ML20205K219

Forwards Response to NRC 990106 RAI Re Util Summary Rept on USI A-46 Program,Submitted 970527.Calculations & Corrected Table 11.1-1,encl

ML20205K219
Person / Time
Site:	North Anna
Issue date:	04/01/1999
From:	Hartz L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205K224	List: ML20205K219 ML20205K235 ML20205K244
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR 99-027, 99-27, NUDOCS 9904130047
Download: ML20205K219 (76)

Text

{{#Wiki_filter:Vincisir EuccTRIC AND POWER CmII%NY Racsiuos >, Vino:NiA 2326: April 1, 1999 U.S. Nuclear Regulatory Commission Serial No. 99-027 Attention: Document Control Desk NL&OS/ETS R0 Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen: VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION ON

SUMMARY

REPORT ON USI A-46 PROGRAM in a May 27,1997 letter (Serial No. 97-246), Virginia Electric and Power Company provided a summary report for the USI A-46 program at North Anna Power Station Units 1 and 2. During a December 22,1998 telephone conference call, the NRC requested additional information to complete their review of our program submittal. The NRC staffs questions were formally transmitted to Virginia Electric and Power Company in a January 6,1999 letter. The enclosure to this letter with Attachments A through D provide the response to the staffs questions. During the development of the responses, we identified administrative errors in Table 11.1-1 of the May 27,1997 submittal. Attachment B of the enclosure, which is Table 11.1-1 has been corrected to eliminate these errors. Please replace Table 11.1-1 in our May 27,1997 submittal with the revised Table 11.1-1 provided in Attachment B. If you need any additional information, please contact us. Very truly yours, v ( ,1 L. N. Hartz jI Vice President - Nuclear Engineering and Services ~ 9904130047 990401 DR ADOCK 0500 3 8 uuw i i

Commitments made in this letter: 1 A seismic housekeeping procedure, which will primarily address interaction concerns in safety-related areas, will be implemented by December 31,1999. Enclosure with Attachments A through D t cc: U.S. Nuclear Regulatory Commission Region 11 Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan i NRC Senior Resident inspector l North Anna Power Stat:on 1 i l

Enclosure Response to Request for AdditionalInformation (RAI) Regarding Virginia Power's Submittal dated May 27,1997 on.he Verification of Seismic Adequacy of Mechanical and Electrical Equipment Unresolved Safety Issue (USI) A-46 Virginia Electric and Power Company North Anna Units I and 2 L

e Question 1: With respect to the comparison of equipment seismic capacity and seismic demand, where Method A in Table 4-1 of GIP-2 was used, and the in-structure response spectra exceed 1.5xBounding Spectrum, identify the components involved, their locations, and the percentage of spectrum acceleration exceedance for these components. Also, provide a technical justification for the acceptability of using Method A for comparing the seismic capacity vs demand in these situations.

Response

l In response to the first part of the question, please note that, as described in section 6.2 (b) of the summary report, Method A of Table 4-1 of GIP-2 was used for capacity to demand comparisons of a handful of components. The components involved (class, description, mark number) are identified in the same section, i.e., page 6-2 of the summary report. The locations (buildings and elevations) of these components are listed in the Screening Verification Data Sheets (SVDS), contained in Appendix D of the summary report. The components and their locations are listed again in Table 1 below for convenience. The maximum percentage exceedances in spectral accelerations, occurring at any frequency in the amplified regions of the horizontal (east-west or north-south) conservative, design in-structure response spectra (ISRS), over the spectral acceleration of the Reference Spectrum (1.5 x Bounding Spectrum) at the same frequency, are provided in Table 1. All the spectra were plotted at 5% spectral damping for this comparison. Also, the percentage margins in the zero period acceleration (ZPA) of the ISRS, in comparison to the ZPA of the Reference Spectrum of GlP-2, are listed in this table. 1 Table 1 Comparison ofISRS With the Reference Spectrum of GIP-2, When Method A is Used l Equip Equipment Mark Building Elev-Maximum percentage Percentage arallable Class Number (s) ation exceedance at any margin in the ZPA of frequency in spectral the ISRS, compared acceleration of the to the ZPA of the ISRS, over the spectral Reference Spectrum acceleration of the (See Note I below) Reference Spectrum 7 1-MS-PCV-101 A, Main Steam Valve 306' 98 % 40% below GIP's 101B,101C llouse. Unit 1 Reference Spectrum 7 1-MS-TV 101 A,101B, Main Steam Valve 285' 37% $8% below GlP's 101C House, Unit i Reference Spectrum l 7 1-MS-TV-1 I I A,111B Main Steam Valve 27 7% 60 % below GlP's l (See Note 2 below) llouse, Unit 1 Reference Spectrum I 7 2-FW4CV-2478,2479, Service Building 286' 35 % 40% below GIP's 2488,2489,2498,2499 Reference Spectrum (See Note 2 below) 7 2-MS-TV 201 A,201B, Main Steam Valve 285' 44 % 36% below GlP's 201C House, Unit 2 Reference Spectrum 18 l-RC-LT-1470 Reactor Containment, 253' 2% 50% below GIP's Internal Structure Reference Spectrum 19 2-RC-TE-2410,2413, Reactor Containment, 256'- 8% 50% below GlP's 2420.2423,2430.2433 Internal Structure 4 Reference Spectrum Note 1: Note that the ISRS ZPAs are considerably lower than the Reference Spectrum ZPA for all these components. Note 2: Calculations have been performed for these valves which show that the stresses in the yoke, which is the weak link in the valves, are within the allowable stress value (<-6000 psi) for a seismic inertia load of 3g. 2

In response to the second part of the question, the technicaljustification for the use of GIP Method A is contained in Reference 5 of GIP-2, Part II, Section 10, Senior Seismic Review and Advisory Panel (SSRAP) report, "Use of Seismic Experience rad Test Data to Show Ruggedness of Equipment in Nuclear Power Plants," February 28, 1991. The description below reiterates the specific requirements in GlP-2 for the use of Method A and describes how these requirements are met at the North Anna Plant. l This response also provides additional technical justification for the use of Methad A for those specific situations at North Anna, where this method was used. This includes a description of many i of the conservatisms that exist in computed ISRS, a discussion which shows that the influence of exceedances in ISRS spectral peaks on line-mounted equipment is inconsequential, and a discussion l to point out that ther:is little safety significance of the differences between computed and actual building responses. GIP-2 Requirements and Ilow They Are Met for Use of Method A: GIP-2, Part II, Section 4.2 defines the requirements and limitation for use of GIP Method A. Table 4-1 and page 4-16 allow use of Method A if: The equipment is mounted in the nuclear plant at a elevation below about 40 feet above the effective grade, and The equipment, including its supports, has a fundamental natural frequency greater than about 8 IIz. As described on pages 4-16 and 4-17 of GIP-2, the above 8 IIz. limitation is not applicable to equipment mounted on a piping system. GlP-2 also limits the use of Method A to equipment mounted in typical nuclear power plant structures, i c., reinforced concrete frame and shear wall structures and heavily-braced steel frame structures. As described below, the above requirements are met for the use of GIP Method A to demonstrate that the seismic capacity of the components in Table 1 exceeds the seismic demand on these items. The effective grade for each of the structures in which Method A components are mounted was provided in Table 3-i., page 3-3 of the summary report. It can be seen that these components are mounted below about 40' above the effective grade. The 8 liz. limitation does not apply to any of the components shown in Table 1 for which Method A is used since all these components are mounted on piping systems. The buildings in which the Method A components are mounted are typical nuclear power plant structures. The Reactor Containment -Internal Structure mW Main Steam Valve llouse for Units I and 2 are reinforced concrete buildings supported on concrete mats or slabs at various elevations. Exterior walls and roofs of these structures are heavy concrete sections to resist missile. No unusual or plant-specific situations were identified which could cause the amplification factors for these buildings to be greater than those of typical nuclear power plant structures. The Service Building lower portion is also constructed of reinforced concrete and the upper portion has steel framing with monolithic concrete floor slabs. IIowever, the valves i 3

in this building for which Method A was used have also been analyzed for 3.0g seismic acceleration in each horizontal direction and the stresses in the yoke, which is the critical valve component, are within the allowable value. j Additional Technical Justification for the Use of Method A at North Anna Plant: This section describes measured building responses subjected to real earthquakes and discusses the reasons why there are conservatisms in calculated ISRS in general, and at North Anna in particular. The most significant sources of conservatism involved in the development ofISRS are listed. The i effect of spectral exceedances for line-mounted components and the safety significance of the ) expected differences between calculated ISRS and actual building response are also discussed. l it is noted that most of the generic information provided below was developed by the SQUG Contractors, and we understand that the NRC staff and SQUG representatives have discussed this topic extensively, including discussions held at one or more meetings. Conservatism in Calculated ISRS: The process of calculating in-structure response spectra (ISRS) is a complicated analytical exercise requiring several approximations, modeling assumptions and engineering judgments. As a result. the historical development of these ISRS has included a significant amount of conservatism, which has typically served two purposes:

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

2. It has reduced the costs associated with a very detailed state-of-the-art analysis, (which would attempt to trim out all the unnecessary conservatisms)

The fact that the conservative, design ISRS show amplifications greater than 1.5 with respect to Ground Response Spectrum is not surprising, nor does it negate the validity of Method A. In fact, as noted in the Senior Seismic Review and Advisory Panel (SSRAP) report, "Use of Seismic Experience and Test Data to Show Ruggedness of Equipment in Nuclear Power Plants", February 28,1991, it was even expected. Areas, which support the fact that ISRS are conservative at U.S. nuclear plants in general, and at North Anna Plant in particular, are discussed below. A. Measurements ofISRS in Actual Earthquakes l SSRAP developed the Method A response estimation technique based on their research of both actual e:rthquake measurements and on recent " median-centered" analyses. They reference (SSRAP report 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 ilumbolt Bay Nuclear Power Plant, and the Fukushima Nuclear Power Plant where amplifications over the ground response spectra do not exceed 1.5 for frequencies above about 6 liz. Other, more recent earthquake data from the Manzanillo Power Plant and Sicartsa Steel Mill in Mexico, as well as several facilities in California and Japan, has been 4

i 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. SQUG l knows of no new measured data that challenges GIP Method A. B. Calculations of Overall Conservatism in Typical ISRS Calculated ISRS have never been portrayed as representing the realistic expected response during an actual earthquake. As previously stated, ISRS typically contain many conservatisnw which make them unrealistically high. The primary reason for the development of Method A was to establish a more median-centered method of defining the l structural response without having to embark on costly new analyses of all the site buildings. It should be noted that even the most modern, state-of-the-art ISRS contain significant conservatisms, even those classified as " median-centered", are often very conservative. An NRC contractor (LLNL) concluded in a study for the NRC (NUREG/CR-1489) that typical calculated ISRS contain factors of conservatism of 1.5 to 8. Recent surveys by SQUG show similar levels of conservatism in calculated ISRS. It was the contention of SSRAP that the ISRS for nuclear structures (considering the 40' and 8 Hz. conditions) would be within about 1.5 times the ground response spectrum (GRS) if the plant were subjected to an actual earthquake. In deriving the Method A criteria they recognized 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. The basis SSRAP gave for 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 in the GIP does not appear to have a significant sensitivity to high frequencies (except possibly for relay chatter, which is addressed separately in the GIP). 1 C. Sources of Conservatism in the Development ofISRS at North Anna Plant The most significant sources of conservatism involved in the development of the ISRS for North Anna Plant include the following: Location ofInput Motion (variation from the free field input location) Ground Response Spectrum Shap Ground Motion Incoherence - Time History Simulation Clipping ofNarrow Peaks Structural Damping Frequency (Structure Modeling) Non-Linear Behavior (e.g., concrete cracking) Peak Broadening and Enveloping The degree of conservatism involved in each of these parameters is specific to the building being analyzed and to the floor level being considered. These conservatisms typically cannot be accurately quantified using conventional calculation techniques, since each parameter contributes to the overall response. Thus, it would take a considerable effort to 5 l l

quantify the exact conservatism inherent in the calculated ISRS at North Anna. However, on the qualitative bases presented below, it is easy to see the origins and levels of this conservatism. The following parameters are the source of the major portions of the excess conservatism: Location ofInput Motion: The defined location of the plant SSE (or DBE) is typically part of the design basis documentation. The SSE should typically be dermed at the ground surface in the free field as defmed in the current Standard Review Plan criteria. The dermed location of the North Anna Plant SSE is considered the ground surface in the free field. But for purposes of generating ISRS, North Anna conservatively dermed the input (currently identified as the " control point" location) at another location - at the building basemat. This conservatism can be significant depending on the specific plant / building configuration. For example, the reactor building is founded on rock at an elevation of 204.5 to 214.5 feet. Plant grade j elevation is 273 feet. In the construction of the reacur building, the site was excavated to ) the foundation level and the buildings constructea. Concrete backfill was placed in the annular space between the compressible materir.i and the vertical line of rock excavation to support the wall of the excavation with a compression ring structure. This fill extends upward to the underside of adjacent foundations or to the top of the rock surface at approximately elevation 246 feet. Virginia Power believes that while full lateral support is provided by the concrete backfill, it has been conservatively assumed for the purpose of the USI A-46 program that the effective grade elevation for the reactor building is at 214.5 feet. Essentially, very little amplification would be expected in the reactor building up to about 246 feet. On this basis, the elevations that may be considered to be within 40 feet of effective grade in the reactor building are up to 286 feet. However, we have used the 40' limitation up to about 255' It should be noted that the SSRAP was in agreement with this type of approach during the trial plant review conducted at a BWR and that the NRC participated in that review and concurred with it. Ground Response Spectrum Shape: The ground SSE dermed within the plant-licensing basis is the one that was used for the USI A-46 program at North Anna. Median-centered and other realistic hazard spectral shapes at the ground level show smaller amplification between peaks and ZPAs. The amount of conservatism is dependent on the spectral shapes at 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. Ground Motion Incoherence: l As has been documented in the EPRI seismic margin report (EPRI NP-6041), there can be an attenuation 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 to around 1.5. Recent studies have documented even 6

greater reduction factors. This ground motion incoherence is applicable to the rock sites structures at North Anna Plant and is particularly appropriate in the frequency range where i exceedances are noted. Time History Simulation: ISRS at North Anna Plant have been generated using time histories, which are intended to approximate the desired earthquake spectra. This process involves the generation of synthetic time l'istories whose response spectra envelop the SSE ground spectra in the horizontal and vertical directions. The amount of conservatism involved in the enveloping process has not been specifically quantified for North Anna Plant, but can range up to a i factor of 1.3 unless significant resources are applied to minimize the degree of enveloping. i Clipping of Narrow Peaks: The SSRAP report and the Generic Implementation Procedure (GIP) recommend procedures for adjusting narrow peaks to reflect two areas of conservatism:

1. Narrow peaks are not as highly amplified in real structures as are predicted by linear l

elast. .athematical models, and,

2. Narrow peaks in ISRS are not as damaging to equipment as are broad frequency input such as the Reference Spectrum.

The GIP procedure recommends an averaging technique over a frequency range of 10% of the peak frequency (e.g.,1 liz range for a 10 IIz peak frequency) using the unbroadened i ISRS. North Anna ISRS have narrow peaks and did not utilize the peak reduction methods l from the GIP. The conservatism involved has been shown to be in ' te range of 1.1 to 1.4 for l typical narrow peaks at several plants. We expect the conservatism for the peaks of the l North Anna ISRS to fall within this range of 1.1 to 1.4. There are several additional sources of conservatism (e.g., structural damping, structural modeling, structural / soil non-linearities, peak broadening and enveloping, etc.) which l add to the overall conservatism in the calculation ofISRS. These additional conservatisms, coupled with those described above, certainly reinforce the overall levels of conservatism in ISRS of between 1.5 and 8 which were referenced by SSRAP (LLNL Report NUREG/CR 1489), and explain why the conservative North Anna Plant ISRS produce exceedance beyond the 1.5 factor.

• The Spectral Exceedances llave Negligible Influence on Line-Mounted Components:

The few components for which Method A was used at North Anna are all line-mounted, and although the 811z limitation does not apply, they are expected to be rigid, i.e., have natural frequencies beyond the frequencies that have predominant energy content in earthquakes. The exceedances in the ISRS over the Reference Spectra occur only in narrow frequency ranges and as shown in Table 1, the ISRS ZPAs are considerably lower than the ZPA of the Reference Spectrum for all these components. The piping systems themselves may be flexible (as they 7

might well have been in the earthquake experience plants), therefore, the required response spectra (RRS) at the location of these components within the piping systems may show peaks that are higher than the 152S at the piping anchor locations. These RRS peaks are expected to be narrow band due to filtering by piping systems, as discussed in IEEE Std. 344-1987 and in other studies. Ilowever, the peaks will quite likely not be in the applicable frequency ranges for the seismic acceptance of these components, consistent with the judgment that the components are rigid. It is expected that the ZPAs of the RRS at the component location will not be amplified as much as the peaks. Therefore, based on the ZPA comparisons of the ISRS with the ZPA of the Reference Spectrum shown in Table 1, which are more meaningful, the use of Method A is considered appropriate. It should also be noted that: Maximum seismic acceleration responses at valves and other in-line component locations from response spectrum analyses of piping systems (which is the same as the ZPA of the RRS at their location) at North Anna are typically low, i.e., less than about 2g in ny direction for the vast majority of components reviewed, and, - More than two hundred valves and many other line-mounted items of equipment were walked down in the USI A-46 program at North Anna. Several valves were analyzed using location-specific seismic accelerations obtained from piping analyses or with generically conservative acceleration value of 3g in each horizontal direction. However, none was determined to be inadequate, or required modification / replacement due to seismic inertial forces. Not a Significant Safety Issue: The expected differences between calculated ISRS and actual building response do not represent a significant safety question. The components in Table 1 are in classes 7,18 and 19. There are no known failures of components in these classes in the earthquake experience database because of high seismic inertial forces alone. The lessons learned from review of hundreds of items of equipment at various sites that have experienced earthquakes which were significantly larger than those for Eastern U.S. nuclear plants are that missing anchorage, seismic interaction hazards, and certain equipment-specific weaknesses (incorporated into the GIP caveats) were the seismic vulnerabilities which cause equipment damage. These areas are appropriately addressed in the GIP and in our implementation of USI A-46 program at North Anna. 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 susceptible to seismic damage." [NUREG-1211, page 16] GlP-2 (page 4-11) cites the SSRAP report as the basis for the Bounding Spectrum, which is used in Method A, and requires users to read and understand it. The SSRAP report clearly explains the limitations and conditions, which appear on page 4-16 of the GIP. SSRAP's report states: l 8

1 "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 possibly at the ftmdamental j frequen.cy of the building where higher amplifications occur when such a frequency is less ) than about 6 ilz."

== Conclusions:== The discussion above leads to several conclusions: The technical justification for the acceptability of using Method A of GIP-2 for comparing equipment seismic capacity vs. demand is contained in the SSRAP report. t The GIP-2 requirements for using Method A are met for the equipment evaluated at the North Anna Plant. The results from actual measured ISRS on " nuclear type" structures support the 1.5 response levels advocated within Method A. Qualitative assessments of the conservatism inherent within the methods utilized to calculate ISRS have been provided. These conservatisms are typically quite sigi ificant (as has been independently verified by median / modern assessments such as the LLNL study) and can/will result in ISRS that show amplifications well beyond the 1.5 factor applied to free-field or ground response spectrum. Virginia Power feels strongly that the specific exceedances beyond the 1.5 factor at North Anna are primarily due to the conservatisms inherent in the ISRS methods and not due to " unusual, plant-specific situations." Therefore, the application of Method A to structures at North Anna is appropriate and valid. For line mounted equipment, the exceedances in ISRS over the Reference Spectrum occur only in narrow frequency ranges of the amplified regions of the spectra. These excursions are not as significant as the ZPAs of the ISRS, which are considerably lower in comparison to the ZPA of the Reference Spectrum for all the components where Method A was used. No issue with inertia foru has been found for more than two hundred valves and many other line-mounted pieces of equipment evaluated at North Anna Plant during the USI A-46 review. Several valves, including some of those where Method A was initially used, were analyzed for a 3g seismic loading or using location specific seismic accelerations. None was determined to be inadequate or required modification / replacement. There is little safety significance in the expected differences between calculated ISRS and actual building response. The largest safety improvements are provided by appropriately reviewing equipment anchorage, seismic interaction hazards, and certain equipment-specific weaknesses where seismic vulnerabilities have caused equipment damage in real earthquakes. Reviews of these areas were a primary focus of the SQUG GIP process, and Virginia Power's implementation of the GIP at North Anna Power Plant, Units I and 2 has resulted in significant seismic safety enhancements. 9

Question 2: Resumes and training records of seismic capacity engineers, lead relay reviewers, and independent reviewers are provided in Appendix E of the referenced submittal. Provide a summary description of background and qualifications for Messrs. P. Ilayes, D. Jacobs and D. Werder, who contributed to the development of Safe Shutdown Equipment List (SSEL). l

Response

l A description of background and qualifications of Mr. P. Hayes, Ms. D. Jacobs and Mr. D. Werder, who contributed to the development of Safe Shutdown Equipment List (SSEL), is provided in Attachment A. Question 3: In Appendix D of the referenced report, the discussion on equipment that did not meet the specific wording of the caveat, but were judged to meet the intent of the caveat, was not sufficiently detailed to enable our understanding of the basis of such determination. For equipment items that met the intent but not the wording of the caveats, provide a brief description regarding the nature of the deviation and how the intent of the caveat was met in each case.

Response

1 There were several instances during the walkdown and review of safe-shutdown equipment where the GIP screening criteria were not met, or not invoked. In some cases, the existing qualification documentation was relied upon (e.g., analyses, seismic qualification per IEEE Std. 344-1975). In other cases, outliers from ae GIP criteria were or are being resolved via acceptable analytical methods or by performing a modification to the equipment item or its anchorage. These situations are discussed in the reference summary report. In a few cases, the equipment item was considered acceptable although the specific wording of a caveat in the screening evaluation worksheet (SEWS) was not met. However, the seismic review teams judged that the intent of the caveat was met and documented the basis of their judgment in the respective SEWS packages. Items in this category were listed in the " Notes" associated with the Screening Verification Data Sheets (SVDS) that are included in Appendix D of the summary report. These equipment items, the nature of the deviation, and a brief description of how the intent of the caveat was met are discussed in Table 2 on the following page. 1 10 i

i Table 2 Equipment Items that Meet the Intent Hut Not the Wording of Caveats of GIP-2 Equip. Equipment Mark Huilding Elevation Description of review which indicates that the Class Number (s) intent, but not the wording of caveat is met 2 1-EE-SS-03 Auxiliary 280' An attached junction box on top of each of these 2-EE-SS-03 Building cabinets was estimated to weigh about 164 lb., whereas Bounding Spectrum caveat no. 5 allows an attached weight of less than about 100 lb. A further evaluation indicated that the effective weight would be smaller because of the cutouts in the box, part of the weight being taken by large, well-supported conduits, and the box extending beyond ' abNt assembly. Thus it was judged that the i ot' the caveat is met. It is also noted that the peer r.. ver, during his walkdown of these components, agreed with this judgment of the Seismic Review Team. 2 1-EE-SS-03,04 Auxiliary 280' The Conservative, Design, in-Structure Responsr 2-EE-SS-03,04 Building Spectrum (ISRS) exceeds the capacity (1.5 Bounding Spectrum) by less than 10% in the 2 to 3.5 liz. range only. This was considered acceptable because the exceedance is small, and in a low frequency band not expected to be applicable since the equipment natural frequency was judged by the Seismic Review 'n eam to be above 3.5 liz. 4 1-EE-ST lill,1J1 Auxiliary 280' The same situation and judgment is applicable, as for 2-EE-ST-2111,2J l Building items 1-EE-SS-03,04 and 2-EE-SS-03,04 above. J 7 1-MS-PCV-101 A, Main Steam 306' These items are located about 44.5' above the 101B,101C Valve liouse, effective grade. Use of Method A of GIP-2 was Unit I considered acceptable for capacity / demand comparison, although about 40' is required. It was judged that for these line-mounted components, a slight deviation is within the approximate bounds of Method A. The ISRS peaks exceed the Reference Spectrum of the GIP in a narrow frequency band, however, the ZPA of the ISRS was 40% below the Reference Spectrum. The adequacy of these components is further discussed in response to request no. I of this RAl. 8A l-SW-MOV-101 A, Quench 256' to The demand (Conservative, Design, ISRS) exceeds 101B,101C,101D, Spray Pump 266' capacity (1.5 x Bounding Spectrum) by less than 10% i 2-SW-MOV-201 A, llouse in the 10.5 to 11.5 liz. range only. This is judged 201B,201C,20lD acceptable because (a) the exceedance is small, and (b) the slight excursion in this frequency band would not be significant since line-mounted valves are generally rigid and the ISRS ZPA is more relevant. Also, the valve's seismic input is below 40' from the effective grade of this structure and the Bounding Spectrum envelops the Ground Response Spectrum. I1

fable 2 continued Equip. Equipment Mark Building Elevation Description of review which indicates that the Class Number (s) intent, but not the wording of caveat is met 18 l-CH-LT-1106 Auxiliary 289' 1he demand (Censervative, Design, ISRS) exceeds Building capacity (1.5 x Bounding Spectrum) by about 14% in the 2 to 3.5 liz. range only. This is judged rn ceptable since tne exceedance is small, and in a low frequency band not expected to be applicable. 'Ihe transmitter support was determined to be rigid, i.e., no amplification of the floor motion would occur at the transmitter mounting location, whereas similar line-mounted, flexibly supported transmitters were found without damage in the seismic exp:rience database. 18 l-RC-PT 1445 Reactor 263' 1he demand (Conservative, Design, ISRS) exceeds Containment, capacity (1.5 x Bounding Spectrum) by about 19% in Internal the 4 to 6.5 liz. range only. This is judged acceptable Structure since the transmitters in a vibration test did not exhibit any significant resonance below 6011z. 19 2-RC-TE-2410, Reactor 256'-4" These components are located about 42' above the 2413,2420,2423, Containment, effective grade. Use of Method A of GIP-2 was 2430,2433 Internal considered acceptable for the capacity / demand Structure comparison, although about 40' is required. It was judged that for these line-mounted components, a slight deviation is within the approximate bounds of Method A. Also, the effective grade was conservatively defined for this structure. The ISRS exceed Reference Spectrism of the GIP l'v only 8% in a narrow band, but the ZPA of ISRS was 50% below the Reference Spectrum. Additional justification regarding the acceptability of these components is provided in response to request no. I of this RAI. 20 2-El-CB-1 I S A Auxiliary 265' The demand (Conservative, Design, ISRS) exceeds 2-El-CB-116A Building capacity (1.5 x Bounding Spectrum) by less than 5% in a very narrou frequency band near 2 liz. only. This was judged acceptable because the exceedance is small. Also, i. is in a frequency band which is not expected to be applicable, i.e., the cabinets are well anchored and were judged to have their natural frequency above 211z. 20 1-EP-CB-10C, l'0F Auxiliary 284' The demand (Conservative, Design, in-Structure 2-EP-CB-10C,10F Building Re:ponse Spectrum) exceeds the capacity (1.5 x Bounding Spectrum) by less, than 10% in the 2 to '.5 Hz. range only. This was judged acceptable since the exceedance is small, and in a low frequency band not expected to be applicable because these are wall-mounted panels and were judged to be rigid by the Seismic Review Team.

== Conclusion:== It is clear from Table 2 above that the deviations from the specific wording of the caveats were very minor and insignificant. Most of the deviations are related to slight exceedances in the spectral peaks over the Reference Spectrum of the GIP in narrow frequency bands, which wasjudged to be 12 1-- -

acceptable. In addition, the response to Request No.1 of this RAI provides a qualitative discussion regarding the potential conservatisms in the In-Structure Response Spectra at North Anna Plant. In summary, for each of the deviations from the GIP caveats, there was a technical basis forjudging that the intent of the caveat is indeed met, and for concluding that the component is seismically adequate. Question 4: On page 6-4, Section 6.5, " Seismic Interaction" of the referenced report, you stated that seismic interaction concerns that were not easily corrected were identified on the SEWS and were reviewed further. You also stated that several of these concerns were subsequently resolved by plant modification. Provide five examples of the significant interactions identified during the walkdown and discuss how these interaction concerns were resolved.

Response

During the walkdown of more than 1000 items of equipment (and many rule-of-the-box items) by our Seismic Review Teams (SRT) for the resolution of USI A-46 at the North Anna Plant, several interaction concerns were noted. Equipment that could not be screened because of an interaction concern was classified as outliers. Section 11, Table 11.1-1 of the summary report lists the major resolved outliers, including those due to interaction concerns. This table provides a description of each major outlier and how the resolution was accomplished. A summary of the interaction caveat of the screening evaluation worksheet (SEWS) is also provided in the " Interaction OK?" column of the Screening Verification Data Sheet (SVDS) in Appendix D of the nference summary report for each item of equipment. Several of the notes labeled O/R (Outlier ' ahed)in this Appendix are related to the resolution ofinteraction issues. Additional discussion of five significant interaction related outliers, and how they were resolved, is provided below. It is noted tint: (a) while most of the interaction related outliers were resolved by making modifications to the components or their anchorages, a few were resolved by performing analytical evaluations or by reviewing existing documentation, and (b) a few outliers are currently outstanding and are in the process of being resolved in accordance with the schedule committed in the summary report. A seismic housekeeping procedure, which will primarily address interaction concerns in safety-related areas, is being prepared and its implementation at the Station is expected sometime in 1999. Examples of significant interaction issues:

1. The most widely noticed interaction concern was the potential of side-to-side or front-to-back impact of cabinets containing essential relays, because of a very small or no gap with an adjacent cabinet or commodity. The concern was that relays within the cabinet might chatter due to the impact. Nineteen Unit I and twenty-four Unit 2 cabinets including Motor Control Centers, relay racks and process cabinets in the Safe Shutdown Equipment List (SSEL) which contain essential relays were connected to adjacent cabinets or walls to resolve this concern via implementation of Design Changes (DCP) at the Station.

Cabinets connected to resolve this outlier are listed in Table 11.1-1 and also in the SVDS, Appendix D, with description provided in the SVDS Notes. It is noted that item 13 of Table 11.1-1 in the summary report contained minor omission; and typographical errors regarding 13 1

l identification of cabinets containing essential relays that were connected to adjacent cabinets. Part of this table is revised (item 9 revised, and item 13 revised / split into items 13 and 14) and Revision 1 is enclosed in Attachment B. Page Il-2 of the summary report should be replaced with the revised page Il-2 contained in Attachment B.

2. The operator for Pressurizer Relief Valve 2-RC-PCV-2455C was approximately 5/8" away from a pipe support (mark no. 2-RC-IISS-144). This wa3 judged potentially unacceptable by the seismic review team. The pipe support was relocated such that a gap of at least 2" exists between the support and the valve operator, via a design change implemented at the station.

{

3. The Unit 2 ASCO solenoid operated pilot valves (2-RC-SOV-2455C-1 & 2) are mounted to a support, which is attached to the adjacent concrete wall.

Valve 2-RC-SOV-2455C-1 is connected to a condulet box, which was very close to the upper diaphragm casing of the operator for pressurizer relief valve 2-RC-PCV-2455C. The solenoid casings were rotated and the flexible conduit restrained with a cable tie, to provide a minimum gap of one and a half inches between the valve operator and the solenoid valves or associated condulet box, via a design change implemented at the Station.

4. A level transmitter (1-FW-LT-1487) was mounted in a rack, and the rack contained a small electrical box to v!hich a redundant, unrestrained, flexible conduit was attached. The conduit had the potential to impact the trans'nitter and was removed by implementing a work order issued to the Station.

5. In some cases, the interaction outlier was resolved by reviewing the existing documentation or by performing an evaluation, rather than a modification. Examples in this category are as follows:

It was unclear from the walkdown whether the side-to-side restraint capacity of the circuit breakers in the Reactor Trip Breaker Cabinets (l(2)-EE-BKR-BYA, BYB, RTA, RTB) was adequate. liowever, the existing documentation (Westinghouse test report WCAP-7817) was reviewed, and it was concluded that the breaker's functional capacity is adequate for the location specific seismic required response spectra (RRS), based on the functional verification performed during testing. Therefore, no modification was necessary for the breakers. Cabinets 1-EP-CB-28G,11, and J, which contain essential relays, were found to have a l minimum side-to-side gap of 3/16" between them. The cabinets are well anchored, and a l calculation showed that with out of phase movement, tne maximum relative displacement between the adjacent cabinets during a design basis seismic event would be less than 1/32". Thus, it was concluded that no impact of the cabinets or relay chatter would occur and a modification was not necessary. 14

Question 5: In reference to Section 8.2.1 of the referenced report, for raceways and cable trays that are outside of the experience data, explain what criteria were used to establish their seismic adequacy, and provide a sample of such configurations.

Response

In one location, a 2 %" conduit coming from containment recirculation fan 1-IIV-1 A had a larger span (18'-0") than acceptable in GIP-2 (16'-0"). T1:e conduit was in the basement of the containment, hence it was judged acceptable due to the low seismic responses at this elevation. In several other locations, as identified in the referenced report Section 11.1.4, conduit spans were found to exceed the GIP-2 requirements. Additional supports were installed in these cases to reduce the span and comply with the GIP-2 requirement. There were no other instances where the cable trays a*,d conduits were outside the experience data. Question 6: In reference to Sectit n rO.3 of the referenced report, provide a sample of the calculations for the limited analytical reviews (LAR) with drawings or sketches for examples 1,6,10,11,13 and 14.

Response

Limited Analytical Reviews (LAR) with sketches for examples 1,6,10,11,13, and 14 are provided in Attachment C. l Question 7: You discussed seismic spatialinteraction on page 8-4. You stated that the seismic interactions were verified during the walkdown. Provide a quantitative discussion as to how l you estimated or calculated cable tray and support displacements (sway) during an carthquake, especially for a long support hung at the ceiling.

Response

l Peak displacement response for cable trays are approximated as a single support with a dead load I tributary span. Using the simple beam theory and including a frequency adjustment term for pendulum effects, the natural frequency m of a rod hanger trapeze support can be estimated as: o e = 1/2n ((K3+ K,)/M )" n Where K is beam bending stiffness of two rods in the hanger trapeze support, and K is the 3 p equivalent stiffness correction term to adjust the frequency for pendulum response. M is the mass i due to tributary dead load weight on the support. If there are multiple tiers, then M will be the tributary dead load of all tiers on the support. M = We / g, where W = dead load tributary weight i i 3 K = 2 x (12El/L ) 3 K is derived from the pendulma frequency af a suspended support as follows: p 15

o> pend = 1/2x ( g/L)U2 = 1/2n ( K /M )" p Solving for K gives p K = We / L p 3 Kb + K = 24 El / L + W / L = Wo / A p i Where Wo is the total lateral load on the support due to earthquake of all tiers and L is the rod hanger length in inches to the first tier from ceiling, including multiple tiers, if any. I is rod root moment ofinertia and E = 29,000,000 psi. From the above equation, the deflection A of the cable tray support is calculated. This deflection is compared with the optimal deflection as follows: 2 A = d f L / 6ER y Where o is the ductility demand, f is the apparent yield stress, and R is rod root radius. y Example 10, provided in Attachment C in response to Request No. 6 of this RAI, shows the cable tray support deflection evaluation. This is a conservative estimate, as the top brace to the cable tray is neglected in the calculation. In this example, a is conservatively assumed to be 1.7 and fy is taken as 90 ksi per EPRI NP-7152-D, to calculate the optimal deflection. The displacement calculated was 0.39", compared to the optimal displacement of 0.49". Further, this displacement is small and no interaction will occur. ( Question 8: In Section 10 "Significant and Programmatic Deviation from the GIP" (Page 10-1 of the referenced report), you stated that the Emergency Condensate Storage Tanks (1/2-CN-TK-1) and the Refueling Water Storage Tanks (1/2-QS-TK-1) were analyzed using the DOE-sponsored methodology in BNL 52361. Our preliminary review of the October 1995 version of report BNL 52361, indicates that your tanks may not be within the scope of the reports since the DOE waste tanks are buried and the report is for double walled shells or concrete outer casings. The report states that application to other types of tank-structures and the suitability of adjusted versions of these concepts to other structural types will be addressed in a future version of this document. Provide the latest revision that addresses the type of tanks at North Anna plant or alternatively, discuss how the tanks are within the scope of the report. Provide the evaluations of the Emergency Condensate Storage Tanks and the Refueling Water Storage Tanks and any other outlier tanks that were resolved using the BNL 52361 methodology together with a technical justification for the methodologies used in the report. Also, the use of the HCLPF concept for the resolution of USI A-46 is not considered appropriate. If the tank evaluations were based on complex computer codes, indicate how these codes were verified and benchmarked. 16 l

Response

In response to the first part of this request, please note that Report BNL 52361, " Seismic Design and Evaluation Guidelines for the Department of Energy fligh-Level Waste Storage Tanks and Appurtenances" October 1995, was developed primarily to provide the seismic design and evaluation guidelines for underground high-level waste storage tanks. However, in Chapter 5 of this report, an approach for the evaluation of seismic capacities of anchored and unanchored above-grormd liquid storage tanks is also presented. Primarily two criteria, related to permissible uplift displacement and fluid hold-down forces, provided in Sections 5.11 and 5.12 respectively of i Chapter 5 of this report, were used to perform USI A-46 seismic evaluations of the unanchored Emergency Condensate Storage Tanks (ECST) and the anchored Refueling Water Storage Tanks (RWST). As indicated in the referenced summary report, these are the only tanks where the methodology from BNL 52361 report was used to resolve the USI A-46 outliers. We are not aware whether a later version or revision of this document has been published, and the October 1995 version of the report was used in our evaluations. However, in addition to the BNL methodology, elements of methods and criteria from other references were also used in the calculations to resolve the outlier conditions for these tanks. Per your request, the USI A-46 seismic capacity calculations of these two tanks, originally prepared by our consultant EQE International but revised by our engineers, are provided in Attachment D. The revision only clarifies some text and adds, for completeness, certain computations which were referenced in the original calculation. The detailed technical justification of the methodology used for these tanks is provided in these calculations and in the respective references used within these calculations. The concept of High-Confidence-of-Low-Probability-of-Failure (HCLPF) was not used in conducting USI A-46 seismic evaluations for these tanks. We have prepared separate calculations for these tanks for the USI A-46 and the Individual Plant Examination of External Event (IPEEE)- seismic programs, although certain elements from EPRI Report NP-6041 were used in the USI A-46 evaluations to resolve the outlier conditions, as follows. In the mse sment of the overturning capacity of these tanks, consideration of the fluid hold-down forces v is included, which takes into account a limited amount of uplift at the base of the tank. The holidown forces resvung from fluid pressure acting on the tank bottom contribute significantly to the overturning capacity of the tank. It is recognized that the GIP methodology does not have any pros ision or guidance to allow the consideration of the fluid hold-down forces. Further, there are no guidelines in the GIP for the evaluation of unanchored tanks. Therefore, the fluid hold-down forces for both these tanks were estimated using the method discussed in EPRI-NP-6041, Appendix H, and in BNL 52361. In this method, a slightly conservative linear expression between the fluid hold-down forces and uplift displacement is used and an iterative procedure is utilized. In addition, the RWST tank is anchored at the base, and a yield line analysis of the chair of the anchor was performed to estimate the limiting anchor bolt capacity, concurrent with a limited amount of uplift at the base of the tank. l It is noted that in Sections 10 and 11 of the summary report, the HCLPF capacity values were only provided to show additional evidence of margin for the ECST and RWST tanks. These tanks have HCLPF capacities, expressed in terms of peak ground acceleration (pga), of 1.33xSSE and 1.5xSSE respectively, when evaluated for the Review Level Earthquake (RLE). These capacities are based on separate evaluations of these tanks performed under the IPEEE-seismic program. Note that the median capacities of these tanks are expected to be more than twice as great as the HCLPF capacities. 17

4 1 In response to the last part of this request, the USI A-46 tank evaluations were not based on complex computer codes. Templates from the MATilCAD software were used to celculate some of the responses, based on the equations provided in the applicable references, and the results were reviewed. Therefore no verification or benchmarking of any computer code is required. The MATIICAD based analyses are included in the two calculations being provided to you in Attachment D. Question 9: In reference to Items (g) and (h) on page 10-2 of the referenced report, you stated that certain valves (i.e.,1-GN-PCV-125A-1,125A-2,125B-1,1258-2,1,2-RC-SOV-1455C-1, etc.) and transmitters (i.e.,1-FW-LT-1475,1484,1485,1486,1487,1494 etc.) were qualified using previously completed efforts documented in Design Change Packages (DCP) because the DCP indicated that these valves and transmitters are qualified to IEEE 344-1975 l requirements. Please indicate if these valves and transmitters (or appropriate surrogates) were actually tested and/or analyzed in accordance with the requirements of IEEE 344-1975 standard and whether you have verified their seismic qualification documentation. l

Response

Most of the components listed under items (g) and (h) on page 10-2 of the referenced summary report were installed via design changes (DCP) implemented at the Station. In response to your l request, we reviewed the available seismic qualification documentation for these components to i determine whether they were actually tested or analyzed in accordance with the requirements of IEEE Standard 344-1975. In a few cases, it was easier to use an alternate approach, i.e., comparison of In-Structure Response Spectra (ISRS) with GIP's Generic Equipment Ruggedness Spectrum (GERS) to show seismic adequacy, and we did not need to locate the original seismic qualification documentation to show compliance with the IEEE Standard. In each case it was con:luded that the component was seismically adequate. Table 3 below provides a summary of the seismic qualification status of these components. TABLE 3 Seismic Qualification Status of Components Listed in Items (g) and (h) of Page 10-2 Equipment Mark No. Equip. Vendor / Design Change Description of Seismic Class Model (DCP) Number Qualification Status 1-GN-PCV-125A-1, A-2 07 FISHER / DCP 78-44 The valves are small and supported to 1-GN PCV-1258-1, B-2 1301F-23, an adjacent wall. The DCP states 2-15200-these valves as seismic Category I but 157 the actual plant specific qualification documentation couk' not be easily retrieved. However, similar to the original evaluation performed for the identical Unit 2 valves, these valves are found acceptable by comparing the In-Structure Response Spectra (ISRS) with the Generic Equipment Ruggedness Spectrum (GERS test data) uvided in the GIP and meeting te GERS caveats. I8

TABLE 3 (Continued) Seismic Qualification Status of Components Listed in Items (g) and (h) of Page 10-2 l Equipment Mark No. Equip. Vendor / Design Change Description of Seismic l Class Model (DCP) Number Qualification Status 1 RC-SOV-1455C-1,1455C-2 088 ASCO/ GlP-2 GERS for ASCO Type 206-1-RC-SOV-1456-1,1456-2 NP8316 381 envelop the Conservative In-2-RC-SOV-2455C-1,2455C-2 Structure

Response

Spectrum 2-RC-SOV-2456-1,2456-2 (ISRS), and the GERS caveats were met. In addition, the valves were qualified via single frequency testing to a level of 10g in ASCO test report no. AQS21678/fR, Rev. A, per the suggestions in IEEE Std. 344-75. I-RC-SOV-1455C-3,1456-3 08B VALCOR DCF 78-44 (For The valves are wall-mounted. Per 2-RC-SOV-2455C-3,2456-3 Series V573 Unit i valves) plant documentation, Valcor Eng. 3-way valves Corp's seismic qualification report no. MR573-5221 for these valves was reviewed and approved by Stone & Webster on 7-14-78. This report could not be readily located, however, the valves are determined to meet the GIP-2 requirements since the GERS envelop the ISRS and the GERS caveats are met. I-SS-TV-107A,109A 08B VALCOR DCP 80-S32 Per vendor report nos. QR-526 and V526568319 QR-52600-5940-2, the valves are 2-way valves qualified by testing to a minimum acceleration of 4.5g and meet the requirements ofIEEE Std. 344-75. 1-FW-LT-1474, 1475, 1476, 18 Rosemount DCP 81S-08A Q.nalified by testing in accordance j 1484,1485,1486,1494,1495, 1153DD4PA with IEEE Std. 344-75 per 1496 Rosemount Qualification Report No. D8300040. Random, multi- ) frequency tests were performed with Test Spectra peaks >20g and ZPA of 6g at 1% damping. 1-FW-LT-1487 18 Rosemount DCP 82-14A Qualified by testing in accordance 1153DD5PA with IEEE Std. 344-75 per Rosemount Qualification Report Nos.108025 and D8300040 (see details above). 1-QS-LT-100A, B, C, D 18 Rosemount DCP 84-47 Qualified by testing in accordance ll53DD5PA with IEEE Std. 344-75 per Rosemount Qualification Report No. D8300040 (see details above). I 19

Question 10: In reference to item (c) on page 10-1, provide a description and identification of the new modified / replaced pumps and discuss how these pumps met the seismic adequacy criteria in GIP-2.

Response

The Service Water Pumps (1/2-SW-P-1 A, IB) were walked down and reviewed in accordance with the GIP-2 criteria by an in-house seismic review team. The pumps are supplied by Johnston Pump Company. Although the impeller shaft is more than 20', the pumps have lateral supports. Other than a couple ofissues for pump 2-SW-P-1 A (a missing nut and an interaction concern) which were found by the seismic review team during the walkdown and resolved expeditiously, all four pumps were considered to be seismically adequate. A calculation was performed for the baseplate and support anchorage and the seismic evaluation was documented in the Screening Evaluation Work Sheet (SEWS) package in accordance with the GIP-2 requirements. Subsequent to the walkdowns by the seismic review team, the Station identified some concerns for pump 2-SW-P-1 A. The pump was placed in the alert range for vibration, based on periodic testing. Also, the pump performance had deteriorated. The other three pumps also exhibited some level of performance degradation. Therefore, a design change (DC No. 95-015) was initiated at the Station j to refurbish all four Service Water Pumps. As part of the design change implementation, pump 2- { SW-P-1 A was replaced by a similar, completely new pump from Johnston Pump Company in December 1995. A Specification (No. NAP-0042) was prepared and issued to the vendor to replace the bowl assemblies of the other three Service Water Pumps. These bowl assemblies have now been replaced for all three pumps via the design change process. The vendor, Johnston Pump Company, performed a seismic analysis of the new/ refurbished pumps to the plant design basis requirements. This analysis included all the critical pump components, motor support, baseplate and the pump support anchorage. A detailed finite element model of the pump was prepared and analyzed using the response spectrum method. Applicable floor response spectra were used in the analysis. The stress and displacement responses in the pump components and supports were found to be within acceptable limits for the design basis earthquake and operating loads. Subsequent to the vendor's analysis, it was discovered that the pumps contained A307 material bolts at the pump cohunn flange connections, which did not meet the design basis allowable value. A Station Deviation Report (N-96-2359) was written to address this deviation condition. In addition, the vendor's analysis did not address the strength of the pump lateral support. To resolve these concerns, Virginia Power performed an operability analysis of the pump with the A307 bolts for design basis seismic and operating loads considered together, taking guidance from NRC Generic Letter 91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability". This analysis, which was similar to the vendor's analysis, was performed using STARDGE computer code. The pump was verified operable following a seismic event. The pump lateral supports were analyzed and determined to be structurauy adequate. Subsequently, the A307 bolts were replaced by high strength A193 B7 bolts in the column flange connections of all four pumps to meet the design basis allowable stress criteria. To allow more operating flexibility in the Service Water System, it was desired that the pump components / supports, including the A193 bolts in the flange connections and the lateral supports, should withstand a pressure of 125 psi (the vendor's and Virginia Power's previous analyses used 90 psi) together with DBE seismic loads. Therefore, Virginia Power prepared a supplemantal calculation, which demonstrates that the pump 20

1 l components, including the A193 bolts and the pump supports, will withstand 125 psi pressure l together with deadweight, nozzle loads and DBE seismic loads and all responses are within the l design basis allowable values. l l In conclusion, it is emphasized that in addition to replacing pump 2-SW-P-1 A and refurbishing the other three pumps, the A307 pump-column flange bolts were replaced with higher strength A193 l bolts as a safety enhancement. The seismic qualification of the new/ refurbished Service Water Pumps has been demonstrated via walkdowns which addressed GIP-2 caveats and interaction i issues, together with vendor's and in-house analyses of the pump, its lateral supports and its i anchorage to meet North Anna's design basis and the USI A-46 requirements. Some of the analyses l and evaluations performed for these pumps are well beyond the GIP-2 requirements and although this issue is listed in the summary report as a minor exception from the GIP, we believe that the new/ refurbished pumps satisfy the seismic adequacy criteria in GIP-2 as well as the design basis criteria of North Anna Plant. Question 11: Provide the current resolution status for the identified outliers and plant installation deficiencies, including, but not limited to, those listed in Table 11.2-1 of the referenced submittal.

Response

The summary report that was submitted to the Staff in May 1997 identified unresolved outliers related to electrical / mechanical equipment, and cable and conduit raceway systems. The current l status of the ri: solution of these outliers and/or plant installation deficiencies is provided below: Status of Outliers for Mechanical / Electrical Equipment: Table 11.2-1, List of Outstanding Outliers and Issues, in the summary report submitted to the staffin May 1997, contained a list ofissues that remained open at the time of the submittal. The current resolution status of each item is shown in bold letters in the following Updated Table 11.2-1. Of the nine issues listed in this table (related to housekeeping, control room ceiling i l concern, and 23 components) which were unresolved at the time of the submittal, three have been completely resolved, five are in progress, and one remains to be addressed. Status of Outliers for Cable and Condait Raceway Systems: l Several condulet covers were miss.1 g in the cable vault and tunnel area and the Service Building. In addition, cables in several locations were hanging loose from the cable tray in l these buildings. Work Orders (WO) were issued to the Station to correct these deficiencies and all the tasks have been satisfactorily completed. As committed in the summary report, all the remaining outliers and installation deficiencies are planned to be resolved no later than the end of the North Anna Unit 1 outage, currently scheduled to commence in March 2000. 21

i Updated Table 11.2-1 ~ Resolution Status of Outstanding Outliers for Mechanical / Electrical Equipment ITEM DESCRIPTION OF EQUIPMENT PREVIOUS RESOLUTION PLAN / NO. OUTLIER / ISSUE MARK NO. CURRENT STATUS (BOLD) { l 1 llousekeeping Issue: Movable CO in the vicinity of This issue is being resolved via an 2 fire extinguisher carts on wheels, severcl Engineering Transmittal No. CEM unanchored tables, trash cans etc. are components - 0032 to the Station. a few inches away from SSEL Refer Appendix cabinets housing essential relays in D, SVDS note A Station seismic honsekeeping the ESGR room and IRR room. O/ UNI procedure has been prepared, and its implementation is expected in 1999. The resolution of this outlier / issue is in progress. 2 Control Room (CR) ceiling panels CR cabinets CR ceiling panels will be reviewed. If may require reinforcement to prevent containing required, they will be tied together with their falling during an SSE essential relays; clips to prevent their falling during SSE. also CR This issue remains to be resolved. d 3 (a) Some of the friction clips used in 1 EE-ST-lil further inspection and review will be transformer cabinet anchorages have 1-EE-ST-lill performed to verify anchorage adequacy a small gap with the base channel. 1-EE-ST-1 J and/or load path. The gap between ne coils are not laterally restrained 1-EE-ST-lJI friction clips and base channel will be at the top, however, the transformer 2-EE-ST-211 closed via modification, as needed. To has beca seismically tested. 2-EE-ST-2111 tighten the nuts, a work order has been (b) He transformer anchorage 2-EE-ST-2J issued to the Station. and/or load-path to the adjacent 2-EE-ST-2Ji cabinets, which are bolted to the A calculation (CE-1394), and a Design transformer

cabinet, require additional review.

Change (DC 98-002) were prepared to g g, ,7 g (c) The nuts for the bolts connecting transformers. The Design Change the transformer coils to the base has been implemented at the Station channels may need to be tightened for Transformers 1-EE-ST-lJ, IJI for transformers 2-EE-ST-23 and 2-EE-SL2J1. and 2-EE-ST-2J,2Jf. The anchorage of the remaining four transformers will be modified in the respective refueling outages, currently scheduled to be completed in April 2000. The resolution of this outlier / issue is 50% complete. i i i i 22 i

ITEM DESCRIPTION OF EQUIPMENT PREVIOUS RESOLUTION PLAN / NO. OUTLIER / ISSUE MARK NO. CURRENT STATUS (BOLf7) 1 4 Cabinets contain essential relays and I-El-CB-08B The cabinets will be bolted with have the potential of side-to-side 2-El-CB-08B adjacent cabir,ets to prevent potential impact with adjacent cabinets 1-EE-EG-01C impact. 1-EE-EG-02C E I An evaluation of the available margins in the capacity of the relays, with the potential of impact, is in progress. If required, the cabinets will be bolted to the adjacent cabinets. The resolution of this outlier / issue is in progress. 5 A space heater is suspended on long 1-EG-B-Ol A The requirement for the space heater in rod hangers over the battery chargers 1-EG-B-03C the EDG room will be reviewed, and if and near the batteries in the EDG 2-EG-B-02B required the heaters will be laterally room. Swaying of the heater during a 2-EG-B-04 D restrained to the adjacent concrete walls. seismic event may break the connecting steam / condensate lin and spray the batteries. A Design Change is being prepared to laterally restrain the heaters with members attached to the adjacent concrete walls. The resolution of this outlier / issue is in progress. 6 A clamp support on an adjacent 1-ilV-FS-1215B Engineering transmittal CEM-97-0012, unistrut is close to the flow switch and Work Order no. 008954 have been and has a potential for interaction written to move the clamp assembly about 1" away. 'Ine clamp assembly has been moved about I", via Work Order 00362829, to prevent the potential interaction. This outlier / issue is therefore closed. i 7 Potential interaction exists between 2-FW-FCV-Further review of the operator valve operator limit switch support 2479 displacement is planned. If required, the bracket and platform support beam. existing suppon beam will be notched to ~lhe valve has cast iron yoke. clear the interference. The operator displacement during a seismic event was calculated to be 0.117". Evaluation of the platform displacement is in progress. The resolution of this issue / outlier is in progress. 23

ITEM DESCRIPTION OF EQUIPMENT PREVIOUS RESOLUTION PLAN / NO. OUTLIER /1SSUE MARK NO. CURRENT STATUS (HOLD) 8 Pressure gage / regulator assembly 2-RH-FCV-2605 Further review of the operator which is mounted to a valve operator displacement is planned. If required, the is approximately 5/8 ' away from handrail will be moved away from valve post for adjacent handrail operator 1 The review of the existing analysis of j the valve / piping system indicated that, because of a simple modeling method used, the displacements at the valve-operator location were not available. However, for other valves in this system, which are heasie.. the maximum displacement was 0.15", based on a response spectrum analysis performed with 1% damped spectra. From this review, it was judged that there is sufficient gap between the pressure gage / regulator and the adjacent handrail to preclude a seismic interaction, Further, as stated in the report, "Use of Seismic j Experience and Test Data to Show Ruggedness of Equipment in Nuclear Power Plants", prepared by the Senior Seismic Review and Advisory Panel (SSRAP), February 1991, failure of air-operated valves from impact is rare in earthquake experience, and credible only when these valves are supported on very flexible piping. Therefore, this outlier / issue is closed. 9 Valve operator cantilever length 1-RC-PCV-Funher review of the valve / yoke is exceeds limits of Figure B.7-1 of the 1455C planned. GIP l-RC-PCV-1456 The valves were determined to be flexible (natural frequency <10 hz). The piping analysis was revised with an updated valve model that reflected the natural frequency, and the acceleration responses at the c.g. of the extended structure of both these valve were obtained. Valve critical components were analyzed using l these accelerations, and the stresses were found to be within the allowable values. This outlier / issue is therefore closed. 24

i l l Request 12: On Page 11-7, with regard to outliers for cahic tray and conduit raceway ss a cm, you stated that these outliers / issues were resolved to enhance the seismic capability the u i raceway systems. Provide a more prescriptive discussion of the approach employed to resolve l the outliers identified under items (6) and (7) of Section i1.1.4 of your summary report.

Response

The following provides a brief description of the approach taken to resolve the outliers identified in l Section 11.1.4 of the referenced summary report for Items (6) and (7): Item 6: a) A %" diameter conduit at containment crane wall Elevation 291' 0" was found to have excessive vertical span. An additional clamp support was added to reduce the span. I b) A 1%" diameter conduit in containment loop room C above entrance door had long span at elevation 241' 0" An additional conduit clamp support was added to reduce the span. c) A 1%" di:aneter conduit in the reactor building annulus on the crane wall at elevation 291' 0" had a missing support. A clamp support was installed. d) A 1%" diameter conduit in the containment annulus on the crane wall at elevation 291' 0" was missing a conduit support. A wall clamp support was added. Item 7: a) A 1%" diameter conduit at the containment crane wall elevation 302' 0" near pressurizer cubicle had 10' 0" cantilever. A wall clamp support was added to reduce the cantilever span. b) Six side by side conduits of 1%" diameter each in the Reactor Building annulus near the crane wall were found to have long cantilever lengths of 8' to 10'. An additional wall bracket support was added to reduce the span. I i 25 l l l l l

1 I l i ATTACHMENT A (Question No. 2) Background and Qualifications of Mr. P. Hayes, Ms. D. Jacobs and Mr. D. Werder l l l l Virginia Electric and Power Company North Anna Units 1 and 2

PAUL W. HAYES, P.E. EDUCATION Virginia Polytechnic Institute and State University, M.S. Aerospace and Ocean Engineering,1988 Oak Ridge School of Reactor Technology,1956 Webb institute of Naval Architecture, B.S. Naval Architecture and Marine Engineering,1954 PROFESSIONAL HISTORY 1954-1988 Nuclear Power Division, Naval Sea Systems Comman' 1988-1990 NUS Corporation i 1990-present MPR Associates,Inc. Since 1954, Mr. Hayes has been involved in nuclear and system engineering and project management for both stationary and Naval propulsion nuclear power plants including responsibility for power plant arrangements, fluid systems designs, operating manuals and initial acceptance and post maintenance test programs. ] MPR EXPERIENCE Mr. Hayes joined MPR in 1990, and has been primarily involved in systems engineering. Projects have included conceptual design of plant modifications, system design assessments to determine ability to satisfy reliability and functional requirements, and development of procedures and processes for documenting design basis information to satisfy regulatory (10 CFR 50 Appendix B) design control requirements. Specific examples of Mr. Hayes' experience include: System Assessment Lead review team in assessment of a reactor plant modification involving replacement of the reactor i protection and process control systems with digital equipment. The object of the review was to determine if the modification had been planned and executed in a manner to support reliable plant operation. The review involved checks of both design and istdware as well as checks of support activities such as operator training, procedure revision, maintenance schedules, and human factors. Participated as a member cf a review board comprised of senior, experienced ongineers who were tasked to oversee review of reactor plant modification packages. The review board was established by a utility as part of a mmedial action plan to correct modification design control problems identified by an NRC enginee. 7 support inspection. Performed design assessr ents of the Emergency Diesel Generator installation at a deferred construction nuciear powei plant in support of an overall evaluation of effort required for plant completion. Safe Shutdown Developed safe-shutdown equipment lists prepared for seismic verification of nuclear plant equipment for both PWR and BWR plants. Instructed EPRI sponsored course in safe shutdown equipment selection for seismic verification of nuclear plant equipment.

P:ul W. Hry:s P:g3 2 Design Basis Documentation Reconstitution Assisted utilities in establishing Design Basis Documentation (DBD) programs. Activities involved surveys of regulatory initiatives and utility practices; development of DBD program control i procedures; supervision of pilot Design Basis Document preparation; and review of draft Design i Basis Documents. Conceptual Design Prepared conceptual designs of plant modifications for Pressurized Water Reactor plants to satisfy station blackout requirements (loss of AC power). OTHER EXPERlENCE While at Naval Sea System Command, Mr. Hayes served as Director, Submarine Systems Division and was responsible for fluid systems engineering for the nuclear propulsion plants of more than 100 submarines. Responsibilities included fluid systems designs and propulsion plant arrangements for Trident Class submarines, USS Narwhal, and USS Glenard P. Lipscomb. Mr. Hayes managed development of nuclear proouision plant acceptance test programs, post overhaul test programs, 1 operating manuals and preventive maintenance programs and seived as Technical Director for initial i sea trials of Trident submarines. He developed procedures and designs for inactivating and disposing of submarine reactor plants and participated in quality assurance audits of shipyards, tenders and submarine bases performing Naval nuclear propulsion plant construction or maintenance. At NUS Corporation, Mr. Hayes evaluated the effects of Station Black Out on the containment system of a BWR. His work included the evaluatian of Heating, Ventilating and Air Conditioning design deficiencies encountered in initial operation of a Pressurized Water Reactor plant. He evaluated valve deficiencies affecting the operational availability of the Post Accident Sampling System at a PWR plant and designed modifications to correct the deficiencies. Mr. Hayes developed a hydraulic model of the Auxiliary Feed System for a PWR using a proprietary computer program and then applied the model to analy::e system modifications and simulate periodic surveillance tests. HONORS Presidential Rank - Meritorious Executive 1987 Navy Distinguished Civilian Service Award,1988 Department of Energy Exceptional Service Award,1988 l

DAWN W. JACOBS, P.E. EDUCATION Virginia Tech, B.S. Mechanical Engineering,1988, Magna Cum Laude MPR EXPERIENCE Ms. Jacobs joined MPR in 1988. She has been involved in projects for the nuclear power industry and the US Navy. Her experience includes on-site technical support for piping system and component replacement, computer piping analyses, preparation of test procedures and field direction of testing. Specific examples of Ms. Jacobs' accomplishments include: Nuclear Power Plant Seismic Qualification Assisted a utility (four units) in developing safe shutdown equipment lists and peitorming the seistnic portion of the Individual Plant Examination of Extemal Events (IPEEE) in response to NRC Unresolved Safety issue (USI) A-46 and NRC Generic Letter 88-20. Received individual training and mentoring from a SQUG instructor on how to select and evaluate the equipment for these projects. Piping System Replacement On Site Engineering Provided on-site technical support at a nuclear power plant for replacement of the Emergency Condenser Piping System. This effort included prompt resolution of technicalissues as they arose in the field, particularly issues involving welding, pipe fit-up, pipe supports, and valve replacement. This effort involved extensive interaction with the utility, the construction engineer, and contractors to provide a general follow and an independent technical review of the pipe replacement. Valve Replacement On-Site Engineering Provided on-site technical support for an unscheduled replacement of two isolation valves in a nuclear power plant. This effort involved preparation of procurement and installation specifications for the replacement valves, and reso'ution of technical issues as they arose in the field. Pipe Stress and Support Analyses Performed pipe stress and support analyses of a nuclear power plant Emergency Condenser System to qualify a replacement piping design. These analyses included preparation of e computer model of the piping system and suppons. The computer results were used to evaluate the design adequacy of the replacement piping, existing supports, and existing nozzles and penetrations. Heat Exchanger Tube Cleaning Evaluated the feasibility of rotary brush tube cleaning equipment for use on shipboard heat exchanger tubes. This effort involved development of a procedure for expnrimentally cleaning a retired shipboard heat exchanger, and development of a laboratory testing specification to evaluate the extent of damage in the cleaned tubes. This effort also involved on-site direction of the expenmental tube cleaning at a US Navy facility. Shipboard Elevator Watertight Seals on Hatches and Doors identified design and maintenance improvements for watertight seals on shipboard elevator hatches and doors. This effort included preparation of sealinspection procedures, trial use of the procedures

e onboard ship, interviews with elevator maintenance personnel, and evaluation of the inspection and ( interview results. Dawn W. Jacobs Page 2 ) l Preparation and Coating of Drywell Surface Developed a test plan to qualify surface preparation and coating processes for use on the drywell exterior surface at a nuclear power plant. This effort included selection of surface preparation tools, specification of a coating process, and direction of testing to qualify the processes for use on the drywell exterior. The objective of this effort was to minimize future corrosion of the drywell surface. Auxiliary Boiler System Upgrade Developed a conceptual design for improvement of an auxiliary boiler system at a nuclear power plant. This effort involved identification of problems affecting maintainability and performance of the 1 existing auxiliary boiler system, and determination of a cost-effective solution to the problems. Ms. Jacobs was particularly involved in design of an automated water chemistry control system to l improve boiler availability and performance. J i

I DAVID J. WERDER EDUCATION University of Maryland, B.S. Civil Engineering,1981 Catholic University of America,1979 Hagerstown Jr. College, A.A. Engineering,1979 MPR EXPERIENCE Mr. Werder joined MPR as a permanent engineer in 1982. He had worked at MPR part-time during 1980-1981 while attending the University of Maryland. Since 1982, Mr. Werder has been involved in a variety of projects in the nuclear and fossil-fueled power industries and the U.S. Navy. Specific examples of Mr. Werder's work include: Nuclear Power Plant Design Basis Verification Assisted three nuclear power stations by performing vertical slice system reviews and system readiness reviews as part of 50.54(f) restart efforts. Systems reviewed include service water, residual heat removal, and makeup & purification systems. Nuclear Power Plant Seismic Qualification Assisted two utilities (five units) in developing safe shutdown equipment lists and performing the seismic portion of the Individual Plant Examination of External Events (IPEEE)in response to NRC Unresolved Safety issue (USI) A-46 and NRC Generic Letter 88-20. Received individual training and mentoring from a SQUG instructor on how to select and evaluate the equipment for these projects. Nuclear Power Plant Service Water Systems Participated in independent system reviews of nuclear power plant service water systems to verify plant conformance to system design bases and licensing bases. System revie.ws include reviews of documentation, system walkdowns, and personnel interviews. In addition, performed material condition assessment of a nuclear utilities' service water system to determine root causes and corrective actions for system leakage problems. Nuclear Power Plant Feedwater Flow Tracer Testing Performed independent reviews and analyses of nuclear power plant feedwater flow tracer calibration tests to provide assessment of test method, conclusions, and uncertainty analyses. Main Condensers Developed a personal computer program for the assessment of plant thermal performance for submarine propulsion plants. The program performs a heat balance to confirm the condenser heat load as determined by steam-side measurements matches the heat load determined by main seawater system measurements. This program is currently used to determine the degree of fouling in all U.S. submarines and to determine the condensers' heat rejection capabilities under various operating conditions. Marine Firemain System and Seawater Service Systems Participated in the redesign of the seawater service system and firemain system of two U.S. Navy surface-ship classes to reduce the number of system regulating valves and increase system reliability. Performed flow tests of seawater service and firemain system mains and branches to validate flow analyses.

f l David J. W:rd:r page 2 Compressed Air Systems j j Developed maintenance programs, inspection procedures, and evaluation guidelines to maintain the 1 l condition of high pressure and low pressure air compressors, receivers, and dehydrators in U.S. l Navy and nuclear power plant installations. l l Submarine Main Seawater and Auxiliary Seawater Systems Developed a Maintenance Assessment Plan to assess and trend the condition of the Main Seawater Systems and Auxiliary Seawater Systems of all current U.S. submarine classes. The Maintenance Assessment Plan includes performance analyses of the main seawater pumps, main condensers, engine room freshwater heat exchangers, system strainers, valves, and valve actuators. The auxiliary seawater system mainbnance assessment plan includes performance analyses of the auxiliary seawater pumps, all auxiliary heat exchangers, system strainers and valves, and valve actuators. Cooling Tower and Circulating Water System Evaluations Performed structural inspections and evaluations of natural draft concrete cooling towers and circulating water systems for nuclear and fossil-fueled power plants. The evaluations included development of repair specifications, evaluation and design of replacement and repair components, and revision of design specifications. The system evaluations covered prestressed concrete, conventionally reinforced concrete, steel, and timber components. Submarine Trim and Drain Systems Developed a Maintenance Assessment Plan to assess and trend the condition of the Trim and Drain Systems of all current U.S. submarine classes. The Maintenance Assessment Plan includes performance analyses of the trim pumps, drain pumps, system strainers, valves, and valve actuators. Marine Distilling Plants Developed personal computer based, user friendly analytical models for several types of marine distilling plants. This software is currently used to assess the performance condition of many classes of U.S. Navy distilling plants. This effort included numerous verification tests for the analytical models on both land-based and shipboard distilling plants. During verification tests, distilling plants were disassembled and modifiv o simulate heat exchanger fouling. Additional work irduded a design review to ensure against d ang plant chloride intrusion and review of a Material Assessment Plan for the assessment of' ang plant performance condition. Fire Protection Systems Contributed to the system reliability reviews and corrective action plans for U.S. Navy fixed fiooding halon and carbon dioxide systems. Also, completed design reviews of various specific installations and components, including Aqueous Film Forming Foam (AFFF), eductor, and bilge and magazine sprinkler components. Performed carbon dioxide fixed-flooding system calculations in accordance with the National Fire Protection Association (NFPA) code. l Nuclear Power Plant Reinforced Concrete Structural Analysis Developed a technical evaluation program to assess the significance of crew and other flaws in reinforced concrete components and structures in nuclear power plants. The program included . development of procedures for inspection, analysis, evaluation, repair, and corrective action implementation. t

r-3 l Dtvid J. W:rdsr i page 3 Stress Analysis at Piping Attachments Utilized elasticity theory methods to determine local stresses at piping connections and attachments, and revised techniques to determine local stresses in cylindrical pressure vessels resting on saddle supports. I i f I l

ATTACHMENT B (Question No. 4) Revised sheet for Table I1.1-1 Note: Page 11-2 of North Anna Power Station USI A-46 summary report should be replaced with the revised page 11-2 contained in this Attachment. i i Virginia Electric and Power Company North Anna Units 1 and 2

I Table 11.1-1 (Continued)- Revision 1 -3/5/99 ITEM CLASS MARK NUMBER DESCRIPTION OF RESOLUTION NO. OUTLIER 5 07 2-MS-TV-201 A,B,C Grating adjacent to air Air lines were not required for t l lines was not properly proper operational mode per secured Systems Engineer 1 6 07 2-RC-PCV-2455C Seismic interaction DCP 94-012 corrected l concern with adjacent interaction concerns structure [ 7 08A l-SW-MOV-101 A,B,C,D Concern over Capacity Capacity exceeds demand by 2-SW-MOV-201 A,B.C,D vs. Demand less than 10% in the narrow frequency range (10.5 to 11.5 Hz), therefore intent of the GIP is met 8 08B 2-RC-SOV-2455C-1,2 Seismic interaction with Interaction resolved per DCP adjacent SOVs 94-012 9 08B 2-RC-SOV-2456-2 Seismic interaction with SOV was replaced with adjacent SOVs Nuclear Grade ASCO and Rev, I fixed to prevent interaction 10 10 1-ilV-AC-1,2 Anchorage has greater Anchorage evaluated and found l 2-IIV-AC-8,9 than 1/4" gap, adjacent acceptable; interaction concern interaction concern. One removed. Missing anchor bolt anchor bolt missing on 2-installed on 2-IIV-AC-9. IIV-AC-9 Ii 14 1 -B P-S W-1,2 Mounted to block wall Anchorage found acceptable 2-B P-S W-l.2,3.4 12 16 1-VB 1-04 Transformer missing two Missing Bolts replaced bolts 13 20 1-EP-CB-28G,il,J Cabinets contain essential A calculation was performed relays and were found w hich demonstrated that the with small side-to-side relative displacement between gap (3/16") with adjacent cabinets would be less than cabinets, therefore they 1/32" during an SSE, and no were considered as impact is likely. interaction outliers Rev.I 14 20 1(2 )-El-CB-23 A,B.C.D.E Cabinets contain essential All these cabinets were 2-El-CB-23 F relays and were found connected together by Design I El-CB-44 with small or no side-to-Change (DCP) No. 93-015-3. l(2)-El-CB-47A,B,E,F side gap with adjacent 1-EP-CB 28A,B,E.F cabinets, therefore they 2-EP-CB-28 A.B.E.F,G,il J were considered as 1(2)-El-CB-48A interaction outliers 1(2 )-El-CB-64 A,B 11-2 t l i 1 l

ATTACHMENT C (Request No. 6) Limited Analytical Reviews (LAR) For Examples 1,6,10,11,13, and 14 l l

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ENGINEERING WORK SHEET g g{ L A*hb Q 9% SHEET NO. CF 4-DOCUMENT TYPE: PROJECT: AlW b DOC. NO. QA O REv. O M WgggNM sunJECT: PREFARED BY: DATE: POWEft SYSTEM: CHECKID BY: DATE: Virginia Power Response Spectra Database MSTN*2275D050B0003 0.701. g ,USTH*X275D050B0003 h 6.60;.. ,.t ' j ~ g 0. 50*.. 6 STH*Y275D050B000 3 g .h / 7.'%.'. s g e. 40,~. - 0. 30 . [/._ s _ _ _- <3 c ..Q. t / is f w 0.20:. .y. ..y.. ~ r J' \\ 0.1 Sk. ....[././ -~-~-- F ., / g%'.r 1 1 1 I l l t f f f I l iI I If f I f 1 I I t ! ,6.te i.ee 1e.ee tee.00 Frequency, Hz LIST OF COMPONENT RESPONSE SPECTRA IN PLOT -USTN*Z275D050B0003, STR:SRV. TURBINE UNT: NAPS, DIR:N-S ELV:275,, EQ :DBE, DMP:5.0%, BRD:YES SRC:5218 -USTN*Y275D050B0003, STR:SRV. TURBINE UNT: NAPS, DIR:VER ELV:275, EQ : DBE, DMP:5.0%,, BRD:YES SRC:5218 l -USTN*X275D050B0003, ST.T-SRV TURBINE UNT: NAPS, DIR:E-W ELV:275, EQ : DBE, P:5.0%, BRD:YES SRC:5218 L.

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6Q /4 %, o Attzu.ArrnrZ 2, Sh.L oP.2 ENGINEERING WORK SHEET SHEET NO.)'L 9 CF 5/ Addendum OA DOCUMENT TYPE: Engineering Calculation DOC NO. CE-1142 REV. No. O PROJECT: North Anna Unlj:1 & 2 PREPARED BY S. DATE UNO E

SUBJECT:

NAS 2016 Rev. 6 S. E. Zinkham M ////V/W POWER CHECKED BY ' DATE SYSTEM: RECTRICE D.P. Madden DP M l~t - / 5 *

• Oualification of CC-5 and CC-6 The following accelerations were used to qualify the CC-5 and CC-6 supports in calculation SEO-369 rev.0 and CE-0676 rev. 20:

H, = 4. 5 6 9 V, = 1.42g + ig (due to gravity) = 2.42g The accelerations used for qualification generally envelope all areas of the plant with the exception of R.C. elev. 313', MSVH elev. 328', SWPH and SWVH. To account for the loads for these areas, a load multiplication factor shown in Attachment I, Table 3-2 of Specificat:.on NAS-2016, and notes on the standard drawings will reduce these loads for these areas. These values represent the acceleration due to 1% damping. Compare them to the 5% values as listed below: H, = 2. 4 4 6 g V, = 1. 0 3 8 g Using the bump factor of 1.3, the following values are obtained: H = 2.446g

• 1.3 = 3.18g

) g V, = 1.038

• 1.3 + ig = 2.35g Ratioing the 1% values to the 5% values:

4.56/3.18 = 1.43 Horizontal Ratio 2.42/2.35 1.03 Vertical Ratio = The tables adjust the previous leading under 1% damping using the lowest ratio, in this case it is vert: _si ratio, to the equivalent loading at 5% damping. The previous loading was multiplied by 1.03 to obtain the values listed in the right column. I 1 l

e@3, W. o AttMnunt~ 2, Sh.h of L ENGINEERING WORK SHEET gggg7 gg qg gj 0 _ DOCUMENT TYPE: Engineering Calculation hhhh"O CE-1 42 REV. No. O PROJECT: North Anna Unit 1 & 2 PREPARED BY .s E ATp // ///4 Y UNEM

SUBJECT:

NAS-2016 Rev. 6 S. E. Zinkham fc A / pggygg CHECKED BY #

DATE, M EM: R E m lC E D.P. Madden BPM i L - i T-H Oualification of CC-5 and CC-6 (continued)

LOADING FOR CC-5 LENGTH (INCHES) PREVIOUSLY UNDER 1% LOADING WITH 5% DAMPING DAMPING ~ 12 220 226 18 180 185 24 150 155 30 130 134 36 110 113 Oualification of CC-5 and CC-6 (cont.) LOADING FOR CC-6 LENGTH (INCHES) PREVIOUSLY UNDER 1% LOADING WITH 5% DAMPING DAMPING 12 350 361 18 300 309 24 250 258 2' 7 l 30 220 i [ 36 180 185 These new 5% loads will be used in the updated version of Rev. 6 of NAS-2016 Drawings for CC-5 and CC-6.

i 7^ '"'"*"' P* yugn Be~nsig %m me au (_ Project Doc. No Rev. No Sneet No E O j of L 6f (Oh k M E {b y}({Lly Al g: h [-13~ l,/ SQo'$f11AC Preparea By Date Sutnec' {& OLJ Syster6 ggJj Checked By Date E'X G.lt casting otat9 4,anu ca't A1.qYod dracbw'- CRef NAs-20l4, Rw. Q Fir >m no ^!A SL l') 5S'Ib c -,~ - o 4 0"(of A*) n 10 n S*. k * // 0 u l ,q s, y 2.> c 'l+Y 09-1 6,, g I o, L '/ D h 3 # 4 l'/[ - _ 4 y o" -@ x %. s l (t/P) 41r i ts t [_ .5ccfloN: A-A y ^1ai:tjLo" and gnaa;au deaJ wGj// cf(,chd;pS L-42 L-q$ (ef. vat.. cort-r/4ea.es-f) c ateasu~M) CI Cr fp4(( h\\r1 SuttWt il ci V fyk.;rrtic . Q 1\\ d t u't j f2 crt Not Aill)(> LenKq cct fastv3 clegalc'oNf 07tc{ padi}. s, masmw~ accaabet Muu tw f,ae n H = a.is 4 4 > i.asW:naan<d & t! hus -fgacJ JM P0C is 3 YLfkl RLffet (116G-ducult,) and XX< $nd is 0 e tduty July ~ is cor\\x'M pc mwr jn 2nZe { s sie tv diqin g ncn 6dit cndwr sygd, sn': n ),mut Jw na cawwa (

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'%5?a2CMnW'iRt~, I or1 } { ENGINEERING WORK SHEET i O SHEET NO. 6 OF 20 DOCUMENi TYPE: Engineering Calculation DOC NO. CE-1142 REV. No. O PROJECT: North Anna Unit 1 & 2 PREPARED BY DATE g NEE

SUBJECT:

NAS-2016 Rev. 6 UPDATE S. E. Zinkham 4/18/94 m CHECKED BY DATE SYSTEM: ELECTRICAL C.C.Ranganath 4/18/94 Calculations:

1. Use of 5% Damped Accelerations:

In accordance with the UFSAR Section 3.10, Conduit Supports can be designed to 5% damped accelerations provided that a multiplication factor of 1.3 is used (the intent was to account for multi-modal participation of the combined systems which-include conduits and supports). The use of 5% damping has been documented by reference 1 and has been based on damping values extracted from shaker table test data and experience data. This is reasonable since conduit spans are usually highly ductile and c hibit damping values 5% and higher. The accelerations used in the original design of the standard conduit supports by Sargeant & Lundy were based on peak 1% SSE accelerations ( references 5 & 8) as follows: Peak 1% Damped Standard Conduit Support Design Accelerations: j Vg = 1.42 g's Hg = 3.52 g's The 5% damped peak acceleration values will be ratioed against the 1% damped peak values using the multiplication f actor of 1.3 required by the UFSAR for 5% damping. Peak equivalent 5% damped accelerations ( from reference 10 ) are as follows: Peak 5% Damped Standard Conduit Support Design Accelerations: Vg = 1.038 g's Hg = 2.446 g's Ratio of 1% to 5% Damped Accelerations: Vertical Refo: (Vg(1%) + 1)/ (Vg(5%)x1.3 + 1) = (1.42 + 1)/(1.038 + 1) = 1.03005 Horizontal Ratio: (Hg(1%))/(Hg(5%)x1.3) =( 3.521/(2.446) = 1.107 Minimum Ratio = 1.03005 The minimum ratio of 1.03005 will be used to multiply against the design loading values in Table 3-1: Example: Support D3 span O'-0" from Rev. 5 of specification NAS-2016: Design Load = 85# x 1.03005(min. ratio) = 87 The other updated values are shown in Table 3-1 located in Attachment B. C

5'E L M /4:, N ' 4An:c -2 E 1 of-E. ENGINEERING WORK SHEET e SHEET NO.15 OF 20 DOCUMENT TYPE: Engineering Calculation DOC NO. CE-1142 REV. No. O PROJECT: North Anna Unit 1 & 2 PREPARED BY DATE WMEM

SUBJECT:

NAS 2016 Rev. 8 UPDATE

5. E. ZinkhttfC, 4/18/94 puggygyr CHECKED 8Y Ob DATE I

SYSTEM: ELECTRICAL n C.C.Ranganath 4/18/94 DETAIL CAPACITY TABLE (4.6) (EQUIVALENT STATIC LOAD LBS) SPAN 0-1 D-2 D-3 CC-1 CC-2 CC-3 CCd CS-1 CS-2 BC-1 TC-1 TS-1 L(2) (3) BC-2 BC-3 0' 0-(1) (1) 87 1

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85 216 422 869 443 624 418 686 836 42 108 229 418 211 624 418 686 836 2' - 3'.0" 27 67 145 294 138 624 418 551 836 47 101 214 100 439 418 418 836 ( 0-34 73 164 76 387 5' - 6' 0-53 134 60 287 r _ 38 109 90 8 c-mmmm. ~m. i. 6 -. 3 TABLE 3-1 NOTES: 1. Detail is adequate for any si::e conduit up to 4" diameter. 2. Span L is the distance to the centerline of the outermost conduit on a support. See Standard Forms NA-S-20 through NA-S-55 for more information. 3. Load capacity is per linear foot, see Form NA-S-12 for more information. 4. Interpolation between given loads is permitted. Extrapolation beyond given loads is not pemitted. 5. The allowable loads are based on S&L STD conduit support manual, DC-SE-03-NS unit conduit weights and allowable wedge anchor '.oads combined linearly with 10% angularity for 3000 psi concrete and 5% damping SSE load. 6. For supports listed in the Detail Capacity Table above, use Table 3-2 multiplication factors for the building listed in it and Table 1-4.

l l I i ATTACIIMENT D (Request No. 8) { Seismic Capacity Calculations of ECST and RWST Tanks I i i 1}}