ML20094P035

NSP PINGP Units 1 & 2 USNRC USI A-46 Resolution Seismic Evaluation Rept

ML20094P035
Person / Time
Site:	Prairie Island
Issue date:	11/20/1995
From:	STEVENSON & ASSOCIATES
To:
Shared Package
ML20094P024	List: ML20094P021 ML20094P035 ML20094P045
References
REF-GTECI-A-46, REF-GTECI-SC NUDOCS 9511280278
Download: ML20094P035 (130)

Text

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O g l NORTHERN STATES POWER COMPANY l Prairie Island Nuclear Generating Plant l

! Units 1 & 2 l i

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USNRC USI A-46 RESOLETION l

! SEISMIC EVALE ATION REPORT

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i j November 1995 i

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i l Stevenson & Associates 10 State Street - Suite 4 Woburn, MA 01801

PINGP A 46 Seismic EvJustion Report November 20,1995 O  !

1. INTRODUCTION AND SEISMIC VERIFICATION METHODOLOGY 1-1

1.1 INTRODUCTION

1-1 )

1.2 SEISMIC VERIFICATION METHODOLOGY 1-1 1.3 SEISMIC QUAUFICATION FOR STATION BLACKOUT EQUIPMENT 1-2 1.4 REPORT ORGANIZATION 1-2 f

2. PRAIRIE ISLAND NUCLEAR GENERATING PLANT SAFE SHUTDOWN PATH 2-1  ;

3. PRAIRIE ISLAND NUCLEAR GENERATING PLANT SEISMIC DESIGN BASIS 3-1 i

3.1 DESCRIPTION

OF lNPUT MOTIONS 3-1

3.2 DESCRIPTION

OF DYNAMIC MODEUNG AND BASES FOR THE SELECTION OF KEY MODEUNG ]

PARAMETERS 3-1  ;

3.3 DESCRIPTION

OF SOIL-STRUCTURE INTGRACTION STUDIES 3-2 i 3.4 IN-STRUCTURE RESPONSE SPECTRA 3-2 l

4. RESULTS OF SCREENING VERIFICATION AND WALKDOWN - EQUIPMENT CLASSES 0 THROUGH 20 4-1 ,

4.1 SEISMIC EVALUATION GUIDEUNES 4-1 4.1.1 SEISMIC CAPACITY Vs. DEMAND 4-2 4.1.2 CAVEAT COMPUANCE 4-2 l

4.1.3 ANCHORAGE ADEQUACY 4-3 l 4.1.4 SEISMICINTERACTION CHECKS 4-6 -

4.2 OUTLIER RESOLUTION 4-6 )

4.3 SEISMIC CAPA81UTY ENGINEERS AND PEER REVIEV/ER 4-6 I 4.4 OTHER TYPES OF SEGMIC EVALUATIONS AND lNTERFACES 4-7 4.5 DOCUMENTATION 4-7 4.6 EVALUATION RESULTS - EQUIPMENT CLASSES 0 THROUGH 20 4-8

5. GIP DEVIATIONS AND COMMENTARY ON MEETING THE INTENT OF CAVEATS 5-1

6. RESULTS OF THE TANKS AND HEAT EXCHANGER REVIEW 6-1 6.1 EVALUATION METHODOLOGY 6-1 6.2

SUMMARY

OF EVALUATION RESULTS 6-2 O 7. RESULTS OF THE CABLE TRAY AND CONDUlT RACEWAY REVIEW 7-1 i

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PINGP A 46 Seismic Evaluation Report November 20,1995 4

7.1 INTRODUCTION

AND PURPOSE 7-1 l 7.2 SCOPE OF ELECTRICAL RACEWAYS ASSESSED 7-1 7.2.1 GENERAL AREAS COVERED 7-1 7.2.2 GENERAL DESCRIPTION OF PRAIRIE ISLAND RACEWAYS 7-2 7.3 SPECIFIC RACEWAY SYSTEMS EVALUATED 7-2 7.3.1 GENERAL APPROACH 7-2 7.3.2 CABLE ROUTING 7-3 7.3.3 CABLE DATA. AND WEIGHT DETERMINATION 7-5 7.4 RACEWAY SEISMIC EVALUATION CRITERIA AND WALKDOWN RESULTS 7-6 7.4.1 GIP INCLUSION RULES RESULTS 7-6 7.4.2 GIP OTHER SEISMIC PERFORMANCE CONCERNS & SEISMIC INTERACTION REVIEW 7-6 7.5 LIMITED ANALYTICAL REVIEW (LAR) RESULTS 7-7 7.5.1

SUMMARY

OF RESULTS 7-8 7.5.2 LOG'C DIAGRAMS FOR CABLE TRAY AND CONDUlT SUPPORT EVALUATIONS 7-8 7.6 RESUL*t %ND CONCLUSIONS 7-11 7.7

SUMMARY

OF CABLE AND RACEWAY OUTUERS 7-11

8. DESCRIPTION OF THE EQUIPMENT OUTLIERS 8-1 8.1 GENERIC OUTLIER ISSUES 8-1 8.2 EOUIPMENT SPECIFIC OUTUERS IDENTIFIED DURING THE FINAL A46 WALKDOWNS: 8-1

9. RESOLUTION OF OUTLIERS 9-1 V

10. REFERENCES 10-1 Appendix A: Peer Review Assessment Appendix B: Seismic Design Basis Spectra Appendix C: Walkdown Personnel Resumes Appendix D: Screening Verification Data Sheets (SVDS)

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, PINGP A 46 seismic Evaluation Report November 20,1995 Listof Acronyms e CB Control Building

CEA Concrete Expansion Anchor 2

CTFR Cable Tray Fill Report EPRI Electric Power Research Institute GERS Generic Equipment Ruggedness Spectra GIP _ Generic Implementation Procedure for the Seismic Vcrification of Nuclear Plant Equipment GL Generic Letter GRS Ground Response Spectrum lAEA Intemational Atomic Energy Agency IPEEE Individual Plant Examination for Extemal Events ISRS . In-structure Response Spectra PINGP Prairie Island Nuclear Generating Plant LAR Limited Analytical Review MCC Motor Control Center OSVS Outlier Seismic Verification Sheet PAB Primary Auxiliary Building

' PASS Plant Area Summary Sheet PSD Power Spectral Density RWST Refueling Water Storage Tank S&A Stevenson & Associates SBO/ESU Station Blackout / Electrical Safeguards Upgrade SCE Seismic Capability Engineer SEWS Screening Evaluation Work Sheet SQUG Seismic Qualification Utility Group SRT Seismic Review Team ,

SSE Safe Shutdown Earthquake i SSEL Safe Shutdown Equipment List SSER Supplemental Safety Evaluation Report SVDS Screening Verification Data Sheet, l USl Unresolved Safety issue i NRC Nuclear Regulatory Commission NSP Northem States Power Company ZPA Zero Period Acceleration i

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PINGP A-46 seismic Evaluston Report November 20.1995 4

1. Introduction and Seismic Verification Methodology 1.1 Introduction i This report provides the final documentation of the seismic adequacy evaluations performed at I Northem States Power Company's (NSP's) Prairie Island Nuclear Generating Plant (PINGP), Units 1 )

and 2, for the resolution of Unresolved Safety issue (USI) A-46, " Seismic Qualification of Equipment in i Operating Plants". USl A-46 was issued by the United States Nuclear Regulatory Commission (NRC) in December,1980 to address the concem with the seismic adequacy of mechanical and electrical equipment in older nuclear power plants. This report describes the methodology used for and the results of the seismic reviews of active mechanical and electrical equipment, selected tanks and heat exchangers, and cable and conduit raceways. -

i 1.2 Seismic Verification Methodology Utilities affected by USl A-46 formed the Seismic Qualification Utility Group (SQUG) in 1982 to develop ,

a consistent industry approach for resolving USl A-46. SOUG utilities, including NSP, with the l technical and financial assistance of the Electric Power Research Institute (EPRI) conducted research and studies regarding this issue in order to formulate a thorough and reasoned program to resolve the identified concem. In February,1987, the NRC issued Generic Letter 87-02, " Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety lasue (USI) A-46," requesting USl A-46 licensees to commit to a detailed approach for resolving USl A-46

[1.]. j Subsequently, further research conducted by SQUG (and its contractors) and r.wiewed by the NRC staff resulted in a detailed procedure developed by SQUG called the " Generic implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment" (2.]. Specifically, the NRC staff .

reviewed Revision 2 of the GIP and accepted (with provisos) the approach in Supplement No.1 to  !

Generic Letter (GL) 87-02 that Transmits Supplemental Evaluation Report No. 2 (SSER #2) on SQUG Generic Implementation Procedure, Revision 2 as Corrected on February 14,1992 (GIP-2) [3.). This GIP version and the clarifications, guidance and additional requirements provided by the NRC in SSER

1. 2 are the basis for the seismic evaluation of mechanical and electrical equipment at Prairie Island for resolution of USl A-46. The GlP Revision 2 referred to as GlP-2 by the NRC is referred to as the GlP in this report.

Separate, but related issues pertaining to methods of analysis for above-ground flexible tanks identified in USl A-40," Seismic Design Criteria" [4.), and seismic adequacy of proximity items above and around

i. Important-to-safety equipment identified in USI A-17 [5.] are explicitly addressed and resolved by implementation of the GIP.

l The GIP approach relies on developing a safe shutdown equipment list (SSEL) which identifies equipment needed to achieve and maintain safe hot shutdown conditions as defined by a nuclear power plant's Technical Specifications. This equipment is then seismically reviewed in accordance .

i 'with the GlP methodology. By means of plant walkdowns to specifically observe and evaluate each equipment item on the SSEL, an assessment can be made concoming its seismic adequacy. By evaluating seismic demand criteria, selected caveats to ensure similarity to the GIP equipment classes, an anchorage evaluation, and a seismic interaction proximity assessment, the trained walkdown O engineer can be satisfied that the equipment will survive the plant's design basis seismic event. The basis for this approach is rooted in detailed observations of representative, if not identical, equipment in industrial facilities that have survived earthquakes of similar or greater magnitude in Califomia and throughout seismically active regions around the world. Each equipment assessment is documented 2

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L PINGP A-46 seismic Evaluation Report November 20,1996 d

} on a Screening' Evaluation Work Sheet (SEWS). Any deficiencies are documented on an Outiler-j Seismic Verification Sheet (s) (OSVS).

I 1.3 Seismic Qualltication for Station Blackout Equipenent in addition to the GlP methodology, the project took credit for the 6pplication of current licensing-requirements to a large number of SSEL components. The Prairie Island Nuclear Generating Plant Station Blackout / Electrical Safeguards Upgrade (SBO/ESU) Program Design Report, Rev. 2 [Ref.

23.) details the design and installation requirements for the SBO/ESU Project. Section 4.6 of the Prairie Island Nuclear Generating Plant Design Basis Document provides the design basis and technical description of the seismic requirements. The DS/D6 Building Seismic Response Spectra (Document No. S-376-S6-005) was used as a design criterion for the seismic design and qualification of systems and components classified as Seismic Category 1 within the D5/D6 building. The plant modification packages listed below document the conformance to these requirements.

1) 89Y973 - Class 1 Bldg for DSL Gen. 1

2) 89Y974 - Safeguard DSL Gen. D5/D6 1

3) 89Y976 - DSL Gen. Pit Interface During the plant walkdowns, the Seismic Review Teams concentra'.ed on spatial interaction  !

considerations. The SRT's judged the equipment and the equipment anchorage acceptable based on i the fact that it is plant design basis documentation performed under Prairie Island's Quality Assurance  !

Program. No explicit review of this documentation was performed by the Seismic Capability Engineers q beyond verifying that the documentation in fact exists and is appropriate.

U 1.4 ReportOrganization The following section of this report discusses the development of the safe shutdown path and the l

resulting Safe Shutdown Equipment List (SSEL) for Prairie Island. The SSEL is provided in Reference 1 20.. The seismic design basis of Prairie Island and the assessment of it by the NRC are discussed in Section 3. The design basis spectra are contained in Appendix B. The Prairie Island equipment  ;

walkdown and results are provided in Section 4. These assessments resulted in summary level  ;

screening verification data sheets, SVDS, (Appendix D)

SSER #2 requires explicit documentation of any deviations from the caveats or their intent in the GIP.

Section 5 provides a detailed listing of exceptions to the rules taken for any equipment item '

assessment. Section 6 discusses the results of the Tanks & Heat Exchangers assessment. They are also summarized under Class 21 - which is the Tanks & Heat Exchangers class - la the SVDS given in Appendix D. Cable Tray & Conduit Raceway assessments are provided in Section 7. Section 8 provides a listing of identified equipment outliers (Classes 0 - 20), with the reasons for which they are outliers.

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i PINGP A46 Seismic Evaluation Report I November 20,1995 O 2. Prairie Island Nuclear Generating Plant Safe Shutdown Path Reference 20. presents the development of the safe shutdown path and equipment on the safe shutdown l equipmentlist. l l

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PINGP A-46 solemic Evaluation Report Novemtw 20.1995

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This section describes the seismic motion used for the Prairie Island USl A-46 study. Based on the i acceptance of the ground (site) design basis resporise spectrum and the associated amplified in-l structure response spectra (ISRS),?rairie island is using plant design basis spectra. The ISRS and

the development of the spectra were presented to the NRC by NSP in response to Generic Letter

! 87-02 as " conservative design" spectra [6.]. The NRC reviewed the design basis ISRS and declared l that they could be utilized as " conservative design" ISRS [7.]. The following sections describe the

basis and development of the design basis spectra.

i h 3.1 Description ofinput Motions

- The input motions used to create the seismic design of PINGP [6.] are based on the Hu iner Ground

i. Response Spectrum (GRS). For the seismic analysis of the Seismic Class I structures, the Housner

! GRS was used. For the seismic analysis of equipment and piping, the original design developed ISRS j based on a time history that was developed to envelope the design basis GRS [22.]. The PINGP j Operating Basis Earthquake (OBE) is defined in the horizontal direction by the Housner GRS scaled to t 0.06g peak ground acceleration (PGA) and the ISRS developed from original design time history developed to envelope the GRS. The OBE in the vertical direction is defined by 2/3 of the Housner GRS with a resulting PGA value of 0.04g. The PINGP Safe Shutdown Earthquake (SSE) is defined by multiplying the OBE acceleration by a factor of 2 resulting in a horizontal direction GRS PGA value of 0.12g.

More recently, PINGP installed the D5/D6 building. The seismic analysis of the D6/D6 building, and O the seismic analysis of equipment and piping in this building, is based on a Regulatory Guide 1.60 ground spectra for 0.06g OBE and 0.12g SSE.

3.2 Description of Dynamic Modeling and Bases for the Selection of Key Modeling l Parameters The PINGP Class I structures consist of: two Reactor Buildings, an Auxiliary Building and a Turbine Building all as one interconnected structure; a Screen House; and, the D5/D6 building. i l

The original design dynamic analysis for the Prairie Island Nuclear Generating Plant modeled the single interconnected structure in a composite model with individual mass points assembled for the following plant structures:

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1. Unit 1 Shield Building, Unit 2 Shield Building

2. Unit 1 Containment Vessel, Unit 2 Containment Vessel

3. Unit 1 Reactor Support Structure, Unit 2 Reactor Support Structure

4. Spent Fuel Tank

5. Auxiliary Building Steel Roof

6. Auxiliary Building Concrete Slabs and Walls

7. Turbine Building

8. Turbine Support The original design also developed a dynamic model of the Screen House structure. Subsequent to the original design, PINGP installed the DS/D6 building designed to a Regulatory Guide 1.60 GRS.

The mathematical models of the above Seismic Class 1 structures were constructed in terms of lumped masses and stiffness coefficients. The damping used is as shown in Table 3-1 shown below:

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, PINGP A 46 Seismic Evaluellon Report j November 20,1995

p Table 3-1 Prairie Island Design Basis Damping Values

V

Structural / Component Type Damping

[

~

Reactor Building Containment Vessel 1.0%

Reactor Building Shield Structure 2.0%

Reactor Building intomal Concrete Construction 5.0%

Steel Frame Structures 2.0%

Reinforced Concrete Construction 2.0%

Piping Systems 0.5%

Mechanical Equipment 2.0%

3.3 Description of Soll-Structure Interaction Studies The soil-structural interaction under seismic motions is represented by the translational and rotational .

springs in the model. The stiffness of these springs was determined by using equations developed for I the case of a rigid plate on a semi-infinite elastic half-space.  !

i 3.4 In-Structure Response Spectra O The horizontal response spectra curves for equipment inside the buildings were generated by the time history technique of seismic analysis [8.]. The input (base motion) time history used developed to envelope the design basis OBE ground response spectra. The amplified time histories were then l developed for each building elevation in the lumped mass building model. From these amplified time 4 histories of acceleration, acceleration floor response spectra were developed for various damping values. The floor spectra developed represent the_ Operating Basis Earthquake (OBE) level input j acceleration for equipment and equipment anchorage design. The spectra. accelerations were increased by a factor of 2.0 to represent the Safe Shutdown Earthquake (SSE). Originally, the ,

response spectra curves were smoothed and broadened to eilminate erratic response and to account j for parameter uncertainties such as building and soil properties. In general, the structural analysis showed the response to be similar in the North-South and East-West directions; therefore, horizontal in-structure response spectra were generated for one horizontal direction only and are considered applicable to both horizontal directions. Separate North-South and East-West spectra were developed for the Turbine Support and the Auxiliary Building Steel Roof, however, no SSEL equipment required the use of spectra at either of these locations. i The vertical in-structure response acceleration is defined by 2/3 of the Housner GRS PGA (OBE -  ;

0.04g, SSE - 0.08g). The PINGP structures, systems and components are considered rigid in the vertical direction. Therefore. 2/3 of the Housner GRS PGA acceleration is used regardless of location.

The project created ISRS curves for 3,4, and 5% or critical damping based on the original design results published for 0.5 and 1%.

Seismic capacity, as described in the GIP by the " Bounding Spectrum" and the " Generic Equipment

. Ruggedness Spectra" (GERS), are presented in a frequency vs. acceleration format. Therefore, in order to facilitate the comparison of seismic demand to capacity, NSP consolidated all of the response spectra into a single format of frequency vs. acceleration [8.]. Appendix B contains the 5% damped floor response spectra for the Seismic Class I structures.

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PINGP A46 Seismic Evaluation Report

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k The NRC staff reviewed the original and subsequent modeling performed by NSP and its contractors l and determined that the building modeling was adequate. The NRC staff concluded that the resulting l In-structure response spectra could be ut<lized as conservative desian ISRS spectra as defined in the  !

GIP [2.] as opposed to realistic, median centered ISRS (7.).  !

The SSE site ground response spectrum and the generated ISRS are provided in Appendix B.

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PINGP A 46 seismic Evaluation Report November 20,1995

4. Results of Screening Verification and Walkdown - Equipment Classes 0 Through 20 )

i The purpose of this section is to describe the Screening Verification and Walkdown performed to verify I the seismic adequacy of active mechanical and electrical equipment identified in the Prairie Island Safe  !

Shutdown Equipment List (SSEL) report (20.]. The guidelines contained in this section were used to screen the equipment fcr seismic adequacy. If the equipment did not pass this screen, it was declared an outlier (see Section 8). Outlier Resolution, described in Section 9, is accomplished by

1) more refined or sophisticated methods for verifying seismic adequacy, or

2) equipment / anchorage modification.  !

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4.1 Seismic Evaluation Guidelines The procedure for performing the Screening Verification and Walkdown is based on the following four ,

seismic screening guidelines, as defined in the GlP [2.]:

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1. Seismic Capacity Compared to Seismic Demand - The seismic capacity of the equipment, based on earthquake experience data, generic seismic testing data, or equipment-specific seismic qualification data, should be greater than the seismic demand imposed on the equipment by the safe shutdown earthquake (SSE).

2. Caveats - In order to use the seismic capacity defined by the earthquake experience Bounding Spectmm or the generic seismic testing GERS, the equipment should be similar to the equipment in the earthquake experience equipment class or the generic seismic testing equipment class and also meet the intent of the specific caveats for that class of equipment. If equipment-specific seismic qualification data is used, then any specific restrictions or caveats for that qualification data apply instead.

3. Anchoraae - The equipment anchorage capacity, installation, and stiffness should be adequate to withstand the sebmic demand from the SSE at the equipment location.

4. Seismic Interaction - The effect of possible seismic spatial interactions with nearby equipment, systems, and structures should not cause the equipment to fail to perform its intended safe shutdown function.

The evaluation of equipment against each of these four screening guidelines at Prairie Island is based upon walkdown evaluations, calculations, and other supporting data.

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PINGP A-46 seismic E.alue6on Report November 20,1995 4.1.1 Seismic Capacity Vs. Demand Prairie Island determined the seismic capacity of safe shutdown equipment using:

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• Earthquake experience data with capacity defined by the Bounding Spectrum, or Reference  !

Spectrum depending of the demand spectrum used;  !

e ' Generic seismic test data which have been compiled into Generic Equipment Ruggedness Spectra

. (GERS); or,

• ' Equipment-specific seismic qualification data.

The seismic demand imposed on an item of equipment depends on whether or not the ground response spectrum or amplified in-structure response spectra were used, and how it is compared to the capacity data.

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Generally, the ground spectrum was compared to the bounding spectrum for equipment within 40' of grade with an estimated fundamental frequency greater than 8 Hz. To a lesser extent, conservative ISRS were compared to 1.5 times the bounding spectrum (i.e., reference spectrum). The GERS were not used in the capacity vs. demand comparisons. Finally, newer, upgraded equipment that had been seismically qualified in accordance with the IEEE 344 Standard,1975 Edition or later, was accepted O based on this testing documentation and anchorage design calculations, and was supplemented only by a seismic interaction review by the SRT.

For purposes of determining the 40' Above Grade elevation, ettective grade for the site and/or each building must be determined. " Effective grade" at a nuclear plant is defined as the average elevetion of the ground surrounding the building along its perimeter. As Prairie Island is a soil site, effective grade was established at El. 695.0 ft. 1 4.1.2 Caveat Compliance The second screening guideline which must be satisfied to verify the seismic adequacy of an item of mechanical or electrical equipment is to confirm that (1) the equipment characteristics are generally

similar to the earthquake experience equipment class or the generic seismic testing equipment class and (2) the equipment meets the intent of the specific caveats for the equipment class. This review is only necessary when the Bounding Spectrum or the GERS is used to represent the seismic capacity of an item of equipment. If equipment-specific seismic qualification data is used instead, then only the specific restrictions applicable to that equipment specific qualification data need be applied.

Another aspect of verifying the seismic adequacy of equipment included within the scope of this procedure is explained by the " rule of the box." For the equipment included in either the earthquake or testing equipment class, all of the components mounted on or in this equipment are considered to be part of that equipment and do not have to be evaluated separately. However, the walkdown engineers did look for suspicious details or uncommon situations which could make the equipment item vulnerable.

An item of equipment should have the same general characteristics as the equipment in the earthquake experience equipment class or the generic seismic testing equipment class. The intent of 1 4-2

PINGP A-46 seismic Evduation Report '

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f this rule is to preclude items of equipment with unusual designs and characteristics which have not i demonstrated seismic adequacy in earthquakes or tests.

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Caveats" are defined as the set of inclusion and exclusion rules which represent specific characteristics and features particularly important for seismic adequacy of a particular class of l' equipment. Appendix B of the GIP contains a summary of the caveats for the earthquake experience equipment class and for the generic seismic testing equipment class.

' The " intent" of the caveats should be met when evaluating an item of equipment as they are not fixed,

inflexible rules. Engineering judgment is used to determine whether the specific seismic concem addressed by the caveat is met. Each item of equipment should be evaluated to determine whether it meets the specific wording of the applicable caveats and their intent. However, if an item of equipment

meets the intent of the caveats, but the specific wording of the caveat rule is not met, then that item is l considered to have met the caveat. At Prairie Island, a small number of SSEL ltems were judged to l meet the intent, but not the exact wording of a caveat, and these cases are reported in Section 5 of this report. ,

4.1.3 Anchorage Adequacy l Prairie Island verified anchorage adequacy with an approach incorporating three elements: l l

• Comparison of the anchorage capacity with the seismic demand. ]

• Evaluation of the anchorage to verify that it is free of gross installation defects.

• Evaluation of the equipment anchorage load path to verify that there is adequate stiffness and O- strength.

The screening approach for verifying the seismic adequacy of equipment anchorage is based upon a combination of inspections, analyses, and engineering judgment. Inspections consist of measurements ar'd visual evaluations of the equipment and its anchorage, supplemented by use of plant documentation and drawings. Analyses compare the anchorage capacity to the seismic loadings (demand) imposed upon the anchorage. These analyses were done using the guidelines in Section 4

, and Appendix C of the GIP. Enaineerina ludament is also an important element in the evaluation of equipment anchorage. As a general rule, all significantly sized equipment was rigorously analyzed using the ANCHOR software package developed by Stevenson & Associates [19.], or by manual calculation. Small equipment, weighing usually 50 lbs. or less was accepted by judgment and a " tug test". The tug test simply involves pulling on the device (say, a wall-mounted transmitter) with a force to exceed 3 times the expected seismic demand for the equipment location. This demonstrates, as a i minimum, a factor of safety of 3 for the equipment anchorage, consistent with the anchorage evaluation criteria in the GIP.

The four main steps used to evaluate seismic adequacy of equipment anchorages at Prairie Island followed the guidance of the GIP and are shown below:

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1. Anchorage installation Inspection

2. Anchorage Capacity Determination i

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3. Seismic Demand Determination

4. Comparison of Capacity to Demand l Q

V The first main step in evaluating the seismic adequacy of anchorages is to check the anchorage installation and its connection to the base of the equipment. This inspection consists of visual checks and measurements along with a review of plant documentation and drawings where necessary, and an anchor bolt tightness and embedment check for anchorage utilizing concrete expansion anchors.

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I PINOP A46 seismic Evaluation Report j November 20,1995 i

All accessible anchorages were' visually inspected. - A check of the following equipment anchorage

attributes was made

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1. Equipment characteristics

2. Type of Anchorage

3. Size and Location of Anchorage

4. Installation Adequacy

5. Embedment Length

6. Gap at Threaded Anchors i

7. Spacing Between Anchorages

8. Edge Distance

9. Concrete Strength and Condition

10. Concrete Crack Locations and Sizes '

11. Essential Relays in Cabinets

12. Equipment Base Stiffness /Psying Action

13. Equipment Base St/ength/ Structural Load Path

14. Embedment Steel and Pads For expansion anchors, a tightness check was pedormed to detect gross installation defects (such as -

oversized concrete holes, total lack of preload, loose nuts, damaged subsurface concrete, and missing ,'

plug for shell types) which would leave the anchor loose in the hole. The tightness check for expansion anchors was accomplished by applying a torque to the anchor by hand until the anchor was " wrench tight," 1.e., tightened without excessive exertion. If the anchor bolt or nut rotates less than about 1/4 tum, then the anchor is considered tight. The tightness check was performed on all accessible O- expansion anchors for floor mounted equipment where the anchorage adequacy is performed by analysis rather than " tug test" as described above. Wall mounted equipment was excluded as allowed i by the GIP because the anchors experience some tensile loading due to gravity. A random (" spot")

embedmont check on selected shell type anchors was performed, inspecting them to ensure that the shell anchor and equipment base are not in contact so as to invalidate the results of the tightness check. Based on the embedment checks and plant documentation, the predominant expansion anchor type at Prairie Island for original equipment is the Phillips Red Head shell anchor which requires no knock-down factor for anchor type. Newer equipment installed since the plant began operation typically utilizes Hitti wedge anchors which also require no knock-down factor.

The second main step in evaluating the seismic adequacy of anchorages is to determine the allowable capacity of anchors used to secure an item of equipment. The allowable capacity is obtained by multiplying the nominal ailowable capacities by the applicable capacity reduction factors. The nominal capacities and reduction factors are obtained from Appendix C of the GIP, based on the results of the anchorage installation inspection checks. The nominal allowable tensile and shear capacities are established in EPRI Report NP-5228-SL [15.]. The nominal allowable capacities incorporate a design safety factor of 3 between the ultimate and allowable (working) capacities.

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l PINOP A 46 seismic Evaluation Report l November 20,1996 i

/ The pullout capacity allowable is based on the product of the nominal pullout capacity and the -

! applicable capacity reduction factors based on identifying the appropriate anchor type and make:

! P = Poa. RT, Rl, RS, RE, RF, RC, RR, Where: Pm = Mowable Eullout capacity of installed anchor (kip) _

P = Nominalallowable Eullout capacity (kip)-

. RT,= Beduction factor for the T.ype of expansion anchors

Rt.,= Beduction factor for short embedmont Lengths RS,= Beduction factor for closely Spaced anchors RE,= Beduction factor for near Edge anchors RF,= Beduction factor for low strength (f ) concrete RC,= Beduction factor for Gracked concrete RR,= Beduction factor for expansion anchors securing equipment with essential Belays The shear capacity allowable is based on the product of the nominal shear capacity and the applicable capacity reduction factors

V, = Voo. RT. RL. RS. RE. RF. RR.

L _ Where: V, = Mowable shear capacity of installed anchor (kip)

Vnom= Ngminal allowable shear capacity (kip) ,

RT.= Beduction factor for the Iype of expansion anchors i RL.= Beduction factor for short embedmont Lengths O

RS.= Beduction factor for closely Spaced anchors i RE.= Beduction factor for near Edge anchors  !

RF.= Beduction factor for low strength (f.) concrete i RR.= Beduction factor for expansion anchors securing equipment with essential Belays Note that the pullout and shear capacities for anchors given above are based on having adequate

- stiffness in the base of the equipment and on not applying significant prying action to the anchor, if Check 12, Base Stiffness and Prying Action, from Part 11, Chapter 4 of the GIP shows that stiffness is not adequate or that significant prying action is applied to the anchors, then the Seismic Capability Engineers lowered the allowable capacity loads accordingly, normally by completely discounting the affected bolt.

The third-step in evaluating the anchorages is to determine the seismic demand imposed on the equipment. The demand load is established based on the type of demand spectrum used. If the amplified ISRS are used, no additional factors of conservatism are used to establish the demand load since the ISRS are deemed " conservative design" by review of the NRC. The demand load is the product of the appropriate spectral acceleration value times the weight of the equipment item. Table '

C.1-1 of the GIP is used, in general, to establish the fundamental frequency and equipment damping for the given classes of equipment. If the SRT deems an item as rigid, the zero period acceleration (ZPA) is used. If the item is deemed flexible, the peak of the response spectrum is used. If the

' fundamental frequency is given in the SEWS, then the largest spectral acceleration in the range from that estimated frequency to the ZPA is used. If the ground spectrum is used for demand, then 1.875 times the appropriate spectral acceleration is used where 1.875 is the product of 1.5, the median amplification factor, and 1.25, the additional anchorage factor of conservatism for non-conservative j demand spectra.

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PINGP A-46 solemic Evaluetkm Report November 20,1995 l

O The fourth and final step to complete the evaluation determines the seismic demand on the equipment anchorage and compares the seismic demand to the anchorage capacity. The demand on the anchorage is calculated by applying the demand load at the equipment center of gravity. If the l demand is less than the capacity, the anchorage is acceptable; otherwise, the equipment item is i declared an outlier.

l l ' 4.l.4 Seismic Interaction Checks

, The fourth and final screening guideline used to verify the seismic adequacy of an item of mechanical i

or electrical equipment was to confirm that there were no adverse seismic spatial interactions with nearby equipment, systems, and structures which could cause the equipment to fall to perform its intended safe shutdown function. The interactions of concem are (1) proximity effects, (2) structural failure and falling, and (3) flexibility of attached lines and cables. Guidelines for judging interaction-effects when verifying the seismic adequacy of equipment are presented in Appendix D of the GIP.

- During the plant walkdowns at Prairie Island, the SRT's identified only one general interaction concem concoming open "S" type hooks used to support overhead lighting. This particular issue and its resolution is discussed in detail in Section 8.

Overhead piping systems and ductwork were closely examined in all plant areas containing A-46 equipment. The SRT's identified no vulnerabilities and noted that the systems were well supported.  ;

, 4.2 OutilerResolution An outlier is defined as an item of equipment which does not meet the screening guidelines noted above. An outlier may be shown to be adequate for seismic loadings by performing additional  ;

evaluations such as the seismic qualification techniques currently being used in newer nuclear power  ;

plants. At the discretion of the Seismic Capacity Engineers, additional evaluations and attemate l methods were documented on the OSVS forms.

4.3 Seismic Capability Engineers and Peer Reviewer The guidelines described in this section were applied by Seismic Capability Engineers as defined in Section 2 of the GIP. These engineers exercised engineering judgment based upon an understanding of the guidelines given in this document, the basis for these guidelines given in the reference l

documents and presented in the SQUG training course, and their own seismic engineering experience.

l The station walkdowns were conducted in several sessions beginning in November 1993 and ending in  !

November 1995. The seismic capability engineers for the Prairie Island walkdown were Messrs. W.

Djordjevic and Frank B. Stille of S&A; Messrs. Mark McKeown and Gerald Gore of Northem States Power Company, and Mr. Gregory Ridder of Wisconsin Public Service Company. All have been l SQUG trained and certified. Their resumes and/or SQUG Walkdown Course Completion Certificates are provided in Appendix C.

O An independent evaluation and peer review of the walkdown process was performed by Dr. Robert P.

Kennedy. As required by the GIP, the review included an assessment of the walkdown and analyses by audit and sampling to identify any gross errors. Dr. Kennedy personally conducted one day of 4-6

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k PINGP A-46 seismic Evaluation Report j November 20.1996

,P walkdowns on July 25,1994 to ascertain completeness and correctness of the A-46 walkdown. Dr.

Kennedy also reviewed Screening Evaluation Work Sheets documenting the seismic verification

conclusions by SRTs on selected components. Dr. Kennedy concluded that the walkdowns were being conducted competently and the findings made were appropriate. Appendix A provides documentation of Dr. Kennedy's peer review. ,

3

4.4 Other Types of Seismic Evaluations andInterfaces In addition to the seismic evaluations covered in this section for active mechanical and electrical equipment, seismic evaluations for two other types of equipment are covered in other sections as follows:

• Section 6 - Tanks and Heat Exchangers Review

• Section 7 - Cable and Conduit Raceways Review A separate Relay Evaluation Report [Ref. 20.) documents the results of the relay functionality review

. required in Section 6 of the GlP.

While these other seismic evaluations can generally be performed independently from those for active

. mechanical and electrical equipment, there are a few areas where an interface with the Relay Functionality Review is appropriate:

. Any cabinets containing essential relays, as determined by the relay review, should be evaluated for seismic adequacy using the guidelines contained in this section.

. A capacity reduction factor should be applied to expansion anchor bolts which secure cabinets containing essential relays. The capacity reduction factor is discussed in Section 4.4 and Appendix C of the GIP.

. Seismic interaction, including even mild bumping, is not allowed on cabinets containing essential relays. This limitation is discussed in Section 4.5 of the GIP.

. in-cabinet amplification factors for cabinets containing essential relays may be estimated, using the guidelines in Section 6 of the GIP, by the Seismic Capability Engineers for use in the Relay Functionality Review.

4.5 Documentation Prairie Island documented the results of the Screening Verification and Walkdown on Screening Verification Data Sheets (SVDS) in Appendix D and Screening Evaluation Work Sheets (SEWS).

As discussed in Section 4.4, the discussion of the review and the description of outliers for Heat Exchangers & Tanks and Cable Tray & Conduit Raceways are given in Sections 6 and 7, respe :tively.

The SEWS also contain the SEWS for Class 21 equipment, Heat Exchangers and Tanks, and Plant Area Summary Sheets (PASS) for the Cable Tray & Conduit Raceway Reviews.

Outliers for all equipment in Classes 0 - 20 are discussed in Section 8. The Relay Funt,'ionality Assessment is given in a separately bound report entitled, "USNRC USl A-46 Resolutin, SSE:. and O Relay Evaluation Report"(20.].

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PINGP A-46 seismic Evaluation Report Novembw 20,1995 l 4.6 Evaluation Results - Equipment Classes 0 Through 20 The seismic review SSEL list contains 621 equipment items excluding: 1) items classified as Rule-of-the-Box; 2) tanks & heat exchangers; and,3) electrical raceways. Of this population,82 items were declared outliers. For a discussion of equipment outliers (Class 0 - 20) see Section 8.

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PINGP A-46 seismic Evaluation Report November 20.1995 0% l V 5. GIP Deviations and Commentary on Meeting The Intent of Caveats 1

No significant or programmatic deviations from the GIP were made while performing the walkdowns i and seismic adequacy evaluations at Prairie Island for resolution of USI A-46. Very few interpretations were made with respect to the wording of the GIP caveats versus the caveat's intent. This section lists those interpretations or measures taken to meet the intent of the caveat in Table 5.1 below. All other equipment not listed in this table met the caveat rules as stated in the GlP.

Table 5-1 Commentary Regarding GIP Deviations Equipment ID and Commentary Description Misc. small relief valves: These small relief valves are mounted on a 3/4" inlet which is less 2AF-29-1, 2AF-29-2, 2CL-25-1, than the 1" diameter required by Bounding Spectrum Caveats 4 2CL-57-3, 2CL-57-4, 2CL-57-5, and 5 for Class 7, Fluid Operated Valves. The valves are all small 2CL-57-6, 2VC-28-1, 2VC-28-2, (12" or less), relatively lightweight, and judged seismically rugged AF-29-1, AF-29-2, CL-25-1, CL- by the Seismic Review Team.

57-3, CL-57-4, CL-57-5, CL 6, SA-54-3, SA-54-6, SA-56-1, SA-56-3, VC-28-1, VC-28-2, p ZH-16-1, ZH-16-2 Q D1 and D2 Diesel Generator These cast iron body valves have both inlet and outlet flanges Cooling Water Supply Control ruggedly mounted directly to the diesel generator skid. The Valves: Seismic Review Team judged that the piping loads applied to the CV-31505 & CV-31506 valve would be low meeting the intent of Bounding Spectrum Caveat 2 for Class 7.

Small Solenoid Valves: These small brass solenoid valves are mounted on copper SV-33242 SV-33245, SV- tubing.(< 1"). Both the valve and tubing are hardmounted together, 33498, & SV-33987 judged by the Seismic Review Team to meet the intent of Bounding Spectrum Caveats 4 & 5 for Class 7.

Small Solenoid Valves: These small colenoid valvos are mounted on small piping and SV-37022, SV-37025, & SV- tubing.(< 1'). The Seismic Review Team accepted the seismic 37460 through SV-37467 ruggedness of these valves by a tug test, meeting the intent of Bounding Spectrum Caveats 4 & 5 for Class 8.

Motor Operated Valves: The weight of these valves exceeds the limits of GlP Figure B.8-1 MV-32060 through MV-32063 for 4" pipe. The product of operator weight times valve offset meets the 30% extrapolation provisions in GIP Rev. 2A for the, thus meeting the intent of Bounding Spectrum Caveat 5 for Class 8.

Cooling Water to Safeguards The Seismic Review Team judged that the piping loads applied to Traveling Screens Solenoid these valves with cast iron bodies would be low meeting the intent Valves: of Bounding Spectrum Caveat 2 for Class 8.

SV-33133 & SV-33134 Racks and Panels: These racks and panels are all mounted by 5/8" cast in place J-1B1,1 B2,1NR3,1NR4,1 PLP, bolts that do not meet the 16D embedment requirement. The 1R1,1R2,1RCS1,1RCS2, anchorage analysis assumed only 1/2" bolts thereby satisfying the J 281,2B2,2NR3,2NR4,2PLP, 2R1,2R2,2RCS1,2RCS2 embedment requirement, and meeting the intent of Bounding Spectrum Caveat 9 for Class 20.

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PINGP A 46 seismic Evrustion Report November 20,1995

]v Table 5-1 Commentary Regarding GIP D9viations (Continued)

Equipment ID and Commentary Description Relay Racks: Various anchors are propped up on shims filling gaps up to 1" 1 AMR1,2AMR1,1 ASG1, below cabinet bases, due to pitched floor leading to drains. The 1 ASG2,1BSG1,1 BSG2, shims meet the intent of Anchorage Caveat 6 for zero gaps under 2BSG1,2BSG2,55800,55300 equipment containing essential relays. The analysis of the anchorage used the methodology of EPRI TR-103960 [Ref.

24.}considering flexural stress resulting from the bolt projections thereby meeting the intent of caveat.

Motor Control Centers: These Motor Control Centers use an anchorage detail that uses MCC 1L1, MCC 1L2, MCC 1X1, pairs of cast in place studs anchoring the frcnt and rear of the MCC MCC 1X2, MCC 2L1, MCC 2L2, bay. At the headed end, the studs are joined by an embeded steel MCC 2X1, MCC 2X2 plate. Taken alone, the studs would not meet the minimum embedment requirement. An analysis of the combined stud and plate embedment satisfies the anchorage demands meeting the intent of Anchorage Caveat 5. i Charging Pumps: These pumps are all anchored by eight 1-1/8" J-bolts. The edge 145-041,145-042, 245-041, distance falls below the 4D minimum for this size bolt. An j 245-042 anchorage analysis treating the bolts as 3/4" satisfied the anchorage requirements thus meeting the intent of Anchorage p Caveat 8.

.( Auxiliary Feedwater Pumps: These pumps are all anchored by 3/4" J-bolts. The embedment 145-201,145-331,245-201, length falls below the 16D minimum for this size bolt. An 245-331 anchorage analysis treating the bolts as 1/2" satisfied the anchorage requirements thus meeting the intent of Anchorage Caveat 8.

Main Control Panels: The main contro! panels are all anchored by 5/8" J-bolts. The A, B-1, B-2, C-1. C-2, E-1, E-2, embedded length falls below the 16D minimum for this size bolt.

F-1, F-2, G-1 An anchorage analysis treating the bolts as 1/2" satisfied the anchorage requirements thus meeting the intent of Anchorage Caveat 8.

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PINGP A-46 Seismic Evolusuon Report i November 20,1995 {

i

{-

6. 'Results of the Tanks and Heat Exchanger Review

, Tanks and heat exchangers were evaluated in accordance with the rules and procedures given in Section 7 of the GIP (2.].

' This section gives the results of the tank and heat exchanger reviews performed. In total,62 tanks and i heat exchangers were evaluated, of which 10 were declared outliers.  ;

a 6.1 Evaluation methodology i

The screening evaluations described in this section for verifying the seismic sdequacy of tanks and heat exchangers cover those features of tanks and heat exchangers which experience has shown can ,

be vulnerable to seismic loadings. These evaluations include the following features:

• Check that the shell of large, flat-bottom, vertical tanks will not buckle. Loadings on these types of .

tanks include the effects of hydrodynamic loadings and tank wall flexibility.

i 1 * ~ Check that the anchor bolts and their embedments have adequate strength ~against breakage and pullout.

• Check that the anchorage connection between the anchor bolts and the tank shell (e.g., saddles, 4 legs, chairs, etc.) have adequate strength.

• Check that the attached piping has adequate flexibility to accommodate the motion of large, flat -

bottom, vertical tanks.

The Seismic Capability Engineers, Messrs. W. Djordjevic, G. Ridder, F. Stille, M. McKeown and G.

Gore, reviewed these evaluations to verify that they meet the intent of these guidelines. This review included a field inspection of the tank, the anchorage connections, and the anchor bolt installation against the guidelines described in this Section 7, Section 4.4, and Appendix C of the GIP [2.].

The derivation and technical justification for the guidelines utilized were developed specifically for:

(1) large, flat-bottom, cylindrical, vertical, storage tanks; and (2) horizontal cylindrical tanks and heat exchangers with support saddles made of plates.

l The types of loadings and analysis methods described in this section are considered to be appropriate l for these types of tanks and heat exchangers; however, a generic procedure cannot cover all the possible design variations. Other design features such as wall mounted heat exchangers, heat exchangers and vertical tanks supported on legs not covered by the GIP were evaluated using the same procedures and loading conditions as given in Section 7 of the GlP.

Other types of tanks and heat exchangers (e.g., vertical tanks supported on skirts and structural legs) which were not specifically covered by the guidelines in Section 7 of the GIP were evaluated by the Seismic Capability Engineers using an approach similar to that described in Section 7 of the GIP.

The other types of tanks covered by the screening guidelines in Section 7 of the GIP are cylindrical steel tanks and heat exchangers whose axes of symmetry are horizontal and are supported on their 6-1

PINGP A 46 Seismic Evaluation Report November 20.1995 e curved bottom by steel saddle plates. The screening guidelines are based on the assumption that the A horizontal tanks are anchored to a stiff foundation, which has adequate strength to resist the seismic loads applied to the tank. All the base plates under the saddles are assumed to have slotted anchor bolt holes in the longitudinal direction to permit thermal growth of the tank, except for the saddle at one end of the tank which is fixed. The saddles are assumed to be uniformly spaced a distance S apart, with the two ends of the tank overhanging the end saddles a maximum distance of S/2.

A simple, equivalent static method is used to determine the seismic demand on and capacity of the anchorages and the supports for horizontal tanks. The screening guidelines contained in Section 7 of the GIP specifically addressed only the seismic loads due to the inertial response of horizontal tanks.

If, during the Screening Verification and Walkdown of a tank, the Seismic Capability Engineers determined that the imposed nozzle loads due to the seismic response of attached piping may be significant, then these loads were included in the seismic demand applied to the anchorage and-supports of the tank. The nozzle loads were obtained from existing NSP piping analysis.

The Refueling Water Storage Tank (RWST)

The PINGP Refueling Water Storage Tank is integral part of Auxiliary Bldg. The Tank was constructed in place with the Auxiliary Building. First the steel liner was errected with external rings for embedment and large anchor bolts into the floor. The Tank was then filled with water and 2' to 1-6* concrete was poured around the entire steel tank, the Tank bottom elevation is 695' and the top elevation is 751'.

There are integral floor joints at the 715',735' and 755' elevations. The Seismic Review Team verifitd the RWST to same design basis value as the safety-related buildings.

6.2 Summary of Evaluation Results The results of the A-46 evaluations are summarized belov/. Table 6-1 lists the original design Tanks and Heat Exchangers analyzed to GIP Section 7 rules. Table 6-2 lists the Tanks and Heat Exchangers installed under the SBO modification to current licensing requirements.

Table 6-1 Tank & Heat Exchanger Evaluation Results, GlP Sec. 7 Evaluations ID Description Type Results 053-201 121 D1 DIESEL GENERATOR FUEL OIL VerticalTank OK - Meets Design Basis in Accordance with DAY TANK (On loos) GlP Section 7 Rules 053-202 122 D2 DIESEL GENERATOR FUEL OIL VerticalTank OK - Meets Design Basis in Accordance with DAY TANK (On loos) GIP Section 7 Rules 053-221 121 DIESEL GENERATOR OIL Buried Tank Outlier. The flexibility of buirted piping could STORAGE TANK not be determined from available documentation 053 223 123 DIESEL GENERATOR OIL Buried Tank Outlier. The flexibility of buiried piping could STORAGE TANK not be determined from available documentation 053-251 121 COOLING WATER PUMP DIESEL Buried Tank Outlier. The flexibility of bulrted piping could STORAGE TANK not be determined from available documentation 053-252 122 COOLING WATER PUMP DIESEL Buried Tank Outlier. The flexibility of bulried piping could STORAGE TANK not be determined from available documentation O 053-321 12 COOLING WATER PUMP DIESEL OIL DAY TANK Vertical Tank (On legs)

OK - Meets Design Basis in Accordance with GIP Section 7 Rules 6-2

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I PINGP A 46 Seismic Evaluation Report November 20,1995 E

p Table 6-1 Tank & Heat Exchanger Evaluation Results, GlP Sec. 7 Evaluations (Continued)

ID Description Type Results l 053-322 22 COOLING WATER PUMP DIESEL VerticalTank OK Meets Design Basis in Accordance with 1

OIL DAY TANK (On legs) GIP Section 7 Rules 1 053-381 121 CONTROL ROOM CHILLED Vertical Tank OK - Meets Design Basis in Accordance with I WATER EXPANSION TANK (On skirt) GIP Section 7 Rules  ;

053-382 122 CONTROL ROOM CHILLED Vertical Tank OK - Meets Design Basis in Accordance with )

WATER EXPANSION TANK (On skirt) GIP Section 7 Rules i i 053-481 121 D1 DlESEL GENERATOR Wall Mounted Tank OK - Meets Design Basis in Accordance with <

EXPANSION TANK GIP Section 7 Rules l 053-482 122 D2 DIESEL GENERATOR Wall Mounted Tank OK Meets Design Basis in Accordance with EXPANSION TANK GIP Section 7 Rules

, 135-021 11 RCP SEAL WATER RETURN HEAT Vertical Tank Outlier The stress in the supports exceed GIP EXCHANGER allowables 135-101 12 CL PUMP DIESEL JACKET CLG HX Horizontal Heat Outlier - This heat exchanger is an outlier Exchanger because it is not secured to the pedestals (mounting cradles are secured to the pedestals but heat exchanger is not attached to the cradles). Further both cradles have slotted mounting holes which further makes this component an outlier.

135-111 REGEN HT EX Horizontal heat OK Meets Design Basis in Accordance with exchangers mounted GIP Section 7 Rules on wall 153-011 11 PRESSURIZER RELIEF TANK Horizontal Tank Outlier The tank is an outlier due to the slotting of holes in both support plates 153-021 11 VOLUME CONTROL TANK VerticalTank OK Meets Design Basis in Accordance with (q) v' 153-081 RFLG WTR STG TK (On legs)

Vertical Flat Bottom Tank GIP Section 7 Rules OK - Meets Design Basis in Accordance with GIP Section 7 Rules. Refer to discussion of the RWST in the preceding section.

235-081 22 CL PUMP DIESEL JACKET CLG HX Horizontal Heat Outlier This heat exchanger is an outlier Exchanger because it is not secured to the pedestals (mounting cradles are secured to the pedestals but heat exchanger is not attached to the cradles). Further both cradles have slotted mounting holes which further makes this component an outlier.

235-111 21 REGENERATIVE HEAT Horizontal heat OK- Meets Design Basis in Accordance with EXCHANGER exchangers mounted GIP Section 7 Rules on wall 235-131 21 SEAL WATER HEAT EXCHANGER Vertical Tank Outlier - The stress in the supports exceed GIP allowables 253-011 21 PRESSURIZER RELIEF TANK Horizontal Tank Outlier The tank is an outlier due to the slotting of holes in both support plates 253-021 21 VOLUME CONTROL TANK Vertical Tank OK- Meets Design Basis in Accordance with (On legs) GIP Section 7 Rules 253-081 21 REFUELING WATER STORAGE Vertical Flat Bottom OK - Meets Design Basis in Accordance with TANK Tank GlP Section 7 Rules l'

O) 6-3

PINGP A46 seismic Evaluation Report November 20.1995 l Table 6-2 Tanks & Heat Exchangers Installed Under SBO Modification

[]m These tanks and heat exchangers were installed to current day requirements (See Section 1.4 of this report) and are considered seismically qualified. SRT walkdowns confirmed the absence of seismic spatial concerns for these items.

ID Description 253-361 21 D5 FO DAY TANK 253-331 21 D5 FO STORAGE TANK 253-362 22 D6 FO DAY TANK 253-332 22 D6 FO STORAGE TANK 253-333 23 05 FO STORAGE TANK 253-334 24 D6 FO STORAGE TANK 246-031 DS 1 A START AIR RECEIVER 246-032 051B START AIR RECEIVER 246-033 DS 2A START AIR RECEIVER 246-034 DS 2B START AIR RECEIVER 253-371 D5 ENG 1 FO LEAKAGE TANK 247-021 D5 ENG 1 HT CLNT PREHEATER 253-401 D5 ENG 1 HT EXPANSION TANK 262-441 D5 ENG 1 HT/LT RADIATOR 235-201 D5 ENG 1 UO PREHEATING HEAT EXCHANGER 253-411 D5 ENG 1 LT EXPANSION TANK 253-372 D5 ENG 2 FO LEAKAGE TANK 247-022 05 ENG 2 HT CLNT PREHEATER 253-402 DS ENG 2 HT EXPANSION TANK

/ 262-442 DS ENG 2 HT/LT RADIATOR

')

'y 235-202 D5 ENG 2 UO PREHEATING HEAT EXCHANGER 253-412 D5 ENG 2 LT EXPANSION TANK 246-035 D61 A START AIR RECEIVER 246-036 D618 START AIR RECEIVER 246-037 D6 2A START AIR RECEIVER 246-038 D6 2B START AIR RECEIVER 253-373 D6 ENG 1 FO LEAKAGE TANK 247-023 D6 ENG 1 HT CLNT PREHEATER 253-403 D6 ENG 1 HT EXPANSION TANK 262-443 D6 ENG 1 HT/LT RADIATOR 235-203 D6 ENG 1 UO PREHEATING HEAT EXCHANGER 253-413 D6 ENG 1 LT EXPANSION TANK 247-024 D6 ENG 2 HT CLNT PREHdATER 253-404 D6 ENG 2 HT EXPANSION TANK 262-444 D6 ENG 2 HT/LT RADIATOR 235-204 D6 ENG 2 UO PREHEATING HEAT EXCHANGER 253-374 D6 ENG 2 LEAKAGE TANK 253-414 D6 ENG 2 LT EXPANSION TANK b

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i l PINGP A-46 seismic EvWation Report l Novernber 20,1995 i

1

{

! 7. Results of the Cable Tray and Conduit Raceway Review i 1

) 7.1 Introduction and Purpose t i

{ The seismic adequacy of electrical raceway systems has been identified as an open issue (Unresolved Safety issue A-46) by the U. S. Nuclear Regulatory Commission (NRC) for older nuclear power plant

~

j facilities.

1 This section gives the results of the USl A-46 evaluation for the electrical raceways at Prairie Island.  ;

{ The evaluations were conducted following Section 8 of the GlP..

The seismic evaluation involves conducting a thorough plant walkdown to identify representative,

" worst case" examples of raceway systems and evaluate their adequacy.

The scope of raceways reviewed is reported in Section 7.2 followed by a discussion of the specific raceway hangers (and systems) chosen for the limited analytical review in Section 7.3. Section 7.4 -

provides the raceway assessment criteria (caveats) and the walkdown results. Section 7.5 presents ,

the limited analytical review results. l l

7.2 Scope of ElectricalRaceways Assessed  ;

i O This section describes the areas of the Prairie Island Nuclear Generating Plant that were assessed and the specific raceway systems chosen for evaluation. Electrical raceways are cable tray and conduit systems that are wall-mounted, floor supported and su:: pended systems.

I l

7.2.1 General Areas Covered All buildings and elevations were surveyed. The station walkdowns were conducted May 24,1994, May 30,1995 and July 10 - 14,1995. Essentially all electrical raceway systems were walked down by the Seismic Capability Engineers conducting the Class of Twenty walkdowns. A list of the buildings in which the walkdowns were conducted is shown below:

1. Containment Buildings (Units 1 and 2)

2. Auxiliary Building, including the Fuel Tank Area

3. Turbine Building

4. Screenhouse

5. D5/D6 Diesel Generator Building All rooms and areas within the buildings as listed above, were considered without exception at Prairie island.

All areas were evaluated against the GIP Inclusions Rules and the Caveats (also known as "Other Seismic Concems" and " Seismic Interaction"). Section 7.5 discusses the evaluation criteria at greater length.

O The surveys are documented on Plant Area Summary Sheets (PASS).

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i PINGP A 46 Seismic Evaluation Report

, November 20.1995 j 7.2.2 General Description of Prairie Island Raceways Prairie Island's raceway systems are primarily of light steel strut frame construction. The strut hangers vary from the very simple wall mounted bracket supporting a few conduits to multi-tier, three-dimensional strut frames supporting cable trays and conduits. The predominant strut hanger type at Prairie Island is the ceiling hung trapeze frame supporting cable trays and/or conduits. The largest number of tray tiers at Pra;rie Island were eight tier systems found in the Relay Room which is located in the Auxiliary Building.

The trays varied in size from 6" width to 30" width, primarily of 12" to 24" ladder and trough type construction. Conduits varied in size from 1/2" to 5" nominal diameter and are of rigid steel material (standard schedule pipe). Trays were sometimes sprayed with fire retardant or covered with insulation  ;

board.  ;

1 The trays and conduits were secured to hangers using standard tray clamps (clips), pipe clamps, or '

bolting. No missing or damaged hardware was noted during the walkdowns.

The hangers are generally constructed of single and double channel member posts and double and triple channel member cross members interconnected with 4-bolt ninety degree bracket type fittings. l Anchorage designs such as welding the posts directly to ovorhead structural steel are common although more frequently seen in the Reactor Buildings. Hangers are also anchored to channels which are embedded into (cast-in) concrete or connected to concrete slabs by base plates and expansion anchors. Hangers are sometimes bolted through connection fittings directly to the reinforced concrete  :

slab or wall.

Lateral (transverse) and longitudinal bracing is used in various systems.

Post-TMI plant construction raceway hangers using substantial light steel strut frame construction including rolled (structural tube) shapes were observed in the D5/D6 Diesel Generator Building.

Photographs of the various types of Prairie Island receways are provided in the PASS forms.

7.3 Specific Raceway Systems Evaluated .

7.3.1 General Approach The goal of the evaluation process is to determine overall plant raceway systems acceptability based  ;

on a detailed examination of a focused review scope. The GIP evaluation procedure requires that  ;

each plant evaluates 10 - 20 raceway supports selected for Limited Analytical Reviews (LAR) to envelop the most heavily loaded of the major different support configurations in use at that plant.

Following the GIP, all of the raceway systems and their supports were first checked against the

Inclusion Rules and Caveats. Then the Seismic Review team (SRT) selected representative, worst-case (bounding) samples of the raceway supports on which LARs were performed. This process allows for the establishment of the adequacy of the plant's raceway systems. The actual supports ,

used for LAR were selected following GlP recommendations and at the discretion of the SRT relying on experience and technical judgment.

4 O A limited number of large junction boxes were observed. The conduit / tray feeding into the junction boxes are well supported in all instances. In addition, the junction boxes are also well supported. No unusual conditions were observed.

7-2 l

PINGP A 46 Seismic Evolustion Report November 20,1995 C Raceways spanning seismically separate buildings were also observed. The raceway trays and supports, including cable and conduit, possess adequate flexibility to absorb relative movement between the buildings. It is also noted that relative seismic movement between seismically separate buildings at Prairie Island is very small.

7.3.2 Cable Routing The power supplies and control stations for the majority of safety related equipment are located in the PINGP Auxiliary Building along the G line central part. The safety related power supplies are the 4160 voit medium voltage switchgear cabinets Bus 15, Bus 16, Bus 25 and Bus 26, the 480 volt low voltage switchgear, Bus 111, Bus 112, Bus 121, Bus 122, Bus 211, Bus 212, Bus 221, and Bus 222; and the 480 voit motor control centers (see Table 7-1). The 4160 voit medium voltage switchgear is located on elevation 715 in the Turbine Building and 718 in the D5/D6 Building. The 480 volt low voltage switchgear is located in the Turbine Building on elevation 715, and Auxiliary and D5/Dg buildings on elevation 735. The control for the safety related equipment is in the control room, located on elevation 735, directly above the Relay Room.

Table 7-1480 Volt Motor Control Centers ID Description Building Elevation MCC 2KA2 MOTOR CONTROL CENTER 2KA BUS 2 AUX 695 MCC 1K1 MOTOR CONTROL CENTER 1K BUS 1 AUX 695 MCC 1K2 MOTOR CONTROL CENTER 1K BUS 2 AUX 695 MCC 1KA2 MOTOR CONTROL CENTER 1KA BUS 2 AUX 695 O MCC 2K2 MCC 2K1 MCC 1X1 MOTOR CONTROL CENTER 2K BUS 2 MOTOR CONTROL CENTER 2K BUS 1 MOTOR CONTROL CENTER 1X BUS 1 AUX AUX AUX 695 695 715 MCC 1X2 MOTOR CONTROL CENTER 1X BUS 2 AUX 715 MCC 2X1 MOTOR CONTROL CENTER 2X BUS 1 AUX 715 MCC 2X2 MOTOR CONTROL CENTER 2X BUS 2 AUX 715 MCC 2J2 MOTOR CONTROL CENTER 2J BUS 2 AUX 715 MCC 1L1 MOTOR CONTROL CENTER 1L BUS 1 AUX 715 MCC 1L2 MOTOR CONTROL CENTER 1L BUS 2 AUX 715 MCC 2J1 MOTOR CONTROL CENTER 2J BUS 1 AUX 715 MCC 2L1 MOTOR CONTROL CENTER 2L BUS 1 AUX 715 MCC 2L2 MOTOR CONTROL CENTER 2L BUS 2 AUX 715 MCC 2LA1 MOTOR CONTROL CENTER 2LA BUS 1 AUX 735 MCC 1LA1 MOTOR CONTROL CENTER 1LA BUS 1 AUX 735 MCC 2LA2 MOTOR C,ONTROL CENTER 2LA BUS 2 AUX 735 MCC 1LA2 MOTOR CONTROL CENTER 1LA BUS 2 AUX 735 MCC 1MA2 MOTOR CONTROL CENTER 1MA BUS 2 AUX 755 MCC 1T1 MOTOR CONTROL CENTER 1T BUS 1 AUX 755 MCC 1M2 MOTOR CONTROL CENTER 1M BUS 2 AUX 755 MCC 1T2 MOTOR CONTROL CENTER 1T BUS 2 AUX 755 MCC 2M1 MOTOR CONTROL CENTER 2M BUS 1 AUX 755 MCC 2M2 MOTOR CONTROL CENTER 2M BUS 2 AUX 755 MCC 1M1 MOTOR CONTROL CENTER 1M BUS 1 AUX 755 MCC 2TA1 MOTOR CONTROL CENTER 2TA BUS 1 D5/D6 718 MCC 1 AB2 MOTOR CONTROL CENTER 1 AB BUS 2 SSCRN 695 O MCC 1 AB1 MOTOR CONTROL CENTER 1 AB BUS 1 MCC 1 AC1 MOTOR CONTROL CENTER 1 AC BUS 1 MCC 1 A2 MOTOR CONTROL CENTER 1 A BUS 2 SSCRN TURB TURB 695 695 695 l MCC 1 AC2 MOTOR CONTROL CENTER 1 AC BUS 2 TURB 695 l 7-3 4

--.,-,,-w-v ----r - y- . = - - - - - 4 . , _ . - . - - - ---u --,-- -

. PINGP A 46 Selemic Evaluelion Report November 20,1995

Table 7-1480 Volt Motor Control Centers (Continued) l

. ID Description Building Elevation MCC 1 A1 MOTOR CONTROL CENTER 1 A BUS 1 TURB 695  ;

MCC1TA2 MOTOR CONTROL CENTER 1TA BUS 2 TURB 695 I MCC 2A1 MOTOR CONTROL CENTER 2A BUS 1 TURB 695 )

MCC 2A2 MOTOR CONTROL CENTER 2A BUS 2 TURB 695 I MCC 2AC1 MOTOR CONTROL CENTER 2AC BUS 1 TURB 695 MCC 2AC2 MOTOR CONTROL CENTER 2AC BUS 2 TURB 695 MCC 1TA1 MOTOR CONTROL CENTER 1TA BUS 1 TURB 695 MCC 1R1 MOTOR CONTROL CENTER 1R BUS 1 AUX 735 MCC 1S1 MOTOR CONTROL CENTER 1S BUS 1 AUX 735 3 MCC 2R1 MOTOR CONTROL CENTER 2R BUS 1 AUX 735 i MCC 2S1 MOTOR CONTROL CENTER 2S BUS 1 AUX 735 l l

As a result, the majority of the control cabling for the safety related components at PINGP is routed )

through the Relay Room, either to equipment in the Control Room, or through the Relay Room west 1 wall penetrations to the Auxiliary Building.' The power cabling is routed in the overhead of the 715 or  !

735 elevations either to loads in the Turbine Building or through the Turbine Building wall penetrations to the Auxiliary Building. j

)

in the Auxiliary Building, the majority of the SSEL equipment is located in the central part. The majority i- of the cabling runs directly west either in the 715' or 735' overheads. For the safety related equipment in containment, the power and control cables are routed through penetrations A, B, C, D, E and F.

Based on the equipment layout and cable routing, the majority of the cable trays are located in Relay j Room, the Control Room, the vital switchgear room, and at the east wall of the Auxiliary Building.  !

As required by the.GlP, the entire plant was inspected, however, the inspection focused on the cable tray and conduit systems located in the Auxiliary and Turbine Buildings, and the Unit 1 and Unit 2 ,

Containments. The most heavily loaded cable tray supports were identified in the Relay. Room. Ten l caMe hanger supports, listed below in Table 7-2, were chosen for limited analytical review. Drawings or sketches of the LAR supports are provided in the respective PASS Forms. )

Table 7-2 Locations of Hangers Chosen for Limited Analytical Reviews ,

Description Location h

1 5 tier cable tray support with a vertical post and horizontal Unit 1 Reactor Building,695' i members to the bottom and side of the shell curvature. elevation at Az. 290 degrees P1001 posts and P1004A horizontals. S' hanger spacing.

2 4 tier trapeze cable tray and conduit support; posts Unit 1 Reactor Building, welded to overhead building steel. P1001 posts and 735' elevation at Personnel P1004A horizontals. 4-3/4' hanger spacing. Airlock 3 6 tier trapeze cable tray and conduit support welded to Unit 1 Reactor Building,  :

overhead 4" WF beam which, in tum, is welded to 735' elevation between i i

building steel. P1001 posts and P1004A horizontals. 6' Personnel Airlock and hanger spacing. Equipment Hatch 7-4

PtNGP A-46 Seismic Evalua6on Report

Novembw 20,1995 ,

i i Table 7-2 Locations of Hangers Chosen for Limited Analytical Reviews (Continued)

Description Location j I 4 3 tier trapeze cable tray and conduit support welded to Unit 2 Reactor Building, j overhead 4" WF beam which, in tum, is welded to 735' elevation AZ 330 deg.

! building steel. P1001 posts and horizontals. 8' hanger (north of Personnel Airlock) 2 spacingc l 5 8 tier multi bay trapeze support - all cable trays, welded Relay Room, Auxiliary

! to overhead 4" WF beams which, in tum, are attached to Building,715' elevation near i overhead concrete ceiling slab and beam structure. west wall l P1001 posts and P1004A horizontals. T hanger spacing.

I 6 4 tier trapeze cable tray and conduit support; posts Turbine Building Mezzanine, l welded to overhead building steel. P1000 posts and 715' elevation Grid Location j P1001 horizontals. 8' hanger spacing. D.0/10.1

. 7 5 tier trapeze cable tray and conduit support; One post is Bus Rocai 16, Turbine l attached to an overhead embedded strut in ceiling. The - Building,715' elevation near l other post is welded to a base plate which is anchored east wall bolted to the ceiling. P1001 posts and horizontals. 8' j hanger spacing.

8 3 tier - 3 bay trapeze cable tray and conduit support; Auxiliary Feedwater Pump i P1001 posts are attached to an overhead embedded Room, Unit 1 side, Turbine strut in ceiling with 4-bolt brackets and knee braces. Building,715' elevation '

! P1004A and P1001 horizontal members. 9-1/2' hanger j i spacing.  !

! 9 3 tier trapeze cable tray and conduit support; P1001 Auxiliary Building,715'  !

I' posts and P1004A horizontal members. Posts are elevation, near Flash Tank i l welded to overhead 12" channels. Channels are anchor j bolted to concrete ceiling beams. 8-1/2' hanger spacing.

i 10 3 tier trapeze cable tray and conduit support; ; P1001 Auxiliary Building,695' posts and P1004A horizontal members. P1001 posts are elevation,in west hallway attached to an overhead embedded strut in ceiling with 4- near Fuel Haul Area

bolt brackets.

}

f j

7.3.3 Cable Data and Weight Determination 1

NSP maintains the cable tray fill data at PINGP in the Cable Tray Fill Report (CTFR) which is 4

accessible on the Plant Information Systems Computer. With few exceptions, the cable trays at PINGP are identified by a unique number which is maintained in the CTFR. The cable tray fill data is l presented in terms of actual fill area available and percent fill area available.

j The CTFR was used to determine the cable tray fills for LAR 005, a multi-tier and multi-bay support located in the Relay hoom. A visual determination was not practical because of the congested nature 4

of the area. Standard GlP recommended cable tray fill weights were used for this evaluation based on j the CTFR data. j i

All other LAR's were based on cable tray fills as visually determined by the SRT.

7-5

PINGP A-46 seismic Evaluation Report November 20.1995 7.4 Raceway Seismic Evaluation Criteria and WsIkdown Results This section discusses the raceway seismic evaluations for the Prairie Island Nuclear Generating Plant.

7.4.1 GIP Inclusion Rules Results As previously stated, all raceway systems in the Buildings noted in Section 7.2.1 were included in the walkdown, it is important to note that a very thorough review of most raceways, raceway supports and supporting concrete was accomplished.

Without exception, no anomalies in design or construction were found. All inspected raceways meet the requirements of Section 8.2.2 of the GlP as follows:

• - Cable tray spans did not exceed the 10' limit between adjacent supports and the 5' limit for cantilevers;

• Conduit spans were within the limits required by Rule 2 of Section 8.2.2 [2.];

• On all cantilever bracket-supported systems cable trays and conduit were found secured to their supports so no tray or conduit sliding can occur,

• Channel nuts used with light metal framing systems were nuts with teeth (ridges) stamped into the nuts ( Fig. 8-1, Ref. 2.);

• No " rigid boot" type connections or similar types (Fig. 8-2, Ref. 2.) were observed during the walkdown inspection; O

• No beam clamps were observed during the walkdown inspection; Cast-iron anchor embedment rule implementation was resolved as follows: To check for cast iron anchorage embedments in a walkdown at all support locations was not feasible; however an SRT

. review of PINGP cabletray hanger detail drawings reveals the use of commonly used and well documented ductile steel anchor types (primarily Phillips shell anchors). No field evidence of cast iron embedments was found by the SRT. Therefore, this issue was judged to have no impact on Prairie Island.

Prairie Island meets the Cable Raceway inclusion Rules of the GlP in their entirety.

7.4.2 GlP Other Seismic Performance Concerns & Seismic Interaction Review in addition to the inclusion Rules the SRT inspected the raceway system's for the Caveats known as "Other Seismic Performance Concems" and " Seismic Interaction Review". The assessment results are as follows:

,Other Seismic Performance Concems

• All raceway anchorages were reviewed for adequacy in accordance with Section 8.2.3 [2.]. No concems were found;

• No concems were found regarding visible cracks, significantly spalled concrete, serious honeycombs or other gross defects in the concrete to which the raceway supports are attached; 4

• No significant corrosion of cable trays, conduit supports or anchorage was noted by the SRT; 7-6

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PINGP A-46 seismic Evaluatkm Report November 20,1995

• No noticeable sag of any conduit or cable tray as defined in Concem 4 of Section 8.2.3 [2.) was observed; '

• No broken or missing cable tray and conduit components were found by the SRT;

• All cables inspected were restrained so they will be kept in the tray during an earthquake with the exception of that noted in vertical cable tray (2CU-T61-1) located near Pen 42D was without ties.

Cabling was found protruding outside of tray envelope.. . Per 6/7/95 telecon with G. Gore (NSP), the cabling was identified as non-safety however the concem will be remedied by the implementation of Maintenance Work Order #950416. No other concems of that type were observed by the SRT.

• A sampling of plastic ties were pull-tested, and no brittle ties of plastic materials were found by the SRT;

• Tho SR7 evaluated the raceways for stiff /short supports and found no instances of this design flaw.

Prairie island's hangers are of uniform height in long flexible runs of cable trays or conduit.

No fiM;ngs were noted with respect to "Other Seismic Performance Concems". l l

Seismic Interaction l

• The raceway systems were reviewed for seismic proximity interaction in accordance with Appendix D [2.). The SRT identihed one seismic interaction concem in an area known as the operators exercise room located in the Auxiliary Building on elevation 735.

• The raceway systems were reviewed for falling hazards in accordance with Appendix D [2.].  ;

• Conduit and cables were reviewed for sufficient flexibility to accommodate differential displacement l between safe shutdown equipment and adjacent equipment and structure. No concems were found by the SRT;

• No Isolated Outfiers (other findings) were found by the SRT.

In conclusion, one findings was noted with respect to " Seismic Interaction."

7.5 Limited Analytical Review (LAR) Results This Umited Analytical Review (LAR), performed within the scope of Unresolved Safety issue (USI) A- l 46, evaluates the structural integrity of cable tray and conduit supports, which have been chosen as representative, worst case examples of the raceway support configurations within the Prairie Island Nuclear Generating Plant.

The hangers (members, connections and fittings) were first evaluated for static, dead load stresses.

They were then evaluated for lateral load ductility to ensure that there were no brittle failure loads.

Finally, the vertical capacity was checked by comparing the support anchorage capacity to 3 times the support deadweight. If any of these evaluations fail, the support is declared an " outlier" and additional evaluations of lateral load capacity are performed. This section describes the criteria and overall results for all ten t ARs.

In all,10 raceway systems (supports) were chosen for LAR evaluation as shown in Table 7-2.

7-7

. . _ _ ~ . . _ _ . . . _ . _ _ . _ _ . _ _ _ _ _ . _ . _ . - . . _ _ _ . _ _ . _ _ _ _ _ _

i PINGP A-46 seismic Evfuation Report November 20,1995 1 7.5.1 Summary of Results The critical interaction value and related comments for each of the raceway support evaluations in this LAR are summarized in Table 7-3 below. Refer to Reference 11. for details of each of the evaluations.

Table 7-3 Critical Interaction Values LAR No. Interaction Value Members Fittings / Anchorage

• Maximum Connections 001 0.16 DL 0.03 DL 0.25 DL 0.25 002 0.32 DL 0.08 DL 0.15 3DL 0.32 003 0.29 DL 0.04 DL 0.05 3DL 0.29  ;

I 004 0.08 DL 0.02 DL 0.04 3DL 0.08 005 0.61 DL 0.13 DL 1.10 3DL 1.10**

]

006 0.42 DL 0.06 DL 0.10 3DL 0.42 007 0.20 DL 0.11 DL 0.50 3DL 0.50 008 0.10 DL 0.15 DL 0.67 3DL 0.67 l 009 0.12 DL 0.06 DL 0.20 3DL 0.20  !

l 010 0.08 DL 0.07 DL 0.67 3DL 0.67 DL - Dead Load 3DL- 3x Dead Load (Vertical Load Check)

• • Acceptable per SRT Judgment

• Support connection to building structure.

7.5.2 Logic Diagrams for Cable Tray and Conduit Support Evaluations j Logic diagrams indicating the evaluation path taken to demonstrate the acceptance of each of the raceway supports are shown below. Note that the particular evaluation path taken for the support in question is defined in heavy outline. As previously noted, the hand calculations are given in Reference 11.

O 7-8

. . . -. .. _ -. ~. ..... . ...-. - . . . .. .-.. . . ... ~ -.. ... -.-.__ _ .- . _ . - -

t PINGP A-46 Seismic Evaluation Report i

November 20,1995 LAR Nos. 002 through 010 START

% /

.8.3.1 Yes* Does the Support Hove No Vertical Dead Load Copacity )

With Eccentricities g 1.0 x DL ?

Yes 4

% / l 8.3.2 Does the Support Have s Vertical Capoch 3.0 x DL ? -

Yes

% /

8.3.3 No 034 Does the Support Hove Ductile ) Does the Support Have No )

Response to Lateral Loading ? Adequate Lateral Load Stren@th

, l Yes Yes s ,

r s 1 O _

is the Support a Fixed - End 8.3.5 Yes )

8.3.5 Does the Fixed - End Rod Honga No )

Rod Honger ? Pass Fatique Evoluotion ?

No .,

Yes

[ PASS

)t ANALYTICAL REVIEW OUTLIER GO TO SECTION 8.4

• (Directly mounted or rigidly cantilevered from structural wall) f O

7-9

l PINGP A-46 seismic Evaluation Report November 20.1995 O l V 7.5.2 Logic Diagrams for Cable Tray and Conduit Support Evaluations (Cont.)

LAR Nos. 001 Note: Evaluation path is defined ir, heavy outline.

START v

8.3.1 Yes* Does the Support Have No Vertical Dead Load Capac b/

With Eccentricitiesj 1.0 x DL 1

)

Yes V

8.3.2 O

Does the Support Have s Vertical Capacity 3.0 x DL ?

Yes O v V 8.3.3 No 8.3.4 N

Does the Support Have Ducined Does the Sup3 ort Have )

Response to Lateral Loading ? Adequate Latera Load Strength Yes (

v 8.3.5 8.3.5 is the Support a Fixed - End Yes ) Does the Fixed - End Rod Hangw No>

Rod Hanger ? Pass Fatique Evaluation ?

No ( Yes v V PASS OUTLIER

' O TO SECTION

' ANALYTidAL REVIEW 8.4

• (Directly mounted or rigidly cantilevered from structural wall)

/~N.

U i

7-10

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h f PINGP A 46 seismic Evaluelion Report

. November 20,1995 i

lO 7.6 Results and Conclusions The electrical raceways were walked down as part of the USl A-46 effort. All areas of the plant were

surveyed and inspected against inclusion rules and caveats for raceways such as maximum spans, .

!~ missing or broken hardware, and good design practices as presented in the GIP,' Section 8. The  ;

i- results were documented in Plant Area Summary Sheets. In addition, bounding and representative

! supports were selected for structural and seismic evaluations called Limited Analytical Reviews (LAR).

l The LAR evaluations checked dead load stresses, ductility, and veitical capacity. Each of the ten j cable tray and conduit supports chosen for the LAR met the guidelines as set forth in Section 8.3 of the -

l GlP.

i .

The electrical raceways at the Prairie Island Nuclear Generating Plant are concluded to meet the Cable  ;

j and Conduit Raceway Rules of the GIP in their entirety based on the plant walkdowns and Limited 1 Analytical Reviews performed by the SRT.

4 i

7.7 Summary of Cable and Raceway Outilers The SRT identified a single incidence of seismic interaction concems in the cable and raceway review.

Unrestrained exercise equipment existed in the operators exersise room in the Auxiliary Building on elevation 735. Some exercise equipment attachments are stored in and against cable tray. Potential

% seismic interactions exist with cable tray 2 ART 141, a pull box, and safety related conduits labeled 2CNB1 and 1CNB1, and also other conduits labeled 1EM 13 and 2EM-13. The SRT observed that dumbells resting on the floor could roll around during a seismic event in this area.

4 7-11

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PINGP A-46 seismic Evalus6on Report j November 20,1995 i

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8. Description of the Equipment Outliers l

j This section discusses the outliers identified during the USI A-46 walkdowns conducted at Prairie i

Island. The outliers are identified from the Twenty Classes of Equipment discussed in Section 4, the l Tanks & Heat Exchangers Review discussed in Section 6, and the Cable Tray & Conduit Raceway i

' Review given in Section 7. Relay outliers are discussed in the Prairie Island Relay Evaluation Report i [20.]. 1 i l l An outlier is an item of equipment which does not comply with all of the screening guidelines provided l l in the GIP. The GlP screening guidelines are intended to be used as a generic basis for evaluating the j seismic adequacy of equipment. If an item of equipment fails to pass these generic screens, it may I

still be shown to be adequate by additional evaluations.

! 4 Section 9 provides a discussion of the disposition or corrective action, as appropriate, for each outlier

! discussed below, i

i

l 1 l 8.1 Generic OutilerIssues l

! The Seismic Review Teams identified only one generic issue during the final walkdowns that affected l A46 equipment items. In numerous locations, overhead florescent lights connect to supports by open S-hooks. During a seismic event, this configuration could possibly release the light fixture casuing l seismic interaction hazards. PINGP issued Work Order #9510032 to crimp the S-hooks on overhead light fixtures. This work will proceed as a general maintenance activity.

8.2 Equipment Specific Outilers Identitled During the Final A46 Walkdowns:

l Total of ninety-six (96) Outliers identified during the walkdowns are listed below. The outlier discussions l below group the equipment with a common type of outlier. The proposed resolution of these outliers are diarmaad in Section 9.

Table 8-1 Descr!ption of Class 0 - 20 Equipment Outilers ID EQUIPMENT OUTLIER FINDING A1. Motor Control Centers

• The anchorage has insufficient capacity to resist the seismic MCC 1 AB1, MCC 1 AB2, demand.

MCC 2R1, MCC 2S1 A2. Motor Control Center: The "RHR Block Lifting Fixtures" stored nearby present an MCC 1K1 interaction hazard to the MCC A3. Motor Control Centers: The MCC's rest on 2.75" shims causing bending in the anchor MCC 1TA1,MCC 1TA2 bolts.

O A4. Motor Control Centers:

MCC 1L2, MCC 2K2 These MCC's contain essential relays and have potential seismic interactions from nearby piping.

8-1

i.

i r i PINGP A46 seismic Evaluation Report j November 20,1995 Table 8-1 Description of Class 0 - 20 Equipment Outilers (Continued)

EQUIPMENT OUTLIER FINDING f

ID A5' Motor Control Center: The SRT observed that this MCC rocks about its weak axis when i j MCC 2LA2 bumped, making the welding at the base suspect.

j A6' Reactor Trip Breakers: An unbraced overhead room chiller supported on rod hangers may I

2-52/RTA,2-52/RTB swing and break nearby water piping which would sptay the room.

Also, a unit heater hung on 10 ft long rods could swing and break its steam heatina pipina.

8 A7' Pressurizer Heater Group A unit rod hung unit heater could swing and break its piping l A Transformer spraying the area. j r 2PZRHTRA/XFMR  ;

A8' Diesel Generator Submersable pumps are not part of GIP's earthquake experience

Submersable Oil Pumps: data base.

045-271, 045-273, 045-301,045-302 A9~

DD CLP: Anchor bolts for these pumps do not meet the least acceptable 145-392,245-392 edge distance (40). Also, the vertical shaft length exceeds the maximum lenath in the bounding spectrum caveat.

A10' Pressurizer Relief Valves: The floor response spectra (demand) exceeds the capacity spectra RC-10-1, RC-10-2, of 1.5 times the bounding spectrum. l 2RC-10-1, 2RC-10-2 l Fan Coil Unit Cooling Contact with conduits could break the solenoid tap connection for

' A11.

Water Control Valves: these valves.

CV-39401, CV-39409 A12* FCU Cooling Water Contact between the valve diaphragm housing and a 1" conduit Control Valve: could cause chatter of the limit switches in the solenoid valve CV-39421, A13. Shroud Cooling Coils The floor response spectra (demand) exceeds the capacity spectra Chilled Water Control of 1.5 times the bounding spectrum.

Valves:

CV-39405, CV-39417,  !

. CV-39419 _

A14. Screenhouse Exhaust The seismic demand exceeds the seismic capacity. Also, the

, Fans: anchorage details for these fans could not be determined.

132-281, 232-281 A15' Diesel Generator Outside This valve has a potential for differential displacement between the Air Demper SV: wall and ductwork that supports the valve's tubing.

l SV-33498 A16. Chilled Water Cooling Seismic demand based on floor response spectra exceeds seismic Water Isolation Valve: demand based on 1.5 times the bounding spectra.

SV-37467 A17. Relay Room Fan Coil Anchorage capacity of the FCU's could not be established.

Units:

074-031,074-032,074-033,074-034 s A18. Auxiliary Feedwater Pump Anchorage capacity of the FCU's could not be established. Also, a Fan Coil Units: potential seismic interaction exists for 174-051 with a ceiling 174-051, 274-051 mounted multi-tier conduit.

8-2

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PINOP A46 seismic Evuluation Report November 20,1995 -

l Table 8-1 Description of Class 0 - 20 Equipment Outilers (Continued)

ID EQUlPMENT OUTUER FINDING A19* Containment Fan Coil The floor response spectra (demand) exceeds the capacity spectra ,

Unit: of 1.5 times the bounding spectrum.

174-013 A20' Fan Coil Unit Control The floor response spectra (demand) exceeds the capacity spectra  !

Dampers: of 1.5 times the bounding spectrum. j CD-34076 through 34079, CD-34084 through 34087, A21. Control Room Water These chillers have unconfined steel isolator springs in the base Chillers: support.

075-011, 075-012  !

A22. Distribution Panels: The panels base is elevated above the floor subjecting the  :

PNL 11,12,21,22 anchorage to bending stresses A23. Batteries: The batteries are over 10 years old.

11 BATT,21 BATT A24. Batteries: Some of the battery spacers are missing in each rack.

12 BATT,22 BATT ,

A25. Battery Chargers: The cabinet bases are ekvated above the floor subjecting the t 11 BATT CHG,12 and 22 anchorage to bending stresses. Also, there is a sliding door counter weight which could swing into 22 BATT CHG.  !

A26. Diesel Generators: A local control panel mounted on the DG skid is supported on very l O A27.

034-011,034-021 Control Room Panels and Racks:

flexible (wobbly) steel springs.

Aluminum diffussers in the control room ceiling pose a personnel hazard.

14MR,1NR3,1NR4, 2NR3,2NR4, A, B-1, B-2,

• C-1, C-2, D-1, D-2, E-1, E 2, F-1, F-2, G-1

' A28.

Diesel Generator Remote A computer table with a loose computer CRT and printer sits Terminal Unit Cabinets: adjacent to the RTU cabinets presenting a seismic interaction D5/RTU, D6/RTU hazard.

l l

l I

l l

l l

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PINGP A 46 Seismic Evduation Report i November 20,1995 l l

9. Resolution of Outliers I

The licensing cover letter transmitting this report describes the strategy for outlier resolution.

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PINGP A-46 Seismic Evaluation Report November 20,1995

10. References

1. Generic Letter 87-02, " Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety issue (USI) A-46", USNRC, Washington, D.C.,-

February 19,1987.

2. " Generic Implementation Procedure (GIP), for Seismic Verification of Nuclear Plant Equipment",-  ;

Revision 2, Corrected,2/14/92, Seismic Qualification Utility Group.

3. " Supplemental Safety Evaluation Report No. 2 (SSER #2) on GIP-2", USNRC, Washington, D.C., l May 22,1992.

4. USl A-40 " Seismic Design Criteria Short-Term Program", USNRC, Washington, D.C.

5. USl A-17 " Systems Interactions In Nuclear Power Plants" USNRC, Washington, D.C.

6. NSP Response to Supplement 1 to GL 87-02 on SQUG Resolution of USl A-46, Prairie Island Nuclear Generating Plant, Letter from Thomas M. Parker (NSP) to USNRC, dated September 21, 1992.

7. USNRC Letter " Safety Evaluation of Prairie Island Nuclear Generating Plant Unit Nos.1 and 2, 120-day Response to Supplement No.1 to Generic Letter 87-02 (TAC Nos. M69474 and M69475)", M. Gamberoni (USNRC) to T. M. Parker (NSP), dated November 30,1992.

8. Stevenson & Associates," Prairie Island Nuclear Generating Plant Seismic Response Spectra for USI A-46 Program",93C2807-C-002, March 7,1994.

9. SPECTRA Software Package, Stevenson & Associates, Version 2, November,1992.

10. EPRI Report NP-7146, " Development of In-Cabinet Amplified Response Spectra for Electrical Panels and Benchboards." Revision 0, Electric Power Research Institute, Palo Alto, CA, prepared by Stevenson & Associates, December,1990.

11. " Cable and Conduit Raceway Umited Analytical Review (LAR) for USI A-46 at Prairie Island l Nuclear Generating Plant" Calc. 93C2807-C-011, Rev. O,10/26/95, by Stevenson and j Associates.  ;

12. SQUG Letter from N.P. Smith, SQUG Chairman to J.G. Partlow of USNRC,

Subject:

Revision 2A ,

to the Generic Implementation Procedure, dated March 26,1993.

13. ' SSRAP Report, "Use of Seismic Experience Data to Show Ruggedness of Equipment in Nuclear Power Plants," Senior Seismic Review and Advisory Panel, Revision 4.0, February 28,1991.

14. Northem States Power Company, Prairie Island Nuclear Generating Plant - Drawings. (Dwg.

numbers are specified where referenced).

15. EPRI Report NP-5228-SL," Seismic Verification of Nuclear Plant Equipment Anchorage (Revision 1)." Electric Power Research Institute, Palo Alto, CA, prepared by URS/ John A. Blume &

O Associates, Engineers, June,1991.

104

. . _ _ . _ . . . . . . _ . . _ _ - - _ _ - _ . - . . _ . _ _ . _ _ _ . - _ _ . . _ _ _ - . _ . _ _ . . _ . . _ ~ _ . _ _ _ _ _ - . _

PINGP A 46 seismic Evalua6on Report November 20,1995 O 16. ACI 318-83," Building Code Requirements for Reinforced Concrete", American Concrete Institute, 1983.

l I

17. EPRI Report NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1).", Electric Power Research Institute, Palo Alto, CA, prepared by JR Benjamin Associates et. al., August,1991.

I

18. NSP," Prairie Island Updated Safety Analysis Report"

19. ANCHOR 3.0 Software Package (with Verification and Users Maruals), Rev. O, 8/16/90 by  ;

Stevenson and Associates.  !

20. NSP, "USNRC USl A-46 Resolution, SSEL and Relay Evaluation Report" for Prairie Island 1 Nuclear Generating Plant, Units 1 and 2", November,1995.

21. " Prairie Island Nuclear Generating Plant Earthquake Analysis: Reactor-Auxiliary-Turbine ,

Building", John A. Blume & Associates Report JAB-PS-02, January 22,1971 1

22. " Prairie Island Nuclear Generating Plant Earthquake Analysis: Reactor-Auxiliary-Turbine Building i Response Acceleration Spectra", John A. Blume & Associates Report JAB-PS-04, February 16, 1971 4

23. " Prairie Island Nuclear Generating Plant Station Blackout / Electrical Safeguards Upgrade (SBO/ESU) Program Design Report", Rev. 2, dated August 18,1993.

. 24. EPRI TR-103960, " Recommended Approaches for Resolving Anchorage Outliers", Final Report-June 1994 t

O 10-2

l PINGP A-46 Seismic Evduation Report November 20,1995 i

' Appendix A: Peer Review Assessment 9

\

l I

l

Structural Mechanics Consulting, Inc.

Robert P. Kennedy 18971 Villa Terrace, Yorba Unda, CA 92686 o (714) 777-2163 O

October 23,1995 Mr. Albeit M. Kuroyama Project Manager Prairie Island Nuclear Generating Plant

'1717 Wakonade Drive East Welch, hN 55089

Subject:

Peer Review of A-46 Walkdown Performed by Stevenson & Associates (S&A)

Dear Mr. Kuroyama:

' On July 25,1994, I conducted a peer review walkdown of the A46 walkdown performed by Prairie Island (PI) engineers and their contractor, Stevenson & Associates (S&A). On August 16,1994, I issued a letter (copy attached) to Mr. Bihari Desai (PI) summarizing my peer review. - I made several specific comments and recommendations which are not repeated herein. My O\ overall finding was that the walkdown was of high quality and consistent with the requirements of the Generic Implementation Procedure (GIP) and that the conclusions being reached were appropriate.

However, final Screening Evaluation Worksheets (SEWS) and Outlier ,

Seismic Verification Sheets (OSVS) were not available at the time of my previous  !

peer review walkdown, and my review of a sample of completed SEWS and OSVS was left as an open issue. I have now completed my review of a sample of SEWS and OSVS. The sample selected is listed in the attached table. All eight reviewed SEWS appear to be complete and appropriately filled out. The conclusions of the 2-Seismic Review Team (SRT) appear reasonabic.

On four (MCC 1 AB2,045-301,075-011, and PNL 11) of the eight SEWS, the SRT concluded that seismic outlier issues existed and completed OSVS forms.

These OSVS forms clearly define the outlier issues involved and I concur with the defined outlier issues. No proposed method of resolving outliers has been provided. Although providing a proposed method of outlier resolution is optional,  !

I strongly recommend that Prairie Island develop and implement a set of proposed methods of resohing these outliers within a reasonable time frame. The outlier issues on MCC 1AB2,075-011, and PNL i1 are seismically significant and are O

,e -

g -- ,--

October 23,1995 Page 2 O

d relatively easily fixed. The outlier issue on 045-301 appears to be primarily a

" paper" issue and can probably be resolved by a " paper" study.

This letter closes all open issues on my peer review of the A46 seismic '

walkdown of Prairie Island. I appreciate the opportunity to have been of service.

Please contact me if you have any questions.

Sincerely, nnedy [

k O

i o

O

I i .

i I

O Table: Peer Reviewed SEWS Component ID Description Class '

MCC 1AB2 Motor ControlCenter 1AB Bus 2 1 MCC iMi ' MotorContr'ol Center 1M Bus 1 1 l 045-301 121 Diesel Cooling Water Pump Oil Storage Tank Submersible Pump 6 CV-31084 11 Steam Generator Main Steam Safety Rollef to ATM Control Valve 7 SV-37036 RCS Vent System PressurizerVent SV  ;

8 '

s 075-011 121 Control Room Water Chiller 11 PNL 11 Distribution Panel 11 PNL 116 14 Non-Interruptable Panel 116 14 l

1 i.

O

Structural Mechanics Consulting, Inc R:b:;rt P. Kennedy 18971 Villa Terrace, Yorba Linda, CA 92686 e (714) 777 216:-

V August 16,1994 Mr. Bihari Desai Northern States Power Company Prairie Island Nuclear Power Plant 1717 Wakonade Drive East Welch, MN 55089

Subject:

Preliminary Peer Review of A-46 Walkdown Performed by Stevenson &

Associates (S&A)

Dear Mr. Desai:

Please let this letter serve as my comments concerning the subject review. This site review is considered preliminary in that final Screening Evaluation Worksheets (SEWS) are not yet completed for Prairie Island (PI). I recommend another site visit near the end of the project when final SEWS are in place.

Walkdown participants were Messrs. B. Desai and M. McKeown of the PI staff, Mr. W.

Djordjevic of S&A, and myself acting as peer reviewer. The walkdown was conducted in all accessible areas of the plant on July 25,1994. All equipment in those accessible areas were reviewed with specific attention given to potential outliers as determined by the Seismic Capability Engineers (SCE). The only areas not reviewed v'ere the Containments since both units were in full operation. I found the work and the assessments made by PI engineers and their contractor, Stevenson & Associates (S&A), to be of high quality and consistent with the requirements of the Generic Implementation Procedure (GIP) and the conclusions to be appropriate.

Specific comments and recommendations are offered hereafter for PI's consideration:

1.1 ARP1 Regarding the Relay Racks in the Unit I relay room, many ,

of them are supported on shims due to the pitched floor design for drainage purposes. I do not consider them outliers so long as the bolt bending is evaluated and found acceptable. This comment applies to all cabinets / panels in the relay room. No open S-hooks were noted in the relay room.

2. D1 & D2 Static Found open S-hooks above the Exciter. Ladder hung on Exciter wall behind D2 Exciter should be chained to prevent ladder from falling into Exciter.

3. MCC ITA2 Scaffolding hung on wall behind MCC and D2 Exciter q should be chained to prevent it from falling into equipment.

O

i? ..

August 16,1994

Page 2 4,22 Battery Charger SCE's identified concern with counter weight on open S-a hook swinging into the charger or falling off. This would F

only occur if the door were open and since the door is i normally closed due to the counter weight, this is not .

considered a concern by me. The counter weight could a swing into the 125 VDC 22 Panel, but this panel contains 2

no relays so this is not a concern.

5. MCC 2AC2 There is an approximately 3/8" gap between the top of the MCC and an adjacent cable tray hanger. Check if they have the potential to collide.

6.31411 & 31410 Limit switches are in contact with one another and thus -

. could break offduring an event.

7. Station Batteries Some battery cells have no spacer above the batten (Room 12) connecting rods. This violates spacer caveat for Class 15.

Check ifIEEE 344 seismic qualification test was .

performed on this same configuration to resolve the issue; otherwise, insert spacers above rods.

'.8. Horizontal pumps . Some pumps have alignment pins or lateral blocks (such ; _

, as TDAFWP and 12MDAFWP), and some do not (such as MDAFWP,1ITDAFWP and S1 pumps). SQUG needs to determine whether this is an issue or not.

- 9. Diesel Generators a) Plant Hatch had similar issue as PI with soft springs on local DG control panel mounted on skid. Check with -

O Don Moore on how they resolved it, b) Check on whether or not fire piping overhead is seismically evaluated and determine if system is charged. If charged, broken piping could leak water on DG.

10. B121 SWGR and Chain step ladderin vicinity XMFR

11. B15 Bus Chain ladder behind the Bus.

12. Buses 122 & 222, Check Westinghouse manufactured cabinets adjacent to

& Bkr 122C MCC IP and 2P for positive anchorage as they may pose an interaction hazard to noted buses / breakers

' 13.1 & 2 NR4 Instrument panels in contact with Panel F (main control board) on both units. Interaction issue only if either c,abinet(s) contain essential relays.

14. Control Room Aluminum ceiling diffuser panels should be ' secured to T-bar drop ceiling framing in some fashion to preclude their falling on operator personnel and injuring them. Seismic housekeeping in control room appears very good and no -

issues were'noted.

15.121 Control Room If glass breaks, what happens to the control room chilled Chilled Water Air water system?

Separator O

Au3,ust 16,1994 -

Page 3 O 16.31205 Valve is directly in contact with wall (no clearance).

Check piping displacements to determine if this is an issue, or investigate changing valve orientation, or support valve and pipe. ,

I will be happy to answer any questions you have regarding this peer review assessment.

Please contact me or Mr. W. Djordjevic of Stevenson & Associates to discuss these matters and any future peer reviews that may be called for.

\

Sincerely, . j Robert P. Kennedy cc:

Mark V. McKeown, PI Walter Djordjevic, S&A

O t

i O

l

} PINGP A 46 Seismic Evaluation R port November 20,1995 i l

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h m Northem States Power Company BUILDING : Auxiliary (Concrete)

Prairie Island Nuclear Generating Plant MASS POINTS:30 Amphfied Floor Response Spectra D!RECTION : HORIZONTAL Safe Shutdown Earthquake (SSE) RADIAL DIST:O ELEVATION : 735.00 10 ........ ... 3... .. .....

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PRACTICAL AND PROFESSIONAL EXPERIENCE

'O - Resume for Gerald Gore EXPERIENCE July 76 to Feb 81 Hanna Mining Co Hibbing MN (The remaining open operation is at Keewatin as National Steel Pellet co.)

Work locations included Butler Taconite, National. Steel Pellet Co.

and Hanna's District office.

Work experience included: Mine planning estimating _and layout,(as Junior Mining Engineer under a Senior Mining Engineer) drilling and blast pattern design, blasting noise and seismic monitoring, evaluation of blast damage claims, drill bit failure / economic analysis, and drilling equipment specification (as a drilling and blasting engineer reporting to the Lead Mining Engineer) .

Oct 1981 to Oct 1995 Northern States Power (Prairie Island Nuclear Generating-Plant).

Latest 18 months as acting superintendent of Civil and Mechanical Engineering for the Nuclear Projects group. Previous 21/2 years as Project Engineer for the Nuclear Projects Group (Projects include:

' Installation of fire penetrations and equipment, 24 and 30" cooling O- water (service water) header replacement and hydraulic modeling),

and 10 Years as system engineer for the Prairie Island Nuclear Generating Plant (As system engineer responsible for operations support, maintenance and modification coordination of the following systems: Safety Injection, Coordinator for Safety Injection suction line replacement, Condensate,. Feedwater, Heater Drains, Reactor Make up Water, Station Air, Snubbers (seismic and event type),

Station Turbine Generator systems and the Motor Operated Valves program).

OTHER Member of Utility User Groups for Prairie Island Nuclear Generating Plant: Snubber user group (SNUG thru 1990) and MOV (Motor Operated Valve group in 1991) .

Completed NSP SRO License certification Program.

Passed Minnesota EIT test.

Completed EPRI A46 Walk Down (SQUG) training for Seismic Capable Engineers.

Also previously certified in Weld Inspection and ASME Visual testing VT 1,2,3,4 level 3. Currently certified in Local Leak Rate Testing and Visual Testing VT 2 level 2.

Present member of the Prairie Island Emergency Response Organization.

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Mark V. McKeown i s

EDUCATION:

Tufts University, Medford,'MA~ ,

May 1982 t Sachelor of Science in Civil Engineering and Geology _ l Part time graduate studies in Civil Engineering at University of New Hampshire, Tufts University, and University of Lowell.

PROFESSIONAL WORK EIPERIENCE:

Northern States Power Co., Prairie Island Nuclear Generating Plant i Senior Civil Engineer _

June 1991 - present  !

Lead project engineer for on-site dry fuel storage activities. Responcible for i fabrication activities, loading and transfer operations.

Provide engineering and project management support to plant construction projects,  ;

including cost assessment, scheduling and coordination of A/E activities. Proj ect responsibilities included upgrading a 125 ton bridge crane ($3M); and developing a  !

structural / seismic verification walkdown process for site implementation. Additional  :

responsibilities include supporting Operation and System engineers by performing design and analysis; esp. structures, piping, pipe supports and component supports and t electrical supports. i e

Civil Engineer June 1599 - June 1991 Responsibilities include developing technical responses to NRC inquiries, provide I engineering support to a variety of plant design changes, participate as a group leader

. in EPRI Technical Working Croup for industry wide procurement evaluations. Daily use

[ of: word processing and spreadsheet applications Telodyne Engineering Services, Waltham, MA

~

Project Engineer June 1982 - June 1989 Responsibilities included project management of jobs from $5K - $1M, sales and marketing {'

at various electric utilities and proposal writing. _ Evaluation and field engineering associated with the design of nuclear and fossil power piping systems under static and i dynamic loads. Stress analyses of frames and components using finite element techniques and requiring knowledge'of various computer codes. i Project work included:

• Interface between Architect-Engineer and electric utility on $1 million piping upgrade project. As responsible engineer I managed scheduling, documentation, and developed construction work packages.

• Design review of a waste to energy facility's major structures which could impact safety of personnel, including columns, reinforced pushwalls & explosion chambers.

• Provided onsite engineering for the structural design and review of cable tray and conduit supports during construction of Seabrook Power Station.

OTHER WORK EXPERIENCE:

Prairie Island SRO Certification - November 1993

?EMBERSHIPS:

Registered Professional Engineer, Minnesota American Society of Civil Engineers - Associate Member

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! GREGORY C. RIDDER l Nuclear Power Deputment l Wisconsin Public Service Corporation i

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l UniversityofNebraska Ihvalm Lincoln, NE i

l Master of Science degree in Agricultural Engineering, Dec.1989.

! (Emphasis of studies were la civil and machaaleal engineerlag).

j Other Training 4

l Completed Shit Technical Advisor training in academics and plant systems at the Kewaunee j Nuclear Power Plant (KNPP). Jan.1991 to Dec.1991.

i RELEVANT WORK EXPERIENCE i

j Wisenasin Public Service Corporation Creon Bay, WI l

i Nuclear Enginaar August 1990 to Present i

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Project management of the USI A-46 project at KNPP. Specific responsibilities include;

planning and scheduling of projaa teks, management and coordiawien of contracted T --ig work, performance of tarhaical evaluations, development of procedures, j Project budgeting, ad preparation of project reports.

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Seismic structural analyses of new and existing equipment installations.

Engineering support for plant design changes.

Applied Power Associates,Inc. Omaha, NE Anhitects, Engineers, Consultants l

Civil / Structural Enginaar May 1939 to August 1990 i

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Structural stasi and reinforced concrete design and analysis for a multi-story office j WM4 4

Consulting engineer to Nebraska Public Power District, Cooper Nuclear Station.

l Responsibilities included seismic analysis and design of equipment installations, provided j on-site engineering support to work crews during installation of new ductwork and j aanduir hanger lastallations at the Cooper plant.

Consulting engineer to Omaha Public Power Didrict, Fort Calhoun Nuclear Plant.

Performed power plant design change work related to building structural modifications.

Performed design and analysis work for modifications to a fly-ash bandling system for the Iowa Power and Light Company.

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WALTER DJORDJEVIC  ;

EDUCATION:

1 B.S. - Civil Engineering, University of Wisconsin at Madison,1974 M.S. - Structural Engineering, Massachusetts Institute of Technology,1976 REGISTRATION:

State of Califomia, State of Wisconsin, Commonwealth of Massachusetts, State of Michigan PROFESSIONAL HISTORY:

Stevenson & Associates,Inc.,Vice President and General Manager of the Boston area office, 1983 - present URS/ John A. Blume & Associates, Engineers, Boston, Massachusetts, General Manager,1980-1983; San Francisco, Califomia, Supervisory Engineer,1979 - 1980 Impell Corporation, San Francisco, Califomia, Senior Engineer,1976- 1979 r Stone & Webster Engineering Corporation, Boston, Massachusetts, Engineer,1974 - 1976 V]_

PROFESSIONAL EXPERIENCE:

Mr. Djordjevic founded the Stevenson & Associates Boston area office in 1983 and serves as Vice President and General Manager. He is currently preforming numerous seismic walkdowns forresolution of the USI A-46 and seismic lPEEE issues, and serving as the Project Manager for the Kewaunee, Point Beach and Palisades projects, alljoint A-46 and Seismic PRA projects.

Mr. Djordjevic is expert in the area of seismic fragiity analysis and dynamic qualification of electrical and mechanical equipment. He has participated in and managed over twenty major projects involving the evaluation and qualification of vibration sensitive equipment and seismic hardening of equipment. As demonstrated by his committee work and publications, Mr. Djordjevic has participated in and contributed steadily to the development of equipment qualification and vibration hardening methodology.

Mr. Djordjevic's previous walkdown experience included all of the SEP plants (8 plants), Nine Mile - Unit 1, D.C. Cook - Units 1 & 2, the Hanford Reservation Purex facility and the Savannah River Plant Reservation L-Reactor. He has personally participated in seismic walkdowns at 26 U.S. nuclear units.

Representative projects include overseeing the SEP shake-table testing of electrical raceways, in-situ testing of control panels and instrumentation racks at various nuclear facilities, equipment anchorage walkdowns and evaluations at various nuclear facilities, principal author of the CERTIVALVE software package to evaluate nuclear service valves, and contributing authorin the development of the ANCHOR

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C/ and EDASP software packages commercially distributed by Stevenson & Associates.

SA-

hd Mr. Djordjevic has been involved extensively in the reassessment of safety-related equipment for commercial nuclear facilities and govemment U.S. Department of Energy facilities, for which he maintains an active Q-clearance status. He has served on advisory groups and review teams touring older existing nuclear facilities to assess safety and has performed earthquake reconnaissance at such installations following seismic events.

PROFESSIONAL GROUPS:

Member, Institute of Electrical and Electronics Engineers, Nuclear Power Engineering Committee Working Group SC 2.5 (IEEE-344)

Chairman, American Society of Civil Engineers Nuclear Structures and Materials Committee, Working Group for the Analysis and Design of Electrical Cable Support Systems Member, American Society of Mechanical Engineers Operation, Application, and Components Commit-tee on Valves, Working Group SC-5 A

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(Q/ FRANK B. STILLE EDUCATION:

B.S.C.E. - Syracuse University - 1957 M.S.C.E. - Syracuse University - 1961 SQUG Walkdown Screening and Seismic Evaluation Training Course - 1993 REGISTRATION:

Commonwealth of Massachusetts PROFESSIONAL HISTORY:

Stevenson & Associates Inc., Wobum Massachusetts,1993-present Northern States Power Company / Prairie Island Nuclear Generating Plant, Nuclear Projects Department- Consulting Engineer, 1992 -1993

( Teledyne Engineering Services, Waltham Massachusetts - Engineering Services Manager,

( Project Manager, Civil and Mechanical Engineer,1974 - 1992 Aerospace Industry,1957 -1974 PROFESSIONAL EXPERIENCE:

Mr. Stille has nineteen years experience in the Civil / Structural / Mechanical Engineering of large nuclear power generation facilities. His experience covers: 1) direct performance of Civil and Mechanical Engineering; 2) project management of power plant design modification projects; and, 3) management of engineering, design and drafting services for an architect - engineering firm. His current work at Stevenson & Associates has included participation in several A-46 / IPEEE projects and the performance of a Q.A. verification of a general purpose finite element program.

He spent 15 months at a two unit nuclear power generation plant as a Civil / Mechanical engineering consultant. During this period he performed a second level review of an A/E's pipe stress and pipe support design modifications for a large intake Cooling Water Header replacement project as the utility representative. His responsibilities also included the provision of new support design, weld control records, and field support to construction. His services to the utility also included participation on the Auxiliary Building Crane Upgrade Project. He solicited bids and purchased fabricated steel forinstallation on a 125-ton crane runway girder modification. This work also included field support to construction.

SA-

As a civil / mechanical engineer he has performed design and analysis of pressure components, piping

% systems and structures including field support to construction on a broad variety of projects. This work included the use of SAP - based piping stress codes and STRUDL based structures programs.

As a project manager he was responsible for nuclear power plant design modification projects ranging in size from $100K to $6M. His responsibilities included administration, scheduling, cost control, technical direction, and client interface. Examples of some of the projects he has managed are:

NRC Bulletin 79-14 Program, Turkey Point - Piping and pipe support analysis, design modification and field support to construction for 90 percent of all safety- related systems on both units.

As - Built /ISI Drawing Program, Turkey Point - Creation of piping stress isometric and pipe support CAD drawings for ali safety- related piping systems on both units. These drawings contained information neccesary for implementation of the utility's Second Ten Year ISI Program.

Reactor Gas Vent Modification, Prairie Island - Piping and pipe support analysis, design modification, and field support to construction on Unit # 1 reactor.

MEMBERSHIP:

American Society of Mechanical Engineers - Member f~'\

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PINGP A-46 Seismic Evaluation Report November 20,1995 O Appendix D: Screening Verification Data Sheets (SVDS) 1 1

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(v) 11/9/951:44 PM Northern States Power Company . Prairie Island Nuclear Generating Plant Page # 1 SCREENING VERIFICATION DATA SHEET (SVDS)

Eq. Eq. ID Rev Sys'Eq. Desc Bk!g. FI EL Rm or Rw/Cl Base EL <40? Cap. Demd. Cap > Caveats Anchor interact Equip Cl No Spec. Spec Demd? OK7 OK7 OK OK7 1 MCC1K1 0 1EB / MOTOR CONTROL CENTER 1K BUS AUX 695.00 G.2/5.2 NEAR RHR PIT 695.00 N/A ABS CRS Yes Yes Yes No No 1

1 MCC 1K2 0 1EB / MOTOR CONTROL CENTER 1K BUS AUX 695 00 G.8/6.5 NEAR CHG 695.00 N/A MIS CRS Yes Yes Yes Yes Yes 2 PUMPS 1 MCC1LA1 0 1EB / MOTOR CONTROL CENTER ILA BUS AUX 735.00 J.2/5.2 NEAR PERS 735.00 N/A ABS CRS Yes Yes Yes Yes Yes 1 AIRLOCK 1 MCC 1R1 0 / MOTOR CONTROL CENTER 1R BUS 1 AUX 735.00 H.2/3.7 U1 ROD DRIVE 735.00 N/A ABS CRS Yes Yes Yes Yes Yes RM 1 MCC1S1 0 1EB / MOTOR CONTROL CENTER 1S BUS AUX 735.00 H.3/3.7 U1 ROD DRIVE 735.00 N/A ABS CRS Yes Yes Yes Yes Yes 1 RM 1 MCC 1T1 0 1EB / MOTOR CONTROL CENTER 1T BUS 1 AUX 755.00 G.4/8.1 121 CONT RM 75iOO N/A ABS CRS Yes Yes Yes Yes Yes CHLR RM 1 MCC 1T2 0 1EB / MOTOR CONTROL CENTER 1T BUS 2 AUX 755.00 G.4/10.0122 CONT RM 755.00 N/A ABS CRS Yes Yes Yes Yes Yes CHLR R 1 MCCITA1 0 1EB / MOTOR CONTROL CENTER 1TA BUS TURB 695.00 K.9/2.9 D1 DIESEL 695.00 N/A ABS CRS Yes No No Yes No 1 ROOM 1 MCC 1TA2 0 1EB / MOTOR CONTROL CENTER ITA BUS TURB 695.00 H.3/2.8 D2 DIESEL 695.00 N/A ABS CRS Yes No No Yes No 2 ROOM 1 MCC1X1 0 1EB / MOTOR CONTROL CENTER 1X BUS AUX 715.00 J.4/5.2 NEAR PENET 715.00 N/A ABS CRS Yes Yes Yes Yes Yes 1 CAB 1134 1 MCC1X2 0 1EB / MOTOR CONTROL CENTER 1X BUS AUX 715.00 J.4/6.1 NEAR 11 VCT 715.00 N/A ABS CRS Yes Yes Yes Yes Yes 2 ROOM 2 1-52/RTA 0 1RP / A - TRAIN REAC TRIP BREAKER AUX 735.00 U1 ROD DRIVE RM 735 00 Yes BS GRS Yes Yes Yes Yes Yes 2 BUS 111 0 1EB / BUS 111480V SWITCHGEAR TURB 715.00 E.0/9.3111 BUS ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 2 BUS 121 0 1EB / BUS 121480V SWITCHGEAR TURB 715.00 F.0/9.3121 BUS ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 3 BUS 15 0 / 4.16KV SFGDS BUS 15 TURB 715.00 715.00 Yes BS GRS Yes Yes Yes Yes Yes 3 BUS 16 0 / 4.16KV SFGDS BUS 16 TURB 715.00 F.5/8.716 BUS ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 4 1PZRHTRB/X 0 1EB /1 PRZR HTR GRP B TRANSFORMER AUX 735.00 H.1/3.9/735 AUX 735.00 Yes BS GRS Yes Yes Yes Yes Yes FMR Certification: Certification:

All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regant.ing systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our kno# edge and belief, correct and conclusion (whether venfied to be seismically adequate or not). accurate.

Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved. (One signature of Systems or Operations Engineer is required if the Seismic Capability an required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)

entries and conclusions. One signatory should be icensed prof sional engineer.)

l Greg Ridder l lIN T6l l l l Pnnt or Type Name Sgnh re ' Date Pnnt or Typ6 Name Sqnature Date l Walter Djordjevic l l [/ }l l l l Pnnt or Type Name Shnature /Datd Pnnt or Type Name Sgnature Date I i l i I I i Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date

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U J v 11/9/951:44 PM Northern States Power Company - Prairie Island Nuclear Generating Plant Page # 2 SCREENING VERIFICATION DATA SHEET (SVDS)

Eq. Eq.10 Rev Sys/Eq. Desc Bldg. FI EL Rm or Rw/Cl Base El. <40'? Cap. Demd. Cap > Caveats Anchor interact Equip Cl No Spec. Spec Demd7 OK7 OK7 OK OK7 5 045-591 0 /121 CONTROL ROOM CHILLED WATER AUX 755.00 G.6/8.7 755.00 N/A ABS CRS Yes Yes Yes Yes Yes PUMP

$ 045-592 0 /122 CONTROL ROOM CHILLED WATER AUX 755 00 G.6/9.3 755.00 N/A ABS CRS Yes Yes Yes Yes Yes PUMP 5 145-041 0 1VC /11 CHG PUMP AUX 695.00 H.5/6.6 695.00 Yes BS GRS Yes Yes Yes Yes Yes 5 145442 0 1VC /12 CHG PUMP AUX 695.00 H.5/7.0 695.00 Yes BS GRS Yes Yes Yes Yes Yes 6 045-271 0 /121 DSL GEN OIL STOR TK FUEL 695.00 LA7/0.3 695.00 N/A ABS CRS Yes No Yes Yes No SUBMERSIBLE PUMP 6 045-273 0 /123 DSL GEN O!L STOR TK FUFL 695.00 KA.5/0.3 695.00 N/A ABS CRS Yes No Yes Yes No SUBMERSIBLE PUMP 6 045-301 0 /121 DSL CLG WTR PMP OIL STOR TK SSCRN 695.00 C1/51.5 695.00 N/A ABS CRS Yes No Yes Yes No SUBMERSIBLE PMP 6 045-302 0 /122 DSL CLG WTR PMP OIL STOR TK SSCRN 695,00 B1.5/51.5 695.00 N/A ABS CRS Yes No Yes Yes No SUBMERSIBLE PMP 7 CV-31084 0 1MS /11 STM GEN MN STM SAF RLF TO AUX 736.00 735.00 Yes BS GRS Yes Yes N/A Yes Yes ATM CV 7 CV-31089 0 1MS /12 STM GEN MN STM SAF RLF TO AUX 756.00 IN 6 UNE J.3/5.6 755.00 N/A ABS CRS Yes Yes N/A Yes Yes ATM CV 7 CV-31098 0 1MS /11 LOOP A MN STM HDR ISOL CV AUX 726.40 735.00 Yes BS GRS Yes Yes N/A Yes Yes 7 CV-31099 0 1MS /12 LOOP B MN STM HDR ISOL CV AUX 739.00 IN 30 UNE J.2/5 8 755.00 N/A ABS CRS Yes Yes N/A Yes Yes 7 CV-31255 0 1RC /1 REAC CLNT LOOP PRZR LTDN LN CNTMT 705.00 IN 2 LINE 28/259 711.50 Yes BS GRS Yes Yes N/A Yes Yes ISOL LCV 2 7 CV-39401 0 1ZX /11/13 FCU CLG WTR SUPPLY CV AUX 704.00 IN 10 LINE J.0/6.0 715.00 Yes BS GRS Yes Yes N/A No No  ;

7 CV-39403 0 1ZX /12/14 FCU CLG WTR SUPPLY CV AUX 702.00 IN 10 LINE J.0/6.0 715.00 Yes BS GRS Yes Yes N/A Yes Yes I

7 CV-39409 0 1ZX /12/14 FCU CLG WTR RETURN CV AUX 704.00 IN 10 LINE J.0/7.0 715.00 Yes BS GRS Yes Yes N/A No No T CV-39411 0 1ZX /11/13 FCU CLG WTR RETURN CV AUX 704.00 IN 10 LINE J.0/7.0 7?5.00 Yes BS GRS Yes Yes N/A Yes Yes 8 CV-39201 0 1CL /11 & 13 FCU CLG WTR RTN B-P CV AUX 736.00 IN 10 LINE J.5/6.4 735.00 Yes BS GRS Yes Yes N/A Yes Yes 8 CV-39203 0 1CL /12 & 14 FCU CLG WTR RTN ORIF B-P AUX 720.00 IN 10 LINE J.5/6.0 735.00 Yes BS GRS Yes Yes N/A Yes Yes CV Certification: Certification:

All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations t

cur knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether ve'ified to be seismically adequate or not). accurate.

Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved:(One signature of Systems or Operations Engineer is required if the Seismic Capability t.re required; there should be atleast two on the SRT. All signa ries should agree with all the Engineers deem it necessary.)

entries and conclusions. One signatory should licensed p s io alengineer.)

l Greg Ridder l -

, l /b/3 -76} l' l l Pnnt or Type Name P S4 nature Dat Pnnt or Type Name Signature Date l Walter Djordjevic l l /'l l l l Pnnt or Type Name Sig'na( Date Pnnt or Type Name Sqnature Date l I I I I I i Pnnt or Type Name Signature Date Print or Type Name Sgnature Date

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Northern States Power Company - Prairie Island Nuclear Generating Plant Page # 3 11/9/951.44 PM SCREENING VERIFICATION DATA SHEET (SVDS)

Eq. Eq.ID Rev Sys/Eq. Desc Bldg. Fi El. Rm or Rw/Cl Base El. <40'? Cap. Demd. Cap > Caveats Anchor interact Equip No Spec. Spec Demd? OK7 OK7 OK OK?

CI 1VC / RFLG WTR EMERG MK-UP TO CHG AUX 699.00 IN 4 LINE H.9/6.9 715.00 Yes BS GRS Yes Yes N/A Yes Yes 8 MV-32060 0 PMPS MV 0 1VC /11 VOL CONT TNK TO CHG PMPS AUX 707.00 IN 4 LINE H.8/6.9 715.00 Yes BS GRS Yes Yes N/A Yes Yes 8 MV-32061 ISOL MV 0 1VC /1 REAC EXCS LTON LINE ISOL MV A AUX 720.00 IN 3 LINE L.5/6.8 735.00 Yes BS GRS Yes Yes N/A Yes Yes 8 MV-32166 8 MV-32199 0 1VC /1 REAC EXCS LTDN LINE ISOL MV B CNTMT 720.00 IN 3 LINE 1/273 733.75 Yes BS GRS Yes Yes N/A Yes Yes 8 SV-37035 0 1RC / RCS VENT SYS PRZR VENT SV CNTMT 760.00 IN 1 LINE 40/10 755.00 N/A DOC RRS Yes Yes N/A Yes Yes SV-37036 0 1RC / RCS VENT SYS PRZR VENT SV CNTMT 760.00 IN 1 LINE 40/10 755.00 N/A DOC RRS Yes Yes N/A Yes Yes 8

8 SV-37037 0 1RC / RCS VENT SYS REACTOR HEAD CNTMT 760.00 IN 1 LINE 40/10 755.00 N/A DOC RRS Yes Yes N/A Yes Yes VENT SV 8 SV-37038 0 1RC / RCS VENT SYS REACTOR HEAD CNTMT 760.00 IN 1 LINE 40/10 755.00 N/A DOC RRS Yes Yes N/A Yes Yes VENT SV 8 SV-37039 0 1RC / RCS VENT SYS TO PRT SV CNTMT 760.00 IN 1 LINE 40/10 755.00 N/A DOC RRS Yes Yes N/A Yes Yes 8 SV-37040 0 1RC / RCS VENT SYS TO CNTMT ATMOS CNTMT 760.00 IN 1 LINE 40/10 755.00 N/A DOC RRS Yes Yes N/A Yes Yes SV 10 076-021 0 /121 CONTROL ROOM AIR HANDLER AUX 755.00 G.5/8.5 755.00 N/A ABS CRS Yes Yes Yes Yes Yes 10 076-022 0 /122 CONTROL ROOM AIR HANDLER AUX 755.00 G.5/9 5 755.00 N/A ABS CRS Yes Yes Yes Yes Yes 10 174-011 0 ZC /11 CNTM FAN Coll UNIT CNTMT 711.00 20/50 711.50 N/A ABS CRS Yes Yes Yes Yes Yes 10 174-012 0 /12 CONTAINMENT FAN COIL UNIT CNTMT 711.00 30/90 711.50 N/A ABS CRS Yes Yes Yes Yes Yes 10 174-013 0 ZC /13 CNTM FAN COIL UNIT CNTMT 755.00 8/310 755.00 N/A ABS CRS No Yes Yes Yes No 10 174-014 0 /14 CONTAINMENT FAN COIL UNIT CNTMT 735.00 12/320 733.75 N/A ABS CRS Yes Yes Yes Yes Yes 10 CD-34049 0 OZG /121/122 DSL GEN RM OUTS AIR CD TURB 725.00 IN DUCT JA.5/1.0 735.00 Yes BS GRS Yes Yes Yes Yes Yes 10 CD-34072 0 1ZC /11 FCU DISCH TO CNTMT DOME CD CNTMT 737.00 IN DUCT 14/73 733.75 N/A ABS CRS Yes Yes Yes Yes Yes 10 CD-34073 0 1ZC /11 FCU NORM DISCH TO GAP & CNTMT 741.00 IN DUCT 13!69 733.75 Yes BS GRS Yes Yes Yes Yes Yes STRUCT CD 10 CD-34074 0 1ZC /12 FCU DISCH TO CiNTMT DOME CD CNTMT 741.00 IN DUCT 17/128 733.75 Yes BS GRS Yes Yes Yes Yes Yes 10 CD-34075 0 1ZC /12 FCU NORM DISCH TO GAP & CNTMT 741.00 IN DUCT 21/117 733.75 Yes BS GRS Yes Yes Yes Yes Yes STRUCT CD 11 075-011 0 /121 CONTROL ROOM WATER CHILLER AUX 755.00 G.7/8.0 755.00 N/A ABS CRS No No No Yes No Cert fication: Certification:

All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and belief, correct and accurate. "A!! information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, corred and conclusion (whether verified to be seismically adequate or not). accurate.

Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Se~smic r Capabihty are required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)

entries and conclusions. One signatory should icensed prof 'onal engineer.)

l Greg Ridder l /' _ l //-/$ Tfl l l l Pnnt or Type Name Egnafu Date Pnnt or Type Name Signature Date

1. M l Walter Djordjevic l ~

l l l l Pnnt or Type Name Signatdref /Dat6 Pnnt or Type Name Sqnature Date I I

/ I l I  ! l Pnnt or Type Name Sgnature Date Pnnt or Type Name Sqnature Date

. . .- = - - . -- - . , , . . . - - ,.

-i 11/9/95 1:44 PM ' Northern States Power Company - Prairie Island psuclear Generating Plant Page # 4 SCREENING VERIFICATION DATA SHEET (SVDS) - {

Eq. Eq.10 Rev Sys/Eq. Desc Bldg. Fi El. Rm or Rw/Cl Base El. <40"? Cap. Demd. Cap > Ca vests Anchor interact Equip i Cl No Spec. Spec Demd? OK7 OK7 OK- OK7 .;

11 075-012 0 /122 CONTROL ROOM WATER CHILLER AUX 755.00 G.7/10.0 755.00 N/A ABS CRS No No No Yes- No 12 046-031 0 1121 D1 DIESEL GENERATOR STARTUP AUX 695.00 D1 DIESEL 695.00 N/A ABS CRS Yes Yes Yes Yes Yes j AIR RECEIVER GENERATOR ROOM 12 046-032 0 I122 D2 DIESEL GENERATOR STARTUP AUX 695.00 D2 DIESEL 695.00 N/A ABS Yes Yes Yes Yes GENERATOR ROOM CRS ' Yes AIR RECEIVER 14 PNL 111 0 11P / INSTR BUS 11 PANEL (WHI) 111 TURB 715.00 G.1/8.2 715.00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 112 0 11P / INSTR BUS I PANEL (RED) 112 TURB 715.00 G.4/8.8 715.00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 113 0 11P / INSTR BUS Ill PANEL (BLUE) 113 1 URB 715 00 G.1/8.0 715.00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 134 0 EX / AC DISTRIBUTION PANEL 134 AUX 695.00 G.1Ms.4 695 00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 15 0 1DC / NUCLEAR DISTRIBUTION PANEL 15 TURB 715.00 G.0/8.3 715.00 Yes BS GRS Yes Yes - Yes Yes Yes 14 PNL 151 0 DC / DISTRIBUTION PANEL 151 AUX 715.00 715.00 Yes BS GRS Yes Yes Yes Yes Yes

, 14 PNL 16 0 1DC / NUCLEAR DISTRIBUTION PANEL 16 TURB 715.00 G.5/8.5 715.00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 161 0 DC / DC DISTRIBUTION PANEL 161 AUX 715.00 715.00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 162 0 DC / DC DISTRIBUTION PANEL 162 AUX 715.00 715.00 Yes BS GRS Yes Yes Yes Yes Yes 14 PNL 191 0 1DC 1 DC DISTRIBUTION PANEL 191 AUX 715.00 J.3/4.1 SW SIDE OF 715.00 Yes BS GRS Yes Yes Yes Yes Yes

- RWST I 14 PNL1EMA 0 1EM / DISTRIBUTION PANEL 1EMA TURB 735.00 H.3/5.3 TRN A EVENT 735.00 Yes BS GRS Yes Yes Yes Yes Yes MON ROO 14 PNL1EMB 0 1EM i DISTRiDUTION PANEL 1EMB TURB 735.00 H.0/12.0 TRN B EVENT 735.00 Yes BS GRS ~Yes Yes Yes Yes Yes MON RO 17 034-011 0 D11121 D1 DIESEL GENERATOR TURB 695.00 K.5/2.4 695.00 Yes BS GRS Yes No Yes .Yes No

17 034-021 0 D2 / D2 DIESEL GENERATOR TURB 695.00 H.3/2.4 695.00 Yes BS GRS Yes No Yes Yes No 18 18036 0 1AF / AUX FW TO 11 STM GEN F1 AUX 700.00 ON W SIDE WALL 695.00 Yes SS GRS Yes Yes Yes Yes Yes GRID H.0/8 l 18 18038 0 1AF / AUX FW TO 12 STM GEN Fl AUX 700.00 ON W SIDE WALL 695.00 Yes BS GRS Yes Yes Yes Yes Yes

)' GRlD H.0/8 f 18 1F1-1158 0 1VCi 11 REAC CLNT PMP SL WTR INJ FI AUX 720.00 ON W SIDli WALL 735.00 Yes BS GRS Yes Yes Yes Yes Yes L5/7.0 Certification: Cettdicahon AM the information contaMed on this Screening Verificabon Data Sheet (SVDS) is, to the best of The informahon provided to the Seesmic Capatuhty Engineers regarding systems and operabons our knowledge and beleef, correct and accurate. *AN informabon" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seesmecaNy adequate or not). accurate.

Approved: (Signatures of as Seesmic Capatsty Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Opersbons Engineer is required if the Seestmc Capatety tre required; there should be attesst two on the SRT. AR signatories should agree with as the Engineers deem it necessary.)

entries and conclusions. One signatory should be licensed al engineer.)

l Greg Ridder l l //D6l l l l I Pnnt or Type Name Sal _

D Pnnt or Type Name Signature Date

/

l Walter Djordjevic l l l l l l Pnnt or Type Name Sogn Date . Pnnt or Type Name Signature . Date

, I I I I I I I

< Pnnt or Type Name Segnature Date Print or Type Name Segnature Dale

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I 11/9/951:44 PM Northern States Power Company - Prairie Island Nuclear Generating Plant Page#5 SCREENING VERIFICATION DATA SHEET (SVDS)

Eq. Eq.10 Rev Sys/Eq. Desc Bldg. FI El. Rm or Rw/Cl Base El. <40"? Cap. Demd. Cap > Caveats Anchor interact Equip Cl No Spec. Spec Demd? OK? OK7 OK OK7 18 1FI-1168 0 1VC /12 REAC CLNT PMP SL WTR INJ F1 AUX 720.00 ON W SIDE WALL 735.00 Yes BS GRS Yes Yes Yes Yes Yes L.5/7.0 18 ILT-426 0 1RP /1 REAC CLNT LOOP PRZR (CHNNL I- CNTMT 720.00 ON N SIDE WALL 11/16 733.75 Yes BS GRS Yes Yes Yes Yes Yes RED) LVL XMTR 18 1LT-428 0 1RP /1 REAC CLNT LOOP PRZR (CHNNL CNTMT 720.00 ON E SIDE WALL 12/30 733.75 Yes BS GRS Yes Yes Yes Yes Yes til-BLU) LVL XMTR 18 1LT-487 0 1EM /11 STM GEN LOOP A WR LVL XMTR CNTMT 716.00 ON SHLD WALL 6/192 733.75 Yes BS GRS Yes Yes Yes Yes Yes 18 1LT-488 0 1EM /12 STM GEN LOOP B WR LVL XMTR CNTMT 716.00 ON SHLD WALL 18/337 733.75 Yes BS GRS Yes Yes Yes Yes Yes 18 ILT-751 0 1EM /11 RX VSL HEAD UPPER RNG TRN A AUX 735.00 ON INSTR RACK GRID 735.00 Yes BS GRS Yes Yes Yes Yes Yes D/P XMTR J.5/ 4.2 18 ILT-753 0 1EM /11 RX VSL HEAD DYNAMIC RNG TRN AUX 735.00 ON INSTR RACK GRID 735.00 Yes BS GRS Yes Yes Yes Yes Yes A D/P XMTR J.5/ 4.2 18 1LT-761 0 1EM /12 RX VSL HEAD UPPER RNG TRN B AUX 135.00 ON INSTR RACK GRID 735.00 Yes BS GRS Yes Yes Yes Yes Yes D/P XMTR J.5/ 4.5 18 1LT-763 0 1EM /12 RX VSL HEAD DYNAMIC RNG TRN AUX 735.00 ON ISNTR RACK GRID 735.00 Yes BS GRS Yes Yes Yes Yes Yes B D/P XMTR J.5/ 4.5 18 1LT-920 0 1EM /11 RWST LVL XMTR AUX 700.00 ON E SIDE WALL 715.00 Yes BS GRS Yes Yes Yes Yes Yes J.3/4.3 18 1LT-921 0 1EM /11 RWST LVL XMTR AUX 700.00 ON E SIDE WALL 715.00 Yes BS GRS Yes Yes Yes Yes Yes J.3/4.3 18 1PT-429 0 1RP /1 REAC CLNT LOOP PRZR (CHNNL l- CNTMT 720.00 ON N SIDE WALL 11/16 733.75 Yes BS GRS Yes Yes Yes Yes Yes RED) P XMTR 18 1PT-431 0 1RP /1 REAC CLNT LOOP PRZR (CHNNL CNTMT 720.00 ON E SIDE WALL 12/30 733.75 Yes BS GRS Yes Yes Yes Yes Yes ill-BLU) P XMTR 18 1PT-468 0 1MS /11 STM GEN LOOP A (CHNNL l-RED) AUX 720.00 ON NORTH SIDE 735.00 Yes BS GRS Yes Yes Yes Yes Yes P XMTR WALL P.0/8 0 18 1PT-478 0 1MS /12 STM GEN LOOP B (CHNNL lil-BLU) AUX 720.00 ON EAST SIDE COL 735.00 Yes BS GRS Yes Yes Yes Yes Yes P XMTR J.1/5.9 19 1TE-451 A 0 1EM /1 REAC CLNT LOOP B HOT LEG RTD CNTMT 723 00 IN 29 LINE 35/352 733.75 Yes BS GRS Yes Yes N/A Yes Yes Certification: Certification:

All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not). accurate.

Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)

entries and conclusions. One signatory should icensed profes i engineer.)

l Greg Ridder l V

^

l #Ni8 l Sqnature l

Date l

Pnnt or Type Name gn yr Dat Pnnt or Type Name l Walter Djordjevic l l [ [_f l l l Pnnt or Type Name Sgn ture Date Pnnt or Type Name Sqnature Date l I I I I I I Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date

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1 V V Northern States Power Company - Prairie Island Nuclear Generating Plant Page # 6 11/9/951:44 PM SCREENING VERIFICATION DATA SHEET (SVDS)

Eq. Eq.10 Rev Sys/Eq. Desc Bldg. FI El. Rm or Rw/C4 Base El. <40'? Cap. Demd.8 Cap > Caveats Anchor interact Equip No Spec. Spec Demd? OK? OK? OK OK?

Cl AUX 735.00 Control Room 735 00 Yes BS GRS Yes Yes Yes No No 20 14MR 0 1MP /14 MISCELLANEOUS RELAY RACK 1AMR1 0 1RP / MISCELLANEOUS RELAY RACK AUX 715.00 RELAY ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1AMR1 0 1RP / SAFEGUARD RELAY RACK 1 ASG1 AUX 715.00 RELAY ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1ASG1 0 1RP / SAFCGUARD RELAY RACK 1ASG2 AUX 715.00 RELAY ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1ASG2 181 0 1RP / PROCESS PROTECTION RACK 181 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes Yes Yes 20 182 0 1RP / PROCESS PROTECTION RACK 182 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes Yes Yes 20 0 1RP / SAFEGUARD RELAY RACK 1BSG1 AUX 715.00 RELAY ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1BSG1 1BSG2 0 1RP / SAFEGUARD RELAY RACK 1BSG2 AUX 715.00 RELAY ROOM 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 20 1NR3 0 1NI / NUCLEAR INSTRUMENTATION RACK AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No 1NR3 20 1NR4 0 1NI / NUCLEAR INSTRUMENTATION RACK AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No 1NR4 20 1PLP O 1BM / PROCESS CONTROL RACK 1PLP AUX 735 00 Control Room 735.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1R1 0 1RP / PROCESS PROTECTION RACK 1R1 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1R2 0 1RP / PROCESS PROTECTION RACK 1R2 AUX 735 00 Control Room 735.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1RCS1 0 1RC / PROCESS CONTROL RACK 1RCS1 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes Yes Yes 20 1RCS2 0 1RC / PROCESS CONTROL RACK 1RCS2 AUX 735.00 Control Room 735 00 Yes BS GRS Yes Yes Yes Yes Yes 20 A 0 CMP / CONTROL PANEL A AUX 735 00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No 20 B-1 0 1MP / CONTROL PANEL B-1 AUX 735.00 Control Room 735 00 Yes BS GRS Yes Yes Yes No No 20 BIS LOGIC-1 0 2EA / BUS 15 LOGIC CAB 1 TURB 715.00 E.3/8.0 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 B15 LOGIC-2 0 2EA / BUS 15 LOGIC CAB 2 TURB 715.00 E.3/8.0 715.00 Yes BS GRS Yes Yes Yes Yes Yes 20 B15/ LOAD 0 EA / BUS 15 SAFEGUARDS LOAD TURB 715.00 715.00 Yes BS GRS Yes Yes Yes Yes Yes SEO CAB SEQUENCER CABINET 20 C-1 0 1MP / CONTROL PANEL C-1 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No 20 0-1 0 1MP / CONTROL PANEL D-1 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No 20 E-1 0 1MP / CONTROL PANEL E-1 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No 20 EM-A1 0 1EM / EVENT MONITORING RACK EM-At AUX 735.00 120 BUS RM 735.00 Yes BS GRS Yes Yes Yes Yes Yes Certification: Certification:

AH the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not). accurate.

Approved: (Signatures of all Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved:(One signature of Systems or Operations Engineer is required if the Seismic Capability tre required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)

entries and conclusions. One signatory should licensed pro sionalengineer.)

l Greg Ridder l l// M.3-7.Sl l l l Pnnt or Type Name Sdna re Date Pnnt or Type Name Srgnature Data l Watter Djordjevic l l N/ l l l l Pnnt or Type Name Sdnat e /Date Pnnt or Type Name Sgnature Date 1 I I I I I l Pnnt or Type Name Sgnature Date Pnnt or Type Name Signature Date

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11/9/951:44 PM Northern States Power C-:- _ _ ., - Prairie Island Nuclear Generating Plant ' Page # 7 SCREENING VERIFICATION DATA SHEET (SVDS)

Eq. Eq.10 Rev Sys/Eq. Desc Badg. FI El. Rm or RwC,1 Base El. <40*? Cap. Demd. Cap > Caveats Anchor interact Equip j Cl No Spec. Spec Demd? OK? OK? OK OK7  !

g 20 EM-A2 0 2EM t EVENT MONITORING RACK EM-A2 AUX 735.00 120 BUR RM 735.00 Yes BS GRS Yes Yes -Yes Yes Yes -[

i^ AUX 735.00 120 BUS RM 735.00 Yes BS GRS Yes- Yes Yes Yes Yes 20 EM-A3 0 OEM t EVENT MONITORING RACK EM-A3 - l 20 EM-81 0 1EM t EVENT MONITORING RACK EM-B1 AUX 735.00 220 BUS RM 735.00 Yes BS GRS Yes Yes Yes Yes Yes

• 20 EM-82 0 2EM / EVENT MONITORING RACK EM-82 AUX 735.00 220 BUS RM 735.00 Yes BS GRS Yes Yes Yes Yes Yes  ;

20 EM-83 0 OEM / EVENT MONITORING RACK EM-83 AUX 735 00 220 BUS RM 735.00 Yes BS GRS Yes Yes Yes Yes Yes -t 20 G-1 0 OMP 1 CONTROL PANEL G-1 AUX 735.00 Control Room 735.00 Yes BS GRS Yes Yes Yes No No j 2

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I Certificaten. CertrReabon.

AM the information contained on this Screening Venficatto i Data Sheet (SVDS) is, to the best of The informabon provided to the Sersme Capatdaty Engineers regarding systerrs and operations our knowledge and belief, correct and accurate. "AE information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seismically adequate or not), accurate,