "Seismic Walkdown Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic," Final, June 2012.5. Davis-Besse Nuclear Power Station, "Unresolved Safety Issue (USI) A-46, Seismic Evaluation Report," Toledo Edison, August 1995.6. "Individual Plants Examination of External Events for the Davis-Besse Nuclear Power Station, Submitted in Response to U.S. Nuclear Regulatory Commission Generic Letter 88-20 Supplement 4," The Toledo Edison Company, December 1996.7. A Methodology for Assessment of Nuclear Power Plant Seismic Margin, EPRI NP-6041-SL, Revision 1, August 1991.8. System [024] Description for "Spent Fuel Pool Cooling and Cleanup System," Revision 4, Sections 1, 2, and 3, September 2005.176 APPENDIX A RESUMES AND QUALIFICATIONS A-1 279 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 2OOl to August 2003: Davis-Besse Nuclear Station Oak Harbor, OH Supervisor Quality Assurance Oversight for Maintenance:

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

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

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

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

Direct oversight of outage preparations and implementation.

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

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

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

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

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

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

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

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

198o -,1984 Structural Engineer Responsible for steel and concrete design and analysis for LaSalle and Fermi Nuclear Power plants. Performed vibrational load and stability analysis for numerous piping systems. Member of the on-site team of engineers responsible for timely in-place modifications to the plant structure at LaSalle.1979-198o Wagner Martin Mechanical Contractors Richmond, IN Engineer/Project Manager Responsible for 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 1 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 ABS 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 COMPAWY 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 Assessment (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 ABS 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 I 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 ABS 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 Dosimetry 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 Society/European Nuclear Society Topical Meeting on Probabilistic Risk Assessment, 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 Gosgen 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-11 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 -February 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 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 -ENECIKEPCO 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 ABS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.PROFESSIONAL HISTORY ABSG Consulting Inc., Contractor, Presently Paul C. Rizzo Associates Inc., Pittsburg, PA, Engineering Associate II, 2010- Present Engineers without Borders, Aquaculture Development, Makili, Mali, Africa September 2007 -December 2009, Southwestern Pennsylvania Commission, Pittsburgh, Pennsylvania, Transportation Intern, May 2008 -August 2008 PROFESSIONAL

SUMMARY

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

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

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

He has authored numerous calculations and reports related to the work conducted for this project, including responding to requests for additional information from the NRC. His analyses for this project included finite element analyses of the stresses within the Bridging Mat under static and dynamic loading and the determination of long-term settlement at the site.A-20 ADS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.Mr. Lucarelli also authored the Work Plan and served as on-site quality control during laboratory testing of RCC block samples in direct tension and biaxial direct shear. His responsibilities included inspection of the testing being performed and control of documentation related to testing activities.

September 2011 -March 2012 Akkuyu NPP Site Investigation

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

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

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

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

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

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

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

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

Previous Experience:

September 2007 -December 2009 Aquaculture Development

-Makili, Mali, Africa: The University of Pittsburgh Chapter of Engineers Without Borders designed and constructed an aquaculture pond in rural Mali, Africa with a capacity of 3.6 million gallons. This pond is designed to maintain enough water through a prolonged dry season to allow for year-round cultivation of tilapia. As the project technical lead, Mr. Lucarelli was involved in developing A-21 ABS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.conceptual design alternatives and planning two site assessment trips. These scope of these site assessment trips included topographic surveying, the installation of climate monitoring instrumentation, soil sampling and characterization, and laboratory soils testing.As the project coordinator, his primary responsibilities included maintaining a project schedule, developing a budget for project implementation, and coordinating technical reviews of project documentation with a Technical Advisory Committee.

May 2008 -August 2008 Southwestern Pennsylvania Commission

-Pittsburgh, Pennsylvania:

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

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

Engineer-in-Training

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

-Associate Member American Society of Civil Engineers (ASCE)Engineers Without Borders (EWB)A-22 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-23 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 I & 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-24 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-25 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 Uniit-I Nuclear Station PROFESSIONAL SOCIETY MEMBERSHIP:

0 Member, American Society of Civil Engineers (ASCE)

REFERENCES:

Provided upon request CITIZENSHIP:

Citizen of the United States of America A-26 I 2 LETIPOE Certificate of Completion John Reddington Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012.Date Robert K Kamawara EPIRI Manager.Strurtural Reliability

& Integrity A-27 U Certificate of Achlievement This is to Certify that i!:i ihln!

has Completedathe T'iiaSQUsqGA46 Walkdown Screening andSeismic Evaluation Training Course 1HeldfJNovember 20-25, 1987 Richard 0. POR Assaciatcs, Inc. Robert P, Kassawara, EPRI Tra ining Program Manager t J A-28 A-29 5I~~f~ II LCtIIC POW!R IUI$ARCH INSJIUIt Certificate of Completion Farzin Beigi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 13, 2012ift 04 644 Dae Robed K, Kesmae.m EPM Muwgr, Structura Reftbilfy

& Integty A-30 SQU(j Certificate ofAchievement to Certify that Eddie M. Guerra has Complted the SQU G Wakdown Screening andSeismic Evaluation Training Course t. 90 June 11-15,2012 GQenAflen, Virginia I .ua 1"3S Coia6a SQUGW=m A-31 Certificate of Completion Tdde Guerra Trainlng on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns Nish R. V a VP Advanced Ert Projects July 6, 2012 A-32 A-33 Certificate of Completion Adaim Yefe(ffic/

Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns Nish R VP Advanced Eng Projects July 6. 2012 A-34 Certificate of Co pletion Brian Lucarefli Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns I154 FS J, July 6, 2012 Nish RL VaIdy.VP Advanced Eng Projects A-35 U!Presents this Certificate ofl chievemen t To Certify That has. Compfetedthe SQ.QG Walkfdo-wn Sceezg andSeismic Evaluation Training Course Hfeld November 4 th -9 th, 1992 Neil P. Smith, (onimion~valth bliron SQUG, Owamnan David A. Freed, MPR Associates SQUG Training Coordmiuitor Robert P. Kassawara.

EPRI SQUG Program Manager--- ------- -- ------- --- ------ ---- ------- ---- -- ---- -A-36 m ELCTR~IC POWEP aaf~ a l IESEARCH INS1170TS Certificate of Completion Mohammed Alvi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012 Date Ift 00-P 444a4146ýRobert K. Kassawara EPRI Manager, Structural Reliability

& Integrity A-37

~. * -@-.. -.4 SQU(Certificate of Achievement 4 4 This is to Certify that Mohammed F. Alvi 4 has Completed the SQUG Training Course for Demonstrating Seismic Adequacy of New and Replacement Equipment and Subcomponents Using GIP and STERI Methods Held September 19-21, 1994 4 Neil P. Smith. Commonwealth Edison SQUG Chairman 4 Patrick Butler, MPR Associates Course Coordinator Robert P. iassawaro, EPRI SQUG Program Manager I A-38

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