ML20153G790
| ML20153G790 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 02/24/1986 |
| From: | Chen C GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT |
| To: | |
| Shared Package | |
| ML20153G706 | List: |
| References | |
| OL, NUDOCS 8602280430 | |
| Download: ML20153G790 (78) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensino Appeal Board In the Matter of )
)
THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY, et al. ) 50-441
)
(Perry Nuclear Power Plant, )
Units 1 and 2) , )
AFFIDAVIT OF DR. CHANG CHEN, County of Berks )
- ss.
State of Pennsylvania )
CHANG CHEN, being first duly sworn, deposes and says as follows:
- 1. I, Chang Chen, am Manager of the Civil / Structural De-partment and Chief Structural Engineer for Gilbert Common-wealth, Inc. (" Gilbert"). My business address is Route 10 and Pheasant Road, Green Hills, Reading, Pennsylvania, 19607. I have personal knowledge of the matters set forth in this Affi-davit, which are true and correct to the best of my knowledge and belief.
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- 2. Gilbert is an international architect-engineering firm specializing in the design of nuclear and fossil power plants in the United States and abroad. Gilbert has designed and engineered 15 nuclear plant units in the United States, Japan, Korea and Yugoslavia. Gilbert is the principal Archi-tect/ Engineer for the Perry Nuclear Power Plant.
- 3. In my present position, I have the overall responsi-
, bility for all civil / structural work associated with all of the power plants (both nuclear and fossil), including Perry, designed by Gilbert. I also am responsible for nuclear power plant equipment seismic qualification on all Gilbert-designed nuclear plants. There are over 100 engineers and designers re-porting to me in my present position. I have been employed by Gilbert for over 16 years. I have supervised the seismic anal-t ysis and design of the Perry Plant, including development of the Perry design response spectra, since Gilbert commenced the engineering for Perry in 1972. While working at Gilbert, I have also been responsible for major seismic design reviews of nuclear plants in other countries designed by other firms, including as a consultant to Kraftwerk Union (KWU) in connec-tion with the seismic design of a 1,300 MW nuclear plant in Iran.
- 4. A statement of my professional qualifications is attached hereto as Exhibit "A". As indicated therein, I hold a Bachelor of Science degree in Civil Engineering (1962) from h
Cheng Kung University in Taiwan; a Master of Science in Civil Engineering from Duke University (1965); and a Ph.D in Engi-neering Mechanics from The Pennsylvania State University (1969). I am a Registered Professional Engineer in Pennsylvania. I have published over 25 articles in the fields of nuclear plant civil / structural design and earthquake engi-neering.
- 5. On January 31, 1986, an earthquake occurred in north-ern Ohio (the "1986 earthquake"). Immediately afterwards, I and other Gilbert personnel under my supervision undertook a number of investigations to assess any impact of the 1986 earthquake on the sei .ic design of Perry as reflected in the FSAR. Our studies were based on data recorded by seismic in-strumentation in the plant buildings (see Affidavits of Kalman Lee Benuska and Paul D. Engdahl) and on equipment qualification data supplied by vendors. The purpose of this Affidavit is to describe the results of these studies and to give my conclu-sions as to the adequacy of the current Perry seismic design.
- 6. This Affidavit first gives general background information on seismic design of nuclear power plants and the development of the seismic design for Perry in particular. The Affidavit then briefly describes the seismic instrumentation installed at Perry. Next, an evaluation of the engineering significance of the 1986 earthquake is made as to Perry
structures, systems, and components, followed by a specific evaluation of equipment margins.
BACKGROUND ON NUCLEAR POWER PLANT SEISMIC DESIGN
- 7. The seismic design basis for nuclear power plants is established by requirements in 10 CFR Part 100, Appendix A, and NRC Regulatory Guida 1.60. These regulations require nuclear plant structures and safety class systems and components to be designed to withstand loads induced by a " Safe Shutdown Earth-quake" (SSE) for the particular site. The SSE is the strongest earthquake in terms of magnitude of vibratory ground motion that is ever expected to occur at a particular site. The SSE is the design basis earthquake considered for plant licensing.
A second seismic event also considered in designing nuclear plants is the " Operating Basis Earthquake" (OBE). The OBE is the strongest earthquake considered likely to occur at a par-ticular site and is at least one-half of the SSE. Operations may resume follcwing an earthquake which exceeds the OBE after demonstrating that no functional damage has occurred to safety-related plant features (10 CFR Part 100, Appendix A).
- 8. The SSE can be described by means of a " response spectrum," which depicts the maximum acceleration, velocity or displacement response to an input excitation (here the SSE) at a_specified damping value for single degree-of-freedom oscilla-tors of varying natural frequencies. The high frequency end of a response spectrum indicates the "zero period acceleration" (ZPA) associated with the event.
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- 9. In the design of any plant, it is difficult to pre-dict the shape of postulated earthquake acceleration time-
> histories and associated ground response spectra. Appendix A of 10 CFR Part 100 therefore requires an expected SSE to be developed by statistically combining the response spectra from multiple historical earthquakes. Following this guideline, the NRC has provided in Reg. Guide 1.60 standardized response spectra that can be used in lieu of spectra developed for each site (see Fig. 1). These standardized spectra were derived by normalizing and combining spectra calculated from numerous sets of historically recorded acceleration time-histories. From these sets of spectra, smoothed response curves (acceleration, velocity and displacement) were generated at a level equal to
- one standard deviation greater than the mean of the responses.
This method provides an 84% level of statistical confidence that responses at any particular frequency will not be exceeded by any future SSE event.
- 10. Thus, in lieu of developing site-specific SSE ground response spectra, the standardized response spectra of Reg.
Guide 1.60 can be used. The standardized spectra need only be scaled up or down to reflect the effective maximum ground ac-celerations (i.e., ZPA's) expected for the SSE at that site.
The SSE design ground response spectra are used to dynamically analyze a lumped-mass model of the power plant structures.
BACKGROUND ON SEISMIC DESIGN OF THE PERRY PLANT
- 11. The Perry design response spectra were derived by using the standard response spectrum of Reg. Guide 1.60 scaled to a ZPA of 0.15 g determined for the Perry site. This was used to generate the design response spectra at the foundation elevations for use in designing the plant buildings.
- 12. From these spectra, a simulated SSE time-history of ground accelerations was developed for each directional compo-nent (N-S, E-W, and Vertical). The conservatism of these simu-lated time-histories was checked and confirmed by assuring that the response spectra generated from the simulated time-histories envelop the Reg. Guide 1.60 design response spectra (see Fig. 2).
- 13. Seismic Category I structures were analyzed by apply-ing the simulated time-histories to a lumped-mass model of the entire structure, as shown in Figure 3. From this analysis, time-history accelerations at each floor elevation were also derived. These time-histories were then used to derive re-sponse spectra for each floor of each main building. The floor response spectra were used in designing the safety class equip-ment, components, and systems.
- 14. In addition to the conservatism included in the deri-vation of response spectra, there were numerous other conserva-tisms included in the overall design of the Perry structures,
systems and components. Examples of some of the more signifi-cant conservatisms are as follows:
- a. Broadenino the Envelope of Floor Response Spectra Frequency bands of floor response spectra were artificially broadened (typically by 15%) to account for possible frequency variations. Responses used for design were thus overestimated for systems having more than one dominant frequency falling into the broadened frequency i
bands of the floor response spectra.
- b. Equipment Qualification by Test Equipment qualified by shake table testing used time-histories simulated from the floor response spectra.
The simulated time-histories were generated in such a way 4 that their calculated response spectra envelop the broad-ened floor response spectra, which in turn already envelop the original floor response spectra. The conservatism of the time-histories was increased by this " envelope on top of an envelope" process. Moreover, this process resulted in simulated time-histories with maximum accelerations much higher than the ZPA's of the original floor response spectra.
- c. Strain Hardenino Not Accounted For and Static Allowables Used for Dynamic Load In equipment design, material is assumed to be-have linearly up to the yield point, then to deform con-tinuously to collapse when the external load is e
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maintained. All material used in equipment design exhibits characteristics of strain hardening. This means that resistance to deformation increases after the defor-mation exceeds the yield point. Furthermore, even if no strain hardening is assumed, the material can resist dy-namic loads having peak values higher than the yield strength through the absorption of energy in the plastic region.
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- d. Loadina combinations The plant was designed to withstand load *ng com-binations with a very low probability of simultaneous oc- .
currence. For example, some load combinations included seismic loads, hydrodynamic loads, and hypothetical loss-of-coolant-accident loads simultaneously. This results in design capability well above the loads associated with seismic alone. I
- e. Primary Versus Secondary Stresses Computed seismic stresses used in design were ,
considered to be primary, non-self-limiting stresses in-stead of secondary stresses with a self-limiting nature.
The actual behavior of seismic stresses is somewhere be-tween a primary and secondary nature. Consideration of seismic stresses as primary stresses resulted in conserva-tive values used for design.
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- f. Damoino Values Conservative damping values were employed at Perry pursuant to NRC Regulatory Guide 1.61. The recent ASME Code Case N-411 (not employed at Perry) would have allowed increased (i.e., less conservative) damping values l to be used. l SEISMIC INSTRUMENTATION AT PERRY
- 15. Three different types of seismic monitoring instru-mentation were used to record the 1986 Ohio earthquake. Table 1 indicates the building and specific location of each of the
< instruments. Figure 4 shows a plan and an elevation view of each instrument location. Figures 5 through 12 show the mount-ing details for each instrument.
- 16. One type of instrument used was the Kinemetrics Model SMA-3 strong motion triaxial time-history accelerograph. This system detects and records three mutually perpendicular compo-nents of acceleration over the entire duration of the earth-quake onto cassette magnetic tape. Power to the unit is supplied by internal rechargeable batteries which are kept in a charged state by 120 VAC line power. Two instruments of this type were used and were located on the Reactor Building founda-tion mat at an elevation of approximately 575 feet. Further information on the Kinemetrics instruments is set forth in the Affidavit of Kalman Lee Benuska.
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- 17. The second type of instrumentation used was the Engdahl PSR 1200-H/V response spectrum recorder. This totally mechanical system also records three mutually perpendicular components of acceleration. The instrument uses twelve reeds l fabricated of varying lengths and weights of spring steel, one for each frequency (ranging from approximately 2 Hz to 25 Hz). !
A diamond-tipped stylus is attached to the free end of each reed to inscribe a permanent record of its deflection on one of twelve record plates. The record plates are made of aluminum and are plated with successive layers of nickel, tin and lead-tin. This system is totally self-contained and requires no outside power source.
- 18. Four instruments of this type were used -- two on the Auxiliary Building foundation mat at an elevation of approxi-I mately 568 feet, one at the Reactor Building foundation mat at an elevation of approximately 575 feet, and one at the Reactor Building Inside Drywell Platform at an elevation of approxi-l mately 630 feet.
- 19. The third type of instrument was the Engdahl PAR 400 peak accelerograph. This totally mechanical system records three mutually perpendicular components of peak local accelera-tion (i.e., the zero period acceleration). A diamond-tipped scriber at the end of an amplifier arm records a permanent mark on a record plate made of aluminum and successive layers of nickel, gold and burnt gold. Again, this system is totally n .
self-contained and requires no outside power source. Two instruments of this type were used and were located on the Aux-iliary Building foundation mat at an elevation of approximately 568 feet and on the Reactor Recirculation Pump at an elevation ,
of approximately 605 feet. A third instrument of this type was out of service at the time of the 1986 earthquake because it was being recalibrated. Further information on the Engdahl instruments is contained in the Affidavit of Paul D. Engdahl.
- 20. The instrumentation installed at Perry as described conforms to the requirements of NRC Regulatory Guide 1.12
(" Nuclear Power Plant Instrumentation for Earthquakes, Rev.
1").
EVALUATION OF THE JANUARY 31 EARTHQUAKE
- 21. Based on data collected by the National Earthquake Information Center of the United States Geological Survey (USGS), the January 31, 1996 earthquake had a magnitude of M bLg = 4.96 with an epicenter at about 11 miles (17.7 Km.)
south of the Perry plant sits, This is of much leso magnitude than the earthquake for which the plant was designed (the SSE) and contained substantially lower total energy than the Perry SSE. Evidence of the low energy content of the January 31 earthquake is shown by a comparison of the acceleration time-histories it induced at various elevations with the corre-sponding design acceleration time-histories (see Figs. 13 through 18). The time-histories used for design are 22 seconds
l long and of sustained high amplitude (strong motion). By con-trast, the January 31 time-histories are about 5 seconds long and contain strong motion in only less than a one-second inter-val (total) of the event.
, 22. A comparison of Figures 1 (Reg. Guide 1.60 response spectra) and 19 (sample response spectra from the January 31 earthquake) gives a further indication of the low energy con-tent of the January 31 event. These figures show that the Reg.
Guide 1.60 spectra used for design have much broader frequency contents than those of the recorded earthquake, which contain strong motion only at high frequencies. The design earthquake i
therefore contains much greater total energy.
- 23. Table 2 compares the structural response ZPA's of the l recorded data with those of the SSE and OBE. The square-root-1 1
of-the-sum-of-the-squares (SRSS) comparison indicates that the recorded values of the 1986 earthquake vary from significantly i
below OBE values to 74% of SSE values, except at elevation 686 4 feet of the Reactor Building Containment Vessel. At that loca-tion, the N-S and Vertical acceleration components exceed SSE
' values, while the E-W acceleration component is less than the l SSE value. In addition, recorded response spectra accelera-tions show that the design response spectra accelerations in certain instances were exceeded at the high frequency end of the spectra. At lower frequencies (at or below approximately 14 Hz) the recorded accelerations are all well under the design
values (see response spectra comparisons, Table 3, Figures 20 through 31).
- 24. The measurement of accelerations outside the pre-dicted responses at the high frequency ends of certain response spectra has no engineering significance. This is explained by the interrelationships among the frequencies, accelerations, velocities, and displacements associated with a seismic event.
In general, high frequency acceleration responses have corre-spondingly low velocity and displacement responses. The 1986 earthquake accelerations occurred at very high frequencies.
Therefore, despite some recorded maximum acceleration responses which exceeded SSE values at higher frequencies, corresponding velocities and displacements (and resulting stresses) were nev-ertheless acceptably low.
- 25. Confirmation of this is shown in Table 4, which indi-cates the maximum relative displacements from the recorded time-histories for the Reactor Building Containment Vessel.
Tne overall SRSS relative displacement shown in the Table is 0.34 cm for the SSE and 0.10 cm for the actual event. Since structural stress is proportional to relative displacement, and the recorded relative displacement was far less than the SSE design value, the stresses induced by the 1986 earthquake at this location were well within design capabilities despite the acceleration exceedances described above. This small relative displacement is consistent with the high frequency nature of the disturbance. The high frequencies combined with the short duration resulted in an earthquake that contained very low total energy compared to the SSE.
- 26. The maximum recorded velocity at the top of the Reac-tor Building foundation mat during the 1986 earthquake was 0.87 inches /sec (2.21 cm/sec). This can be compared with the Bureau of Mines ("BOM") velocity threshold for no di) mage to non-encineered buildings, which is 1 inch /sec (2.54 cm/sec). This shows that the BOM considers it acceptable for blasting work or pile driving operations to induce velocity waves in nearby res-idential housing foundations that are greater than the maximum velocities induced by the 1986 earthquake at the Perry plant.
I This example provides perspective on just how low the 1
velocities and energy content of the 1986 event were. .
- 27. As described in the Affidavit of Robert A. Stratman, extensive plant inspections and operability checks have indi-cated_that no structural or equipment damage resulted from the i
1986 earthquake. This is as expected based upon the low ener-gy, short duration, and low velocities and displacements of the event. The inspections and operability checks that were per-
. formed.were adequate to detect any structural damage. For these reasons, it is unnecessary to perform any further inves-
, tigations such as containment integrated leak rate testing, hy-drostatic testing of the reactor coolant pressure boundary, nondestructive structural testing, or any other type of testing.
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- 28. Although some hairline cracks in the structural con-crete were documented during plant walkdowns, this does not constitute damage. Reinforced concrete structures are expected to show hairline cracks. Regardless of their cause, such cracks have no effect on the strength and integrity of the structures. Moreover, in my judgment, such cracking is not attributable to the 1986 earthquake because of the low magni-tude of the event.
- 29. Section 7.5 of IEEE 344, " Recommended Practices for Seismic Qualification of Class lE Equipment for Nuclear Power Generating Stations," was employed at Perry. Confirming the above discussion, this standard recognizes that short duration /
high frequency / low energy input motions will not cause signifi-cant structural stresses, and thus prohibits the use of such input motions to qualify equipment. Instead, it requires qual-ification by long duration / broad-band frequency /high energy testing to provide conservatism.
EVALUATION OF SPECIFIC DATA
- 30. In light of the above discussion, recorded responses at particular locations can be evaluated. At all four instru-ment locations recording response spectra, SSE design spectra are well above the recorded spectra in the frequency range of 1 Hz to 14 Hz (see Figures 20 through 31). These figures com-pare recorded data with the appropriate design spectra at adja-cent elevations. These figures also compare the data from
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different types of seismic instrumentation at the same elevation.
- 31. At high frequencies, the design spectra are exceeded by recorded values in certain cases. However, the corre-sponding displacements based on recorded data are all extremely small (on the order of several one-hundredths of an inch) at 20 Hz (where peak acceleration exceedances occur). These ex-tremely low displacements conform to the above analysis, de-monstrating that the stresses at higher frequencies are insig-nificant despite acceleration exceedances.
- 32. In evaluating the spectra data recorded at the vari-ous locations, it was noted that the acceleration responses at the Reactor Building Platform outside the Biological Shield Wall varied from the general pattern of responses recorded at the other three locations. The recorded N-S and E-W accelera-tion components for this location are all well-enveloped by the entire range.of the SSE spectra, while the recorded vertical acceleration component exceeds the SSE spectra at the high fre-quency end (see Fig. 28). This response may be due to the fact that this particular Engdahl PSR-1200 instrument is located near multiple supports and piping system snubbers and compo-nents. Actuation of snubbers or local loads induced by nearby components may have influenced the recorded vertical response.
Such impacts would be of a local, secondary nature. Regard-less, the low energy, short duration, high frequency nature of A
the event indicates that these accelerations had no engineering significance. The recorded displacement spectrum value is only l 1
0.023 inches (0.06 cm) at 25 Hz at this location.
- 33. In general, the high frequency acceleration content of ground motion will be filtered out by buildings and thus will not appear at higher elevations. This is due in part to the low participation factor generally associated with modes at the higher frequencies. This phenomenon is exhibited by the responses recorded at the Reactor Building mat and elevation 686 feet of the Reactor Building Containment Vessel. A very high frequency p-wave was recorded at the Reactor Building foundation mat. The time-histories shown in Figures 13 through 18 indicate that this p-wave (appearing during the first second or so of the time-histories) was filtered out by the building f and did not appear at elevation 686 feet.
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! 34. There was a response in the range of 20 Hz that was transmitted to the higher elevations. The explanation for this involves the structural characteristics of the buildings on the Reactor Building foundation mat. The Reactor Building consists i of multiple structures sitting on a common foundation mat -- a concrete shield building, steel containment vessel, concrete drywell wall, and biological shield wall. The structural re-sponse of each building influences the responses of the others. :
The mode shapes and participation factors of the two most domi-nant vibration modes -- roughly 4 Hz and 18.4 Hz -- are shown 4 , . - - . _ _ - - - . , . -
in Figures 32 through 34. These two dominant frequencies cor-respond to the peaks at 4 Hz and 20 Hz on the recorded spectra for the Reactor Building at the mat and elevation 686 feet.
The input motion at 20 Hz (corresponding to the s-wave) was am-plified by this latter mode with some rigid body motion. The 20 Hz input was thus not filtered out but did appear at the higher elevation. As discussed, the acceleration peaks at 20 Hz at this location correspond to very small relative dis-placements and thus are not significant in an engineering sense.
l 35. I have reviewed the Motion to Reopen and to Submit a New Contention submitted by intervenor Ohio Citizens for Re-sponsible Energy ("OCRE") dated February 3, 1986 (the "OCRE Mo-tion"). OCRE refers in its Motion to a news account " stating that accelerations from the (1986] earthquake were estimated to range from 0.19 g to 0.25 g. Perry is designed to withstand 0.15 g (safe shutdown earthquake)." OCRE Motion at 2. OCRE relies on this news account as a basis for calling the Perry seismic design basis into question. Id. The news account, however, compares two different types of measurements. The 0.19 g and 0.25 g values referred to apparently were prelimi-nary readings of the ZPA's at two basemat locations. The value of 0.15 g, on the other hand, represents the postulated maximum vibratory ground motion (SSE) in the free-field. To compare like quantities, the recorded ZPA's should be compared against SSE ZPA's at the same locations derived by analysis, as is done
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in Table 2. That Table shows that the recorded SRSS ZPA's at these two basemat locations were well under their design values (0.18 vs. 0.33; 0.23 vs. 0.31). OCRE's citation to this matter thus is not a basis for calling the Perry seismic design basis into question. In any event, exceedances above the design basis response spectra which occurred in the January 31 earth-quake are of no significance to the plant's seismic design for the reasons set forth above.
REEVALUATION OF EQUIPMENT QUALIFICATION
- 36. As indicated in the Affidavit of Robert A. Stratman, all energized plant equipment functioned during this event as designed. To confirm the design adequacy of the active equip-ment, the qualification data for equipment listed in Table 5 has been compared against recorded response spectra. The eval-uation shows that the original conservatism in the equipment qualification was more than adequate to accommodate the recorded event.
- a. Selection of Eauipment to be Evaluated As described above, there are four sets of recorded response spectra at the following locations:
(1) Reactor Building Mat elevation 574'-10:
(2) Reactor Building Platform elevation 630' (3) Containment Vessel elevation 686' (4) Auxiliary Building Mat elevation 568'
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There is no equipment at location 1 because of the sup-pression pool. At the Reactor Building 630' and 686' elevations, the single records available at each location may be biased by secondary effects of adjacent equipment on the building response. The Auxiliary Building Mat elevation 568' has two seismic instruments which provide confirmation of the measured responses. Thus, the Auxil-iary Building Mat elevation 568' was selected as the most appropriate location for comparison of equipment data.
- b. Method and Results of the Marcins Evaluation An envelope of the records from the two Engdahl response spectrum recorders at the Auxiliary Building Mat was used to represent the recorded response spectra. The highest frequency of the recorded data from these instru-ments is at 25.4 Hz. The recorded spectra were extended to higher frequencies by extrapolating to ZPA values at 40 Hz as recorded by Engdahl PAR-400 instruments No.
D51-R120 and No. D51-R140, as shown in Figures 35, 36 and
- 37. The peaks of the 3% damping spectra were obtained by reducing the peaks of the 2% damping spectra by 12%. The 12% reduction factor was derived by examination of the ratio of 2% and 3% spectra obtained from the Kinemetrics instruments.
Active components that were evaluated are listed in Table 5. The results of the comparisons are as fol-lows:
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(1) Instrument Racks Instrument racks were originally qualified by testing. The response spectra from testing far exceed the spectra from the 1986 earthquake. An example of this is shown in Figure 38.
(2) Pressure Transmitters and' Flow Transmitters l
To compare the recorded spectra with the original spectra from testing, the recorded spectra were first amplified to represent spectra at the transmitter locations inside the racks. The test response spectra were found to envelop the amplified recorded spectra with ample margin. An example of this comparison is shown in Figure 39.
(3) Pumps and Motors Pumps and motors supplied by General Elec-tric were originally qualified by analyses. These analy-ses'were rerun with recorded spectra from the 1986 earth-quake as input. A dynamic finite element analysis of each piece of equipment was performed using the response spectra method. The SAP finite element program was used to analyze these dynamic models. The earthquake loads de-rived from the dynamic modeling were combined with previ-ously determined static loads such as piping nozzle loads, deadweight, maximum operating pressure, and pump operating loads. The resulting equipment stresses were found to be under the design allowable valuen.
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- c. Marcins of Other Ecuipment The above comparisons were made for equipment at the foundation level of the Auxiliary Building. Equipment and components at other locations are similarly deemed to have adequate design capability to accommodate events such as the 1986 Ohio earthquake for the following reasons:
(1) The typical comparisons of the response spectra from testing with the recorded response spectra s indicate that margins are ample, as shown in Figures 38 and 39.
(2) The pumps and motors that were analyzed have natural frequencies at 18.7 Hz, which is in resonance with the peak region of the recorded response spectra after 15% broadening. This analysis therefore included the most critical response spectra comparisons in terms of the resulting stresses.
(3) Floor response spectra at higher elevations will have higher peak values compared to spectra at lower elevations when the frequency of the earthquake input co-incides with the fundamental structural mode, which domi-nates the building response. The mode at Perry corre-sponding to the 20 Hz peaks in the recorded spectra is not a fundamental mode, and its mode shape is not one that would contribute to significant amplification at the higher elevations (see Figure 33). Therefore, the floor response spectra at upper elevations are not much higher than those at lower elevations for the 1986 earthquake.
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(4) The BWR 6 equipment and components used at Perry are over-qualified in the high frequency region be-catne of the conservative assumption of simultaneous oc-currence of seismic and hydrodynamic loads.
(5) The majority of the equipment was qualified by the vendors for generic applications, enveloping much higher SSE values for other sites.
(6) I-was involved with applicable equipment margin studies for the V.C. Summer nuclear plant in 1982 with regard to high frequency content earthquakes. Those evaluations concluded that equipment margins in the high frequency region were sufficient. The average margin be-tween seismic response spectra and qualification response spectra was a factor of approximately 2.5.
- 37. To summarize, equipment margins were evaluated by comparing the recorded floor response spectra of the 1986 Ohio earthquake with the original spectra from either testing or analysis. The comparisons demonstrate that, both for equipment directly analyzed and equipment at other locations, the origi-nal design is more than adequate to accommodate events such as the 1986 Ohio earthquake.
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CONCLUSION
- 38. The 1986 Ohio earthquake was a low energy, high fre-quency, short duration, low velocity, and " mall displacement event. As a result of these characteristics and the above dis-cussions, the 1986 earthquake had no adverse effects on the Perry structures, systems, or components, and no changes to the Perry seismic design are required.
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g W ' l (.d >3 V ' CHAb G"CHEN Subscribed and sworn to before me this l * day of
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JANICE M. DC380$EY. I?OT.WI PU8 tic CUWrJ TOWNSHIP NRKS C01',9TY My Commission expires: m m u mr . . . . . . . . . . _
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EXHIBIT A PROFESSIONAL QUALIFICATIONS OF
- CHANG CHEN Sixteen years of extensive experience in the application of structural mechanics theory to the design and analysis of nuclear and fossil power plants in the U.S., Japan, Korea, Yugoslavia, Germany, and Iran. Specialization in seismic resistant design, thermal stress analysis, vibrational analysis and design for impact and impulsive loading, hydrodynamic
- sloshing problem, and experience in the design and analysis of ocean thermal energy conversion systems, plus project management and personnel administration.
EXPERIENCE: GILBERT / COMMONWEALTH since 1969
, 1982 to Manager, Civil / Structural Department and Chief Structural Engineer -
Present Responsible for technical supervision and personael administration in the area of structural drafting, layout and models, architecture, civil engineering and structural engineering for nuclear plants, fossil plants, and continuing services.
4 1979-82 Section Manager, Specialty Structures - Responsible for technical supervision and personnel administration in the areas of continuing services of all operating nuclear power plants, computer applications, applied research, engineering mechanics, and special projects.
Supervised engineering work in the area of NRC I&5 Bulletins 79-02, 79-14, and 80-11, and systematic evaluation program. Seismic resistent design of reference fossil power plants. Project manager of design review of TVA Browns Ferry Units 1,2 and 3, Long Term Torus
, Integrity program. Participated in the investigation of reservoir induced seismicity effects on South Carolina Gas and Electric Company's V.C. Summer Nuclear Station structural and equipment design. Technical presentation before Advisory Committee on
! Reactor Safeguards (ACRS) and testimony before the Atomic Safety Licensing Board (ASLB).
1978-79 Supervising Structural Engineer - Responsible for technical supervision and personnel administration in the areas of structural mechanics and computer applications. Project Manager of the Kraftwerk Union (KWU) project for the seismic design review of the 1,300 MW nuclear
- power plants in Iran, and for providing technical support to the KWU Engineering Department. Provided technical supervision on the Safety Relief Valve Discharge problem of Boiling Water Reactor System.
1974-78 Supervisor of Applied Research in Structural Mechanics - Supervision of tile analytical aspects of PWR, BWR and fossil power plant designs, seismic resistant design of structures and equipment, missile protection design, pipe whip restraint design, compartment pressurization, jet impingement design, finite element stress analysis, --
, and thermal stress analysis of reinforced concrete structures.
Aircraft impact resistant design using soft shell concepts and pipe rupture restraint design for Brown Boveri Reaktor (BBR) in Germany, and hydrodynamic sloshing of water tank due to seismic disturbance.
l Shrinkage and creep of concrete, effect of coarse aggregates on the l
rdert/Cammeneesen (Continued)
CHANG CHEN (Cont'd) ,
crack propagation of concrete structure. Behavior of concrete structure under multiaxial stresses. Platform and cold water pipe analysis of the ocean thermal energy conversion system under random waves and current effects.
1972-74 Senior Research Engineer - Seismic resistant design of PWR and HTGR, preparation of equipment seismic qualification specifications, seismology study, fluid sloshing study, low-tune turbine foundation design, pipe whip restraint design, standard plant design, and PSAR, FSAR write-ups.
1973 Consultant to Atomic Power Department of Taiwan Power Company -
Seismic resistant design of nuclear power plants.
1969-72 Research Engineer - Seismic resistant design of nuclear power plant facilities, computer programming for dynamic analysis, aircraft impact analysis of containment, and stress analysis.
1969 Institute of Building Research, He Pennsylvania State University, University Park, Pennsylvania Engineer - Heat transfer and thermal stress analysis of multistory steel frame structures.
1965-69 Department of Engineering Mechanics, ne Pennsylvania State University, University Park, Pennsylvania Teaching Assistant - Class lecturing in statics, dynamics, and material testing.
1963-65 Department of Civil Engineering, Duke University, Durham, North Carolina Teaching Assistant - Class lecturing in material testing.
EDUCATION: B.S.C.E., Cheng Kung University,1962
_ M.S.C.E., Duke University,1965 Ph.D., he Pennsylvania State University,1969 REGISTRATION: Profess,lonal Engineer - Pennsylvania (1973)
SOCIETIES: Member, IEEE Working Group 2.5 on the " Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations" Member, ASCE Working Group of Dynamic Analysis Committee Member, ASME Working Group-Shells Member, AISC Geert /Conunenuese (Continued) -
CHANG CHEN (Cont'd)
PUBLICATIONS: "Aseismic Design of Asymmetric Structures and the Equipment Contained," First International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, September 1971.
" Dynamic Analysis of Vital Piping Systems Subjected to Seismic Motion," First International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, September 1971.
"Some Considerations in the Aseismic Analyses of Nuclear Power
, Plants," Symposium on Structural Design of Nuclear Power Plant Facilities, University of Pittsburgh in cooperation with ASME, ASCE, April 1972.
" Comments on Floor Response Spectra," Second International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, September 1973.
" Seismic Resistant Analysis of Heavy Equipment," Second International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, September 1973.
Discussions on " Interaction of Soil and Power Plants'in Earthquakes,"
Proceedings of ASCE, Journal of Power Division, November 1973.
" Seismic Resistant Design of Safety Class Structures and Equipment,"
ASCE Specialty Conference on Structural Design of Naclear Power Plant Facilities, Chicago, December 1973. Also published in NUCLEAR ENGINEERING AND DESIGN, Vol. 30, No.1, July 1974.
Discussion on " Modal Damping for Soil-Structural Interaction,"
Proceedings of ASCE, Journal of the Engineering Mechanics Division, December 1974.
4
" Definition of Statistically Independent Time Histories," Proceedings of ASCE, Journal of the Structural Division, February 1975.
" Analytical and ExperimentalInvestigations of the Martins Creek Low Tuned Concrete Turbine Pedestal," Presented at the Pennsylvania Electric Association, Structures and Hydraulics Committee. Winter Meeting in Bethlehem, Pennsylvania, February 1975, i
" Vertical Responses of Nuclear Power Plant Structures Subject to Seismic Ground Motions," Third International Conference on Structural Mechanics in Reactor Technology, London, England, September 1975.
" Correlations of Artificially Generated Three Component Time Histories," Third International Conference on Structural Mechanics in Reactor Technology, London, England, September 1975.
l (Continued) l __L. - _ _ _
^ CHANG CHEN (Cont'd)
PUBLICATIONS: " Simulation of 3ree Component Spectra Compatible Time Histories,"
(ContM) Presented at the 2nd ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, December 1975.
" Effects of Uplift on Soil StructuralInteraction and Toe Pressure Calculation," published in Vol. H of 2nd ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, December 1975.
" Moment-Shear Interaction Effect on the Ultimate Capacity of Wide Flange Beams," Presented at the 2nd ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, December 1975.
" Artificial Earthquake Generation for Nuclear Power Plant Design,"
Sixth World Conference on Earthquake Engineering, New Delhi, India, January 1977.
" Structural Design for Aircraft Impact Loading,4th International Conference on Structural Mechanics in Reactor Technology, San Francisco, August 1977. -
" Experimental Verification of U-Bolt Connection for Pipe Whip Restraint Design," 4th International Conference on Structural Mechanics in Reactor Technology, San Francisco, August 1977.
"The SRSS and the Static Coefficient Method for Seismic Resistant Design of Equipment and Structures," ASME Energy Technology Conference and Exhibit, Houston, September 1977.
" Seismic Resistant Design of Heavy Equipment," Proceedings of the Conference on Structural Analysis, Design and Construction in Nuclear
. Power Plants, Porto Allegre, Brazil, April 1978.
" Reinforced Concrete Structural Design for nermal Effect,"
Proceedings of the Conference on Structural Analysis, Design and Construction in Nuclear Power Plants, Porto Allegre, Brazil, April 1978.
" Soft Shell Hard Core Concept for Aircraft Impact Resistant Design,"
Proceedings of the Conference on Structural Analysis, Design and Construction in Nuclear Power Plants, Porto Allegre, Brazil, April 1978.
"Research Needs and Improvement of Standards for Nuclear Power Plant Design," Nuclear Engineering and Design, Vol. 50, Number I, October 1,1978.
"The Uncoupling Criteria for Subsystem Seismic Analysis," Sth International Conference on Structural Mechanics in Reactor Technology, Berlin, Gerrnany, August 1979.
Geert/Cammenmese (Continued) ,
CHANG CHEN (ContM)
PUBLICATIONS: " Seismic Qualification of Equipment - Research Needs," 5th (ContM) International Conference on Structural Mechanics in Reactor Technology, Berlin, Germanys August 1979.
"'Ihe Steel Containment Design and Analysis for High Seismic Zone Application," 1980 Symposium on Nuclear Power, sponsored by the Chinese-American Engineering & Management Institute, New York, October 1980. Also presented at 6th International Conference on Structural Mechanics in Reactor Technology, Paris, France, August 1981.
m e
M Geert /Commonweep 2/86*
TABLE 1 p.,9e l PERRY NUCLEAR POWER PLANT UNIT NO.1 SEISMIC MONITORING INSTRUMENTATION I" "I Location Type Manulutuser / Model Number Rextor Building
, D58 N101 (l) Kanemette(s t SMA-3 7 10" Azemuth 175' Rextot Buildeng D5I Nill (I) Kanemetrics / 5MA-3 f,"'[,]" .V'$ _O*
Azimuth 174*
Rex or Reciscu' ssonPump (Inside Drywell, Rextor Suelding)
D51Al20 (2) Engdahl / PAR-400 Elevasson 605 -0* ( Approssmately)
Azemuth 145*
D51 Rl30 (2) Engdahl / PAR-400 OUT O F 5 E R VIC E Aumalsary Bucidang D58-Rl40 (2) Engdahl / PAR-400 fouMatson Mat (ltPCS Pump Hoom) tievation %H'-4*
I Is..s. I lessie elasses y At t riesuaja.spis
/ l e .a....I Prab A4 e rece naja apte j leianial Hespoone Spettruen Hetusder w
TABLE 1 Page2 PERRY NUCLEAR POWER PLANT UNIT NO.1
) SEISMIC MONITORING INSTRUMENTATION
," Type Manufmluses iModel Number Location l Rextor Sueldeng fou atm Mas I
i D51-R l60 (3) Engdahl t PSR- 1200-H1V-12A Elevation 574*-10*
Asimuth 225*
Rea< tor Buildeng 630* Platform D51 Al70 (3) Engdahl/PSR 1200-H/V II"'N AMII
{. Elevation 630*-l*
l Asimuth 238' l
l.
Auaeleafy Buildeng D51 RISO (3) Engdahl / PSR-1200-H f V foumiatsu Mat
( HPCS Pump Room)
Elevation 568*-4*
4 Auseliary Sueldeng D51-Rl90 (3i Engdahl t P5R-1200 H s V founciation Mat
' ( RCIC Pump Room )
Elevation 568*-4*
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i TABLE 3 l i
FLOOR R ESPO N S E S PECTR A DE SIGN VERSUS R ECOR D ED lastrument Damoina m m Direction Q1L131 Percentaae 051 R180 Awailaary and Suilding N5 $5E 2 051 R190 Foundation Mat 051 4100 Aumliary and Suilding EW SSE 2 051 4190 Foundation Mat 051 R190 Auxiliary Building vtRT SSE 2
\ Fowndation Mat 051 N101 Reacor and Guilding N5 55E 2 051 4160 Foundation Met 051 N101 Reactor and Suilding EW SSE 2 051 R160 Foundation Mat 051 N101 Reacor and Building VERT SSE 2 051 R160 Foundation Mat
. 4
TABLE 3 -
testrument camniaa m m Direction M Perteatane 051 R:70 lesice Qafweit N5 $5E 2 Reactor tuilding Matform-430' 051 R170 -
Inside Oweil EW $5E 2 Reactor Building Platform-430' 051 R170 inside cryweil vER? SSE 2 Reactor Building Matform-430' 051 N111 Reactor Building N5 55E 2 Containment vesset-446*
051 N111 Reactor Building EW '55E 5 Containment vessel-646' 051 N111 Reactor Building VERT 55E 2
. Containment vessel-446' 0
i TABLE 4
' Comparison of Design Desplacements' V5 Recorded Displacements' 1
' ( Empressed in centimeterstone inch = 2.54 <m)
{
i' COLUMN 1 COLUMN 2 C O L U M N 2 minus C O L U M N 1 1'
1 l Reactor Building Reactor Building
! ' Foundation Mat Containment Vessel Relative Displacements Elevation 574*-10* Elevation 686* for the i
SMA-3 ( Kinemetri<s ) SMA-3 ( kinemetrics ) Containment Vessel D51 N101 DSI-NilI i
i Re<orded 0 09 0 17 0 08 1
l N5 SSE O044 0 28 0 24 06E 0 023 0 17 0 15 j
Recosded 0 16 0 21 0 05 l EW SSE O 044 . 0 28 0 24 .
- t i
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a Re<orded 0 05 0 07 0 02 g
.! VERT. SSE 0 02 8 0 37 0 017 08E O 013 0 022 0 009 Socorded - -
8.1 5855a 55E - -
8.34 i OSE - -
S.21 1
l 1 (Mpt.ndements ti.ised on same time step to detesmane sel.itive da9 14(ements 1 Square soot ol tric Sum of Ilie squases v
TABLE 5 EQUIPMENT LIST AT AUXILIARY BUILDING ELEVATION 568' 152270001 LPCS Isotrument Rack 152270017 RCIC Instrument tack 152270014 RER Instr e nt Back A 152270021 RER Instrument tack B 152270055 RER Instrument tack C 1C61N0001 Differential Press Transmitter 1E12N0007A,3 Differential Press Transmitter 1E12N0015A,B,C Differantial Preee Tranesitter IE12N0026A B Pressure Transmitter 1E12N0028 Pressure Tranesitter 1E12N0050A,3 Pressure Tranesitter 151250051A,3 Pressure Tranesitter
- 1812N0051A,5,C Difforential Press Transmitter 1E1250055A,5,C Pressure Transmitter 1512N0056A,3,C Pressure Transmitter 1E1250058 C Pressure Tranesitter 1121N0003 Pressure Tranesitter 1E21N0050 Pressure Tranesitter 1521N0051 Plow Transmitter 1E21N0051 Pressure Tranesitter 1E21N0053 Pressure Transmitter 1E2150054 Pressure Transmitter 1131N0075A Pressure Tranesitter 1131N0077A Pressure Transmitter 1E31N0083A,3 Pressure Transmitter 1E51N0003 Differential Press Transmitter 1E51N0050 Pressure Transmitter 1E51N0051 Differential Press Transmitter 1151N0053 Pressure Traaesitter 1E51N0055A,B,E,7 Pressure Tranesitter 1E51N0054A, E Pressure Transmitter 1512C002A RER Pump & Motor 1512C0013 REE Pump & Motor 1E12C002C RER Pump & Motor
, 1121C001 LPCS Pump & Motor l 1E22C001 EPCS Pump & Motor
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- 2. #D51 N111 R/8 Containment Vessel, El. 686', Ar.174*
- 3. #D51 R120 Reactor Recirc Pump, El. 605', Az.145' . . .
- 4. #D51 R140 A/B Foundation Mat, El. 568'
- 5. #D51 R160 R/8 Foundation Mat, El. 574' A2. 225'
- 6. #051 R170 R/8 Platform, El. 630' Ar. 238'
- 7. #D51 R180 A/B Foundation Mat, El. 568'
- 8. #051 R190 A/B Foundation Mat, El. 568' FIGURE 4
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