ML13008A031

Davis-Besse Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Cover Through Appendix a, Resumes and Qualifications, Page A-39

ML13008A031
Person / Time
Site:	Davis Besse
Issue date:	08/10/2012
From:	Lucarelli B A, Wakefield D, Beigi F, Guerra E, Helffrich A L, Reddington J ABS Consulting, FirstEnergy Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
References
L-12-283
Download: ML13008A031 (194)

Text

Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2,3 Seismic Walkdown Report August 10, 2012 Prepared by: Dontild WaRSf~DO (AUS ConSufling)

Fanzl *Ogl 1.-6ifA-C'Edcde Gudrra (A13"OnsfIng).

Reviewed by: Approved by: MTohammfted AMV (FENOGQ--TIM Rldlon (FENUC)Greg Ich FENOC)*I; $ectlons 1, 3,4, 5, 6, ahd 10 have been prepared by ABS Consulflng, Sections 2, 7, 8, and 9 have be n prepared by FENOC.2.. Thie r6VIW Pand approval ofthils doeument by FENOC personnel constitutes the owner acceptance.

ofwbk per6rrmed by ABS'Consulling.

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

2 4.0 SELECTIO N O F SSCs ...............................................................................................

3 4.1 Development of the SWEL 1 List (Related to Key Safety Functions)

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

3 4.2 Development of SWEL 2 for Spent Fuel Pool Related Items ...................................

6 5.0 SEISMIC WALKDOWN AND AREA WALK-BYS ..............................................

146 5.1 W alkdown Preparation

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

146 5.2 NTTF 2.3 W alkdowns ................................................................................................

146 5.3 Post W alkdown Activities

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

147 6.0

SUMMARY

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

148 6.1 W alkdown Item s and W alk-By Areas ........................................................................

148 6.2 W alkdown and Area W alk-By Findings .....................................................................

154 6.2.1 Seism ic W alkdown Findings ..........................................................................

155 6.2.2 Area W alk-By Findings .................................................................................

159 6.3 Configuration Checks .................................................................................................

163 7.0 LICEN SIN G BA SIS EV A LU A TIO N ........................................................................

163 8.0 IPEEE V U LN ERA BILITIES .......................................................................................

164 9.0 PEER REV IEW ..............................................................................................................

164 10.0 REFEREN CES ...............................................................................................................

176 ii List of Tables Table 4-1: Base List 1 -The Equipment Coming Out of Screen #3 and Entering Screen #4, for Five Safety Functions

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

.8 Table 4-2: SWEL 1 -Selected Equipment for Five Safety Functions

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

130 Table 4-3: List of Equipment Enhanced due to Vulnerabilities Identified During the A -46/IPEEE Program s ..................................................................................................

139 Table 4-4: Base List 2 -List of All SSCs for Spent Fuel Pool ..........................................

143 Table 4-5: SWEL 2 (Spent Fuel Pool) ..................................................................................

145 Table 6-1: Davis-Besse NTTF 2.3 Walkdown Items (SWEL 1+2) ...................................

148 Table 6-2: Davis-Besse NTTF 2.3 Walk-By Areas .............................................................

152 Table 6-3: Davis-Besse NTTF 2.3 Inaccessible Items on SWEL ................

153 Table 6-3a: Davis-Besse NTTF 2.3 Cabinets to be Opened ..............................................

153 Table 6-4: Davis-Besse NTTF 2.3 Components Categorized by EPRI Classes ............

154 Table 6-5: Potentially Adverse Seismic Conditions Identified from A rea W alk-Bys ..............................................................................................................

160 List of Figures Figure 2-1: Figure 6-1: Figure 6-2: Figure 6-3: Figure 6-4: Figure 6-5: Figure 6-6: SSE Response Spectrum ...................................................................................

1 Unprotected Fluorescent Light Tubes .............................................................

156 Wooden Scaffold in Battery Room 429B ........................................................

157 Missing Nuts along Cooler Fan Housing ........................

158 Support Conditions for Component CS1530 .................................................

159 Typical Wall Mounting for Fire Extinguishers

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

162 Unrestrained Maintenance Equipment

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

162 List of Appendices A. RESUMES AND QUALIFICATIONS B. SEISMIC WALKDOWN CHECKLISTS (SWCs)C. AREA WALK-BY CHECKLISTS (AWCs)D. COMPONENT LIST FOR ANCHORAGE CONFIGURATION CHECK E. MASONRY BLOCK WALLS VERIFIED UNDER IE BULLETIN 80-11 F. DAVIS-BESSE DESIGN CRITERIA MANUAL G. DAVIS-BESSE A-46/IPEEE VULNERABILITIES iii List of Acronyms AWC Area Walk-By Checklist BWST Borated Water Storage Tank COLA Combined Construction and Operating License Applications DB Davis-Besse Nuclear Power Station DHR Decay Heat Removal EPRI Electric Power Research Institute FENOC FirstEnergy Nuclear Operating Company GIP Generic Implementation Procedure IPEEE Individual Plant Examination of External Events LERF Large Early Release Frequency LOCA Loss of Coolant Accident MCC Motor Control Center MWO Maintenance Work Order NPP Nuclear Power Plant NSSS Nuclear Steam Supply System NTTF Near-Term Task Force PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor RAW Risk Achievement Worth SEL Seismic Equipment List SQUG Seismic Qualification Utility Group SRO Senior Reactor Operator SSC Structures, Systems, and Components SWC Seismic Walkdown Checklist SWE Seismic Walkdown Engineer SWEL Seismic Walkdown Equipment List iv

1.0 INTRODUCTION

This Report presents the results of the Seismic Walkdown conducted for the Davis-Besse Nuclear Power Station (DB) 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 Updated Safety Analysis Report and implemented through the "Davis-Besse Design Criteria Manual."Section II.E "Seismic Design" of the"Davis-Besse Design Criteria Manual" is attached to this report in Appendix F. In summary, the maximum possible earthquake or Safe Shutdown Earthquake is a horizontal peak ground acceleration of 0. 15g (zero period acceleration).

An example of the 0. 15g SSE at 5% damping is given in figure 2-1.0 4-, U U Z 0.7 0.6 0.5 0.4 0.3 0.2 0.1 Davis-Besse SSE Spectra (5 % Damping)0 4-0.01 0.1 1 10 Frequency

[Hz]100 Figure 2-1: The SSE response spectrum for Davis-Besse is created based on Table 2C.3-6 of Davis-Besse FSAR 1

3.0 PERSONNEL

QUALIFICATIONS The following personnel worked together to formulate the list of selected equipment for the Davis-Besse Nuclear Power Plant NTTF Recommendation

2.3 Seismic

Walkdown: " J. Reddington" D. Wakefield* F. Beigi The ABS Consulting Walkdown Team for the DB 2.3 Seismic Walkdown consisted of the following individuals:-

• F. Beigi" E. Guerra" A. Helffrich* B. Lucarelli Additionally, J. Reddington served as the reviewer of the DB Licensing Basis and of the DB Individual Plant Examination External Events (IPEEE), and M. Alvi served as the peer reviewer for the walkdown.All of these individuals possess technical degrees from accredited universities and have been trained in the application of seismic experience data for seismic verification of nuclear power plant (NPP) structures, systems, and components (SSC). In addition to completion of the NTTF 2.3 training provided by EPRI, five of these individuals (J. Reddington, M. Alvi, F. Beigi, E. Guerra, and A. Helffrich) have also completed the EPRI Seismic Qualification Utility Group (SQUG) training.

Resumes and certificates of the walkdown team members are presented in Appendix A of this report.All of the individuals have experience in earthquake engineering and seismic analysis.

This experience relates to seismic qualification of equipment and seismic evaluations supporting recent Combined Construction and Operating License Applications (COLA). Additionally, the team collectively represents previous NPP 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 DB IPEEE evaluation, these individuals also supported equipment selection, walkdown 2 planning, equipment location determination, and selection of walk-by areas for the DB 2.3 Seismic Walkdown.4.0 SELECTION OF SSCs Consistent with the guidance in EPRI 1025286, "Seismic Walkdown Guidance," (Reference 3)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

3) 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 Davis-Besse.

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 Davis-Besse internal events PRA (i.e., model PRA-DB 1-AL-R05) and the Davis-Besse IPEEE SEL list of SSCs (Reference 5). Duplicate SSCs, caused by (1) overlap between the two lists and (2) because the PRA contains basic events for each failure mode of the SSC, were removed.Information about the original source of the remaining SSCs was retained.

In short, we have gone beyond the requirements in the EPRI walkdown guidance document in preparing the SSC SEL list. 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, that would be useful for PRA purposes, were also chosen 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.

Again, duplicate SSCs were eliminated.

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

3) were also reviewed for completion.

Some SSCs were added as a result of this review.The list of SSCs, evaluated in the original Diablo Canyon seismic PRA, was also reviewed as a comparison with another pressurized water reactor (PWR), for completeness.

Parts of the emergency diesel generator system were considered as additional SSCs so as to avoid excessive screening using the rule of the box. Nuclear steam supply system (NSSS) related SSCs were added to the complete list, though they are not required for this application.

Also excluded are 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, new equipment, added to the plant since the performance of the IPEEE and the last Davis-Besse internal events PRA update are noted in a separate column of the developed lists.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 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" 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 4 guidance.

Panels not previously evaluated for their associated safety functions were assigned the designator, "operator," and retained for the selection process.Internal event PRA risk achievement worth (RAW) and Fussell-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 3)used for fragility analysis.

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

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

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

All of the EPRI categories were later employed as part of the SSC selection process.Except for EPRI Category 13 (motor generators) at least one SSC was selected from each EPRI category.Base List 1, as defined in the EPRI walkdown guidance is attached as Table 4-1 for Davis-Besse.

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 I 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 Davis-Besse.

This is also indicated in the SSC source column.SWEL 1, as defined in the EPRI walkdown guidance (Reference

3) 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. Table 4-3 summarizes the IPEEE and related A-46 program (Reference

4) findings and cross references them to equipment IDs.5 SWEL 1 includes representative items from some of the variations within each of the above attributes.

A total of 109 SSCs were selected.

Davis-Besse plant operations staff was consulted in the SSC selection process. The selected list of SSCs is largely located in the auxiliary building, but selections from the intake structure and containment are included.

Many of the selected SSCs are from support systems, but there are also SSCs selected from each frontline system. A total of 101 SSCs came from the original IPEEE or current internal events PRA model. SSCs are selected from each of the safety functions, including 10 related to the containment function.

There were eight SSCs selected that are located in relatively high radiation areas and seven that are in damp areas. Most SSCs selected are in cool and dry areas.However, 14 are chosen from warm areas and three from hot areas The column in Table 4-2 labeled "Reason for Selection into SWEL 1" summarizes the basis for selecting the chosen SSCs. The screens referred to for each SSC are associated with the screen numbers listed across the top of the table. SSCs which are new or subject to a major replacement are assigned a screen of 4d. Also, SSCs subject to an enhancement resulting from the A-46 program or to an 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 FVI>.005).

A representative set, but not all, of such 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 walkdown.4.2 DEVELOPMENT OF SWEL 2 FOR SPENT FUEL POOL RELATED ITEMS For spent fuel pool related items, there was no starting list of SSCs with which to begin. Instead, the functions of the spent fuel pool systems were reviewed and equipment related to pool cooling and make up were included on a new list. The functions included were normal spent fuel pool cooling, spent fuel pool makeup from demineralized water, spent fuel pool makeup using gravity feed from the borated water storage tank (BWST), and spent fuel pool makeup from the BWST using a recirculation pump. Further, the spent fuel pool cooling valves related to makeup from the BWST using a decay heat removal (DHR) loop were included.

The BWST and DHR system were not included in the spent fuel pool list of SSCs as those systems are included in Base List 1;i.e., see Section 4.1. Noticeably absent from the list is equipment for spent fuel pool cleanup.This equipment is not Seismic Category 1.6 Base List 2 is attached as Table 4-4. The equipment coming out of Screen #2 and entering Screen #3 in Figure 1-2 of the EPRI walkdown guidance report (Reference

3) 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. Note that none of the spent fuel pool cooling pumps and heat exchangers are Seismic Category 1 at Davis-Besse.

For this reason, supporting systems for this equipment are not included.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 Davis-Besse, there is no equipment of any kind connected to the spent fuel pool in such a way that failure could result in a rapid drain-down (Reference 7). All equipment is connected at least 9 feet above the top of the fuel assemblies.

Therefore, the rapid drain-down list of SSCs is empty for Davis-Besse.

SWEL 2, as defined in the EPRI walkdown guidance is attached as Table 4-5.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.7

5.0 SEISMIC

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

It also presents the results and findings of the walk-down, and documents the checklists utilized to record the walk-down data.It is concluded that the approach implemented to conduct the seismic walk-downs and area walk-bys satisfies the characteristics and recommendations outlined in EPRI Report 1025286.Therefore, by following these guidelines, the walk-down 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 WALK-DOWN PREPARATION The overall procedure directly implements the EPRI guidelines.

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

EPRI guidelines recommend that a minimum of 50 percent of the equipment considered in the walk-down be examined to document the existing anchorage configurations, and assess this configuration relative to the design basis. It also recommends that the block wall maps be retrieved to document previous evaluations in support of NTTF 2.3. However, Perry NPP does not possess any safety related masonry block walls associated with Seismic Category 1 components, thus the process to verify block wall adequacy per LE 80-11 has been omitted for this walk-down.

Prior to the walk-downs, the Seismic Walk-down Engineers (SWE) examined available plant documentation associated with anchorage design and correlated this to. relevant SWEL components, the respective Seismic Walk-down Checklists (SWC), and Area Walk-By Checklists (AWC). This pre-walk-down 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 for use in verifying, if required, any anchorage configuration during the walk-downs.

Radiation and High Radiation Maps were also incorporated into the SWEs electronic tablets prior to the walk-down proceedings.

This action helped the SWEs conduct briefs with Radiation Protection (RP) prior to entering any radiation, high radiation, or contaminated areas. This 118 practice resulted in lower than expected doses for the walk-down team, and were noted by the site as good ALARA practices.

5.2 NTTF 2.3 WALK-DOWNS The NTTF 2.3 walk-downs at Perry were performed over a duration of five days from August 6 to August 10, 2012. During the walk-downs, the SWEs completed the walk-down checklists as SWEL components were inspected.

Selected anchorage configurations were verified for 50 percent of the floor or wall mounted components on the SWEL. The anchorage was compaired to design documentation, including anchorage design drawings and IPEEE calculations.

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

Some of the findings from the walk-down, which could not readily be dispositioned during the walk-downs, were evaluated further through additional calculation/modification package reviews for proper disposition.

Furthermore, the post walk-down activity also developed this walk-down report.6.0

SUMMARY

OF THE WALK-DOWN RESULTS 6.1 WALK DOWN ITEMS AND WALK-BY AREAS The SWEL 1 included a total of 109 components, and SWEL 2 included a total of 11 components.

From this total of 120 components, 110 components were walked down and 10 components were inaccessible and will require walk-down during the next refueling outage. The SWEL 1 included nine items located in the Drywell and SWEL 2 included one item in the Intermediate Building, which were inaccessible and therefore were not walked down. These items will be walked down during the next refueling outage. Table 6-1 and Table 6-2 identify the walk-down items and walk-by areas, respectively.

The area walk-bys and the walk-down items are cross-correlated on the respective SWCs and AWCs. Table 6-3 lists the components that will be walked down during next refueling outage and Table 6-4 provides the total number of walked down components arranged by their respective equipment classes.119 Table 6-1: Perry NTTF 2.3 Walk-down Items (SWEL 1+2) *Equipment ID 1 C [Floor No Equipment Class Bldg El OG41AO002A

21. Tanks and Heat Exchangers 1B 599 OG41CO003A

5. Horizontal Pumps IB 574 OG41CO003B

5. Horizontal Pumps 1B 574 0G41F0542A 0D. Other-Check Valve or Manual Valve IB 574 0G41F0542B 0D. Other-Check Valve or Manual Valve 1B 574 OG41FO545A 0D. Other-Check Valve or Manual Valve 1B 574 OG41FO545B 0D. Other-Check Valve or Manual Valve IB 574 0G41F0546A 0D. Other-Check Valve or Manual Valve IB 574 OG41FO546B 0D. Other-Check Valve or Manual Valve IB 574 OH51P0039

20. Instrument and Control Panels CC 574 OH51PO164

20. Instrument and Control Panels CC 679 OH51PO193

20. Instrument and Control Panels CC 574 OH5 1P0318 20. Instrument and Control Panels CC 574 0H51P1310

20. Instrument and Control Panels EW 586 OM23CO001A

9. Fans CC 679 OM23CO002A

9. Fans CC 679 OM23FO100A

9. Fans CC 679 OP47BOO01A

11. Chillers CC .574 OP47CO001A

5. Horizontal Pumps CC 574 OP49DO001A

0. Other EW 586 0P49F0502A 0D. Other-Check Valve or Manual Valve EW 586 0R24S0020

1. Motor Control Centers CC 620 0R42S0011

16. Battery Chargers and Inverters CC 620 1B21N0067C

18. Instrument (on) Racks CO 620 1B21N0073C

18. Instrument (on) Racks CO 620 iC11F0083 8A. Motor-Operated Valves FH 620 1Cl1F0110A 8B. Solenoid Valves CO 642 1C11FO160A 8B. Solenoid Valves CO 642 1C22POO01

20. Instrument and Control Panels CC 654 1C41A0003

21. Tanks and Heat Exchangers IB 620 IC41CO001A

5. Horizontal Pumps CO 642 lC41FOO01A 8A. Motor-Operated Valves CO 642 1C41F0029A

7. Pneumatic-Operated Valves CO 642 1E12BOO01B

21. Tanks and Heat Exchangers AX 599 120 Table 6-1: Perry NTTF 2.3 Walk-down Items (SWEL 1+2) *Equipment ID Equipment Class Bldg Floor No El 1E12BOO01D

21. Tanks and Heat Exchangers AX 599 1E12C0002B

6. Vertical Pumps AX 574 1E12F0003B 8A. Motor-Operated Valves AX 574 1E12FOO04B 8A. Motor-Operated Valves AX 574 1E12F0029B 0D. Other-Check Valve or Manual Valve AX 574 1E12F0048B 8A. Motor-Operated Valves AX 599 1E21CO001

6. Vertical Pumps AX 574 1E21FOO01 8A. Motor-Operated Valves AX 574 1E21N0050

18. Instrument (on) Racks AX 574 1E22B5003

21. Tanks and Heat Exchangers DG 620 1E22C0001

6. Vertical Pumps AX 574 1E22F0004 8A. Motor-Operated Valves AX 620 1E22F0010 8A. Motor-Operated Valves AX 574 1E22F0024 0D. Other-Check Valve or Manual Valve AX 574 1E22N0005

18. Instrument (on) Racks AX 574 1E22S0001

17. Engine Generators DG 620 1E22S0005

15. Battery Racks CC 620 1E51B0002

21. Tanks and Heat Exchangers AX 574 1E51FOO45 8A. Motor-Operated Valves AX 574 lE51NO055A

18. Instrument (on) Racks AX 574 1G41F0145 8A. Motor-Operated Valves IB 620 1G43N0060A

18. Instrument (on) Racks AX 574 1G43N0060B

18. Instrument (on) Racks AX 574 1H13P0701

20. Instrument and Control Panels CC 654 1H5 1P0037 20. Instrument and Control Panels AX 599 1H51P0871

20. Instrument and Control Panels CC 620 1H51P0975

20. Instrument and Control Panels AX 599 1H51P1421

20. Instrument and Control Panels DG 620 IM16FOO20A 0D. Other-Check Valve or Manual Valve CO 642 IM39BOO01B

9. Fans AX 574 1M39B0003

9. Fans AX 574 1M39B0006

9. Fans AX 574 1M43CO001A

9. Fans DG 620 1M43C0002B

9. Fans DG 620 1M43F0030A

7. Pneumatic-Operated Valves DG 620 121 Table 6-1: Perry NTTF 2.3 Walk-down Items (SWEL 1+2) *Equipment ID Equipment Class Bldg Floor No El 1M43F0220B 8A. Motor-Operated Valves DG 620 1P42A000lB

21. Tanks and Heat Exchangers 1B 665 1P42BOO01B

21. Tanks and Heat Exchangers CC 574 1P42C0001B

5. Horizontal Pumps CC 574 1P42F0665A

7. Pneumatic-Operated Valves CC 574 1P42F0665B

7. Pneumatic-Operated Valves CC 574 1P43F0055 8A. Motor-Operated Valves 1B 599 1P45CO001B

6. Vertical Pumps EW 586 1P45F0130B 8A. Motor-Operated Valves EW 586 1P45F0660 0D. Other-Check Valve or Manual Valve EW 586 1P57A0003A

21. Tanks and Heat Exchangers FH 620 1P57F0015A 8A. Motor-Operated Valves lB 599 1P57F0015B 8A. Motor-Operated Valves AX 620 1R22S0006-E04

4. Transformers CC 620 1R22S0007

3. Medium Voltage Switchgear CC 620 1R22S0009

2. Medium Voltage Switchgear CC 620 1R23S0012

2. Low Voltage Switchgear CC 620 1R24S0028

1. Motor Control Centers CC 620 1R24S0032

1. Motor Control Centers EW 586 1R25S0014

20. Instrument and Control Panels CC 620 1R25S0016

20. Instrument and Control Panels CC 620 1R25S0018

14. Distribution Panels CC 620 1R25S0020

14. Distribution Panels CC 620 1R25S0022

14. Distribution Panels CC 620 1R25S0025

4. Transformers CC 620 1R25S0033

4. Transformers CC 620 1R42S0002

15. Battery Racks CC 638 1R42S0008

16. Battery Chargers and Inverters CC 638 1R42S0012

14. Distribution Panels CC 638 1R42S0013

14. Distribution Panels CC 638 1R42S0014

14. Distribution Panels CC 638 1R42S0024

2. Low Voltage Switchgear CC 638 1R42S0037

2. Low Voltage Switchgear DG 620 1 R42S0038 1. Motor Control Centers CC 638 1R43SO001B

17. Engine Generators DG 620 122 Table 6-1: Perry NTTF 2.3 Walk-down Items (SWEL 1+2) *Equipment ID Equipment Class Bldg Floor No EaEl 1R45A0003A

21. Tanks and Heat Exchangers DG 620 1R45C0002B

5. Horizontal Pumps DG 620 1R45C0003A

5. Horizontal Pumps DG 620 1R45N0140A

18. Instrument (on) Racks DG 620 1R46BOO01A

21. Tanks and Heat Exchangers DG 620 1R46B0002B

21. Tanks and Heat Exchangers DG 620 2R42S0037

2. Low Voltage Switchgear CC 620* Does not include items in Table 6-3, which will be walked down during the next plant outage.Table 6-2: Perry NTTF 2.3 Walk-By Areas Room Bldg Floor El 1AB-la AX 574 lAB-Ic AX 574 1AB-le AX 574 1AB-le (2) AX 599 1AB-if AX 574 lAB-If (2) AX 620 lAB-Ig AX 574 lAB-2 AX 599 1AB-3a AX 620 1CC-3a CC 620 1CC-3b CC 620 1CC-3c CC 620 1CC-4c CC 638 1CC-4g CC 638 1CC-4h CC 638 1CC-5a CC 654 1DG-la DG 620 1DG-lb DG 620 IDG-Ic DG 620 2CC-3a CC 620 2CC-6a CC 679 123 Table 6-2: Perry NTTF 2.3 Walk-By Areas Room Bldg Floor El CC-1 EW 586 CO-620 CO 620 CO-642 CO 642 EW-1 EW 586 FH-2a IB 599 FH-3a FH 620 IB-1a IB 574 IB-2 IB 599 IB-2 (2) IB 599 IB-3a LB 620 IB-4 IB 665 Table 6-3: Perry NTTF 2.3 Inaccessible Items on SWEL 1+2 *T IColumn/Row Equip. ID Description Bldg El j orDeg IB21AO003A Accumulator For F0041A DW 630 58 DEG 1B21F0022A IB MSIV B21-F0022A DW 630 1 DEG IB21FOO41A ADS Valve DW 630 55 DEG 1B21FO410A SV For F041A DW 630 55 DEG 1B21FO410B SV For F041A DW 630 55 DEG 1B33CO001A RX Recirculation Pump A DW 599 145 DEG 1E51F0063 1 E5 1 -F0063 Motor Valve DW 626 0 DEG 1H22P0011 Standby Liquid Control Instrument Rack DW 630 306 DEG 1H51P1335 SDV Local Instrument Rack DW 620 105 DEG 0G41F0581 Fuel Pool Cask Pit Drain Isolation Valve IB H/07* These items will be walked down during the next refueling outage 124 Table 6-4: Perry NTTF 2.3 Components Categorized by EPRI Classes EPRI TComponents Cat No. Equipment Description WalkedeDow Cat No.Walked Down 0 Other 13 1 Motor Control Centers and Wall-Mounted Contactors 4 2 Low Voltage Switchgear and Breaker Panels 4 3 Medium Voltage, Metal-Clad Switchgear 2 4 Transformers 3 5 Horizontal Pumps 8 6 Vertical Pumps 4 7 Pneumatic-Operated Valves 8 8 Motor-Operated and Solenoid-Operated Valves 19 9 Fans 5 10 Air Handlers 2 11 Chillers 1 12 Air Compressors 0 13 Motor Generators 0 14 Distribution Panels and Automatic Transfer Switches 8 15 Battery Racks 2 16 Battery Chargers and Inverters 2 17 Engine Generators 2 18 Instrument Racks 8 19 Temperature Sensors 0 20 Instrumentation and Control Panels 12 21 Tanks and Heat Exchangers 13 Total 120 125 6.2 WALK DOWN AND AREA WALK-BY FINDINGS The examination of walk-down items and observations in area walk-bys confirms the general seismic robustness of the design and installation of SSCs at Perry The Plant is well maintained and no major issues related to potentially adverse conditions were uncovered.

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

Generally, the nature of the potentially adverse conditions was related to credible interaction effects and conformance with plant control process.Several relatively minor findings are reported here. Observations in this respect are organized on the basis of potentially adverse seismic conditions identified during both Seismic Walk-downs and Area Walk-Bys.6.2.1 Seismic Walk-down Findings The following section presents potentially adverse seismic. conditions and findings identified during the Seismic Walk-downs.

A total of 11 potentially adverse seismic conditions were identified during the Seismic Walk-downs.

Table 6-5 provides a summary of all 11 adverse finding conditions identified.

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

Justifications for findings for which a Licensing Evaluation was not required are provided in the Component's respective SWC provided in Appendix B.Table 6-5: Potentially Adverse Seismic Conditions Identified from SWC's Ds i o Licensing Reference Equipment Equipment Class descriptiof Basis for ID No Condition Evaluation Justification Required IR22S0006-E04

4. Transformers Unrestrained storage y CR 2012-locker. 12222 Shims-Embed Weld SWC for IR24S0028

1. Motor Control Centers Dtil N S00fo Detail 1 R24S0028 126 Table 6-5: Potentially Adverse Seismic Conditions Identified from SWC's Description of Licensing Reference Equipment Equipment Class Adverse Seismic Basis for IDNo Condition Evaluation Justification Required 20. Instrument and Control Shims-Embed Weld SWC for 1 R25 S0016 N Panels Detail I R25S0016 Shims-Embed Weld SWC for 0R24S0020

1. Motor Control Centers Detail N 0R24S0020 20. Instrument and Control Shims-Embed Weld N SWC for Panels Detail 1R25S0014 Shims-Embed Weld SWC for 1R25S0018

14. Distribution Panels Dtil N S00fo Detail 1 R25S0018 Distribution Panels Shims-Embed Weld SWC for Detail 1 R25S0020 Shims-Embed Weld SWC for 1R25S0022

14. Distribution Panels Dtil N S002o Detail 1 R25S0022 Centered trolleys on SWC for IR42S0024

2. Low Voltage Switchgear toptofedwirolgear onN_____for-top of switchgear 1IR42S0024 Centered trolleys on SWC for 1R23S0012

2. Low Voltage Switchgear top tcogeas on N it e for top of switchgear 1 R23S0012 20. Instrument and Control Scaffold storage CR 2012-1 H51 P0037 Pal rack close to HVAC Y12375 control panel Conditions which were noted but subsequently resolved are briefly described below.127

• Shim -Embed Weld Details for MCCs and Panels During the walk-down proceedings, several MCCs, including 1R24S0028, were apparently missing part of the stitch welds at the base. A portion of the base is stitch welded directly to the embedment and the rest is welded to shims that appear to only bear on the embedded angle. This finding was addressed adequately in calculation SQ-0178 for ECP 02-0184 rev 0 and all referenced Design Inputs (i.e., the shims are slot welded to embed angles). No further action was required.Figure 6-1: Missing shim plates on MCC embed 128

• Trolleys on top of Switchgears SWEs identified trolleys with hoists that were on top of switchgear SWEL components.

These trolleys with hoists that are free to move could create a potential adverse interaction by banging into end stops during an earthquake.

These conditions were resolved by ensuring the hoist was properly tagged and restrained within the trolley. In addition, it was confirmed that the issue was addressed during the IPEEE program and no further action was required.Figure 6-2: Centered trolley on top of switchgear 129 e Unrestrained storage locker An unrestrained storage locker approximately 76 inches tall was identified in the Unit 1 Division 2 switchgear room. This was considered as a potential interaction hazard due to the proximity with the Class 1E electrical bus EH12. The electrical bus EH12 powers Division 2 safety systems. The SWT confirmed that there was an existing IPEEE vulnerability resolution associated to this condition, and that its intent was to replace the locker with a shorter locker.The resolution was not implemented.

A condition report (CR 2012-12222) was written to document the issue and operations personnel were notified.

The storage locker was subsequently removed from the room.Figure 6-3: Storage locker near bus EH12 130

• Scaffold storage rack close to HVAC control panel During the seismic walk-downs, a scaffold storage rack was identified close to an HVAC control panel located in Auxiliary Building on floor elevation 599'. The top storage section of the rack was found with inadequate restraint, posing a potential impact from sliding of scaffold planks and tubes during an earthquake, thus hitting the HVAC control panel for the Division 1 and Division 3 pump rooms. A condition report (CR 2012-12375) was issued to address this condition.

Figure 6-4: Scaffold storage rack close to HVAC control panel 131 6.2.2 Area Walk-By Findings The following section presents potentially adverse seismic conditions and findings identified during the Area Walk-Bys.

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

Table 6-6 provides a summary of all 7 adverse findings identified.

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

Justifications for findings for which a Licensing Evaluation was not required are provided in the Area's respective AWCs provided in Appendix C.Table 6-6: Potentially Adverse Seismic Conditions Identified from Area Walk-Bys Floor Desciiption of Adverse Seismic Licensing Basis Reference for Area Bldg El Condition Evaluation Justification Required CC-i cc 586 Flexible branch lines for fire N AWC for CC-I protection piping 1CC-5a CC 654 Control room ceiling grates Y CR 2012-12335 CC-1 CC 586 Unanchored Instrument Air Dryers N AWC for CC-I CC-i CC 586 Seismic violation tags on cast iron N AWC for CC-i CC I cc 586 pipes IB-2 (2) 1B 599 Scaffold structure for chiller unit Y CR 2012-12373 1CC-5a CC 654 Validation of IPEEE commitments Y CR 2012-12331 1AB-2 AX 599 Flooding Source in AX at El 599' N AWC for lAB-2 132

• Flexible branch lines for fire protection piping During the area walk-bys, SWEs identified fire protection piping with rather flexible configurations and Victaulic fittings in area CC-I on the CC 574'. This could expose the ECCW system to seismic induced flooding from the fire protection piping system in the area. It was verified that the Victaulic couplings are Type 77, and that they are grooved couplings with good seismic performance.

The actual field routing and support configuration of the fire suppression system, as well as potential targets, were evaluated and determined to be satisfactory.

Figure 6-5: Branch lines for fire protection piping located in CC-I at elevation 574'* Control room ceiling grates While inspecting the Control Room area, it was noticed that ceiling grates were not tied off vertically such that they could potentially dislodge and fall during an earthquake.

In order to assess this finding, calculation 36.75.2.2 was reviewed.

The upper ceiling assembly had a finite element analysis and a manual calculation of the support rods for the egg crate drop-ins.

These documents identified no seismic concerns.

Even though the seismic design basis requirements were met, an enhancement condition report (CR 2012-12335) was written to identify an opportunity to improve the standard by tie wrapping the egg crate tiles to ceiling's cross members.133

• Unanchored Instrument Air Dryers Instrument Air Dryers (1H51P0097) located in area CC-I on CC 574' were identified with missing anchorage.

This equipment could break free and strike nearby air receiver tanks during a seismic event. SWEs identified this unanchored equipment as a non-safety related component and it was installed in that location per the original design. A subsequent walk-down was performed by a FENOC structural engineer, and it was confirmed that there are no safety related components located in the near vicinity of this rack, as well as no seismic 11/1 concern. It was concluded that this unanchored condition was acceptable and no further action was required.Figure 6-6: Unanchored instrument air dryers located in CC-1 on CC 574'134

• "Seismic Violation" tags on cast iron pipes Some cast iron pipes were identified with tags reading "Seismic Violation" while walking down area CC-1 on CC 574'. Tags were found with unique identification numbers. After post-walk-down discussions, it was confirmed that these tags related back to a seismic walk-down from the past, which had identified issues, and that Design Engineering previously resolved these issues.The 36-series structural calculations document these analyzed conditions.

It was determined that no further action was required.Figure 6-7: Cast iron piping identified with "Seismic Violation" Tags 135 o Scaffold structure for chiller unit While performing an area walk-by walk-down, a scaffold structure was identified as being used as a support system for a chiller unit in area IB-2 on Intermediate Building at elevation 599'. The scaffold was located next to a motor operated valve (1P43F0055).

It was judged that there was no potential interaction with the MOV. However, a condition report (CR 2012-12373) was issued to address what appeared to be an unauthorized modification to the plant for a chiller unit on IB 599' that is supported by scaffold oles.Figure 6-8: View of scaffolding structure for chiller unit SValidation of IPEEE commitments During the walk-down proceedings, it was observed that spatial interaction and relay chatter vulnerabilities identified during the 1990's IPEEE program have not been adequately resolved.The spatial interaction vulnerability originated from control room furniture that could impact cabinets containing seismic sensitive relays. A confirmatory walk down was performed, and there remained one furniture spatial interaction with cabinets P872 and P865. There was a table and a podium stored in close proximity to these cabinets.

The furniture was subsequently moved to Unit 2.136 Calculation 2:13.7 identifies certain relays related to the High Pressure Core Spray System (HPCS) system that would require operator action (OA) in the case of a seismic event.According to the calculation, there are 16 GE HFA Low Ruggedness Relays (LRRs) in the HPCS control circuitry.

11 of these relays are classified as "chatter acceptable" while the remaining five require OA after a period of strong shaking.A condition report (CR 2012-12331) was issued to identify the IPEEE commitment associated with these five relays and the aforementioned control room spatial interaction.

e Flooding Source in AX at elevation 599'A potential seismically induced flooding source was identified from the TB Water Chiller (1P46BOOA) located in area lAB-l on AB 599'. After further review, it was noted that the component is a Non-Category I system, and therefore, it was not considered as a design or licensing basis issue. Furthermore, flooding consequences are non significant, based on the limited volume of water and no Category I components in the vicinity of the component in question.Figure 6-8: Potential flooding source on AX 599'137

6.3 CONFIGURATION

CHECKS The SWEL 1+2 included 79 items, which were not in-line components such as valves. The process of verifying the anchorage configuration focused on 40 SWEL components selected prior to walk-down proceedings.

Appendix D provides a list of the 40 components which comprise the anchorage configuration.

The list is linked to the specific references used for verification purposes; i.e., IPEEE calculations, design drawings, etc.The anchorage configuration for each of the 40 SWEL components listed in Appendix D was verified based on IPEEE calculations and/or Plant Design documentation.

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

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

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

These configuration checks verified consistency of as-installed conditions to that of the design drawings/calculations in all 40 instances.

7.0 LICENSING

BASIS EVALUATION Six condition reports (CR) were generated as a result of these walk-downs

.The following is a list of the condition reports written as a result of the walk-downs:

CR 2012-12222, CR 2012-12331, CR 2012-12335, CR 2012-12373, CR 2012-12375 and CR 2012-12466.

The following summarizes the condition and resolution to the condition reports written.CR 2012-12222 This condition report documents an unrestrained storage locker which could overturn and strike the safety related electrical bus EH 12. This was previously identified during the IPEEE walk-downs as a vulnerability.

The storage locker was removed from the location and relocated to a maintenance shop.CR 2012-12331 This condition report documents that Operator Actions may be required to manually reset relays after a seismic event as they are susceptible to relay chatter during a seismic event. This was 138 previously documented in Calculation 2:13.007 but was not reflected in the Alarm Response Instruction (ARI) procedures.

Corrective Actions 2012-12331-1 and 2012-12331-2 were initiated to update the ARIs associated with the relays.Additionally, the condition report documents a spatial interaction with control room furniture located near control room panels P872 and P865. The furniture was removed and relocated to the Unit 2 Control Room.CR 2012-12335 This condition report documents the control room lower ceiling tiles do not meet industry standards for seismic considerations.

While Perry is not a Seismic Qualification Utility Group (SQUG) plant, plants utilizing SQUG methodology have used plastic tie wraps to tie adjacent ceiling tiles. This ensures that during a design basis seismic event, if the tiles get dislodged from the framing, they will not fall directly onto personnel or on any cabinets below. This modification will help ensure the operability of the components in the control room. This is not a deviation from any design basis requirements.

The control room ceiling consists of an upper and lower ceiling. The upper ceiling has been-seismically analyzed and the lower ceiling is rod hung from the upper ceiling with 3/8" diameter threaded rods. Preliminary calculations concluded that the 3/8" diameter threaded rod with approximately 3' length has a buckling capacity of around 140 pounds, which is more than enough to resist any upward load from the ceiling tiles during a-design basis seismic event. Therefore, the lower ceiling is capable to perform its intended design function, and thus does not pose any operability concern. Notification 600769396 was initiated to install plastic tie wraps to tie adjacent ceiling tiles.CR 2012-12373 This condition report identifies air conditioning/cooler units in a scaffolding enclosure without proper evaluation.

The investigation into this condition report documented that the components had been evaluated by FCR 22300 and Calculation 36:01.3.2.16.84.

As the condition had been evaluated and properly documented no adverse condition existed. Notification 600784453 was initiated to install a storage label near this location to inform of the evaluation.

CR 2012-12375 This condition report identifies a potential seismic 11/1 concern due to a scaffold storage rack located next to a Division 1/Division 3 HVAC panel. The investigation into this condition report documented that the scaffold is installed in accordance with all applicable plant procedures and guidance, and has adequate clearance to the aforementioned equipment.

Additionally, Corrective 139 Action 2012-12375-1 was initiated to eliminate the temporary scaffold storage location from this location.CR 2012-12466 This condition report was written to summarize the potential issues that were identified during the walk-downs.

This condition report does not identify any previously unidentified concerns or issues.8.0 IPEEE VULNEARBILITIES The IPEEE submitted by the Perry Nuclear Power Plant in June (later resubmitted on July 22, 1996) was reviewed for identified seismic vulnerabilities.

The submittal stated on page 8-1, that four enhancements to reduce the threat of spatial interactions were identified and in the process of being implemented.

Section 3.1.4.3.1 "Spatial Interaction Results" documents the spatial interaction problems as housekeeping issues. This task, 2.3 seismic walk-down, treated the spatial interaction problems as commitments and wrote two condition reports documenting the failure to resolve the issues. Calculations were reviewed and the "enhancements," their reference, and final resolution are documented in Appendix F.Section 3.1.4.2.3 documents the relay chatter summary of results. The analysis is treated in section 3.1.4.2.2 and states that there are 5 relays that require manual operator actions to reset post seismic event. The results of this evaluation are contained in Appendix F, and a condition report was generated for failure to modify the operation procedures to alert the operator that post seismic event, certain relays may need to be reset.140 9.0 PEER REVIEW A peer review of the Submittal Report for the Near Term Task Force Recommendation 2.3"Seismic Walk-downs" was performed using the guidance provided in Section 6 of EPRI Document 1025286, "Seismic Walk-down Guidance." The following are the peer reviewers for the Perry Nuclear Power Plant:* Mohammed Alvi (Team Leader)" John Reddington The peer review process included the following activities: " Review the selection of the SSCs included on the SWEL* Review a sample of the checklists prepared for the seismic walk-downs and area walk-bys" Review the Licensing Basis Evaluations" Review the decisions for entering the potentially adverse conditions into the Corrective Action Program (CAP).* Review the submittal report* Summarize the results of the peer review process in the submittal report A. Review the Selection of the SSCs Included on the SWEL: The peer review concluded that the selection of Seismic Walk-down Equipment List (SWEL)was performed in accordance with guidance provided in Section 3 of EPRI Document 1025286"Seismic Walk-down 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 Perry Nuclear Power Plant 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: 141 Screen 1: Seismic Category I Screen 2: Equipment or System Screen 3: Support for the five safety functions Screen 4: Sample Considerations The plant did use the existing documentation that resulted from IPEEE program in identifying the components.

A matrix/spreadsheet was prepared that identifies all the selected components on SWEL 1 and SWEL 2. It was confirmed that these two lists did include a variety of types 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 Perry Nuclear Power Plant included 109 components.

The peer review also confirmed that the plant used the following EPRI screens in the development of SWEL 2: Screen 1: Seismic Category I Screen 2: Equipment or System Screen 3: Sample Considerations Screen 4: Rapid Drain-Down Similar process was used in the development of SWEL 2 as for SWEL 1. SWEL 2 for the Perry Nuclear Power Plant included 11 components.

==

Conclusion:==

No major concerns were identified by the peer review team in the selection process for SWEL 1 or SWEL 2.142 Peer Review Checklist for SWEL Instructions for Completing Checklist This peer review checklist may be used to document the review of the Seismic Walk-down Equipment List (SWEL)in accordance with Section 6: Peer Review. The space below each question in this checklist should be used to describe any findings identified during the peer review process and how the SWEL may have changed to address those findings.

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

Y ON-1 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 [N-]See Attached Comments b. Major new and replacement equipment?

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

Y ZNE]See Attached Comments d. Various environments?

See Attached Comments e. Equipment enhanced based on the findings of the IPEEE (or equivalent) program?"YZNL1"YZNEI See Attached Comments f. Were risk insights considered in the development of SWEL I?Y ZN[-]See Attached Comments 143 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 ONE1 Y ONO--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 ON[-Peer Reviewer Date: 10- Date: l!t Peer Reviewer #2: c~4~144 Peer Review Checklist for SWEL Comments on Question 1: A peer review of the SWEL selected for the Perry Nuclear Power Plant 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. AC Power System B. Controlled Complex Chilled Water C. Control Rod Drive Hydraulics D. Containment Isolation E. Diesel Generator Building Ventilation System F. DC Power System (Essential and Normal)G. Onsite Electric Power (Division 1 and 2 Diesel Generators)

H. Onsite Electric Power (Division 3 Diesel Generator)

I. Drywell Vacuum Relief J. Emergency Closed Cooling Water System K. Emergency Core Cooling System Pump Room Cooling L. Emergency Service Water System M. Emergency Service Water Screen Wash System N. High Pressure Core Spray 0. Instrument Air P. Low Pressure Core Spray System Q. MCC, Switchgear, and Misc. Electrical Areas HVAC R. Nuclear Boiler 145 S. Nuclear Closed Cooling Water System T. Reactor Core Isolation Cooling System U. Safety Related Instrument Air V. Standby Liquid Control System W. Suppression Pool Make up X. Reactor Pressure Vessel Y. Remote Shutdown Comments on Question 2b: The selected components represent many new and replacement equipment based on the following modifications:

A. ECP 96-5089: Installed Torque Limit Switch B. ECP 09-0821: Motor Replacement C. ECP 94-0027: New 3-Way Valve for bypass around ECCW Hx D. ECP 02-0184-011:

Replacement of MCC EF1D08, Compartment D (1R24S0028-00D)

E. ECP 99-5010: Equipment Upgrades F. ECP 03-0358: Upgrade Valve Actuator G. ECP 09-0828-001:

Motor Replacement H. ECP 02-0184-019:

MCC Compartment Replaced I. ECP 00-5003: Reconfigure Stop/Auto/Start Switches Comments on Question 2c: The peer review concluded that the selected components represent various type of equipment installed in the plant. The various equipment types are indicated as follows: A. 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. Tanks and Heat Exchangers 146 K. Medium Voltage Switchgear L. Transformers M. Horizontal Pumps N. Pneumatic Operated Valves 0. Motor Operated Valves P. Solenoid Valves Q. Fans R. Chillers S. Check/Manual Valves 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. Warm C. Humid D. Hot E. Cool F. Dry G. Dry/Humid H. Damp I. Wet Comments on Question 2e: Based on the review, the selected components represent equipment enhanced based on findings of the IPEEE.Comments on Question 2f: The risk insights were considered in the development of SWEL 1. Specifically, Risk Achievement Worth (RAW) and Fussel-Vessley (FV) were considered.

147 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.

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

The sample review should be between 10 percent and 25 percent.The following comments were identified during the early stages of peer review and were successfully resolved: A. In some cases, statements regarding minor anomalies (not resulting in a condition report)identified during the walk-downs 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, a minor anomaly stated but no justification provided.D. Editorial and typographical errors The above comments were discussed with the Seismic Walk-down Engineers (SWEs) and were successfully resolved in the final signed version of the checklists.

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

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

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

Questions asked were successfully answered during the interview as well as during the walk-downs.

Four SWEs participated in the walk-downs.

See their resumes for experience and background training.Conclusion:

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

All comments were successfully resolved.

Adequate documentation has been provided in the checklists for the components that were walked down.C. Review of the Licensing Basis Evaluations The walk-downs identified several minor anomalies, however six of them resulted in generating condition reports as follows: 1. CR 2012-12222 This condition report documents an unrestrained storage locker which could overturn and strike the safety related electrical bus EH 12. This previously identified during the IPEEE walk-downs as a vulnerability.

The storage locker was removed the location and moved to a maintenance shop.2. CR 2012-12331 This condition report documents that Operator Actions may be required to manually reset relays after a seismic event as they are susceptible to relay chatter during a seismic event. This was 149 previously documented in Calculation 2:13.007 but was not reflected in the Alarm Response Instruction (ARI) procedures.

Corrective Actions 2012-12331-1 and 2012-12331-2 were initiated to update the ARIs associated with the relays.Additionally, the condition report documents a spatial interaction with control room furniture located near control room panels P872 and P865. The furniture was removed and relocated to the Unit 2 Control Room.3. CR 2012-12335 This condition report documents the control room lower ceiling tiles do not meet industry standards for seismic considerations.

While Perry is not a Seismic Qualification Utility Group (SQUG) plant, plants utilizing SQUG methodology have used plastic tie wraps to tie adjacent ceiling tiles so that during a design basis seismic event if the tiles get dislodged from the framing they will not fall directly on personnel or on any cabinets below thus safeguarding the operability of the components.

This is not a deviation from any design basis requirements.

The control room ceiling consists of an upper and lower ceiling. The upper ceiling has been seismically analyzed and the lower ceiling is rod hung from the upper ceiling with 3/8" diameter threaded rods.Notification 600769396 was initiated to install plastic tie wraps to tie adjacent ceiling tiles.4. CR 2012-12373 This condition report identifies air conditioning/cooler units in a scaffolding enclosure without proper evaluation.

The investigation into this condition report documented that the components had been evaluated by FCR 22300 and Calculation 36:01.3.2.16.84.

As the condition had been evaluated and properly documented no adverse condition existed. Notification 600784453 was initiated to install a storage label near this location to inform of the evaluation.

5. CR 2012-12375 This condition report identifies a potential seismic 11/1 concern due to a scaffold storage rack located next to a Division 1/Division 2 HVAC panel. The investigation into this condition report documented that the scaffold is installed in accordance with all applicable plant procedures and guidance and has adequate clearance to the aforementioned equipment.

Additionally, Corrective Action 2012-12375-1 was initiated to eliminate the temporary scaffold storage location from this location.150

6. CR 2012-12466 This condition report was written to summarize the potential issues that were identified during the walk-downs.

This condition report does not identify any previously unidentified concerns or issues.The plant 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 walk-down results. Specifically, Section 6.2.1 discusses seismic walk-down findings associated with SWEL 1, and Section 6.2.2 discusses seismic walk-down findings associated with area walk-bys.

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

Table 6-6 identified seven potentially adverse seismic conditions.

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

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

151

==

Conclusion:==

The peer reviewers agree with the decisions taken for entering or not entering the identified potentially seismic walk-down 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 walk-downs.

The peer reviewers participated in sample walk-downs, observed the conduct of walk-down team members, and discussed issues while remaining independent.

The Seismic Walk-down 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.

10.0 REFERENCES

1. NRC letter 50.54(f), March 17, 2012.2. NRC letter endorsing EPRI document, May 31, 2012.3. 'EPRI 1025286, "Seismic Walk-down Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic," Final, June 2012.4. "Individual Plant Examination of External Events for Severe Accident Vulnerabilities," Perry Nuclear Power Plant, Submitted in response to U.S. Nuclear Regulatory Commission Generic Letter 88-20 Supplement 4," The Cleveland Electric Illuminating Company, June 1996.5. A Methodology for Assessment of Nuclear Power Plant Seismic Margin, EPRI NP-6041-SL, Revision 1, August 1991.152

6. RG 1.29, "Seismic Design Classification." 7. RG 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants." 8. RG 1.61, "Damping Values for Seismic Design of Nuclear Power Plants." 9. RG 1.100, "Seismic Qualification of Electrical and Mechanical Equipment for Nuclear Power Plants." 10. IEEE 344-1975, "IEEE Guide for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Plants." 11. ASME Boiler and Pressure Vessel Code Section III 1974 including Winter Addenda 1975.153 APPENDIX A RESUMES AND QUALIFICATIONS A-I 279 Dorchester Rd, Phone 234-678-8262 Akron Ohio 44313 E-mail jreddman@aol.com JOHN E. REDDINGTON Work experience January 2007 to present: Principal Consultant, Probabilistic Risk Analysis:

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

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

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

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

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

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

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

Direct oversight of outage preparations and implementation.

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

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

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

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

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

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

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

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

198o -1984 Structural Engineer Responsible for steel and concrete design and analysis for LaSalle and Fermi Nuclear Power plants. Performed vibrational load and stability analysis for numerous piping systems. Member of the on-site team of engineers responsible for timely in-place modifications to the plant structure at LaSalle.1979-198o Wagner Martin Mechanical Contractors Richmond, IN Engineer/Project Manager Responsible for 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 FARZIN R. BEIGI, P.E.PROFESSIONAL HISTORY ABSG Consulting Inc., Oakland, California Senior Consultant, 2004-Present Technical Manager, 2001-2004 EQE International, Principal Engineer, 1990-2001 TENERA L.P., Berkeley, California, Project Manager, 1982-1990 PROFESSIONAL EXPERIENCE Mr. Beigi has more than 29 years of professional structural and civil engineering experience.

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

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

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

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

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

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

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

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

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

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

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

Substantial cost savings were achieved by maximum application of the experience data procedures for seismic qualification.

• Assessment of earthquake risk for Genentech, Inc., in South San Francisco, CA. The risk assessments included damage to building structures and their contents, damage to regional utilities required for Genentech operation, and estimates of the period of business interruption following a major earthquake.

Provided recommendations for building or equipment upgrades or emergency procedures, with comparisons of the cost benefit of the risk reduction versus the cost of implementing the upgrade. Project included identification of equipment and piping systems that were vulnerable under seismic loading and design of retrofit for those components, as well as, providing construction management for installation phase of the project." Fault-tree model and analysis of critical utility systems serving Space Systems / Loral, a satellite production facility, in Palo Alto, CA." Seismic evaluation and design of retrofits for equipment, tools and process piping, as well as, clean room ceilings and raised floors at UMC FABs in Taiwan.* For LDS Church headquartered in Utah, performed seismic vulnerability assessment and ranked over 1,200 buildings of miscellaneous construction types for the purpose of retrofit prioritization." Seismic evaluation and design of retrofits for clean room ceilings at Intel facilities in Hillsborough, Oregon.* Assessment of programmable logic controls as part of year 2000 (Y2K) turn over evaluation at an automatic canning facility in Stanislaus, ca." Seismic evaluation and design of retrofits for equipment and steel storage tanks at the Colgate-Palmolive plant in Cali, Colombia.* Design of seismic anchorage for equipment and fiberglass tanks at the AMP facilities in Shizouka, Japan." Evaluation and design of seismic retrofits for heavy equipment, and piping systems at Raychem facilities in Redwood City and Menlo Park, CA." Assessment of the seismic adequacy of equipment, structures and storage tanks at the Borden Chemical Plant in Fremont, CA.A-7 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E." Design of seismic bracing for fire protection and chilled water piping systems at the Goldman Sachs facilities in Tokyo, Japan.* Design of seismic retrofits for low rise concrete and steel buildings and design of equipment strengthening schemes at AVON Products Co. in Japan.* 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-8 ABS Consulting AN ABS GROUP COMPANY FARZIN R. BEIGI, P.E.U. Klapp, F.R. Beigi, W. Tong, A. Strohm, and W. Schwarz, ,,Seismic PSA of Neckarwestheim 1 Nuclear Power Plant," 19th International Conference on Structural Mechanics in Reactor Technology (SMIRT 19), Toranto, Canada, August 12-17, 2007.A. P. Asfura, F.R. Beigi and B. N. Sumodobila.

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

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

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

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

SUMMARY

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

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

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

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

Puerto Rico -2009 MEMBERSHIPS American Society of Civil Engineers (ASCE)American Concrete Institute (ACI)Network for Earthquake and Engineering Simulation (NEES)U.S. Dept. of Labor (OSHA)Society of Hispanic Professional Engineers (SHPE)A-11 ABS Consulting AN ABS GROUP COMPANY EDDIE M. GUERRA, E.I.T.HONORS AND AWARDS 2010 Recipient of the Thornton Tomasetti Foundation Scholarship Golden Key International Honor Society Tau Beta Pi Engineering Honor Society University of Puerto Rico at Mayaguez Dean's List PUBLICATIONS Guerra, Eddie M., "Impact Analysis of a Self-Centered Steel Concentrically Braced Frame," NEES Consortium, May -July 2007.A-12 ABS Consulting AN ABS GROUP COMPANY BRIAN A. LUCARELLI, E.I.T.PROFESSIONAL HISTORY ABSG Consulting Inc., Contractor, Presently Paul C. Rizzo Associates Inc., Pittsburg, PA, Engineering Associate II, 2010- Present Engineers without Borders, Aquaculture Development, Makili, Mali, Africa September 2007 -December 2009, Southwestern Pennsylvania Commission, Pittsburgh, Pennsylvania, Transportation Intern, May 2008 -August 2008 PROFESSIONAL

SUMMARY

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

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

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

He has authored numerous calculations and reports related to the work conducted for this project, including responding to requests for additional information from the NRC. His analyses for this project included finite element analyses of the stresses within the Bridging Mat under static and dynamic loading and the determination of long-term settlement at the site.A-13 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. Lucareli has also performed a settlement analysis for the Makeup Water Intake Structure.

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

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

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

Previous Experience:

September 2007 -December 2009 Aquaculture Development

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

May 2008 -August 2008 Southwestern Pennsylvania Commission

-Pittsburgh, Pennsylvania:

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

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

Engineer-in-Training

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

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

SUMMARY

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

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

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

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

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

• Master of Science (Structural Engineering), University of Illinois, Chicago (1977)" Bachelor of Engineering (Civil), Bhopal University, India (1976)PROFESSIONAL LICENSES/CERTIFICATIONS: " Registered Professional Engineer, State of New York* Boiling Water Reactor (BWR) Plant Certification Course for Nine Mile Point Unit-1 Nuclear Station PROFESSIONAL SOCIETY MEMBERSHIP:

• Member, American Society of Civil Engineers (ASCE)

REFERENCES:

Provided upon request CITIZENSHIP:

Citizen of the United States of America A-19 FENOC RICHARD C. SIEMBOR Education:

Case Western Resenre University Cleveland,.

Ohio Bachelor of Science -Mechanical Engieering January 1991 Nuclear License: Work History: 2008 -Present 2006 -2008 2003 -2006 2001- 2003 Case Western Reserve University Cleveland, Ohio Bachelor of Science -Pluid and Thermal Engineering Sciences January 1991 Edinboro University Edinboro, Pennsylvania Bachelor of Arts -Physics January 1991 FirstEnergy-Perry Nuclear Power Plant Senior Reactor Operator (SRO) License Number 31758 Operator Docket Number 55-32836 February 2003 Perr, Nuclear Power Plant January 1991 -Present Probabilistic Risk Assessment Perry Nuclear Power Plant PRA Lead. Regulatory Guide 1.200 model compliance efforts- Post Fukushima analyses and mitigative strategy development.

Concentration in event modeling, accident analysis/strawegies, and human reliability analysis aspects.Responsible System Engineer.

Systems included:

Residual Heat Removal, Control Rod Drive, Suppression Pool, Suppression Pool Cleanup.Senior Reactor Operator License holder/ Shift Engineer.Training for Senior Reactor Operator License.A-20 1996- 2001 1991- 1996 Other: Probabilistic Risk/Safety Assessment Instrumental in PSA product preparation for License Amendment 99 Diesel Generator Allowed Outage Time Submittal including establishment of the Configuration Risk Management Program.Applied Engineering Analysis Generic Letter 89- 10 Design Basis calculations, Improved Technical Specifications, Emergency Operating Procedure calculations/verification, USAR Chapter 15 accident analyses and radiological assessments, Probabilistic Risk Analysis.Fuel Handling Supervisor, Refueling Supervisor (SRO). Project Manager (Feedwater Check Valves, Residual Heat Removal System), Safe Shutdown Advisor.A-21 Alexandra Zelaski azelaski(d firstenergycarp.com FirstEnergy Nuclear Operating Company 1392 Oakwood Trail Perry Nuclear Power PLant Paainesville, 011 441077 10 Center Rd (440) 666-4317 Perry, OH 44081 EDUCATION The Ohio Stide University, M.S. CLams of 2011 Cohlmbus, O1 Nuclear EngMeering Program Institute of Nuclear Poawr Opetions:

National A mcadry ofNuclear Tini Mng Fellow Passed Doctoral Qua lifing Exam M aster's Thesis related to SiC alpha prticle detector to take actinide imnentories in pyropucessing strearns Swarihmore Caokge, Class of 2009 Swarthmcre, PA Bachelor of Arts Major Physics M inar: Education Midpark High School, Class of 2005 Middleburg Heights, O1 GPA -4.04.0, Cass rank -3ff in class of276 Honors: Phi Beta Kappa, Bausch and Lomb Science Scholar EXPERIENCE Nuclear Engineer, Analytical Methods Fleet Design Engineering-Analytical Methlds, FirstEnergy Nuclear OpenitingCompany August 2011 -Current* Developmig, updating and maintaining current PRA model for Perry Nuclear Power Plant" Supported data collectiom and documentation of internai flooding model" Currently supporting model development forseismic and external flooding PRA* Qualified to do PRA on-line risk managment Nuclear Engineering Graduate Fellow The Ohio State University Department of Mechanical Engineering, The Ohio State University August 2009 -Current* Working todevelop SiC detectorto determine actinide concentrations in pyroprocessing molten salt" Developing TCAD model to simulate SiC detector damage" Wrote MATLAB con to analyze SRIM program files* Developed and machined parts for alpha particle detection experimental set-up" Certified to work in c Lam 100 cleanroom Physics Research Intern Oberlin College Department of Physic and Astronomy, Obeilin College Summer 2007, 2008* Synthesizedand characterized Fe 3 0 4 nanoparticles using VSM, XRD, SANS* Performeda small angle neutronscattering experiment at NIST Center for Neutrn Research Science Assadak Sarthmare College Department of Physics and Astronomy, Swarthmore College Spring 2007-Fall 2008* Led problem session kir Physics 4: General Physics 1i (E&M}" Led problem session for Physics 3: General Phys-ics I (Mechanics)

Physics Instructional Aide Swarthmore College Department of Physics and Astronomy, Swarthmore College. Fall 2006 -Fall 2007" Graded 1br PhysicsiAstrooomy 5: Spacetime, Quanta, and Cosmology" Graded fIr Astronomy 1: lntroductory Astronomy* Graded for Physics 8: Introductiom to Electricity and Magnetism A-22 American Sign Language Drill Instructor Swarthmore College Department of Lm-MguLtics.

Swarthmonr College Fall 2006" Creatodcwmiulum and lesson plans and lead drill section* Led drill section of a First Year Seminar. Language and Deafness SKILLS* Able to operate Vibrating Sample Magetometer (VSM). X-ray diffracttoneter(XRD)

• Knowledge of MCNP, MATLAB, Mathematica, Igor, SRIM., Kaleidagraph, LabVie,, Genie2000* Physics machine shop and electronics experience, familiarity with milling machine, lathe. tabl sa, soldering* Language proficiency in Spansh and American Sign Language" Able to use both Mac intush and Windows AC.rIITI ES The Ohio State Universky Student Chapter of the American N uclear Society Columbus, Oil President Spring 2010-C"-rrent

• Given tours of OSU Nuclear Reactor Lab to Sen. Voinovich, Commissioner Magwood ofNRC* hostoed movie i.soving November 15, 2010 to OSU student body* Plamning outreach wctivitits for surrounding Columbus high schools, teacher workshops and for Boy and Girl Scouts SWAP: Swarthmore Women in Astronomy and Physics Swarthmore.

PA Co-Leader 2007-Sprmg 2009* Planned Liquid Nitrogen Ice Cream Social for incoming freshmen to interest them in Physics. Astronomy and Engineering I N07, 2008)EYli: Expanding Your Horizons Conference Swarthmore, PA Conference Planner and Recla-iter 2007-2008* Helped plan conference to get middle school aged girls more interested in ,iencc" Gathered cumtact information to send to mintrity schools in suburhs of Philadelphia Swarthmore College Women's Varsity Fasopiklh

-%f(all Swarthmore, PA Varsity Pitcher Fall 2005 -Spring 2009* National Fastpitch Couches Assuciation Academic All-American (2009)" All Centennial Conference Academic Team (2008, 2009)" Member of firs sofiball team in college history to make Centennial Conference playoffs 12006, 2007, 20081 and East Coast Athletic Conference Playoffs (2008)" ECAC South Region Pitcher of the Week* First pitcher in Swarthmore College softball history to win 10 or more Wimes, N" most career wins mi Swarthmore College softball history Softball Pitching Instructar Cleveland, 011 Sumnmer 2003 -Current" Teach younger girls the art of fastpitch softball pitching" Teach pitching camps to diverse (skill, age) groups of girls* Cawhed RBI (Reviving Baseball in the Inner City) and other travel teams PRESENTATIONS Materials and Energy: The Building Blocks for Ohio's Economic Future Spring 2010" Poster Title: The Ffect of Neutron Irradiation and Low Temperature Annealing on the le.tr"ial Propeities of Highly Doped 411 Silicn Carbine* Presented posterrelated to current and past researxh conducted at The Ohio State Univer-sity OWLU Poster Presentation Surrmner 2007* Presented poster about summer research symposi umn hosted by Ohio 5 Research Consortium A-23 AS Consulting DANIEL A. RENY PRO FESSION AL HISTORY ABSG Consulting Inc, [rvine, CA, Senior Consultant, _2008-Present ARES Corporation, Los Angeles, CA, Senior Consultant, 2002-2008 SCIENTECH, Inc., Kent, WA, Senior Consultant, 1997-2002 Jason Aswciates Corporation, Idaho Falls, ID, Senior Consultant, 1993-1997 TENERA, L. P., Idaho Falls, ID, Senior Consultant, 1991-1993 EG&G Idaho, Inc., Idaho Falls, ID, Senior Engineering Specialist, 1987-1991 Rockwdl Internaliona), Anaheim, CA, Senior Engineer, 1985-1987 PLG Inc., CA, Engineering Analyst, 1981-1985 PROPESSION AL

SUMMARY

Senior Nuclear Power Plant PRA, Risk and Reliability Consultant and Project Manager with over 31 years professonal experience providing services to Nuclear Power Operators in USA and former Soviet Union countries, USNRC, LS DOE, and NASA.Mr. Reny has over 31 years experience as technical lead and project manager conducting probabilistic risk analyses (PRA), IPEs, IPEEEs seismic and fire PRAs, human reliability analyses (HRA), safety and risk informed management applications for domestic and international commercial nuclear power utilities and DOE nuclear facilities.

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

A-24 ABS Consufting DAN ILL A. RLNY ARES Corporation, Senior Consultant, Los Angeles, CA (2002 -May 2008)Mr. Reny served as Project Manager and Senior Consultant performing PRA, external events including fire PRA, and risk-informed analyses for Diablo Canyon Nuclear Power Plant. He performed PRA analyses in support of upgrades and improvements of plant PRA model for fire PRA uses and other risk-informed applications.

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

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

He also conducted Human Reliability Assessment (HRA) for Ukraine nuclear power plants.Performed PRA updates, fault tree/event tree modeling, data analysis, Human Reliability Analysis, IPEEE, seismic and fire PRAs, quantification, and consequence analysis of nuclear power plant PRAs at Indian Point 2, Wolf Cieek, Columbia, Diablo Canyon, San Onofre and other nuclear facilities.

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

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

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

Mr. Reny served as Idaho Falls Office Manager responsible for 35 employees and $6 to $10 million worth of contracts with DOE.A-25 ABS Consulting DANIEL A. RENY Project Safety engineer responsible for development of SARs for Pit 9 buried waste rerediation facility at INEL Radioactive Waste Management Complex.Program Manager for DOE technical support contract (-'$5M anrnually) responsible for all tasks supporting DOEIlD office in preparation and review of environmental and safety profects.Mr. Reny was the project safety analyst for Spent Fuel Project at Hanford responsible for safety design requirements and SAR preparation for Cold Vacuum Drying Facility.TENERA, L. P., Senior Consultant, Idaho Falls, ID (1991-1993)

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

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

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

Mr. Reny applied PRA models in risk management/decision analyses, publication of NUREG report and testimony in support of NRC Generic Safety Issues resolution.

Principal Investigator with project responsibility for Air Force RAM analyses on over 22 Air Force bases. Authored and implemented RAM analysis guidelines to be used for Air Force facilities emergency power and utilities systems.Mr. Reny also served as safety engineer for PREPP experimental TRU waste incinerator at I NEL responsible for preparation of facility design requirements and SAR.Rockwell International, Senior Engineer, Anaheim, CA (1985-1 987)Senior Reliability Engineer for Naval Shipboard Weapons Systems RAM Engineering, Reliability Testing, and MIL Standard Programs for the development of naval shipboard A-26 ABS Consufting DANIEL A. RENY systemvs, guidance control systems for missile programs, and advance avionics for aircraft.Program Reliability Engineer responsible for development and implementation of reliability programs in conjunction with development and procurement of naval electronics equipment.

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

Analytical responsibilities included performing reliability modeling, predictions, reliability testing and growth analysis, failure reporting, failure anal-sis, and corrective action, and EEE parts program.PLO Inc-, Engineering Analyst, rvhne, CA (1981-19&5)

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

Probabilistic Safety Assessment, and Reliability/

Availability/

Maintainability Assessment.

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

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

Mr. Reny performed RAM analyses and design decision input for procurement and design of a coal fired power plant, geothermal power plant, and various I&C systems designs.EDUCATION B.S., Applied Physics, University of California, Irvine 1982 A-27

-I Certificate of Completion John Reddington Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012to 4 ýDate Robert K Kassawara EPRI Manager.Structural Reliability

& Integrity A-28 I cm=: Certificate ofA chievement This is to Certify that 5 John E. Reddington has Comp1~ted the Th'afSQUGgA46 Walkdown Screening andSeismic Evaluation Training Course HfeldNovember 20-25, 1987 kihard 0. Starckn, Awdat¢m Inc. Rober P. Kamawsr EPRI Traiing Cowihnawtvr Progm ManamrýjU A-29

~'.1 I crtifiratr of hi Arti'rur Mnt IIU4is is to Mplrtifg tI!ad)1 1((ffarzin R TA ri qi ii.4a5 Tomplrtrb thr §,(QU(S Watkbown#ýrrrrninq anb 'uismir TIVatuation T~rainting Qoursr TrIfb Maq 3-7 1993 4~?~Pi? ~ekI Neil P. Smith, Commonwealth Edison SQUG Chairman l m~3~~ --David A. Freed. MPR Associates SQUG Training Coordinator Robert P. Kassawara, EPRI SQUG Program Manager 4-I A-30 ELECTRIC POWER aar=rei I 015fARCH INSTITUTE Certificate of Completion Farzin Beigi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 13, 2012 Robert K, Kessawara EPRI Manager, Struttural Reliability

& Integrity Date A-31 A-32 Certificate of Completion Tddie Guerra Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns Nishe ER V o VP Advanced Et~Prjects July 6, 2012 A-33 Certificate of Completion Brian Lucarelli" Training on Near Term Task Force Recommendation

2.3 Plant

Seismic Walkdowns Nish R. VP Advanced Eng Projects July 6, 2012 A-34 Presents this Certificate ofA chiievement IT& That IN M.. o h d ed F. Alvi,, P.E.h.haSc "nm fl-aooftfl Cffi )C 2a i'tb'civ (croaning andSeismic Evaluation Training Course Weld November 4 tfi- 99f, 1992 j9,4,,b~VflNeil P. Smith. Commonwahth E&Kmo SQIJG (Thartman-~ David A. F-reed. MPR Av~ociates SQUJ6 Training Coordinator900 Robert P. KM~aWstaw, EPRI SQUG Program Manager A-35 ILECTPIC FOWEV RESEARC-14 IN511TUIE Certificate of Completion Mohammed Alvi Training on Near Term Task Force Recommendation 2.3-Plant Seismic Walkdowns June 27, 2012 Date Robert K. EPRI Manager.Structural Reliability

& Integrity A-36 p A-37 A-38 S Quc.....+ ... .MM~ iN OW Il&4TITftflttC of Ad~ir I*JX1 1 is is It (U2rriif~j U 1 at oi4mmcb Ati 4 Ms (comtploleb.

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~'drction QJour~-i: 15r-t 4 N Paul W. Hayes. -MPR 6sociates.

Richard G. Starck II. MPR Associatea II A-39