ML17265A536
| ML17265A536 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/02/1999 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Vissing G NRC, NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR NUDOCS 9902100108 | |
| Download: ML17265A536 (31) | |
Text
CATEGORY 1 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSXON NBR:9902100108 DOC.DATE: 99/02/02 NOTARIZED: NO DOCKET FACXL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244
,,AUTH. NAME, AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas S Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION VISSING,G.
SUBJECT:
Forwards response to NRC 981203 RAI re resolution of unresolved safety i'@sue USX A-46.
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TITLE: Seismic Qualification of Equipment in Operating Plants A GL-87' NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72) 05000244 E
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Ar)rn ROCHESTER GAS AND ElECTRIC CORPORATION ~ 89 FAST AVENUE, ROCHESTER N.Y. Id6d9.0001 AREA CODE 716 546-2700 ROBERT C. MECREDY Vice President Nuclear Operating Group February 2, 1999 U.S. Nuclear Regulatory Commission Document Control Desk ATTN: Guy Vissing Project Directorate I-1 Washington, D.C. 20555-0001
Subject:
Response to NRC "Second Request for Additional Information" (RAI) on the resolution of Unresolved Safety Issue (USI) A-46.
R.E. Ginna Nuclear Power Plant Docket No. 50/244
Reference:
A. Letter from Robert C. Mecredy (RGEE) to Document Control Desk (NRC), dated January 31, 1997, "Resolution of Generic Letter 87-02, Supplement 1 and Generic Letter 88-20, Supplements 4 and 5 (Seismic Events Only)."
B. Letter from Guy S. Vissing (NRC) to Dr. Robert C.
Mecredy (RGRE), dated April 6, 1998, "Request for Additional Information on the resolution of Unresolved Safety Issue (USI) A-46."
C. Letter from Robert C. Mecredy (RG&E) to Document Control Desk (NRC), dated May 27, 1998, "Response to RAI on USI A-46."
D. Letter from Guy S. Vissing (NRC) to Dr. Robert C.
Mecredy, dated December 3, 1998, "Second Request for Additional Information".
Dear Mr. Vissing:
1 This letter provides responses to the NRC's "Request for Additional Information" (RAI), dated December 3, 1998 (Ref. D) . Enclosures 1 and 2 along with the paragraphs below respond to Question 1 parts a and b regarding the use of GIP "Method A" at Ginna Station. Responses to (/
questions 2, 3 and 4 regarding specific SQUG screening methods and testing data are provided in Enclosure 3. QO 9902i00108 990202 PDR ADQCK 05000244 P PDR
The use of "GIP Method A" is described in the Generic Implementation Procedure, Revision 2 (GIP-2), the Supplemental Safety Evaluation Report No. 2 (SSER No. 2), and the documents referenced in GIP-2 upon which GIP-2 is based. RG&E used Method A to estimate seismic demand for certain equipment within 40 feet of effective grade at Ginna. The NRC has questioned RG&E's use of Method A on the basis that Method A may be used only if the amplification factor between the free-field ground response spectrum (GRS) and the calculated in-structure response spectra (ISRS) being used by the plant is not more than about 1.5. The NRC position is based on their interpretation of the language on page 4-16 of the GIP which says that "the amplification factor between the free-field response spectra and the in-structure response spectra will not be more than about 1.5...".
ItElectric does not agree with the Rochester Gas and NRC's interpretation. is RG&E's position that the approach used for applying and implementing GIP Method A for estimation of the seismic demand on equipment at Ginna for resolution of the USI A-46 program is appropriate and technically justified. Detailed bases are provided in .
With respect to the NRC's question regarding differences between the in-structure response spectra and the 1.5x ground response spectra, RG&E notes that these spectra were generated using conservative methods and assumptions (typical of most nuclear plant response analyses) which artificially increased the amplifications over those which would be expected in an actual earthquake. A detailed qualitative assessment of these conservatisms are provided in .
Based on the above, and the information in Enclosures 1 and 2, we believe that RG&E has properly interpreted the conditions on use of Method A, and that these conditions appear to have been understood and accepted by the NRC staff until recently, after RG&E completed'the USI A-46 reviews at Ginna. To change this interpretation at this stage in the program for resolution of A-46 would be inconsistent with the spirit and intent of A-46 and would also require rework of equipment or additional analyses and evaluations without a commensurate safety benefit.
Please contact George Wrobel at (716) 771-3535 additional questions.
if you have any Very truly yours, Robert C. Mecredy
Enclosures (3) xc: Mr. Guy S. Vissing (Mail Stop 14B2)
Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. P. Drysdale U.S. NRC Ginna Senior Resident Inspector
Enclosure 1 Bases for Interpretation and Implementation of GXP-2 Rules for Method A It is Rochester Gas and Electric's position that Rochester's Ginna Station has properly interpreted and implemented the rules for use of GIP Method A as previously reviewed and accepted by the NRC. The bases for this position are as follows:
SQUG and Rochester Gas and Electric's Interpretation of the GIP The caution given on page 4-16 of GIP-2 lists two limitations on use of Method A:
Equipment should be mounted in the nuclear plant below about 40 feet above the effective grade, and Equipment should have a fundamental natural frequency greater than about 8 Hz.
The introductory wording in GIP-2 for these two limitations provides the bases or purposes for imposing them, namely (1) to limit amplification to no more than about 1.5 and (2) to avoid the high-energy frequency range of earthquakes. The specific limitations which are intended by the SQUG/NRC expert panel (SSRAP) and SQUG to satisfy these bases are included in the two bulleted items listed above.
The statement on page 4-16 that "the amplification will not exceed about 1.5" is the expected result of meeting the above limitations, not a third condition.
At no time was a comparison of Method A amplification with that of calculated ISRS ever intended. In fact, the entire context of the caution on page 4-16 of GIP-2 makes clear that the advantage of Method A is that equipment meeting the two bulleted limitations above "can be evaluated without the need for using in-structure response spectra..."
- 2. The Intent of the GIP is Clear and SSRAP Agrees The GIP (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:
"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 fundamental frequency of the building where higher amplifications occur when such a frequency is less than about 6 Hz. Thus, for equipment with fundamental frequencies greater than about 8 Hz in the as-anchored condition it was judged that floor spectral amplifications within 40 feet of grade would be less than 1.5 when reasonably computed using more median centered approaches."
[SSRAP Report, page 102]
The SSRAP Chairman and developer of Method A, Dr. Robert Kennedy, was contacted by SQUG and concurs with the interpretation given in item 1 above.
The NRC Was Aware of SQUG's Interpretation When It Was Developed The NRC backfit analysis in NUREG-1211, which justifies implementation of the USI A-46 program by affected licensees, relies on the conclusions reached by SSRAP in their review of seismic experience data. NUREG-1211 states the following:
"The NRC staff has closely followed the SSRAP work and is in broad agreement with its conclusions. The staff has concluded that if the SSRAP spectral conditions are met, is generally unnecessary to perform explicit seismic it qualification on the eight (1) classes of equipment studied."
[NUREG-1211, page 17]
(1) The eight classes of equipment cited in NUREG-1211 were later expanded to 20 equipment classes.
Note that this quotation specifically makes reference to the SSRAP "spectral conditions." The spectral conditions are described in SQUG's position given above and were included in GIP-2.
The use of Method A was previously reviewed and accepted by the NRC and SSRAP representatives during two pilot plant reviews conducted in 1987 and 1988. These reviews are documented in GIP-2 References 16 and 25. The specific material presented to the NRC representatives on use of Method A is described in the report of the BWR pilot review as shown in Figure 1. Note that the seismic demand criteria described during this trial plant review are the same as described in item 1 above. NRC and SSRAP representatives raised no objections to the approach used by SQUG in conducting these trial plant reviews. The topics discussed with and comments made by NRC and SSRAP representatives during the BWR pilot'eview are included in Figure 2; note that seismic demand information was discussed in some detail.
The Rochester Gas & Electric/SQUG interpretation of the rules for applying Method A is also consistent with the SQUG training course on use of the GIP methods'igure 3 is an excerpt from the class notes used during this course. It shows, in Slide 26, several screening methods for comparing equipment capacity to demand. Screen 52 illustrates uses of GIP Method A as described in Item 1 above'hat is, if equipment is below 40 feet and above 8 Hz, and the Bounding Spectrum envelopes the ground response spectrum, the equipment is acceptable.
This training material was used during the first session of the SQUG training course held during the week of June 22, 1993. Two NRC staff members (P.Y. Chen, Michael McBrearty) and a NRC contractor (Kamal Bandyopodhyay) attended this initial session and later provided comments on the training course in a letter dated August 28, 1992. The NRC did not raise any objections to the approach taught by SQUG in this course for applying Method A. Subsequent to this initial session of the course, 11 additional NRC staff members and contractors attended other sessions of this course; similarly, none of them raised objections to how SQUG was teaching use of GIP Method A.
NRC Interpretation Renders Method A Not Useful The..NRC interpretation is that Method A can be used only when calculated ISRS are less than 1.5x GRS. This interpretation negates the value of using Method A because when it it could only be used produces higher seismic demand than Method B where calculated ISRS are used. Under this interpretation, Method A would never be used. This is inconsistent with Method A' development and use, and was never the intent.
FIGURE j.
(Figure 1 contains an excerpt from GIP-2, Reference 25, which shows the seismic demand criteria used during the BWR Trial Plant Review.)
SEISMIC DEMAND CRITERIA APPLICATION DEMAND CRITERIA Equipment in experience data 1. Compare ground Spectra with base and less than 40'bove bounding spectrum (Figure 3.1 243', and fundamental frequency in SSRAP report).
greater than 8 Hz.
Equipment in experience data Compare amplified floor base over 40'bove 243'over response spectra with 1.5 x 281'levation) or fundamental bounding spectrum (Figure frequency less than 8 Hz. R1....,Rn, TI,...,Tn).
Equipment covered by GERS (any Compare amplified floor spectra elevation, frequency). (median-centered) with 2/3 x SERS for specific equipment class.
4 ~ Anchorage evaluation and equipment-specific stress checks (excluding valves):
Equipment within 40'f "grade" (elevation 281'nd Utilize accelerations from (1.5 x ground spectra) x below) and fundamental 1.25.
frequency less than 8 Hz.
Equipment at any elevation. Utilize accelerations from median-centered amplif ied floor response spectra x 1.25.
Equivalent static load factor Using appropriate spectra for all equipment (except with multiplier, use:
valves).
Peak acceleration for flexible equipment.
ZPA for rigid equipment.
Acceleration at calculated fundamental frequency.
Static load check for valve 3G, Weak direction.
operator/yoke checks.
Note: In general, for equipment. with fundamental frequency greater than 8 Hz and within 40'f grade. 1.5 x ground spectra may be used as an estimate of median-centered amplified floor spectra.
FIGURE 2 (Figure 2 contains an excerpt from GZP-2, Reference 25, which summarizes the SSRAP and NRC comments on the BWR Trial Plant Review)
Section 8 SENIOR SEISMIC REVIEW AND ADVISORY PANEL (SSRAP) AND NUCLEAR REGULATORY COMMISSION (NRC) REVIEWS Representatives of SSRAP and the NRC attended the NMP-1 walkdown on February 1st through 3rd (Days 8 through 10) . On February 1st, following radiation protection training and dosimetry issuance, the SSRAP and NRC representatives were briefed on the organization and conduct of the NMP-1 walkdown. The indoctrination and pre-walkdown materials covered by SQUG for the walkdown participants were also reviewed with SSRAP and the NRC. The indoctrination/training materials are given in Appendix C and include information on the organization and schedule of the walkdown, the rules of conduct in the plant, plant-specific data on the seismic demand levels for the walkdown, and summary information on GIP requirements for review of seismic demand versus capacity, equipment caveats, anchorage evaluation and evaluation of interactions.
The NMP-1 seismic demand information used for this walkdown was discussed in some detail. SQUG representatives explained that the seismic ground motion used as a basis for the walkdown is a plant-specific, uniform hazard, ground-motion spectra developed by A. Cornell and R. McGuixe and is anchored at 0.13 G. This ground-motion spectra envelopes the NMP-1 FSAR licensing basis SSE spectra which is anchored at 0.11 G. The NMP-1 uniform hazard ground-motion spectra is shown in Appendix C. Also in this Appendix are amplified floor response spectra developed for NMP-1 using modern reactor and turbine building models and the 0.13 G uniform hazard ground-motion spectra. Mr. Djordjevic (Stevenson s Associates) reviewed the bases for the amplified floor response spectra and indicated that they are being used as mean-centered, realistic spectra. Dr. Kennedy (SSRAP) expressed the view that he believes the floor response spectra are conservative and generally in accordance with current Standard Review Plan criteria. As a result, SSRAP considers that it is not necessary to utilize the additional factors of safety recommended by SSRAP for use with mean-centered spectra (1.5 for use of GERS and 1.25 for anchorage evaluation) in using the NMP-1 floor response spectra during this walkdown.
A second area discussed regarding the seismic demand was the effective grade level at NMP-1. At this site, the tuxbine building is founded on rock at elevation 243 feet above sea level. The reactor building is founded on rock at 198 feet. Grade elevation is 261 feet. In the construction of the buildings, the sites were excavated to the foundation level, the buildings constructed, and the annular space between the building and the rock excavation was backfilled with crushed stone up to the 251 foot grade elevation. An elevation view of the plant is included in Appendix C. SQUG and NMPC representatives explained that while they believe lateral support is provided by the crushed stone backfill, it has been conservatively assumed for the purpose of this walkdown that the effective grade elevation is at about 240-243 feet. This elevation corresponds to the foundation of the turbine building and the elevation in the reactor building where the structure changes from an essentially monolithic concrete block structure (including the reactor base mat) to that of reinforced concrete walls and floors. Essentially no amplification is expected in the reactor building up to about 243 feet. On this basis, the elevations which are considered to be within 40 feet of effective grade, are those elevations in the reactor and turbine buildings up to and including the 281 foot elevation. SSRAP was in general agreement with this approach.
Prior to walkdown of the plant by SSRAP and NRC reviewers, the three SRTs described their progress to date, highlighting areas they particularly wanted the reviewers to evaluate. SSRAP and NRC representatives spent most of February 2nd performing independent walkdowns of NMP-l. Essentially all safe shutdown equipment was seen by them with the exception of the emergency condensers and related equipment, several reactor coolant system instruments, several reactor coolant system isolation valves, core spray and containment spray pumps in the basement corner rooms and the equipment in the drywell, all of which were inaccessible due to the need for radiation work permits (RWPs) . Following this walkdown, Dr. Kennedy provided a summary of SSRAP's observations and conclusions:
The SSRAP walkdown was performed to determine how the seismic review teams (SRTs) were operating, to assess how the SRTs were evaluating and dispositioning the safe shutdown equipment, and to obtain a general sense of the seismic ruggedness of NMP-1.
SSRAP did not observe many seismic concerns and no serious seismic issues.
The expected outliers identified by the SRTs were considered by SSRAP to be typical. Dr. Kennedy remarked that, in fact, there were fewer outliers than would be expected for a plant of this vintage. He believes that this is result of the numerous seismic upgrades performed by NMPC over the years which were apparent to SSRAP during their walkdown.
It is SSRAP's judgment, based on their walkdown, that the SRT members received adequate training to perform the walkdown and that they were doing an adequate and qualified job of evaluating the seismic adequacy of the safe shutdown equipment. SSRAP generally expressed the opinion that when the SRTs reached different conclusions than SSRAP, the SRTs'onclusions were more conservative (i.e., the SRTs may have identified more outliers than would SSRAP). SSRAP is uncertain if the utility SRTs used during the trial plant walkdown are representative of the SRTs other utilities might use for their walkdowns, since SSRAP believes that the utility SRT members at the trial plant walkdown have considerable seismic experience. As a result, SSRAP continues to believe that it is essential that the SRT members have adequate qualifications and experience in seismic engineering.
Following Dr. Kennedy's summary report, NRC representatives presented their observations and conclusions. Dr. T.Y. Chang, USI A-46,Program Manager, reported the following:
The NRC generally agrees with the SSRAP review findings. The NRC believes that the walkdown has shown that the use of utility engineers is a viable approach provided the SRT members have the proper level of experience. The NRC still strongly believes that the qualifications of the SRT members are very important, irrespective of whether the members are utility employees or contractors. Further, the NRC believes that the training program is not enough to make an engineer a seismic expert. The SRT members should have the requisite seismic experience prior to their selection for training and the walkdowns .
The conduct of the NMP-1 walkdown was very smooth. The NRC commented is clear that the lessons learned from the Trial Plant 1 walkdown were that it factored into this walkdown in that there was a considerable amount of pre-walkdown planning which contributed to the smoothness of the walkdown.
The NRC was impressed with the layout of NMP-1. The plant is open and has considerable space which contributes to both good maintenance and a lack of seismic interaction hazards.
The NRC observed during their walkdown (as did the SRTs and SSRAP) that the quality of the anchor welds in some electrical cabinets was marginal.
The NRC noted that the relay review for NMP-1 was performed for a sample of typical safe shutdown circuits and did not cover every safe shutdown circuit and relay in this plant. They noted that the remaining circuits and relays need to be reviewed before the seismic review for NMP-1 is complete.
There was some discussion of the uniform hazard ground-motion spectra used for this walkdown. Since this spectra bounds the licensing basis ground-motion SSE spectra for NMP-1, the NRC concluded that this ground-motion spectra is acceptable and meets the requirements of USI A-46. The NRC also noted that they concur that the amplified floor spectra used for this walkdown are conservative spectral
(Figure 3 contains an excerpt from the SQUG Walkdown Training Course class notes which shows the screening process for comparing equipment capacity to demand.)
Equipment Capacity vs. Demand Screening Process Reference Spectrum > IRS Screen 1 Below 40' Above S Hz 8 8ounding Spectrum > GRS Screen 2 GERS > IRS Screen 3 QualiTication Documentation >
Screen 4 IRS Ou5iers Resolve-Capacity >
Derrt and Slide 26
ENCLOSURE 2 Position Pa er on the Use of Method A at Ginna
~Pur os e The purpose of this position paper is to provide supporting information for application of Method A at Ginna as requested by the NRC in question 1 of a second RAI on the USI A-46 program.
This enclosure describes many of the conservatisms that exist in computed in-structure response spectra and the safety significance of the difference between computed and actual building response.
- 1. Conservatism in Calculated ISRS The process of calculating in-structure response spectra (ISRS) is a complicated analytical exercise requiring a significant number of approximations, modeling assumptions and engineering judgments. As a result, the historical development of these ISRS has included a tremendous amount of conservatism which has typically served two purposes:
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) .
As a part of the A-46 program resolution methodology, the SSRAP developed and SQUG subsequently endorsed an alternate ISRS estimation technique (referred to as Method A within the GIP) which was much more median centered and realistic than the typical design practice. RG&E's'position is that the application of Method A at Ginna was appropriate and technically justified.
The fact that design ISRS may show amplifications greater than 1.5 is not surprising, nor does it negate the validity of Method A. In fact, as noted in the SSRAP report it was even expected.
"Secondly, most unbroadened computed in-structure spectra have very narrow, highly amplified peaks at the resonant frequency of the structure. In most cases these narrow, highly amplified peaks are artificially broadened to account for uncertainty in the structure's natural frequency. This process simply increases the emphasis on these highly amplified peaks. SSRAP is also of the opinion that these narrow peaks will not be as highly amplified in real structures at high ground motion levels as if predicted by linear elastic mathematical models, nor are such narrow peaked in-structure spectra likely to be as damaging to equipment as is a broad frequency input which is represented by 1.5 times the Bounding Spectrum."
As described below, three areas are presented to support the application of Method A at U.S. nuclear plants in general, and at Ginna in specific:
A. Measurements of ISRS in Actual Earthquakes B. Calculations of Overall Conservatisms in Typical ISRS C. Description of the Conservatisms in ISRS in General and Ginna ISRS in Particular A. Measurements of ISRS in Actual Earthquakes SSRAP developed the Method A response estimation technique based on their research of both actual earthquake measurements and on recent "median centered" analysis. They reference (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 Humbolt 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 Hz. 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 recently reviewed by SQUG. This data also shows that stiff buildings (similar to typical nuclear structures) amplify very little at elevations less than 40 feet above grade and frequencies over 8 Hz. SQUG knows of no new measured data that challenges 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 conservatisms 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 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). A NRC contractor (LLNL) concluded in a study for the NRC (NUREG/CR-1489) that typical calculated ISRS contain factors of 1.S to 1.8.
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'nd 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' applicability:
"It is SSRAP's firm opinion that the issue of potential amplifications greater than 1.5 above about 8 Hz for high frequency input"is of no consequence for the classes of equipment considered in this document except possibly for relay chatter'."
[SSRAP Report, Page 106]
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 does not appear to have a significant sensitivity to high frequencies (except possibly for relay chatter, which is addressed separately in the GIP).
C. Description of Conservatisms in ISRS in General and Ginna ZSRS in Particular The most significant sources of conservatism involved in the development of the ISRS for Ginna include the following:
Location of Input Motion (variation from the free field input location) 0 Ground Response Spectrum Shape 0 Soil-Structure Interaction (Soil Damping, Wave Scattering Effects) 0 Ground Motion Incoherence 0 Frequency (Structure Modeling) 0 Structural Damping 0 Time History Simulation 0 Non-Linear Behavior (e.g., soil property profile variation, concrete cracking) 0 Peak Broadening and Enveloping 0 Clipping of Narrow Peaks
'Because of the SSRAP concern related to possibly relay chatter at frequencies above 8 Hz, the SQUG methodology specifically addresses relay which are sensitive to high frequency vibration.
Such relays are included on the Low Ruggedness Relays list in Appendix E of EPRI Report, NP-7148.
~
~
The degree of conservatism involved in each of these parameters is specific to the building being analyzed, to the floor level being considered, and often, to the equipment location within the specified floor level. These conservatisms typically cannot be accurately quantified using simplistic calculational techniques since each parameter contributes to an overall set of highly non-linear responses. Thus, it would take a considerable effort to quantify the exact excess conservatisms inherent in the calculated ISRS at Ginna. However, on the qualitative level 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 of In ut Motion - The defined location of the plant SSE is typically part of the design basis documentation.
The SSE should typically be defined at the ground surface in the free field as defined in the current Standard Review Plan criteria. The defined location of the Ginna SSE is considered the ground surface in the free field. But for purposed of generating ISRS, some plants conservatively defined the input (currently identified as the "control point" location) at another location, such as the embedded depth of a building basemat. This conservatism can be significant depending on the specific plant/building configuration. The Ginna plant site geology consists of a thin layer of natural or compacted granular soil (30 to 40 feet in depth) immediately above bedrock. The bedrock is a mixture of sandstone and fissile shale with shear wave velocities calculated to be 7000 feet per second or greater.
Prior to construction of the plant, the soil over burden (30 to 40 feet of glacial drift) was removed.
All Ginna Station Category 1 buildings, except for the control building and diesel generator building, are founded on solid bedrock. The foundations of the control and diesel generator buildings were excavated to the surface of bedrock. Lean concrete or compacted backfill was placed on the rock surface to a depth whereby the elevation of the top of the fill material was coincident with the elevation of the bottom of the concrete foundation of that particular building.
Sections 2.5.2.1 (Seismicity) and 2.5.2.2 (Maximum Earthquake Potential) of the Ginna FSAR describe the original investigation which was performed to develop estimates of the maximum expected (OBE) and maximum credible (SSE) earthquakes for the site. It was judged that the maximum credible earthquake would be one of Richter magnitude 6.0 with an epicenter 30 miles from the site or one of magnitude 7.0 at a 90-mile epicentral distance.
A procedure developed by Dames & Moore, using the results of research at the Earthquake Institute of Tokyo, was used to estimate ground motion at a given location if the earthquake magnitude, epicentral distance, and elastic properties of foundation soils and rock are known. The FSAR contains the following description of the location of ground motion:
"Using this method and the assumed maximum credible earthquakes discussed above, maximum acceleration on the site was calculated to be 8'.
of gravity for soil surface and 7% for bedrock surface. Plant structures, systems, and components designated as Seismic Category 1 are designed to remain within applicable stress limits for the operating-basis earthquake (0.08g) and the safe shutdown earthquake (0.20g)."
Based on the above licensing basis descriptions, the design earthquakes (OBE E SSE) were clearly defined at the soil surface. Since the ISRS for Ginna were generated using a conservative model defining the input motion at the foundation level, significant conservatism exists due to the location of input motion. The level of conservatism involved in this assumption is difficult to estimate without performing additional analyses, but past studies have proved it can be considerable.
Ground Res onse S ectrum Sha e - The SSE defined within the plant-licensing basis is the appropriate review level for the A-46 program. Some utilities utilized alternative (conservative) spectral shapes for the earthquake levels utilized for their A-46 resolution (i.e., submitted as part of their 120-day response letters). The amount of conservatism is directly related to the difference between these two spectral shapes at the frequencies of interest for the structures being reviewed. This factor can range from 1.0 to around 2.0 depending on the differences between the spectra.
The licensing basis safe shutdown earthquake for Ginna is characterized by a site-specific horizontal ground response spectrum anchored to a PGA of 0.17g. However, ISRS were never generated in the original seismic design of Ginna and this earthquake was not used for the USI A-46 program. A more conservative earthquake anchored to a PGA of 0.2g and with a Reg Guide 1.60 shape (broader band) was used for the generation of ISRS in the A-46 program. The use of this alternate earthquake input is conservative for 3 reasons:
The 4: damped spectra were used instead of the S:
damped specific for the A-46 program. The conservatism is typically quantified by taking the square root of the damping levels, which would result in a 1.12 (12%) factor of conservatism.
- 3) The Reg Guide Shape and the site specific shape are both broad banded, but their levels of amplification are different and their differences vary as a function of frequency. Depending on the building in question and the frequency range of interest, there can be additional conservatisms due to the differences in shape.
Soil Structure Interaction SSI - Typical design analyses do not account properly for the phenomena of SSI, including the deamplification with depth that really occurs for embedded structures and for the radiation damping effects inherent at soil sites. Fixed-base analyses have been performed in typical design analyses, both for structures founded on rock and for structures founded on soil columns.
For rock foundations, the fixed-base model has been shown to be slightly conservative depending on the rock/structure characteristics. For soil founded structures this assumption can vary between conservative and very conservative, depending on the level of sophistication of the modeling of the soil-structure system. The simplified analyses that used the frequency-independent soil springs were typically very conservative in that radiation and/or material damping were either conservatively eliminated or artificially limited during the analysis. Soil properties were also typically not adjusted to reflect anticipated soil strain levels. Significant reductions have been demonstrated over design type analyses using more modern techniques. These reduction factors are highly dependent on the specific soil conditions and structure configurations, but values of around 2 to 4 have been seen in past studies.
The Ginna analyses have ignored any reduction in foundation motion due to embedment effects, wave scattering effects and radiation of energy from the structure into the surrounding media. These effects are less for rock founded structures (Standby Auxiliary Feedwater Buildings and Intermediate Building) than they are for the soil layer founded structures (Control Building and Diesel Building), but they are not negligible. This assumption is commonly made for rock sites because it greatly simplifies the analysis even though it introduces conservatism.
The Ginna analyses also ignored any constraint that surrounding rock or soil placed against exterior side walls of embedded structures. Without considering lateral support from the rock or soil against embedded structures, one computes structural responses at grade that are greater than the free field motion. However, the structure at grade could not respond significantly greater than the free-field motion if the embedded portion of the structure is laterally supported by the stiff soil or rock.
As was the case for the very first conservatism described (location of input motion), it would require some reanalysis to estimate the degree of conservatism involved in the SSI modeling of Ginna structures.
non-trivial It is obvious, however, that some degree of conservatism exits.
Ground Motion Incoherence - As has been documented in the EPRI seismic margin report (EPRI NP 6041) there can be a deamplification effect on nuclear type structures due to the incoherence of ground motion over the relatively large dimensions of typical nuclear structures. Conservative reduction factors as a function of frequency and building footprint have been documented within NP 6041 to account for the statistical incoherence of the input wave motion. These
conservative values range from a factor of 1.1 to around 1.5. More recent studies have documented even greater reduction factors. This ground motion incoherence is applicable to rock sites like Ginna.
Time Histo Simulation - ISRS at Ginna have been generated using a time history which is intended to approximate the desired earthquake spectrum (0.20g, Reg. Guide 1.60 shape).
This process involves the generation of an artificial time history whose response spectra envelops the SSE. The amount of conservatism involved in the enveloping process has not been specifically calculated for Ginna but can range up to a factor of 2 or more unless significant resources are applied to minimize the degree of enveloping.
Cli in of Narro~ Peaks - The SSRAP report and the Generic Implementation Procedure (GIP) recommend procedures for adjusting narrow peaks to reflect two areas of conservatism:
Narrow peaks are not as highly amplified in real structures as are predicted by linear elastic mathematical 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 Hz range for a 10 Hz peak frequency) using the unbroadened ISRS. The Ginna ISRS have narrow peaks and did not utilize the peak reduction methods from the GIP. The conservatism involved has been shown to be in the range of 5% to 20% for typical narrow peaks at several plants. We expect the conservatism for the peaks of the Ginna ISRS to fall within this range based on a sampling for a couple of peaks showing a 10% effect.
There are several additional sources of conservatism (e.g.,
structural damping, structural modeling, structural/soil peak broadening and enveloping, etc.) which addnon-'inearities, to the overall conservatism in the calculation of ISRS. 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 Ginna ISRS produce exceedance beyond the 1.5 factor.
- 2. Not a Si ificant Safet Issue The expected differences between calculated ISRS and actual building response do not represent a significant safety question.
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 conservatively addressed in the GIP.
~
The NRC staff I
acknowledged the seismic ruggedness of nuclear power plant equipment in the backfit analysis for USI 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]
Method A is only applicable to stiff equipment with fundamental frequencies over about 8 Hz. As noted earlier in Section 1 of this paper, SSRAP and SQUG have agreed that excitations over 8 Hz have little damage potential due to low spectral displacements, low energy content and short duration. This judgment is supported by industry and NRC guidance for determining whether an operating basis earthquake (OBE) is exceeded following a seismic event at a nuclear power plant. EPRI Report NP-5930 and NRC Regulatory Guide 1.166 recognize that damage potential is significantly reduced for earthquake ground motions above 10 Hz.
In other words, the question of what is the precise value of building amplification over 8 Hz has very little safety significance.
- 3. Ginna Buildin s are ical Nuclear Structures As requested, RGEE is also providing detailed description of the power block. building construction. The Ginna power block structures are typical nuclear power plant structures which were designed to resist lateral loads with reinforced concrete shear walls or braced structural steel frame systems. A summary description of the buildings and their foundations are contained in the attached Table 1.
- 4. Determination of "Grade Elevation" "Grade Elevation" determinations for Ginna Station power block building were described in Section 2.3 of the January 1997 submittal:
"Grade Elevation The power block structures at Ginna are built on the side of a hi'll. Grade elevation on the north (lake side) of the power block is 253'. Grade on the south side of the power block is 271'. For the A-46 project, a grade elevation of 253'as used for the structures on the north side of the power block (DG, IB, SH TB), and a grade elevation of used for the structures on the south side of the power 271'as block (AB, AF, CB). The containment (RC) is founded on rock at elevation 235'; 235'as used as the grade elevation for A-46.
It should be noted that CB 289's the highest elevation at which seismic SSEL equipment are located, and that the great majority of seismic SSEL equipment are at elevation 271'r lower. Therefore, for equipment outside containment, whether 253'r40'f271's used as grade would not impact the "within about grade" criterion commonly used in the GIP."
It should be noted that CB 289's the highest elevation at which seismic SSEL equipment are located, and that the great majority of seismic SSEL equipment are at elevation 271'r lower. Therefore, for equipment outside containment, whether 253'r40'f271's used as grade would not impact the "within about grade" criterion commonly used in the GIP."
In addition to previous discussions in Enclosure 2 describing the power block structures and corresponding "grade elevations", a general North-South site cross section is provided.
Conclusions The discussion above leads to several conclusions:
Cl All of the Ginna structures are large reinforced concrete shear wall or braced steel frame structures. They are typical of the structures designed for nuclear plants of.
the Ginna vintage and are "typical nuclear structures".
CI The results from actual measured ISRS on "nuclear type" structures support the 1.5 response levels advocated within Method A.
0 Qualitative assessments of the conservatism inherent within the methods utilized to calculate ISRS have been provided above. These conservatisms are typically quite significant (as has been independently verified by median/modern assessment such as the LLNL study) and can/will result in ISRS which show amplifications well beyond the 1.5 factor from Method A. RGEE feels strongly that the specific exceedances noted by the NRC (beyond the 1.5 factor) on Ginna are due to these high conservatisms inherent in the ISRS methods and not due to "unusual, plant-specific situations". Therefore, the application=of Method" A to the structures at Ginna is appropriate and valid.
CI 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; therefore RG&E's implementation of the GIP at Ginna resulted in significant seismic safety enhancements.
Table 1 Building Detailed Description of Building Construction DG The diesel enerator buildin (DG) is a one-story reinforced-concrete (Rc) structure that has two cable vaults underneath the floor. The building roof consists of a RC slab supported by four shear walls that sit on concrete spread footings. Zt is a relatively stiff structure typical of most diesel buildings at nuclear plants.
ZB The intermediate buildin (ZB) is located on the north and west sides of the containment building, and is founded on rock. The west end has a retaining wall where the floor at elevation 253 ft 6 in. is supported. The bottom of the retaining wall footing is at elevation 233 ft 6 in. Rock elevation in this area is at approximate elevation 239 ft 0 in. Foundations for interior columns are on individual column footings and embedded a minimum of 2 ft in solid rock.
SH The screen house-service water (SH) building is comprised of two superstructures, one for the service water (SW) system and one for the circulating water system (the screen house portion).
The service water (SW) portion of the building (both below and above grade) is a Seismic Category I structure. The service water (SW) portion houses four Seismic Category I service water (SW) pumps and Seismic Category Z electric switchgear. The screenhouse portion houses the traveling water screens and circulating water pumps. The entire screen house-service water (SH) building is founded in or on bedrock with the exception of the basement of the electric switchgear portion which is founded approximately 4 ft above bedrock. Since the building is founded in bedrock the basement will not realize any spectral acceleration and the seismic loading is equivalent to the ground motion of 0.08g and 0.20g. The building is constructed of RC below grade and has a structural steel superstructure.
TB The turbine buildin (TB) is a 257.5-ft by 124 .5-ft rectangular building on the north side of the power block. Zt has a concrete basement at elevation 253.5 ft, two concrete floors (a mezzanine floor at elevation 271 ft and an operating floor at elevation 289.5 ft). The building is a heavily braced steel structure.
The auxilia buildin (M)) is a three-story rectangular structure, 70 ft 9 in. by 214 ft 5 in. It is located south of the containment and intermediate buildings and adjacent to the service building. The structure has a concrete basement floor that rests on a sandstone foundation at elevation 235 ft 8 in.,
and two concrete floors--an intermediate floor at elevation 253 ft and an operating floor at elevation 271 ft. Construction below grade is (RC) with a structural steel su erstructure.
The standb auxilia feedwater buildin (AF) is a reinforced-concrete seismic category I structure with reinforced-concrete walls, roof, and base mat. The building is supported by 12 caissons which are socketed into competent rock.
CB The control buildin (CB)is located ad)acent to the south side of the turbine building and is a 41-ft 11-3/4 in. by 54-ft 1-3/4 in three-story structure with concrete foundation mat at elevation 253 ft. The foundation of the control building is supported on lean concrete or compacted backfill. The rock elevation in this area is at approximate elevation 240 ft. 0 in.
The foundation of the control building was excavated to the surface of the bedrock. The fill material was placed on the rock surface to a depth coincident with the control building foundation. The bottom elevation of the deepest portion of the foundation mat is at elevation 245 ft 4 in., with a structural slab supported at elevation 250 ft 6 in. with a thickened slab for column footings. The building consists of both RC and structural steel.
The containment buildin (RC) is a vertical right cylinder with a flat base and a hemispherical dome. The building is 99 ft.
RC high to the spring line of the dome and has an inside diameter of 105 ft. The cylindrical concrete wall, which is prestressed vertically and reinforced circumferentially with mild steel deformed bars, is 3.5-ft. thick. The concrete dome is a reinforced concrete shell 2.5-ft. thick. The concrete base slab is 2 ft thick with an additional thickness of concrete ft fill over the bottom liner plate. The containment cylinder is of 2 founded on rock (sandstone) by means of post-tensioned rock anchors which ensure that the rock then acts as an integral part of the containment structure.
4 Screen house Diesel generator annex Turbine bldg.
Intermediate bldg.
Control Service I uildin building 1 1
Reactor I containment ~Facade building 1
~ Sl Auxiliary building Aux. bldg, addition ROCHESTER GAS AND ELECTRIC CORPORATION R.E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 3.7-6 Containment Building and Complex of Interconnected Seismic Category I and Nonseismic Structures, Flan View
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ENCLOSURE 3 SECOND RE UEST FOR ADDITIONAL INFORMATION R. E. GINNA NUCLEAR POWER PLANT uestion ¹2 In your response to the staff's RAI Questions ¹3 and ¹4, for a number of equipment items, the equipment frequencies were stated to have been-judged by SRT to be greater than 8 Hz by inspection. Provide the basis for the SRT judgement regarding equipment naturaI frequency, especially when the estimated magnitude for natural frequency is relied upon to determine the applicability for the use of GIP-2 Method A.l. You are requested to provide further justification for the frequency estimation, or provide analytical calculations to justify such estimation, for the equipment items identified as FT-4084, FT-4085, PSF01A a B, SAFVPCIP, SAFVPDIP, BVSZ4, DCPDPAB01A&B and DCPDPAB02AaB.
Res onses to uestion ¹2 The bases for the SRT judgement was:
II
- a. The guidance provided in the EPRI report, "Guidelines for Estimation or Verification of Equipment Natural Frequency", Research Project 2925-2, Final Report, August 1992.
- b. The experience of the SRT. The SRT was composed of RGEE staff and staff from an outside consultant, Stevenson 6 Associates,(S&A). The SaA staff members on the walk downs were Dr. John Stevenson, Mr.
Walter Djordjevic, and Mr. Stephen Anagnostis. Dr. Stevenson has 35 years of experience in the nuclear power industry; Mr. Djordjevic and Mr. Anagnostis each have 15 to 20 years of experience in the industry. In addition, Mr. Djordjevic and Mr. Anagnostis each have extensive experience performing in-situ modal (frequency response) tests of nuclear power plant equipment. This testing involves dozen of pieces of equipment at more than ten nuclear power stations. The equipment tested includes control cabinets, motor control centers, switchgear and instrument racks. This test data is the basis of EPRI Report NP-7146, "Guidelines for Development of In-Cabinet Amplified Response Spectra for Electrical Benchboards and Panels", and the procedures for calculating in-cabinet spectra described in GIP-2 Section 6.4.2 (Screening Level 3) .
FT-4084 and FT-4085 Standb AFW Pum Flow Transmitters These flow transmitters are individually mounted to a steel base plate and anchored to a reinforced concrete wall with four (4) 3/8" concrete expansion anchors, as shown in picture below. The SRT judged this equipment to have a fundamental frequency greater than 8 Hz based on the small size, small weight, and stiff support.
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The EPRI guidelines do not discuss individually mounted pressure switches, but do discuss steel frame instrument racks housing a number of pressure switches and related equipment. Section 3.3 of the guidelines state that "Often, braced racks will have frequencies greater than 8 Hz. The walkdown team need only be cautious of very large, heavily weighted, very weakly-braced racks, or very low braced racks". Based on this guidance, it is reasonable to conclude that an individual transmitter, securely mounted to a reinforced concrete wall, will have a fundamental frequency greater than 8 Hz.
PSF01A PSF01B Standb AFW Pum s C and D These are 300 HP electric-motor horizontal pumps. Each pump is mounted on an approximately 3'-6" wide by 10'-6" long steel skid. The skid is anchored to a 27" high continuous concrete pedestal with twelve (12) 3/4" cast-in-place bolts. The pedestal is well reinforced and doweled into the floor slab.
There are no vibration isolators.
Section 2.2 of the EPRI guidelines states:
"Further, the following classes of mechanical equipment are considered to be sufficiently rugged that the walkdown team may assume, without further justification, that their natural frequencies are above 8 Hz:
Pumps Engine and Motor Generators Air Compressors Fans and Air Handlers Chi l,lers Testing on shake tables and in the field has shown that equipment in these classes have natural frequencies greater than 8 Hz given that they have direct anchorage to the floor and that appendages such as very flexible control panels are not present. One possible exception is that deep well pumps may have unsupported cantilever columns and suction bowls with natural frequency below 8 Hz (Note that this addressed by a separate SQUG caveat) .
SAFWPCIP SAFWPDIP Standb AFW Pum Instrument Panels These are wall mounted instrument panels. Each panel is 30" wide x 54" high x 32" deep and is welded to continuous angles running along the top and bottom of the panel. The angles are secured to a reinforced concrete wall with 5/8" concrete expansion anchors (Both top angles and one of the bottom angles are secured with three anchors; the other bottom angle is secured with two anchors.)
The EPRI guidelines indicate that the presence of well-engineered top bracing on electrical equipment is sufficient to support the judgement that the equipment has a fundamental frequency above 8 Hz. The SRT experience is that typical (30" wide x 30" deep x 90" tall), properly anchored, floor mounted instrument cabinets have a fundamental frequency in the range of 10 Hz - 15 Hz. Smaller, wall mounted cabinets of similar construction (such as the subject panel) have at least as high a fundamental frequency.
BUS14 480V Switch ear This is a Westinghouse Type DB low voltage switchgear, 160" wide x 58" deep x 76" high. The switchgear is anchored with ten (10) 3/4" Hilti Kwik bolts.
Section 3.7 of the EPRI guidelines states:
"As long as a lineup has six sections or more, the natural frequency of swi tchgear units may be assumed to be above 8 Hz".
S&A has tested a similar Westinghouse low voltage switchgear at the Connecticut Yankee nuclear station. That unit (designated BUS4 at CY) is 408" wide x 54" deep x 90" high. A fundamental frequency of 9.0 Hz was measured.
DCPDPAB01A and B DCPDPAB02A and B Aux Buildin DC Power Distribution Panels These are wall mounted panelboards (power distribution panels). Each panel board is 30" wide x 40" wide x 12" deep, and well anchored, near the top and bottom, to a reinforced concrete wall or a reinforced concrete column.
Section 2.3 of the EPRI guidelines states that panelboards (wall-mounted distribution panels), if anchored to a substantial floor or wall, will have a fundamental frequency of at least 12 Hz.
uestion 3 In your response to the staff's RAI, Question ¹4, you indicated that the seismic capacity vs. demand evaluation for the Undervoltage Relay Cabinet Bus 24 QRAZRC24) was based on shake-table testing. You are requested to provide a detailed discussion of the testing and to justify the adequacy of such testing.
Res onse to uestion ¹3
References:
American Environments Company, Inc., Report No. STR-142280-1, 11/4/80 (RG&E Project EWR-1444)
American Environments Company, Inc., Report No. STR-142280-2, 12/31/80 (RG&E Project EWR-1444)
American Environments Company, Inc., Report No. STR-142280-3, 1/5/81 (RG&E Project EWR-1444)
There are eight (8) relay/control cabinets of this make on the SSEL. Their equipment designations and locations are listed below:
ARA1CC14 Control Building 271.00 ARA1RC14 Aux Building 271.00 ARA2CC18 Screenhouse 253.00 ARA2RC18 Screenhouse 253.00 ARB1CC16 Control Building 271.00 ARB1RC16 Aux Building 253.00 ARB2CC17 Screenhouse 253.00 ARB2RC17 Screenhouse 253.00 The four cabinets outside the screenhouse are each stand-alone and 24" wide x 24" deep x 70" high. The four cabinets in the screenhouse consist of two pairs of attached cabinets; each pair is 48" wide x 24" deep x 70" high. Each 24" x 24" x 72" cabinet is anchored to a reinforced concrete floor with four (4) Hilti Kwik-Bolts.
The cabinets were shake-table tested as documented in References 1 through 3.
The shake-table tests were random, multi-frequency, and biaxial (one horizontal and the vertical direction); the tests were repeated with the specimen rotated 90 degrees about the vertical axis. The "RRS vs. FRS" plot below shows the tests'RS (Required Response Spectra) compared to the envelope of the FRS (Floor Response Spectra) for all locations listed above.
The RRS envelopes the FRS,'ut note that the RRS is 3% damped, while the FRS is 4% damped. However, the RRS is well above the FRS in the peak range, and the actual test response spectra exceeded the RRS by a substantial amount, particularly for frequencies above the peak range. Section 5.0 of the test reports state:
"The test specimen continued to function before, during and after exposure to the Seismic Qualification Test Program. There was no evidence of physical damage, or reported electrical malfunction observed
. as a result of the stresses of this test program."
10 Required Response Spectrum, 3o/o damping Envelope of Floor Response Spectra, 4'4 damping p
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0.1 10 100 Frequency (Hz) uestion 44 For the 48D VAC Motor Control Center (MCC)-, you indicated that the MCC can withstand a single frequency test consisting of a 1.35g, 5 beat, 5 cycle/beat input, performed at the signficant structural frequencies.
It is known that single-axis, single frequency sine beat tests'ostly performed prior to the issuance of IEEE Standard 344-1975, are considered inadequate for equipment seismic qualification due to their inability to excite multi-axis, multi-frequency responses of equipment (the very reason that plants are included in the USI A-46 program). You are requested to justify the seismic adequacy of this motor control center.
Res onse to uestion 4
Reference:
Letter from G.R. Geertman (Gilbert/Commonwealth) to C.J.
Mambretti (RG&E) dated Au<fust 11, 1976.
The subject MCC (equipment designation MCCL) is a 4 section Westinghouse MCC, 66" wide x 19" deep x 90" high.
expansion anchors. The MCC is located on elevation It is anchored with eight (8) 1/2" concrete 271'f the Auxiliary Building.
The following plot contains several response spectra to illustrate the 10 discussion that follows.
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Rersponse spectrum tor 5 cycrbeat wylae Lab Test {4% damping)
Response spectrum used by westinghouse tor Analysis (ass damping)
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The reference contains a summary report from Westinghouse documenting the MCC's seismi'c qualification. The MCC was originally tested in 1972 at Wyle Labs following the requirements of TEEE 344-1971. The test. used a single frequency, 5 cycle/beat sine beat input dwelled at the significant structural frequencies of the MCC (by test, the fundamental frequency was found to be 8.5 Hz) . The motion was simultaneously applied in the-horizontal direction at 1.35g and in the vertical direction at 0.95g. The corresponding response spectrum is shown in the above figure as the thinner solid line.
Subsequently, Westinghouse performed a multi-frequency, multi-directional dynamic analysis using the response spectrum shown in the above figure as a dashed line. Westinghouse termed this response spectrum the nGinna Station SSE Required Response Spectrum". The analysis showed that the in-structure accelerations induced by this response spectrum were about. 1/2 those induced by the test. Based on this, Westinghouse concluded that the MCC was seismically qualified.
The basis of Westinghouse's analysis spectrum (the dashed line) is not known.
For comparison, the A-46 floor response spectrum for Aux Building 271'the location of the MCC) is shown in the above figure as the thicker solid line.
The A-46 spectrum is substantially below both the Westinghouse analysis spectrum and the Wyle test spectrqm) .particularly in the fundamental frequency range at 8.5 Hz. The ZPA of the A-46 spectrum is 0.3g, or less than 1/4 of the test input level.
While the original sine beat test was not as sophisticated as current-day multi-frequency, multi-axis tests, the SRT concluded that the high acceleration levels used in the test were more than adequate compensation.
Note that the test data documented a fundamental frequency greater than 8 Hz and that Aux Building 271's within 40'f effective grade, therefore the MCC meets the screening requirements for GIP Method A. The SRT chose to base its acceptance on the test report, rather than Method A, because test report to be a more compelling argument.
it found that the