Text

Forwards Response to NRC 981109 RAI Re Resolution of USI A-46 for Pbaps.Proprietary Excerpts from GIP-2,Ref 25 Results of BWR Trial Plant Review Section 8 Also Encl. Proprietary Excerpts Withheld
	ML20217C414
Person / Time
Site:	Peach Bottom
Issue date:	10/06/1999
From:	Hutton J; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20138F988	List: ML20217C414;
References
	REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR NUDOCS 9910130255
	Download: ML20217C414 (48)
	v • d • e

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v PECO: NUCLEAR acco emerav comae"v 965 Chesterbrook Boulevard

' A Unit of PECO Energy--

Wayne.PA 19087-5691-l October 6,1999 1

l Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 U.S. Nuclear Regulatory Commission Attn: Document Control Desk -

L Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station, Units 2 and 3

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Response to Request for Additional Information Regarding

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Unresolved Safety issue (USl) A-46, Seismic Adequacy of Mechanical and Electrical Equipment j

References:

(1) Letter from Mohan C. Thadani, U.S. Nuclear Regulatory Commission f(USNRC), to Garrett D. Edwards, PECO Energy Company (PECO Energy) dated November 9,1998.

(2) Letter from Garrett D. Edwards, PECO Energy, to USNRC dated December 29,1998..

~

Dear Sir:

The reference (1) letter requested additionalinformation regarding the resolution of USI A-46 for Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3. Our reference (2) i' letier deferred responding until methodology issues are resolved between the USNRC and the Seismic Qualification Utility Group. provides a restatement of the questions followed by our response.

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If you have any questions, please do not hesitate to contact us.

. Very truly yours, ll}

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13p021 ll mes A. Hutton irector-Licensing L

Enclosure f

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H.1 Miller, Administrator, Region 1, USNRC fl025 i

A. C.' McMurtray, USNRC Senior Resident Inspector, PBAPS PDR. ADOCK 05000277 h

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ATTACHMENT 1 PEACH BOTTOM ATCMIC POWER STATION UNITS 2 AND 3 DOCKET NOS. 50-277 -

50-278 LICENSE NOS. DPR-44 PPR-56

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Docket Nos 50-277 50-278 j

License Nos. DPR-44 DPR-56 Response to Reauest for Additional Information Peach Bottom Atomic Power Station. Units 2 and 3 Unresolved Safety issue A-46 Question No.1 In regard to the in-structure response spectra (IRS) that was used to resolve the relay outliers for host cabinets located in the radwaste/ turbine building at the 150-foot floor c!svation, Reference 1 indicated the use of clipping of spectra in accordance with Appendix Q of EPRI NP-6041. As stated in the staff's previous request for additionalinformation (RAl) dated June 5,1997, the methodologies described in EPRI NP 6041 Report have not received approval from the staff for the analysis of safety-related systems and components, including the resolution of USl A-46 mechanical, electrical, and structural component outliers. The clipping of spectra leads to a reduction of the amplified spectrum at the elevation of interest and is, therefore, not acceptable to the staff. You are requested to reevaluate the relay outlier resolutions that have involved the use of this methodology.

' Response No.1 In response to your RAI dated June 5,1997 (Reference 1.1), a total of 13 relay outliers were

. resolved using clipped IRS for the host cabinets. All 13 relays are Agastat Model ETR14D non-energized relays with the contacts normally open. The generic equipment ruggedness spectra (GERS) for these relays were obtained from Reference 1.2, which used the lowest GERS for

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either the normally open or normally c^osed contacts.

l PECO Energy has now performed an evaluation to resolve the relay outliers without the use of

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clipping of the IRS for the host cabinets located in the Radwaste/ Turbine building at the 150-foot floor elevation.

Further review of the latest industry technical literature / testing results (Reference 1.3) shows that the subject Agastat relays have peak and zero period acceleration (ZPA) seismic capacities of 9.0g and 5.4g respectively for the normally open contacts; The unclipped peak and ZPA seismic demands are 4.66g and 1.699 respectively. Therefore, based on the latest relay testing information, the subject Agastat reley outliers meet the capacity versus demand screen and are now considered resolved without using the peak clipping methodology described in EPRI NP-6041.

REFERENCES:

1.1 Letter to NRC Document Control Desk, from G. A. Hunger, Jr., PECO Energy Company,

" Peach Bottom Atomic Power Station. Units 2 and 3, Request for Additional Information Regarding Generic Letter 87-02 on the Resolution of Unresolved Safety issue A-46,"

dated September 3,1997.

1.2 EPRI Report NP-7147-SL, Projects 1707-15,2925-2, Final Report, August 1991, Seismic Ruggedness of Relays

i Paga 2 of 26

, n' L1.3 l EPRI Report N_P-7147-SL, V2, Addendum 2, Project 2925-2, Final Report, April 1995,

, Seismic Ruggedness of Relays J Question No. 2 In regard to the adequacy of seismic demand determination, it appears that there are several

buildings (emergency cooling tower, diesel generator building and circulating water pumping structure) where for elevations less than 40 feet above the ground level, the corresponding IRS appears to be higher than 1.5 times the GIP-2 Bounding Spectra, and, hence, is higher than 1.5 times the ground response spectrum (since, as indicated in Reference 1, the ground response spectrum is completely enveloped by the Bounding Spectrum)E You stated in Reference I that Method A of GIP-2 was used to verify the seismic adequacy of the equipment in these buildings.' We note, however, that if the IRS for elevations less than about 40 feet above ground levelis greater than 1.5 times the ground response spectrum, then the use of Method A being not consistent with page 4-16 of GIP-2, under " Adequacy and Limitations," which states, "The irestrictlons and limitations on use'of the ground response spectrum for comparison to the Bounding Spectrum and the GERS (generic equipment ruggedness spectra) are based on the condition that the amplification factor between the free-field response spectrum and the ~in-structure response spectrum will not be more than 1.5,..." You are requested to provide a technical justification for the use of Method A for the above buildings where the IRS at i

elevations less than about 40 feet above grade is greater than 1.5 times the ground response spectrum.

Resoonse No.2

The approach that we used for applying and implementing GIP Method A for the estimation of the seismic demand on equipment at PBAPS for resolution of the USI A-46 program is -

appropriate and technicallyjustified. PECO Energy's interpretation of the GIP-2 rules for the use of Method A, which it, consistent with SQUG's interpretation,' is correct and the

-implementation of the method was proper. The bases for this are provided in Attachment 2.

The PBAPS IRS for the ' elevations where Method A was used exceed 1.5 times the ground response spectrum; however these spectra were generated using conservative methods and

assumptions (typical of most nuclear plant response analyses) which artificially increase the amplifications over those which would be expected in an actual earthquake.

The following is a detailed description of the typical conservatisms normally found in the

analytical methods used for calculating IRS, at nuclear plants in general and at PBAPS in particular. Please note that PECO Energy does not consider providing this additional supporting information as a necessary requirement for the application of the GIP Method A. It is provided as additional evidence of the validity of the use of Method A as originally developed by the iSenior. Seismic Review and Advlsory Panel (SSRAP).

_QalguiniggLQgalgDjgsis IRS Compared to 1.5 times the GIP Boundina Saactra

<-The equipment located in the following areas have used Method A and have calculated design

- basis IRS greater than 1.5 times the GIP-2 Bounding Spectra:

Diesel Generator Building El 127'

-Diesel Generator Building El 151' Circulating Water Pump Structure El 116' E ul

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Paga 3 of 26 The following graphs show the relationship between the design response spectra and 1.5 times the GIP-2 bounding spectra.

I Diesel Generator Building El 127' 5% damping i

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Pags 4 of 26 Circulating Water Pump Structure El 116' 5% damping

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The following table shows the maximum exceedances above the 1.5 Bounding Spectra for

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frequencies 8 Hz and above for the various locations shown above:

Location Direction Freq.

Design 1.5 Bounding (Design Spectra) /

Spectra Spectra (1.5 Bounding Spectra)

D.G. Building El 127' E-W 12 Hz 0.90g 0.80g 1.13 D.G. Building El 151' N-S 8 Hz 2.26g 1.13g 2.00 D.G. Building El 151' E-W

.12 Hz 1.37g 0.80g 1.71 Circulating Water Pump N-S 16 Hz 1.15g 0.689 1.69 Structure El 116' Circulating Water Pump E-W 20 Hz 0.69 g 0.599 1.17 Structure El 116' Conservatism in Calculated IRS The process cf calculating IRS is a complicated analytical exercise requiring a significant number of approximations, modeling assumptions and engineering judgments. As a result, the historical development of these IRS has included a tremendous amount of conservatism which l

has typically served two purposes.

I

1. It has reduced the technical debate as to the correct modeling of the many parameters which are intrinsic to the IRS calculational methodology, and;

2. It has reduced the costs associated with a very detailed state-of-the-art analysis, I

(which would attempt to trim out all the unnecessary conservatisms).

a 1

Paga 5 of 26

' L As a part of the A-46 program resolution methodology, the SSRAP developed and SQUG subsequently endorsed an altemate IRS estimation technique (referred to as Method A within the GlP-2) which was much more median-centered and realistic than the typical design.

practice. PECO Energy's application of Method A at PBAPS was appropriate and technically justified The fact that design _lRS may show amplifications greater than 1.5 is not surprising,

nor does it negate the validity of Method A. In fact,'as noted in the SSRAP report it was even

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expected: '

~

- Secondly, most unbroadened computed in-structure spectra have very narrow, L highly amplified peaks at the resonant frequency of the structure. In most cases -

these narrow, highly amplified peaks are artificially broadened to account for uncertainty in the structure's natural frequency. This process simply increases the emphasis on these highly amplified peaks.... SSRAP is also of the opinion

that these narrow peaks will not be as highly amplified in real structures at high l ground motion levels as is predicted by linear elastic mathematical models, nor are such narrow peaked in-structure spectra likely to be as damaging to equipment as is a broad frequency input which is represented by 1.5 times the Bounding Spectrum."

~ (Ref. 2.1, pages 19 and 20.]

As described below, three areas are presented to support the application of Method A at U.S.

nuclear plants in general, and at PBAPS in particular:

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' Measurements of IRS in Actual Earthquakes :

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J Calculations of Overall Conservatisms in Typical 1RS -

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Description _of the Conservatisms in IRS in General and PBAPS IRS in Particular -

1 Measurements of IRS in Actual Earthquakes ;

SSRAP developed the Method A response estimation technique based on their research of both actual earthquake measurements and on recent " median centered" analysis. They reference

" (Ref. 2.1 page 102) the measured floor response spectra at elevations less than 40 feet above the grade for moderately stiff structures at the Pleasant Valley Pump Station, the Humbolt Bay Nuclear Power Plant, and the Fukushima Nuclear Power Plant where arnplifications over the c ground response spectra do not exceed 1.5 for frequencies above about 6 Hz. Other, more

'recent earthquake data from the Manzanillo Power Plant and SICARTSA Steel Mill in Mexico, as

well as several facilities in Califomia and Japan, has been recently reviewed by SQUG. This data also shows that stiff buildings (similar to typical nuclear structures) amplify very little at elevations less than 40 feet above grade _and frequencies over 8 Hz. SQUG knows of no new measured data that challenges GlP Method A.

Lil : Calculations of Overall Conservatism in Typical IRS -

1 Calculated IRS have never been portrayed as representing the realistic expected response during an actual earthquake. As previously stated, IRS typically contain many conservatisms which make them unrealistically high. The primary reason for the development of Methoa A was to establish a more median-centered method of defining the structural response without having to' embark on costly new analyses of all the site buildings. (It should be noted that even I

r the most modern,' state 4f-the-art IRS contain significant conservatisms; even those classified as " median-centered", are often very conservative.) NUREG/CR-1489, "Best Estimate Method vs Evaluation Method: A Comparison of Two Techniques in Evaluating Seismic Analysis and

e yJ Pags 6 of 26 x

' Design", stated that typical calculated IRS contain factors of conservatism of 1.5 to 8. Recent i surveys by SQUG show similar levels of conservatism in calculated IRS.

< ;lt was the contention of SSRAP that the IRS for nuclear structures (considering the 40' and 8

[Hz conditions) would be within about 1.5 times the ground response spectrum (GRS) if the Lplant were subjected to an actual earthquake. In deriving the Method A criteria they recognized s

that due to the variety of ground motions, soil characteristics and structure characteristics there could be some possibility.of exceedances to the 1.5 amplification, but still strongly justified Method A's applicability;

?lt Is SSRAP's firm opinion that the issue of potential amplifications greater than 1.5 above about 8 Hz for high frequency input is of no consequence for the classes of equipment considered in this document except possibly for relay chatter'." '

[Ref. 2.1, Page 106]

The. basis SSRAP gave'foh drawing this conclusion was that high frequency ground motions do not have much damage potential due to low spectral displacement, low energy content, and short duration; They further noted that the equipment covered does not appear to have a -

significant sensitivity to high frequencies (except possibly for relay chatter, which is addressed separately in the GIP).

'lli Description of Conservatisms in IRS in General and PBAPS 1RS in Particular LThe most significant sources of conservatism involved in the development of IRS for PBAPS

.-include the following:

- Location of Input Motion (variation from the free field input location) e

- e : 1 Ground Response Spectrum Shape Soil-Structura interaction (Soil Damping, Wave Scattering Effects)

' Ground Motion incoherence J.-

e Frequency (Structure Modeling)'

Structural Damping e:

. Time History Simulation

. Non-Linear Behavior (e.g., soil property profile variation, concrete cracking)

Peak Broadening and Enveloping 1

e

. The degree of conservatism involved in each of these parameters is specific to the building being analyzed, to the floor level being considered, and often, to the equipment location within the specified floorlevel; These conservatisms typically cannot be accurately' quantified using simplistic calculational techniques since each parameter fits into an overall set of highly nonlinear equations.f Thus, it would take a considerable effort to quantify the excess conservatisms inherent in the calculated IRS at PBAPS. However, on the qualitative ievel J presented below,'it is easy to see the origins and levels of this conservatism. The following l discussions provide the details regarding the inherent conservatisms.

- T Because of the. SSRAP concern related to possible relay chatter at frequencies above 8 Hz, the SQUG methodology specifically addresses relays which are sensitive to high

. frequency vibration. Such relays are included on the Low Ruggedness Relays list in Appendix E of EPRI Report NP-7148.~

4 s

i, Pags 7 of 26 Location 'of input Motion The defined location of the plant SSE is typically part of the design basis documentation. The SSE should typically be defined at the ground surface

_ in the free field. The PBAPS site is a rock site. With the exception of the Diesel Generator Building, the structures were founded directly on an excavated rock surface.

The Diesel Generator Building is partially founded on piles. The lateral loads (including seismic) are resisted by shear walls which carry lateral loads to the rock. The piles only

- support gravity loads. No credit was taken for the influence of the piles on the lateral characteristics of the structure or equipment response.

The defined location of the PBAPS SSE is at the ground (rock) surface.' But for purposes oi generating IRS, some plants, including PBAPS conservatively defined the input (currently identified as the " control point" location) at another location, such as the embedded depth of a building basemat. This conservatism can be significant i

depending on the specific plant / building configuration. However, since the PBAPS structures are founded on rock, this level of conservatism is not very significant.

Ground Response Spectrum She The SSE defined.within the plant-licensing basis i

is the appropriate review level for the A-46 program. Some utilities utilized alternative (conservative) spectral shapes for the earthquake levels utilized for their A-46 resolution (i.e.,' submitted as part of their 120-day response letters). The amount of conservatism is directly related to the difference between these two spectral shapes at the frequencies ofinterest for the structures being reviewed. This factor can range from 1.0 to around 2.0 depending on the differences between the spectra.

The licensing basis safe shutdown earthquake for PBAPS is characterized by a site-specific Housner-type shape for the horizontal ground response spectrum anchored to a PGA of 0.12g. This type of spectrum is considered reasonable for Eastern U.S.

locations. Since this same site spectrum was used for the USl A-46 program, the GRS shape did not contribute much conservatism to the original licensing basis IRS at PBAPS.

' Soll Structure interaction (SSI) Typical design analyses do not account properly for the phenomenon of SSI, including the deamplification with depth that really occurs for l

embedded structures and for the radiation damping effects inherent at soil sites. Fixed-base analyses have been performed in typical design analyses, both for structures founded on rock and for structures founded on soil columns. Fcr rock foundations, the fixed-base model has been shown to be slightly conservative depending on the rock / structure characteristics. For soil founded structures, this assumption can vary between conservative and very conservative, depending on the level of sophistication of the modeling of the soil-structure system. The simplified analyses that used frequency-independent soi! springs were typically very conservative in that radiation and/or material damping were either conservatively eliminated or artificially limited during the analysis.

. Soil properties were also not adjusted to reflect anticipated soil strain levels. Significant reductions have been demonstrated over design type analyses using more modern techniques. These reduction factors are highly dependent on the specific soil conditions and structure configuration, but values of around 2 to 4 have been seen in past studies.

The PBAPS structures are founded on rock; thus, the amount of SSI conservatism inherent in the foundation input will be small (but not insignificant). The PBAPS seismic models are all based on stick models where any lateral restraining provided by the embedment is neglected. The offect of using a fixed-base model, instead of more accurate SSI analyses, was estimated in EPRI NP-6041 (Ref. 2.2, p.2-48) to have the equivalent effect of using 7% structural damping instead of 10% (i.e. a more accurate h

Pags 8 of 26 representation of the response would have been to use 10% damping in the analysis instead of the standard 7% value). This would have the effect of about a 16% (1 -

(.07/.10) ) reduction in the seismic response.

Furthermore, the modeling techniques used at PBAPS do result in some additional conservatisms, which are not easily quantifiable, when ignoring the restraining effects of the embedment against the surrounding backfill. The seismic modeling is especially conservative for the PBAPS Diesel Generator Building whose base elevation is at El 127'.0". The slab at Elevation 127'is built on compacted backfill, with piles and concrete shear walls extending 15 to 20 ft. below the base slab to solid rock. The seismic model of the Diesel Generator Building assumes that the horizontal seismic loads are transferred to the rock through the concrete shear walls alone, with no contribution from the lateral restraint of the compacted backfill on both sides of the shear walls. This simplified modeling results in large amplification factors between the ground response spectrum and the in-structure response spectra.

Ground Motion incoherence As has been documented in the EPRI seismic margin report (EPRI NP-6041) there can be a deamplification effect on nuclear type structures due to the incoherence of ground motion over the relatively large dimensions of typical nuclear structures. Conservative reduction factors as a function of frequency and building footprint have been documented within NP-6041 to account for the statistical incoherence of the input wave motion. These conservative values range from a factor of 1

1.1 to around 1.5. More recent studies have documented even greater reduction factors. This ground motion incoherence is applicable to rock sites like PBAPS and is particularly appropriate in the high frequency range.

I Structural Dampina Structural damping is one of the parameters of dynamic analysis I

to which the seismic analysis results are quite sensitive. It is a physical property of the different materials included in the dynamic model. Values used in current analyses and licensing bases are controlled by Regulatory Guide 1.61 (R.G.1.61). Values specified in R.G.1.61 have been shown by several studies to underestimate actual response of steel and concrete structures. Damping values recommended in NP-6041 (Ref. 2.2) are more realistic, and are suggested for use in median centered analyses. Damping values specified in PBAPS licensing basis are compared to those in R.G.1.61 and NP-6041 below:

Structure or R.G.

R.G.

NP-6041 at NP-6041 PBAPS OBE PBAPS SSE Component 1.61 1.61 about 1/2 Beyond or just Licensed Basis Licensed Basis OBE SSE Yield below Yield Welded Steel 2%

4%

3%

7%

1%

2%

Bolted Steel 4%

7%

10%

2%

5%

Reinforced 4%

7%

5%

10%

2%

5%

Concrete As can be seen, the damping values for both the licensing basis OBE and SSE at l

PBAPS are much lower than values allowed by the regulatory guide for bolted steel structures and reinforced concrete structures, which are known to already be conservative as indicated in NP-6041. These two types of structures encompass all of the structures at PBAPS housing the A-46 SSEL components.

Pag 3 9 of 26 A second area of conservatism, which constitutes a larger level of conservatism associated with the generation of the original licensing basis IRS at PBAPS, is the fact that the original plant design IRS curves were only generated for the OBE load case, using 2% structural damping for all structures. For design activities and in the A-46 evaluations, IRS curves for the SSE were obtained by a linear increase of the OBE curves using the conservative correction factor of 0.12/0.05=2.40. This practice, of linearly increasing the OBE results to obtain the SSE response, results in the conservative application of the OBE structural damping values of 2%, for both steel and concrete, to the SSE case where a value of 5% is allowed by the license basis and values of 7% and 10% are allowed by R.G.1.61 and recommended by NP-6041, respectively.

The effect of all this conservatism can result in peak values of the IRS to be overestimated by factors of between approximately 1.6 and 2.3.

Time History Simulation IRS are typically generated using a time history which is intended to approximate the desired earthquake spectrum defined by the design basis SSE. This process typically involves the generation of an artificial time history whose response spectrum envelops the SSE. The amount of conservatism involved in the enveloping process is variable plant to plant, but can range up to a factor of 2 or more unless significant resources are applied to minimize the degree of enveloping.

The July 12,1952 Taft, California earthquake time history record normalized to 0.05g was used for the PBAPS seismic analysis OBE analysis. The response of this record is compared to the design response spectrum for 2% damping in PBAPS UFSAR figure C.3.12 and reproduced in Appendix 1. The amount of conservatism for this record is higher than a typical artificially generated time history. As shown by the comparison in Appendix 1, the conservatism can be as high as a factor of 2.5 (0.35g vs. 0.14g at approx. 3 Hz.)

Overall Conservatism. There are several additional sources of conservatism (e.g., structural modeling, structural / soil non-linearities, peak broadening and enveloping, etc.) which add to the overall conservatism in the calculation of IRS.

Ther additional conservatisms, coupled with those described above, certainly reit wce the overalllevels of conservatism in IRS of between 1.5 and 8 which were referenced by SSRAP (LLNL Report NUREG/CR 1489), and explain why the conservative PBAPS IRS result in exceeding 1.5 times the SOUG Bounding Spectrum.

The following describes the results of recent median-centered IRS calculations at PBAPS, which confirm the level of conservatism in the licensing basis IRS discussed above.

New median centered response spectra were generated by PECO in 1994 for Elevation 165' of the PBAPS Radwaste Building. These spectra were approved for use in USI A-46 by the NRC in the SER dated 12/01/94 (ref. 2.3). The figure below shows the median centered spectrum in the N-S direction compared to the j

design basis spectrum and 1.5 times the SQUG Bounding Spectrum.

Pag)10 of 26 Radwaste Building El 165' Conservative vs Median Centered Spectra 5 % damping 2.5 l

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The peak of the median centered 5% damped horizontal SSE lRS of 0.85g at El 165 ft is 42% of the corresponding conservative design basis IRS peak of 2.02g.

This is indicative of a level of conservatism of at least 2.4 for a reinforced concrete structure.

Similar levels of conservatism exist in the Design Basis Analysis for other nuclear plants. Enclosed, as Appendix 2, is a comparison evaluation of overall seismic mar 0 ins between Median Centered Analysis and Design Basis Analysis for nuclear power plant structures at other facilities similar in construction, building frequency and damping to those at PBAPS. This information has been

- developed and compiled by EOE International, Inc., and is meant to demonstrate that factors of safety in the original design basis analysis can be shown to be in the range of 2.5 to 5.

Similar reduction in conservatism would be seen if median centered IRS were generated for buildings where Method A was used such as the Diesel Generator Building. It is anticipated that the level of conservatism for the Diesel Generator Building will be higher than the 2.4 factor calculated for the Radwaste Building, due to the conservative seismic modeling which ignored the restraint of the backfill. (For more information refer to the discussion presented earlier under

'SSI Interaction'). Note that while these spectra were termed " median centered,"

they still contain many of the conservatisms inherent in calculated in-structure response spectra as described above.

PBAPS Buildinas Are Typical Nuclear Structures The Diesel Generator Building is a 5 bay reinforced concrete Seismic Class I, tornado resistant structure. It measures approximately 64 ft by 136 ft in plan. The base elevation is at El 127'-0",

and the roof elevation is Elevation 161'-11". Horizontal seismic loads are transferred to solid rock through 2 ft and 3 ft thick reinforced concrete shear walls. Vertical loads are transferred to the rock through the shear walls and piles. The exterior and interior walls are 2 ft thick reinforced concrete shear walls with 2 ft thick concrete floors and roof supported by steel beams.

Paga 11 of 26

~

.The Circulating Water Pump Structure is a reinforced concrete structure, and is approximately -

.80 ft by 280 ft in plan.. The center portion of the building houses the safety related equipment

including the emergency service water pumps and RHR service water pumps for both units.

The structure is founded on solid rock with the top of the slab at Elevation 86', Grade Elevation for the site is Elevation 116'; However, since the east side of the structure is not embedded, the effective grade for' A-46 was conservatively taken at El 86'. The operation floor slab where most of the SSEL components are located is at Elevation 112'. In the center portion, the substructure

' is composed of pump bays separated by reinforced concrete shear walls. The superstructure over the pumps is constructed with reinforced concrete walls and a reinforced concrete roof supported at Elevation 130'-6" on structural steel beams.

Therefore; the buildings at PBAPS in which Method A'was used are stiff structures consisting of reinforced concrete frames and shear walls. Therefore, these buildings represent typical nuclear plant structures as defined in the SSRAP report and Page 4-16 of the GlP-2.

' Not a Significant Safety issue The expected differences between calculated IRS and actual building response do not represent a significant safety question. The lessons learned from review of hundreds of items i

. of equipment at various sites that have experienced earthquakes which were significantly larger than those for Eastem U.S. nuclear plants are that missing anchorage,~ seismic interaction hazards, and certain equipment-specific weaknesses (incorporated into the GlP caveats) were the seismic vulnerabilities which cause equipment damage. These areas are conservatively addressed in the'GlP.

The NRC staff acknowledged the seismic ruggedness of nuclear power plant equipment in the

' backfit analysis for USl A-46 in which they stated the following:

)

... subject to certain' exceptions and caveats, the staff has concluded that equipment installed in nuclear power plants is inherently rugged and not sus' eptible to seismic damage."

c

[Ref 2.4, page 16]

Method A is only_ applicable to relatively. stiff equipment with fundamental frequencies over

, about 8 Hz. 'As noted earlier, SSRAP and SQUG have agreed that excitations over 8 Hz have

little damage potential due to low spectral displacements, low' energy content and short duration. : This judgment is supported by industry and NRC guidance for determining whether an operating basis earthquake (OBE)is exceeded following a seismic event at a nuclear power plant. ' EPRI Report NP-5930 and NRC Regulatory Guide 1.166 recognize that damage i

potential is significantly reduced for earthquake ground motions above 10 Hz; In other words, the question of what is the precise value of building amplification over 8 Hz has very little safety significance.

Conclusions

-.The discussion above leads to several conclusions:

. ! The results from actual measured IRS on " nuclear type" structures support the 1.5 response levels advocated within Method A.

e ' Qualitative assessments of the conservatism inherent within the methods utilized to calculate IRS have been provided above.' These~conservatisms are typically

. quite significant (as has been independently verified by median / modem

r-l l

Page 12 of 26 assessments such as the LLNL study) and result in IRS which show amplifications well beyond the 1.5 factor from Method A. Specific exceedances noted for PBAPS (beyond the 1.5 factor) are due to the conservatisms inherent in the IRS calculation methods, and do not invalidate the application of Method A.

The development of more realistic median centered IRS for the PBAPS Radwaste Building support the above conclusion that the original licensed basis IRS are very conservative. If the reduction factors that were obtained for the Radwaste Building are applied to the Buildings which used Method A, the resulting IRS within 40' from 2

effective grade would be enveloped by 1.5 times the Bounding Spectrum. In effect, Method B of the GlP could have been used on these buildings (As was done on the Radwaste Building)if median centered IRS were generated fcr these buildings.

REFERENCES:

2.1 Senior Seismic Review and Advisory Panel (SSRAP), "Use of Seismic Experience and Test Data to Show Ruggedness of Equipment in Nuclear Power Plants," Revision 4.0, February 28,1991.

2.2 "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1)," EPRI NP-6041-SL, Rev.1, August 1991.

2.3

' J.W. Shea (NRC) letter to PECO Energy Company, dated December 1,1994.

"Use of Realistic Median Centered in-Structure Response Spectra for PBAPS Units 2 & 3, Radwaste Building".

2.4 U.S. Nuclear Regulatory Commission, " Regulatory Analysis for Resolution of

{

Unresolved Safety issue A-46, Seismic Qualification of Equipment in Operating i

Plants," NUREG-1211, February 1987.

Question No. 3 For cases where the IRS exceeds 1.5 times the ground response spectrum, provide a list of equipment, or an appropriate number of equipment items, in terms of percentage, whose seismic verifications were based on Method A of GIP-2.

' Response No. 3 As stated in our response to Question 2 from Reference 3.1, the following structures have in-J structure response spectra for elevations within about 40 feet above grade which are higher in j

amplitude than 1.5 times the SQUG bounding spectra:

j Emergency Cooling Tower - Floor Elevation 153'-0" e

Diesel Generator Building - Floor Elevations 127'-0",151'-0" and 161'-0" l

Circulating Pump Structure (CWPS) - Floor Elevation 116'-0" Below is a list of equipment located at these floor elevations whose seismic verifications are based,in part, on Method A of GlP-2. Method A was used to compare the equipment seismic f

capacity to seismic demand since the conservative spectra exceeds 1.5 times the bounding a

spectra. However, the anchorage evaluations for these items used the appropriate in-structure response spectra to calculate the seismic demand on the anchorages.

}

Page 13 of 26 Emergency Cooling Tower El 153'-0" No equipment at this location used Method A as a basis for seismic verification.

Diesel Generator (D/G) Building Elevation 127'-0" Emergency Service Water (ESW) Outlet Block Valves AO-33-0241A/B/C/D l

D/G Air Start Solenoid Valves AO-52C-7231A/B/C/D AO-52C-7232A/B/C/D f

ESW Return to Discharge Pond Valve (at El 121')

MO-33-0498 High Pressure Service Water (HPSW) Return Valves MO2-32-2486 l

(at El 121')

MO2-32-2803 i

MO3-32-3486 MO3-32-3803 ESW Booster Pumps (at El 121')

0AP163/0BP163 D/G Air Coolant Auxiliary Pumps -

0AP164/0BP164 0CP164/0DP164 D/G Jacket Coolant Auxiliary Pumps -

0AP166/0BP166 0CP166/0DP166 D/G Lube Oil Auxiliary Pumps 0AP167/0BP167 OCP167/0DP167 Fuel Oil Transfer Pumps OAP60/0BP60 0CP60/0DP60 I

Diesel Generators 0AG12/0BG12 OCG12/0DG12 125V Distribution Panels 0AD13/0BD13 0CD13/0DD13 CO System Electrical Control Panels OAC216/ OBC 216 2

OCC216/ODC216 Diesel Generator Building Elevation 151'-0" D/G Ven'ilation Supply Fans OAV064/OBV064 OCV064/ODV064 D/G Building Vent Supplement Supply Fans 0AV091/0BV091 OCV091/0DV091

.Outside Air Dampers PO-00272-1,2/PO-00273-1,2 PO-00274-1,2/PO-00275-1,2 E

l

Paga 14 of 26 Supplemental Supply Fan Damper PO-00540-1,2,3,4

'Cardox Heat Detectors HD-550A(B - HD-557A(B) f HD-796A(B) - HD-819A(B)

D/G Room Supply Ternp Controls 0AC479/0BC479 OCC479/0DC479 O'esel Generator Building Elevation 161'-0" No equipment at this location used Method A as a basis for seismic verification.

Circulating Pump Structure (CWPS) El 116'-0" HPSW pumps (Anchored El 111'-0")

2AP42/3AP42/2BP42/3BP42 2CP42/3CP42/2DP42/3DP42 ESW pumps (Anchored El 111'-0")

OAP5//0BP57 Motor Operator for Sluice Gates MO2-30-2233A(B)

MO3-30-3233A(B)

HPSW Crosstie Valves MO2-32-2344/MO3-32-3344 HPSW Pump Room Supp!y Fans 2AV060/2BV060 3AV060/3BV060 HPSW Pump Room Exhaust Fans 2AV083/2BV083 3AV083/3BV083 HPSW Pump Room HVAC Dampers PO20233-1,2,3,4 PO30223-1,2,3,4 D HPSW Pump Room HVAC DPS20224-1,2,3,4 Differential Pressure Sensors DPS30224-1,2,3,4 HPSW Pump Room HVAC Pressure Switches PS30224-1,2 HPSW Pump Room Temperature Controllers TIC 20223/ TIC 30223 TIC 30224-1,2 HPSW Pump Room Temperature Sensors TS20224-1,2 HPSW Pump Room Temperature Transmitters TT20223/TT30223 HPSW Pump Room HVAC Control Panels 20C139/30C139

Pag 315 of 26

REFERENCES:

' 3.1 Letter from G. A. Hunger, Jr. (PECO Energy Company) to USNRC Document Control Desk, " Peach Bottom Atomic Power Station. Units 2 and 3, Request for Additional

. Information Regarding Generic Letter 87-02 on the Resolution of Unresolved Safety issue A-46," dated September 3,1997 Question No.~ 4' con page ia of Calc. No. PS-0912 (Reference 1), you stated that the specific work was not

.. performed as nuclear safety-related, but was performed, reviewed, snd approved in accordance with the preparer's (Duke Engineering and Services) Quality Assurance Program. Discuss the

. calculational method or assumption and demonstrate that it meets the GIP-2 criteria for cable trays.

Response No. 4-Calculation PS-0912 was prepared in accordance with the Duke Engineering & Services (DE&S) Quality Assurance program which complies with 10CFR50, Appendix B and ASME NOA-1-1989 " Quality Assurance Program Requirements for Nuclear Facilities", through NOA-1b-1981 addenda.

The subject calculation was performed to document the seismic adequacy of the Peach Bottom Atomic Power Station (PBAPS) cable and conduit raceway in accordance with Section 8.2.4,

" Selection of Sample for Limited Analytical Review" requirements of GIP-2 (Reference 4.1).

The Gir-2 guidelines, for selecting representative worst-case samples of raceway supports on which Limited Analytical Reviews (LARs) will be performed, were followed by the Seismic j

Capability Engineers (SCEs). Prior to performing the walkdowns, the SCEs chose the

" representative worst-case samples" to perform the LAR. The SCEs were very familiar with the LAR guidalines in Section 8.3 of GIP-2 and the sample evaluations. During the plant 1

walkdowns, the location of the selected sample was noted, and detailed sketches and i

photographs of the as-installed supports were made once the worst-case samples were identified. The SCEs identified the most heavily-loaded raceway supports for each configuration by taking into account cable fill, long spans, multiple tiers systems, top supports at I

vertical runs, fire protection coatings and for rod hung raceway short rod lengths for fatigue evaluations. Additional walkdowns were performed by the SCEs to obtain missing information.

The LAR evaluations were performed by qualified experienced structural engineers under the direction of the SCEs. Each engineer was familiar with all the GIP-2 requirements and participated on other USI A-46 projects in the same capacity. Any assumptions that were made in the calculations were of a conservative nature and do not require future verification.

It should be noted that the SCEs in addition to performing the walkdowns to select the " worst-case samples" for LAR evaluations, also performed walkdowns in accordance with Sections 8.2.2 and 8.2.3 of GIP-2 to, identify the important limits of the earthquake experience and shake table test data bases and certain undesirable details which, if violated, could significantly compromise the seismic adequacy of a raceway system. The results of these walkdowns were reported in the USl A-46 submittal report (Reference 4.2) for the PBAPS SQUG program to the NRC.

B Pags 16 of 26

[In summary, the walkdowns and LAR evaluations that were performed at PBAPS meet the guidelines and requirements of Section 8 of GIP-2.

p

REFERENCES:

4.1.'- SQUG Generic implementation Procedure (GlP) for Seismic Verification of Nuclear Plant

Equipment, Revision 2, corrected 6/28/91. (GIP-2)

-4.2.

G.A. Hunger (PECO) letter to NRC, dat'ed May 7,1996, Attachment 3, Seismic Evaluation Report for Peach Bottom Atomic Power Station in Response to NRC Generic Letter 87-

02/USl A-461 Question No. 5 In reference to Sample 1 (page 2 of Attachment A of Reference 1), the sketch indicated that the supports have only two bolts, one at the top and one at the bottom. Discuss the lack of

? moment resistance capacity in the direction of the cables, and for this support, indicate why it is

- adequate with no such moment resistance capacity.

1The sketch also indicated that the upper horizontal strut members and the lower strut members are connected by rods (there is no indication in the sketch about what this llne represents).

Discuss how the lateral load is resisted. Discuss whether the rod hanger type indicated in the sketch is within the scope of the database referenced in GlP-2.

Also in the above Sample 1, t'he tensile and shear capacities are provided only for anchor bolts.

~

Discuss the anchorage capacity of the concrete and the amount of a safety margin against

)

' spillage (shear) at the interface between the anchor bolts and the concrete. Explain whether or not the capacity of the bolts is limited by the bolt itself or the capacity of concrete.

Response No. 5~

1. Detailed information conceming GlP-2 (Section 8.3 of Reference 5.1) limited analytical

~

review guidelines is presented in Section 4 of Reference 5.2. Item.3 on page 4-1

(Reference 5.2) states in part the following: "The majority of the data base raceway supports have no conventional lateral load-resisting structural system....". In addition, per J Reference 5.2, limited analytical review guidelines address structural integrity by correlating

. raceway support systems that performed well in past earthquakes. Reference 5.4 provides examples of longitudinal cable tray runs from the seismic experience data base. Section

2.9 of Reference 5.4 shows a cable tray system at the Fertimex Auxiliary Power Plant. The supports on this long run are similar to Sample 1 in that the supports are attached to the wall using two bolts with insignificant longitudinal load moment resistance. There were no observed seismic effects to any of the supports or cable tray sections following the 1985

' Mexico Earthquake;

'The referenced Sample i support of Calculation PS-0912 is considered a rigid wall mounted L support per Section 5.3 of GIP-2 Reference 4. Per Section 8.3 of GlP-2 and also Section l 5.3 of GIP-2 Reference 4, the only check required for the subject support, in performing a LAR review is the " Dead Load Check". If the " Dead Load Check" is satisfied, the support is seismically rugged and there is no need to evaluate support / anchor bolts for lateral loads.

Pages 11 to 13 of Calculation PS-0912 show the Sample 1 support is seismically acceptable. Therefore, the support meets the GIP-2 requirements and is seismically rugged, even'with the apparent lack of moment resistance capacity in the direction of

cables.

c t

Pago 17 of 26 ljg

.1,

2.' As explained above, there is no need to evaluate support / anchor bolts for lateral loads 0 4 because the " Dead Load Check" is satisfied.

'D

. The rod hanger support of Sample 1 LAR of Calculation PS-0912 is similar to rod hanger support, example 2, of Section 5.4.2 of Reference 5.2. Therefore, it is within the scope of the earthquake database referenced in GIP-2. The rod capacities in this case were not z ! evaluated since the seismic capability engineers determined that this condition was not the

" 3" worst-case" rod hung support for LAR evaluation (i.e., see LARs for Sample supports 4,7, L8 and 9).

3.. The nominal anchor bolt capacities were obtained from Table C.2-1 (Nominal Allowable Capacities for Expansion Anchors) of GlP-2 (Reference 5.1) and were evaluated for all the

requirements of GIP-2 Appendix Section C.2 to determine the allowable tension and shear

loads in the LAR evaluation. These allowable loads were obtained from Reference 5.3. The allowab!e loads were obtained from expansion bolt tests and include the effects of concrete

. strength. Typically expansion anchor bolts fail in the anchor slip mode under tension and I

bolt material failure mode under shear. A factor a safety of 3 was applied to the mean

- strengths. For more information refer to Section 2 of Reference 5.3.

The shear load transfer between the wall bracket and the bolts is by bearing, therefore, slippage of the bracket support is not a concern in this case.

REFERENCES:

5.1.

SQU' G' Generic implementation Procedure (GIP) for seismic verification of nuclear plant equipment, revision 2 corrected,6/28/91. " GIP-2".

5.2.

EPRI Report NP-7151-D " Cable Tray and Conduit System Seismic Evaluation Guidelines", Project SQ01-1 Final Report, March 1991.

5.3.

EPRI Report NP-5228-SL " Seismic Verification of Nuclear Plant Equipment Anchorage (Revision 1), Volume 1: Development of Anchorage Guidelines, Project 2925-1, Final Report, June 1991.

5.4.

EPRI Report NP-7153-D " Longitudinal Load Resistance in Seismic Experience Database Raceway Systems," Project SQ01-1, Final Report, March 1991.

' Question No. 6

. In reference to Sample 2 (pages 3 & 4 of Attachment A), it appears that there are two distinct

. cable supports. However, the calculation does not appear to address the support on page 4, which is a ceiling hung support. Provide an explanation for this.

Response No. 6 The sketches that appear on pages 3 and 4 of Attachment A do represent 2 distinct supports.

The Limited Analytical Review (LAR) for support Sample 2 field sketch is shown on page 3 of Attachment A and the support is evaluated on pages 14 through 16C of Calculation PS-0912.

LAR support Sample 3 field sketches are shown on pages 4 and 5 of Attachment A and pages 3 and 4 of Attachment B.' The evaluation of Sample 3 is shown on pages 17 and 18 of Calculation PS-0912.

re 4

Paga 18 of 26 Support Sample 3 is a three tier cable' tray supported by rods from above. Page 4 of '

4 Attachment A shows the typical cable tray details for a hung system. However this support has

- a unique cantilever support at the top which is shown on page 5 of Attachment A (later modified by page 4 of Attachment B) The applicable Attachment A and B pages identify the support as I

Sample 3.'

A

' Question No.~ 7 in reference to Sample 3 (pages 17 and 18 of Cal. PS-0912 and the corresponding sketch on page 5 of Attachment A), there appears to be an inconsistency in the welded connection detail'

. provided.- On pages 17 and 18 of Cal. PS-0912, it is indicated that the strut is welded into the iflange of the structural steel beam,'whereas on page 5 of Attachment A, the strut is indicated as welded to the web. You are requested to provide a clarification as to which one is correct

- and how it was actually analyzed. Explain also how you concbded that the welded connection j

is acceptable.

. Response No. 7

. The LAR evaluation for. support Sample 3 is shown on pages 17 and 18 of Calculation PS-0912. The actual weld configuration is as analyzed on page 18 of the subject calculation. The original walkdown sketch of the weld detail (i.e., page 5 of Attachment A) which shows unistrut member welded to the beam flange was incorrect. This sample support was re-walked down to get additionalinformation to complete the LAR evaluation. During the second walkdown it was determined that the unistrut member is welded to the web of the structural beam.' This configuration is shown on page 4 of Attachment B.

J

'On page 18 of Calculation PS-0912 the actual weld stress is 2291 pounds per inch and the

' minimum allowable weld stress is 2356 pounds per inch. Therefore, since the allowable weld j

stress is greater than the actual weld stress the welded connection is acceptable.

Question No. 8 On the cover sheets of calculations 0912 and 0939, you checked the box which corresponds to "non-safety-related" as opposed to " safety-related," implying that the affected cable tray supports are not safety-related. Confirm that supports with a non-safety-related classification meet the GIP-2 criteria.

IResoonse No. 8 The response'to Question 4 above for calculation PS-0912 also applies to Calculation PS-0939.

The checked box indicates that the work was not performed as safety-related but as noted in

? the calculation, the work was performed, reviewed and approved in accordance with the preparer's Quality Assurance Program. The cable tray and conduit walkdowns and LAR L evaluations (both safety-related and non-safety-related) that were performed at PBAPS meet the guidelines and requirements of Section 8 of GlP-2.

' Question No. 9 Discuss.how spatial interaction between the cable trays and supports and other objects were evaluated. In particular, there are many cable trays supported by suspended rod hangers whose moment resisting capacity may be insignificant for lateral seismic motion (free to swing).

1 i

l Pagn 19 of 26 Discuss how the potential for interaction was addressed for such cases. Provide a drawing for the longest rod hanger or other cable tray supports together with the location of objects that may interfere with the cable tray supports. Demonstrate that there are sufficient spaces for the cable supports to swing freely without impacting with adjacent equipment or structures such as j

walls or columns.

Response No. 9 The evaluation of the effects of spatialinteraction between the cable trays and supports and other objects is deendent on the potential object being impacted.

4 If the cable tray could impact fragile or sensitive equipment on the Safe Shutdown Equipment

)

List (SSEL), the interaction was identified during the SQUG walkdowns. Section 4.2.3,

(

" Seismic :nteractiun" of Reference 9.1 states; "The SRT seismic walkdowns included evaluation j

for potential seismic interaction concerns, per GIP 11.4.5 and Appendix D. Seismic interaction 1

concems are documented on the equipment SEWS". In addition, the SRT placed additional emphasis during the walkdowns, on reviewing spatialinteractions to assure that no potential interactions were caused by permanent plant commodities or structures, since the original plant design basis did not specifically address compliance with RG 1.29. Tbcrefore, any spatial interaction concerns between cable trays and supports with equipment on the SSEL for the PBAPS USl A-46 program are identified on the Screening Evaluation Work Sheets (SEWS) and if the interaction was a concern the component was listed as an outlier and documented on the Outlier Seismic Verification Sheet (OSVS).

If the cable tray displacements could impact hard objects such as building columns or large bore piping, minor damage to the trays were considered to be acceptable, as long as the damage would not affect the ability of the safety related cables to perform their function. Before performing the plant walkdowns to evaluate the seismic adequacy of the raceways, the seismic capability engineers reviewed existing plant raceway documents and details and industry backup information that supported the requirements of GIP-2 Section 8. Reference 9.2 and relevant sections of the GlP were reviewed in detail by the SCEs in order to become thoroughly familiar with guidelines addressing plant walkdowns and LAR evaluations for cable and conduit raceway systems. It should be noted that the SCE team had over 80 years experience addressing seismic issues on nuclear plants and in addition Mr. Robert P. Kennedy who is a recognized industry expert on seismic issues and was the chairman of the Senior Seismic Review and Advisory Panel (SSRAP) was retained as a consultant and participated in plant walkdowns to specifically review concerns that the SCEs had with overhead systems seismic adequacy and seismic interactions.

In the final analysis, the SCEs and Mr. Kennedy concluded that spatial interactions between the raceway systems and adjacent _ structures such as walls or structural columns was not a concern for PBAPS and thus, the overall seismic adequacy of the raceway systems (with the exception of the outliers that identified interactions with equipment on the SSEL) is acceptable per the GIP requirements. This conclusion is further strengthened when Reference 9.2 is reviewed in detail.' Some of the Reference 9.2 conclusions and guidance provided in performing the cable tray and conduit system seismic evaluations is excerpted below:

Section 1.1 " Cable tray and conduit systems have consistently performed well at conventional power and industrial facilities subjected to past strong-motio.. earthquakes....... even though the systems are typically not designed for earthquake loading".

Section 2 "In many cases of earthquake-resistant design the focus is on developing an adequate lateral force-resisting system. This does not appear to be the primary factor for designing a raceway system that will perform adequately in an earthquake.....What does

Pagn 20 of 26 appear to be important is a support system that will maintain primary overhead vertical support.

c

if the structural elements challenged by lateral movement are ductile, the lack of any significant lateral load resistance capability appears to be a relatively insignificant consideration".

(Section 2, Footnote 1. " Past earthquake experience shows that most spatial interact ons involving raceways are not harmful (the impacts do not cause damage that would be

. considered to be safety-related). The impacts that do occur seem to introduce another nonlinearity in the structural system that prevents the resonant build-up of motion. This is -

' based on the observed numerous spatial Interactions in piping systems. There are few known cases'of spatial interaction in raceway systems. Nevertheless it is important to determine if there are fragile _ equipment proximate to raceways that could become damaged".

In summary,' the above discussion shows that the SCEs in performing the plant walkdowns, for review of the seismic adequacy of cable tray and conduit systems, including spatial interaction of raceway systems and other plant components or structures had the knowledge and experience required by the GIP to determine if the raceway system was seismically adequate.-

and if spatialinteractions was a concern with adjacent equipment or structures. The final conclusion of the SCEs was that spatialinteraction of cable trays with hard objects was not a concern and was judged acceptable by the SCEs.

Subsequent to the receipt of this RAl,' PECO Energy performed additional plant walkdowns by two SCEs. The SCEs specifically walked down cable trays throughout the plant that are supported by long rods from the ceiling above and located near building hard spots such as

. walls or columns.. Appendix 3 contains the typical support detail for rod hung cable trays at PBAPS. In most cases when long rod-supported trays were identified, the trays were laterally supported at or near the long supports from the building walls or columns. In cases were lateral a

bracing was not provided sufficient space exists either to prevent impact with the walls or columns or the impacts would be minor. The following is the basis for this conclusion:

. The maximum rod hanger length is approximately 20 ft.

Treating the rod hung cable tray as a pendulum, the natural frequency of a 20 ft long a

pendulumis f =

. Using g = 32.2 ft/sec and L=20 ft; f=0.202 Hz.

For 5% damping, the spectral accelerations from the PBAPS design response spectra are l

very low at frequencies below 1Hz.

The corresponding spectral displacements at 0.2 Hz are on the order of 4 to 7 inches.**

-in conclusion, the recent plant walkdown confirm the results of the PBAPS USl A-46 program

' that spatial interactions with cable trays and other plant structures is not a concem and thus meet the_ screening requirements of the GIP.

REFERENCES:

j 9.1

G. A. Hunger (PECO) letter to NRC, dated May 7,1996, Attachment 3, Seismic Evaluation Report for Peach Bottom Atomic Power Station in Response to Generic Letter 87-02/USl A-46

~

The maximum displacement from the SSE ground response spectra at 5% damping is 4 inches. The

. maximum spectra acceleration from the radwaste/ turbine building floor response spectra for 0 2 Hz,5%

a+386.4 damping is 0.028g. Converting this to spectral displacement: d =

= a+ L = 6.7 in.

I

c Pago 21 of 26 9.2 EPRI Report NP-7151-D " Cable Tray and Conduit System' Seismic Evaluation Guidelines", Project SQ01-1 Final Report, March 1991.

- Question No.1'O Provide an updated status and estimated schedules for the resolution of all the identified outliers.

Response No.10 m

[7

.'A) Outilers identified in the Seismic Evaluation Report:

Table 4.2-4 of Reference 10.1 provided a summary of the screening and walkdown results for each equipment class. The following table provides an update of the status of the unscreened components that were identified in Table 4.2-4.'

Equipment Class Unscreened Unscreened issue Status Class Description Component ID 1-Motor Control 30D11;-

Evaluation of Resolved: Anchorage Centers 00853,54,55,56 anchorage required adequate 00B97/98/99 Interaction Concerns Resolved - Breaker hoists have been restrained 20B36 Demand exceeds Resolved: Anchorage capacity, Evaluation adequate of anchorage required 2

Low Voltage 20B10/11/12/13; Anchorage concerns Adjacent Oil filled Switchgear 30810/11/12/13 and interaction transforrners have been issues replaced and anchored Breaker hoist restraints have been installed.

Anchorage enhancements completed for 20B10 and 20B12 Anchorage enhancements for the remaining switchgear are scheduled for.1999 and 2000 00B94/95/96 Anchorage concerns Resolution of and interaction anchorage scheduled issues for 1999. Interaction issues are resolved.

l i

Paga 22 of 26 i

Equipment Class Unscreened Unscreened issue Status

{

Class Description Component ID 3

Medium Voltage 20A15/16/17/18; Door latching Top latch provided for Switchgear 30A15/16/17/18' mechanism not 20A15/17 adequate Others scheduled to be completed by end of 2000 Interaction Concerns Spare breakers removed.

4 Tran: formers 20X133/150; Transformer Coil Resolved: Anchorage 30X133/150 Anchorage adequate 20X30/31/32/33; Unanchored Modification to replace 30X31/33 Transformer transformer and anchorage is complete 00X103; Verification of proper Resolved: Anchorage 30X135 coil anchorage is adaquate required 0AX26; OBX26; Unanchored and Demand vs Capacity OCX26 Demand exceeds issue resolved, Capacity Anchorage issue scheduled to be resolved in 1999 6

Vertical Pumps 2A(B,C,D)P42; Pump casing and Resolved by analysis 3A(B,C,D)P42; impeller shaft greater 0AP57/0BP57 than 20ft in length Interaction concern Resolution scheduled with gantry crane to be completed by end of 2000.

7 Fluid Operated A02 Distance from Resolved by analysis Valves 080A/B/C/D; A02-centerline to top is 01-086A/B/C/D; outside of experience A02-03-33;

database, A03-03-33; A03 080A/B/C/D; A03-01-086A/B/C/D j

A02-03-35A/B; Seismic Demand Resolved by analysis

{

A03-03-32A/B; exceeds Capacity j

A03-03-35A/B RV2-02-071A/B/C/

D/E/F/G/H/J/K/L; RV3-02-071A/B/C

/D/E/F/G/H/J/K/L; A02-03-32A/B

Pags 23 of 26 Equipment

. Cess Unscreened Unscreened issue Status Class Description Component ID t

8A Motor Operated M03-12-018 Seismic Demand Resolved by analysis Valves' exceeds Capacity M0-33-0498 Interaction concern Resolved - Support has with support been modified M02.

Valve Operator

. Resolved by analysis 016A/B/C/D M02-- weights and/or 23-019; M03-centerline distances23-019;-

M02-are outside of the 23-025; M03-experience database 23-025; M02-10-025A/B; M03-10-016A/B/C/D M03-12-015;.

M03-10-025A/B M02-30-2233A/B; Cast Iron Yoke Resolved by analysis M03-30-3233A/B 8B Solenoid SV-8130B Component not Resolved - Component Operated Valves located was located and walked down. No issues 9

Fans 0AV035/36; Overhead ducts Modification to ducts 0BV035/36 need to be reviewed scheduled to be and seismic completed by end of interactions were 2000.

identified 2 over 1 concern of overhead fan / motor has been resolved by restraining the fan / motor 0AK032; Seismic Demand Resolved by analysis 0BK032; exceeds Capacity OCK032 -

10 Air Handlers -

00F043; OAV034 Interaction concerns Modification to ducts identified for scheduled to be overhead HVAC completed by end of system 2000.

0AV034 Nozzle load from Resolved-nozzle loads attached piping are acceptable.

needs to be evaluated PO20223-1; Interaction concerns Clarnps schedu!ed to PO20223-3 with conduit be repaired by end of suppoded by beam 2000.

clamps 14 Distribution 20D21/22/23; Anchorage concerns Top of panels are Panels -

30D22/23/24 schedul6d to be l

anchored to walls by

]

the end of 1999.

'l 4

Page 24 of 26 Equipment Class Unscreened Unscreened issue Status Class Description Component ID 15 Batteries on 2AD01/ 2BD01/

Batteries are outside Resolved Racks 2CD01/ 2DD01; of Database

- 0 3AD01/ 3BD01/

3CD01/ 3DD01 Some end rails not Scheduled to be l

snug tight resolved by end of J

2000.

Interaction concerns Scheduled to be with overhead resolved by end of fluorescent lights 2000.

16 Battery-20D37 Interaction Concerns Scheduled to be Chargers and resolved by end of j

Inverters 2000.

]

Drip Shield required Drip shield installed

{

17 Engine-0AG12/0BG12/

Interaction concern Scheduled to be Generators.

OCG12/0DG12 with overhead crane resolved by end of controller 2000.

i Local panel on Restraints have been vibration isolators are provided, without lateral capacity 1

18 Instruments on DPS 20224-1 Interaction concern Scheduled to be l

Racks DPS 20224-3 with overhead resolved by end of

)

. conduit supported by 2000.

beam clamps TIC 30223 Cover to panelis Scheduled to be I

loose resolved by end of 2000.

PT-2508B/

Seismic demand Resolved by analysis PT-3508B exceeds capacity 2AC65/2BC65; Seismic demand -

Rescived by analysis 3AC65 exceeds capacity,-

Anchorage did not screen out 20 Control and 0AG13/0BG13/

Seismic demand Resolved by analysis Instrumentation OCG13/0DG13; exceeds capacity Panels and 0AC097/0BC097 Cabinets 0CC097/0DC097-Interaction concern Scheduled to be with overhead crano resolved by end of controller 2000.

00C29A/B/C/D;

, Interrction concern Resolved: Items have 20C124/30C124 with housekeeping been removed or issues res' rained 20C32/33; interaction concerns-Cabinets are scheduled 30C32/33; Adjacent cabinets to be tied together by 20C722A/B; not tied together the end of 1999.

30C722B i

i:

Pags 25 of 26 Equipment Class Unscreened Unscreened issue Status Class Description Component ID 20C139 Interaction concern Scheduled to be resolved by end of 2000.

21 Tanks and Heat 2AE24/2BE24/

Review of Resolved: Anchorage is Exchangers 2CE24/2DE24; calculations required adequate 3AE24/3BE24/

to evaluate 3CE24/3DE24 anchorage capacity 22 Electrical Cable Pipe Stanchion Lateral Load criteria Unit 2 Reactor Building trays and Supports Rx of GIP not met stanchion support has

. Conduit Building El 195',

been modified. Others RW Bldg. El 165' are scheduled to be completed by the end of 2000.

Conduit in Room Conduit supported by Modification to supports T2-171 beam clamps which scheduled to be rely on friction complete by end of 2000.

Conduit in RW Interaction concern Modification to supports Bldg El 165' Fan and conduit scheduled to be room supported by beam complete by end of clamps which rely on 2000.

friction N/A HCUs HCU's for Unit 3 Top horizontal frame Modification to supports for HCU piping scheduled to be requires bracing complete by end of 1999.

B) Outilers identified in the Relay Evaluation Report A total of 226 relay outliers were identified in the USl A-46 relay evaluation report. Section 5.0 l

of the report described the proposed outlier methodology for these relay contacts. The following is the status of the outlier resolution for these relay contacts.

Forty-two (42) relay contacts have been resolved by analysis. The methods as described in Section 5.0 of the Relay Evaluation Report, with the exception of peak clipping, were used. The Relay Evaluation Report states that 55 relay contacts were resolved by analysis. This number is

. revised, as the 13 relay contact outliers that were originally resolved by peak clipping have been resolved with the use of existing test data. (See response to Question 1 of this RAI).

Eighteen (18) relay contacts have been resolved by existing operator actions Thirty three (33) relay contact outliers have been resolved by the evaluation of existing

]

documentation. The Relay Evaluation Report states that 20 contact outliers were resolved using j

this method. This has been updated to 33 as described above and by the response to Question 1 of this RAl.

S!xteen (16) rel'ay contact outliers were previously reported to be resolved by a previous (post-walkdown) modification. A review of this modification indicated that 8 of these contacts required replacement. The K2 and K3 relays in panels 0AG13, OBG13, OCG13, and ODG13 have no GERs. These relays are scheduled for replacement as follows:

1 1

Pags 26 of 26 Panel 0AG13 Completed Panel OBG13 Scheduled for 2000 e Panel OCG13 Completed e Panel ODG13 -

Scheduled for 2000

Section 5.0 of the Relay Evaluation Report identified an additional one hundred seventeen

- (117) relay contacts to be reso!ved by relay replacement. These 117 contacts are located in a total of 24 GE PVD relays which are classified as low ruggedness relays. The schedule for replacement is as follows:

' BUS E12 - Relays 2-187-15A,B and C Completed e

BUS E32 - Relays 2-187-17A,B and C Completed BUS E22 - Relays 2-187-16A,B and C Scheduled for 2000 BUS E42 - Relays 2-187-18A,B and C Scheduled for 2000 BUS E13 - Relays 3-187-15A,B and C Completed.

BUS E33 - Relays 3-187-17A,B and C Completed

. BUS E23 - Relays 3-187-16A,B and C Scheduled for 1999.

BUS E43 - Relays 3-187-18A,B and C, Scheduled for 1999.

REFERENCES:

-10.1 G. A. Hunger (PECO) letter to NRC, dated May 7,1996, Attachment 3, Seismic Evaluation Report for Peach Bottom Atomic Power Station in Responsv to NRC Generic Letter 87-02/USl A-46

e I

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UNIT 213 SH.113 84 ggq_

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Bases for Interpretation and implementation i

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4

Bases for Interpretation and Implementation of GlP-2 Rules for Method A

References:

1. Seismic Qualification Utility Group (SQUG), " Generic implementation Procedure (GlP) For Seismic Verification of Nuclear Plant Equipment,"

. Revision 2, Corrected 2/14/92.

2. Senior Seismic Review and Advisory Panel (SSRAP),"Use of Seismic Experience and Test Data to Show Ruggedness of Equipment in Nuclear-Power P;mns," Revision 4.0, February 28,1991.

3. Nuclear Regulatory Commission, " Seismic Qualification of Equipment in

' Operating Nuclear Power Plants," NUREG-1030, February 1987.

4. Nuclear Regulatory Commission, " Regulatory Analysis for Resolution of Unresolved Safety issue A-46, Seismic Qualification of Equipment in Operating Plants," NUREG-1211, February 1987.

5. Nuclear Regulatory Commission, " Supplement No.1 to Generic Letter (GL) 87-02 That Transmits Supplemental Safety Evaluation Report No. 2 (SSER No.2) on SQUG Generic implementation Procedure, Revision 2, As Corrected on February 14,1992 (GIP-2)," May 22,1992.

Method A of GlP Table 4-1 (Ref.1) provides a methodology to evaluate the seismic adequacy of equipment by comparing equipment capacity based on earthquake experience ground response spectra at database sites with the plant's SSE ground response spectrum (GRS). The composite earthquake experience GRS from the database sites (reference spectrum) was reduced by a factor of 1/1.5 to account for possible additional amplification of motion in nuclear plants compared to database plant?'and is referred to as the " Bounding Spectrum" in the GIP.

The seismic capacity of equipment defined by the Bounding Spectrum is compared to the seismic demand at the effective grade using the plant licensing basis SSE

' GRS. The GIP Method A conservatively limits use of this approach to equipment which has a fundamental frequency above about 8 Hz and is located lower than i

about 40' above the effective grade of the building. These restrictions prohibi %e use of GlP Method A for equipment with low fundamental frequencies and fcr equipment located at high elevations in buildings where the structure seismic response is known to be typically high.

--Additional details justifying the use of GIP Method A may be found in the Senior Seismic Review and Advisory. Panel (SSRAP) report (Ref. 2). This report, included as Reference 5 in GIP-2, summarizes SSRAP's judgment on this subject:

"..... the use of very conservative floor response spectra should be avoided when, Page 2 of 17 E

assessing the seismic ruggedness _of floor mounted equipment...... Only for cases of equipment mounted more than 40 feet above grade or equipment with as-anchored frequencies less than about 8 Hz is it necessary to use floor spectra."

[Ref. 2, pages 102 and 103]

We believe that we have interpreted and implemented the rules for use of GIP Method A

- as they'were previously reviewed and accepted by the NRC in the " Supplemental Safety Evaluation Report (SSER-2)" of Reference 5. Note that ? Nre Method A was used at Peach Bottom Atomic Power Station (PBAPS), it is My juMW's based on the requirements of GIP-2 and its basis documents, such m t% SS%P Report and SQUG's guidelines on the use of Method A. The basis for this pedor is as follows:

q 1.

SQUG and PECO Enerov's Interpretation of the GIP 1

The caution given on page 4-16 of GIP-2 lists two limitations on use of Methed A:

1 Equipment should be mounted in the nuclear plant below about 40 feet above j

the effective grade, and

]

Equipment should have a fundamental natural frequency greater than about 8 Hz.

1 I

The introductory wording in GIP-2 for these two limitations provides the bases or

. purposes for imposing them, namely (1) to limit amplification to no more than about j

1.5 and (2) to avoid the high-energy frequency range of earthquakes. The specific limitations which are intended by the SQUGINRC expert panel (SSRAP) and SQUG to satisfy these bases are included in the two bulleted items listed above.

Indeed, Table 4-1 of GIP-2 which describes the two methods (Method A and Method B) in detail, includes the criteria which need to be met for each of the two methods. The table includes the above two limitations, but does not include the rement for checking the amplification between the in-structure response and r-ee field.

The statement on page 4-16 that "the amplification will not exceed about 1.5" is the expected result of meeting the above limitations, not a third condition which is required to be met.

The caution on page 4-16 of GlP-2 makes it clear that the advantage of Method LAis:

"TheidvantaDe of using yound response comparisons is that with the applicable rostrictions and limitations (i.e. the two bullet items listed above), all the equipment covered by the Bounding Spectrum or the GERS can be evaluated for seismic adequacy without the need for using, Par. 3 of 17

l in-structure response spectra which are often based on very conservative modeling techniques or may not be available." [Ref.1, page 4-16]

l 2.

The Intent of the GIP The GIP cites the SSRAP report (Ref. 2) as the basis for the Bounding Spectrum which is used in Method A (Ref.1, page 4-11). The SSRAP report explains the limitations and conditions which appear on page 4-16 of the GIP. SSRAP's report states:

"Thus, it is SSRAP's judgment that amplifications greater than a factor of 1.5 are unlikely in stiff structures at elevations less than 40 feet above grade except post,ibly at the fundamental frequency of the building where higher amplification:: occur when such a frequency is less than about 6 Hz. Thus, for equipment with fundamental frequencies greater than about 8 Hz in the as-anchored condition it was judged that floor spectral amplifications within 40 feet of grade would be less than 1.5 when reasonably computed using more median centered approaches."

[Ref. 2, page 102]

This judgment by the SSRAP was based on : ;merous studies and actual earthquake measurements which led them to conclude that:

"Thus, amplification of the horizontal free-field ground spectra by factors greater than 1.5 are considered to be generally unlikely for elevations less than 40 feet above grade."

[Ref. 2, page 104]

The SSRAP was aware that many nuclear plants were originally licensed based on very conservative ISRS and that the use of conclusions based on earthquake experience and more median-centered approaches would be more appropriate. A detailed discussion of some of the sources of this conservatism is presented in item 5 below. With reference to this topic, the SSRAP report states:

"It was judged by SSRAP that the use of very conservative floor spectra should be avoided when assessing the seismic ruggedness of floor mounted equipment. It was also the opinion of SSRAP that many of the operating plants may only have these very conservatively computed floor spectra available. To avoid the burden of having to compute more realistic floor spectra, SSRAP

~ decided to anchor its conclusions to ground spectra at the nuclear plant sites in those cases where this was judged to be reasonable."

[Ref. 2, page 102], Page 4 of 17

1 l

i This was the basis for SSRAP's recommendation (and included in GIP-2 as 1

e methods A and B):

"Thus, for the case of equipment with fundamental frequencies greater than about 8 Hz mounted less than 40 feet above grade, SSRAP's conclusions are based upon comparing the bounding spectra with nuclear power plant ground spectra. Only for the case of equipment mounted more than 40 feet above grade or equipment with as-anchored frequencies less than about 8 Hz is it necessary to use floor spectra."

[Ref. 2, page 102]

The SSRAP Chairman and developer of Method A, Dr. Robert Kennedy, was contacted by SQUG and concurs with the interpretation given in item 1 above.

- 3.

The NRC Was Aware 'of SQUG's Interpretation When it Was Developed The NRC staff has been closely following and involved in the development of the GIP methods from their inception. Even though the shape and number (there originally were three bounding spectrum curves) of the " Bounding Spectrum" may have been slightly different, and the application only covered eight classes of equipment, the concept of Method A has been part of the original GIP development since its inception. NUREG-1030, Ref. 3, presents the following salient points of Methods A and B as developed by the SSRAP:

"The comparison of these seismic bounds with design horizontal ground response spectra is judged by SSRAP to be acceptable for equipment mounted less than 40 feet above grade (the top of the ground surrounding the building) and for moderately stiff structures (fundamental frequency greater than 2 Hz).

For equipment mounted more than 40 feet above grade, comparisons of 1.5 times these spectra with horizontal floor spectra is necessary."

(Ref. 3, page 2-60]

The NRC backfit analysis in NUREG-1211 (Ref. 4), which justifies implementation of the USI A-46 program by affected licensees, relies on the conclusions reached by SSRAP in their review of seismic experience data. NUREG-1211 states the following:

I "The NRC staff has closely followed the SSRAP work and is in broad agreement with its conclusions. The staff has concluded that if the SSRAP spectral conditions are met, it is generally unnecessary to perform explicit seismic qualification on the eight classes of equipment stud!ed" 1 The eight classes of equipment cited in NUREG-1211 were later expanded to 20 classes., Page 5 of 17

[Ref. 4, page 17]

Note that this quotation specifically makes reference to the SSRAP " spectral conditions" which were described in the previous bullet and reference the 40 foot criteria only. The three spectral curves were later refined to the present day single curve through the introduction of the 8 Hz requirement on the fundamental frequency of the equipment. At no time was a check on the building amplification a

- requirement for the application of Method A.

The use of Method A was previously reviewed and accepted by the NRC and SSRAP representatives during two pilot plant reviews conducted in 1987 and 1988. These reviews are documented in References 16 and 25 of GIP-2. The specific material presented to the NRC representatives on use of Method A is described in the report of the BWR pilot review as showr, in Appendix 1. Note that the seismic demand criteria described during this trial plant review are the same as described in item 1 above. The topics anscussed with and comments made by NRC and SSRAP representatives during the BWR pilot review are included in

{

' Appendix 2. Note that seismic demand information was discussed in some detail i

with no reference to the 1.5 amplification screening criteria.

. ~ The GIP-2 was fully reviewed and endorsed, with some " Clarifications, interpretations, Exceptions and Positions," by the NRC staff through their SSER-2 of Reference 5. This document included a detailed presentation of the staff review of each section of GIP-2 and presented all clarifications, interpretations, exceptions and positions taken by the staff for each individual section of GIP-2.

SSER-2 included a lengthy discussion on the section for " Seismic Capacity Compared to Seismic Demand." None of the items presented in SSER-2 indicated that a clarification of the application of Method A was needed. The method was presented and described in GIP-2 as found in Table 4-1_which does not require a check of the building amplification.

The PECO Energy /SQUG interpretation of the rules for applying Method A is also consistent with the SQUG training course on use of the GIP methods. Appendix 3 is an excerpt from the class notes used during this course. It shows, in Slide 26, several screening methods for comparing equipment capacity to demand. Slide 26 illustrates uses of GIP Method A as described in item 1 above. That is, if equipment is below 40 feet and above 8 Hz, and the Bounding Spectrum -

envelopes the ground response spectrum, the equipment is acceptable.

This training material was used during the first session of the SQUG training course held during the week of June 22,1992. Two NRC staff members

.(P. Y._Chen, Michael McBrearty) and a NRC contractor (Kamal Bandyopodhyay) attended this initial session and later provided comments on the training course in a letter' dated August 28,1992. The NRC did not raise any objections to the, Page 6 of 17 l

1 I

i l

approach taught by SQUG in this c7;tse for applying Method A. Subsequent to this initial session of the course,11 additional NRC staff members and contractors attended other sessions of this course. Similarly, none of them raised objections to l

how SQUG was teaching use of GIP Method A.

t 4.

Proposed Interpretation Effectivelv Eliminates Method A An interpretation that Method A can be used only when calculated ISRS are less than 1.5 times the licensing basis ground response spectrum negates the value of ever having, developing or using Method A. With such an interpretation, Method A could only be used when it produces higher seismic demand than Method B.

Under this interpretation, the user would always be using Method B which is inconsistent with Method A's development and intended use as described in item 2 I

above.

k

), Page 7 of 17

1 Appendix 1 l

1 Excerpt from GlP-2, Reference 25 Results of BWR Trial Plant Review Seismic Demand Criteria, Page 8 of 17 i

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