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{{#Wiki_filter:. Pege 1 of 8 CONDITION REPORT EVENT INVESTIGATION cR_-A6 _01 O_6 Event Descrap.taon f
. On Friday, January 31, 1986 at approximately 1148 hours, a step l increase of approximately one and one half inchea (1 1/2") was I observed on both the Suppression Pool wide and' narrow range levelg recordera, 1G43-ROO93A/B and 1G43-RO73B. Each instrument loop la totally independent in that they have separate power supplica, censing and reference lega, and are mounted in different locations. This step change was therefore assumed to.be related to the seismic event which was experienced at approximately the came time. In an attempt to ascertain the cause of the observed otep change, as found date only was taken on the Suppression pool level instruments by the performance of Surveillance Instructions G43-T1305 A thru F. This work won accomplished following prior approval from both plant management and the NRC resident inspector. The as found data indicated that there was a ,
measurable positive 2ero shift of approximately three eights to f three quarters of an inch (3/8" to 3/4") of pool level from the previous calibration date taken approximately two months prior. g m. Lnvestigation of Potyntial Causes The cause of the recorded level increase was investigated at first for possible sources of water being dumped or leaked into the cuppression pool as a result of the seismic event. The step increase of 1 1/2 inch corresponds to.approximately h ! of water. Prior to the event there had been indication of a small I but continuous pool level increase of approximately 800 ge11ona per day due to a known source. Engineering evaluation after the , event produced an extremely low probability that the level increase was due to an actual level increase in the suppression j pool, as no potential source showed a corresponding loan of water ' inventory. r The measured instrument zero offset of 3/8" to 3/4" taken from the Surveillance Instruction data did not correspond directly wAth the recorded increase of 1 1/2 inch. As a result it was decided to investigate the following areas of potential cause: l Potential meismic damage to the Rosemount differential pressure transmittera Installation of the level transmitters - Calibration procedure or technique 8605000700 860321 PDR FOIA ' HIATT86-91 PDR i, _ __. . _ __ _ _ _ _ . . _~ _
J Page 2 of 8 Air entrapped in the variab13 monaing lines due to inedequate filling and venting after the recent suppression psol outage Recorder pen aticking To aaaast in the evaluation, ND&AS Called in factory representatives from Rosamount, the manufacturer of the instrumentation. The overall investigation covere from Januery 31 to February 14,1986. Analypyy Several possible causes of the observed anomaly were considered, however only those diacumsed below dimpley e reasonable dagree of prodottlity or collaborating evidence: P_qf,_eIttle_1 A ijote pgman - At first, it was assuied that there could have possibly been some damage to the Rosemount differential trenamitter resulting from the meismic event. ND&AS has researched the environmental quellfication of the transmitter as compared with the actual seismic event data, and report that the test response spectra of the Rosemount tronamitter enveloped the ceismic event. In a further attempt to confirm potentiel seismic damage, three similar differential pressure instruments, listed f below, were tested by taking ta found data only using the corresponding Surveillence Instruction. All of the selected instruments i.ed been previously calibrated in e time frame similar to that of the suppression pool level trenpaitters. The results of the data taken from these instruments showed no seesurable offset, and the es found data compared favorably with the sa left dote from the previous calibration. Lne_tyuamnt Number Surveillance Numb _sy 1E12-N053A E12-T1193 1E12-N053B E12-T1195A i 1E12-N053C E12-T11958 i Close observation of the suppression pool level transmitter zero ) offset showed that if the value of the zero shift was subtracted 1 from the dote points taken, that all values would then fall within l the leave-aa-ia zone for the loop collbration. This observation l eliminated oli other loop componente except the level transmitter J es a potential source of deviation. It wem then necessary to take j e closer look at the transmitter installation. j Tr_a_np m i tier. , l n a t e l l a t ion - The Rosemount factory representatives 1 I could find no problems with the installation of the transmitter that could potentially cause a zero offset or step change in transmitter output. A review of each instrument's history file I chowed some previous occurrence of zero offset as a result of l ! i 1
Page 3 of a l changing mounting hardware, but not evident on all transmitters. A trend shalysis shows that efter the obearved shift, subsequ4nt rccalibration showed no trend of zero offast, nor any comacn cause ; could be identified for the offset. The factory representatives I did nffer some observations with reap.act to the celibration ' tcchnique used and installation of the senaang lines. ! l Egikbration_Teqhn1R_ue - The suppression pool leveA inst rumenta are l domighed auch that they have a " dry" reference leg. vented to l conteJnment atmosphere, and a wet verleble leg senaing pool level. This confaguration requires the use of a " heed tank" on only the , veraable side. A head tank is used to allow the instrument to be calibrated with water. It is crucia3 that the head tank be it the ocze elevation aoch tioe the calibrr; ion is performed. The cathods to accomplish ttaa are for the moet part left up to the tGehnician with the probable result being that the reproducibility of the calibration is queutionable. Dry calibration techniques connot be used because not all water can be removed from the instrument, and as a result, an offset would be experienced. An cdditional problem occura in that it the instrement bypsea valve ' io opened, water would be admitted to the reference leg. Water in the raference leg would result in a positive offset. During the - toking of as found data, n6 water was observed in the reference 109 Additionaliyr it La relatively impossible to cause water to Antar both reference legs et the some time, unless the containment to flooded to above the 674.0 foot elevetion. When calibrating with one head tank, extreme care aunt be observed when adyusting the elevation of tre tank. The current procedure io to match the elevation of the head tank with the center of the vent screw on the transmitter. If the tsnh is above the center line, a negative offset will occur, and below the center will rosult in a positive offast. It should be noted that there le no counting pechanias currently provided to align and hold the position of the head tank assembly dur$ng calibration. IrmicMIb - 5- wy V'A lV<- \ T* Press
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t I I Page 4 of O l Sensino Line_InaJellation - The sensing lines for the narrow range ) tronemittera are extreacly long ae compared with the wide range. There are neveral high pointa that have installed. There are et leant three bloh pointe in common with both narrow and wide range tranamittera, Dut only one of thpae points is vented. The suppr9aston pool transmitters had been filled and vented after the suppreeston pool outage, epproximately two weeks prior to the event. Our current program does not require a formal procedure for filling and venting every inetrument in the plant. There are I however, generic Itatrument Maintenance Instr'uctions (IMI's) providsng methodology for the filling and venting of various types of 2nstrumentn. ( It was previously observed that the filling and venting of the suppremelon pool lines wee difficult due to the short section of three quarter inch pipw followed by a long acetion of- small instrument line, The method used was to open the instrument vent lines, and allow the water to flow until the voriable line no IQncer showed magna of asa. The other h2gh pointa would be vented an a camilar aanner. This method done appear to be inadequate in that .i t remains uncertain that the variable sensing line may not be fully vented of all entrapped air due to the low maas flow rate l of the water through the instrument linea an6 ven't valves. The geometry of the eenuing lines is such thnt the amount of residual entrapped air would be roughly equal in both loopa in the secticn upstrean of the narrow range tap. 2 Recorder Pen Movement - It han been observed that when a recorder l pen receives 11ttle or no . movement, the pen Say tend to " stick" in place untsi a force, auch as a shake or tap, causes the pen drivera to overcome the alldewire resistance (friction). At (Arat, it was assumed that this may have been a major con tributing coune of the observed step increaac. The momentary sh'Jke, combined with the actual momentary change in pool level could have contributed to the discontinuity of the chart trace. Further investigation shows that no other active chart recorders showed a oimilar pen movement et the time of the event. Addittor.olly, the cxpected step change would be much less than observed at the current gain setting of the recorder amplifier. It may then be purmiaed that the observed change in level was not a result of a "atuck" recorder pen. i Probable Cause NDGAS has performed an analysia of the earthquake and how it chould have affected the Rosemount dit'ferential transmitters. l They have found that the event was within the environmental qualification of the instrument, and that they responded as expected. { This evaluation la substantiated in that there were no ' other instruments other than the suppreaeton pool level that showed an abnormal indication during or after the event. No other recorders or differential pressure instruments were observed to
i Page S of 4 omhibit similar or any other anomalies. The actual level of the suppression pool was surveyed, and the i results show that all of the instruments are currently indicating opproximately thirteen mixteenths Anch (13/16") lower then actual level. The instrumenta do have a 3/8-3/4 anch positive zero offset which make the instruments sensing actually about an inch ! to inch and one half (1 - 1/2") lower than actual level. This difference is moet likely due to some amount of air atA11 being g , ontrapped in the canalng lines. The air entroppent is due to the . Inability to completely fill end vent the existing lines from the l ~ consing tap to the instrument.
- Ao mentioned earlier, there currently exists three points in the ocnsing line, between the sensing top and the tee where the narrow range instrument line tops off, where the high pointe cannot be vented. Two of these points are in the three quarter-inch line, where a high mese flow rate would be required in order to completely remove any entrapped air. Most important to note is that the sensing line inside the suppression pool has en elevated offset Just prior to the elbow wl.ich directs the sensing line down ;
towards the pool floor (see sketch below). The opening is directed towards the floor to minimize the turbulent effects when RHR is running, as both the RHR auction line and the level sensing line share the same penetration. 1 \
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This offset can, and is expected to stili contein some entrappvu ) air. Entrapped ear in this location will have equal effect on ! both the wide and narrow ronge level channels. Both the A and B instrument loops are installed on a similar manner. l l Acauming that the sensing 1ine contained air in the location deceribed, prior to the event all of the suppression pool level l instruments may have been indicating about two and one half inches (2 1/2") lower than actual pool leveJ. At the time of the scrthquake, the pool was upset, causing a " wave" effect. This upset can he clearly seen en the suppreanton pool level recorder chcrts. When the level went momentarily low at the sensing tep, l/ l thore was e momentary vacuum senmed et the and of the pipe tap. . This low pressure caused some of the entrapped air to be pulled out the end of the pipe, resulting in a restoration of about 1 1/2
1 Page 6 of 8 inch of levtl. As mentioned before, it la expected that there still mey be as much as one inch of air still entrapped in the sensing linea. This phenomena occurred at both sensing tapa in nearly equal magnitude, and almost simultaneously. It te not expected t h a t. the zero shift had anytting to do with the vbaetved step chance in level, but exaated prior to the event, and k was due to the improper elevation of the water head tank during calibration. Similar zero shifts were observed prior to the event on the suppreaaion pool level channela and the cor.densate storage tank level instruments which utilize a similar installation of both a wet and a dry sensing line. At this time, the epontaneous venting of some entrapped air in both of the suppression pool level instrument sensing lines is the only auspected cause for the observed pool level change at the tacc of the seismic event. Several corrective aationa are therefore recommended. Pr,o,p o ned Corrective retions Serto.i.ng 1 & n_e_ E1_1_11 rLq., a rtd_ Ve_n_tiriq. - The fac111t.ies currently installed are inadequate to provide complete filling and venting of the sensing linea. Additionally, due to the complex nature of the proceas being measured and the diff2culty in assuring the , proper and repeatable venting and filling, the following actions will te taken: i ID_atallation of__Sensinq Linea - An EDCR has been prepared to provide at least two additional high point vents. One of the vents shall be in the three quarter inch line upstream of the containment isolation valve to assure adequate flow rate through the unventable high point inside the suppression pool. Installation of the high point vanta should be completed prior to enteritig Operational Mode 3 C A 106 -01 To approve EDCR, Instrument & Controla Section. Due date March 1, 1986 [illinq a n_d Venting Procedure - A detailed Instrument , Maintenance Instructicn' will be prepared to provide step by step instruction to assure the adequate and repeatable filling and venting of the sensing lines. The I&C supervisors will aneure that personnel who are ammigned the task of filling and venting the suppression pool level sensing lines w211 have been properly trained in this activity priot t6 starting. CA-86-106-02 To prepare and apporve IM1 for filling and ve.nting sensing linen. Instru4ent & Control Section. Due date March 15, 1986. I
Page 7 of 8 [natrument Calibration - The methodology used for the celibration of differential pressure instruments with one wet and one dry sensiftg line is difficult to implement. portable test equipment la difficult to ad J ust to assure proper repeatable positioning of the head tank assenh3y. The following corrective actions will be taken to assure future calabretions do not introduce zero offset - et t ors: r i 8eo_d. Te_nk. Ausemb11ee - An EDCR has been submitted to
- provide for the anstallation of petmonently head tank :
assemblies on all instruments which heye one wet and one ! dry sensing line. This will include the suppression pool and condenaste storage tank level transmitters. Use of permanently installed head tanka will assure repeatable ! results due to the occurate positioning of the test l' equipment. The head tank ensemblies should be installed prior to the end of the first refueling outage. ; i CA-86-106-03 To approve EDCR for head tonk assemblies, f Instrument & Control Section. Due date March 1, 1986. ! t Instrument Calibration Instructions - The Surveillance l Instruction for the suppression pool level channel } calibration will be revised to incorporate appropriate j changes to the calibration machodology so as to assure l proper positioning and use of portable head tanks until l x permanent tanks are installed. When permanent tanks are installed, the procedure will be revised accordingly as i part of the DCP process. Additionally, a generic ; Instrument Calibration Instruction will be written to ; provide specifically for transmitter calibration when l there are dissimilar fluid sensing lines. CA-86-106-04 To prepare and approve TCN for Surveillance ! Instructions, Technical Section. Due date February 26, ! 1986. . i CA-86-106-05 To prepare and approve ICI for transmitter l calibration, Instrument & Control Section. Due dote April J 1, 1986. 1 Instrument Calibration Frequency. - The Technical Specifications r equire that the suppression pool level ; instruments be calibrated on a frequency of once per le ! montha (ref. 4.3.7.4 Table 4.3.7.4-1 & 4.3.7.5 Table 4.3.7.5-1). As a conservative measure, the frequency for Surveillance Instructione G4Q-T1 F will be accelerated to once per 9p*, days ,3,05 A thruuntil auch time as three consecutive as-found data measurements are within the currently established leave-as-is zone. At that time the Lead Engineer E/I&C may relax the frequency concurrent with technical specifications.
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Page 8 of 8 CA-86-106-06 To revise the frequency of SVI G43-71305 A thru F to once per 9f days, Technical Section. Due date March 1, 1986. g wu si t' ATTACHMENTS I i i I I t I 4 i ! k
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DRAFT ATTACHNENT V PRELIMINARY REPORT ON KINEMETRICS DATA 39
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DRAFT ATTACHMENT VI
/
NRR TRIP REPCRT FROM T. Y.. Cliang' . Leon Reiter' 5, a 1 e T$ 40 l /
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+#. .. .. . og UNITED STATES aft ! n NUCLEAR REGULATORY COMMISSION % E WA$mNGTON, D. C. 20555 \...../ res 2 s ass MEMORANDUM FOR: Robert J. Bosnak, Chief Engineering Issues Branch DSR0, NRR Frank J. Congel, Chief Reliability and Risk Assessment Branch DSRO, NRR FROM: T. Y. Chang, Task Manager Leon Reiter, Sr. Reliability and Risk Analyst DSRO, NRR
SUBJECT:
REPORT OF QUICK RESPONSE TRIP TO OBSERVE EFFECTS OF THE NORTHERN OHIO EARTHQUAKE ON THE PERRY NUCLEAR POWER PLANT
Background
An earthquake occurred at 11:48 a.m. on January 31, 1986 near the Perry Nuclear Power Plant in Northern Ohio. Preliminary infonnation from USGS indicated that the earthquake had a body-wave magnitude (m of 5.0, and the epicenter is located approximately 12 miles south of the PNr)ry site. The Perry Plant is a General Electric BWR-6 with a Mark III containment. It is not yet licensed. Nuclear fuel is on site but has not been loaded into the reactor core. The owner of the plant is Cleveland Electric and Illuminating Company (CEI). The utility was performing pre-operational tests and according to CEI, most of the safety related systems were running when the earthquake occurred. Although Perry is not an operating plant and is not required to do so, CEI immediately activated the emergency plan in order to assess facility damage and personnel injuries. No injuries occurred. In addition, NRC and other relevant federal and local agencies were notified of this event. The plant was switched to the recovery mode in the afternoon of January 31, 1986. The recovery organization is shown in enclosure 1. A detailed walkdown inspection involving about 40, and at times 65, people was conducted for the whole plant, which covered buildings, equipment, instrumentation, piping systems, etc. Only minor damage was found (see preliminary findings below). The inspection was completed at 2:00 a.m. on February 1,1986.
~
An Augmented Inspection Team (AIT) from NRC Region III and staff from NRR were sent to the Perry site on the morning of February 1. The team from NRR consisted of John Stefano (Perry Project Manager), Leon Reiter and T. Y. Chang (both from DSRO). They left Perry on the evening of February 2. The Region III group was led by Carl Paperiello (Director of the Region III Reactor Safety Division). The NRC team held several meetings with the CEI engineering and licensing staff, the Architect / Engineer (Gilbert Associates), CEI consultants (Weston Geophysical) and suppliers of the seismic monitoring instrumentation (Kinemetrics, and Engdahl Enterprises). A list of meeting
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i attendees is attached as enclosure 2. The NRC team members reviewed preliminary seismic recordings and conducted a walk-through inspection of buildings ind equipment. PRELIMINARY FINDINGS
- 1. Seismic Records:
The plant had nine sets of seismic monitoring instrumentation on the site when the earthquake occurred. All but one were operating and recorded the earthquake. They are of three different types: triaxial time-history accelerographs (made by Kinemetrics), triaxial peak response spectra recorders, and triaxial peak accelerographs (both made by Engdahl Enterprises). A list of these instruments and their locations is attached as enclosure 3. There were some difficulties in interpreting data obtained from the Engdahl peak rc;ponse recorder. Each of these instruments has 12 tuned reeds, each tipped with a diamond stylus which makes scratches on a recording plate. Some of these instruments are wired to annunciators in the control room. Each annunciator has one set of 12 amber lights and one set of 12 red lights, with each set of lights marked with the corresponding 12 tuned reed : frequencies. These amber lights and red lights are pre-set to be activated at FSAR values of 2/3 OBE spectral accelerations and full OBE spectral accelerations respectively for the correspondir.g tuned reeds. The recording plates have been in these instruments for about a year, and spurious annunications have occurred previously when there were no earthquakes. Paul Engdahl, President of Engdahl Enterprises, was present at Perry during the visit and helped interpret the data obtained from the scratched plates. Enclosure 4 shows the preliminary data from the Engdahl instruments in the ' reactor building and the auxiliary building as plotted on the FSAR spectra. The 2% damping spectra curves should be used for comparison since the reeds have a built-in 2% damping. There were some concerns about the way the Kinemetrics instruments were installed at the reactor base mat level. The instruments are not located on the mat, rather they are positioned on brackets bolted on the containment wall 9" above the mat. A " ring" test done by Kinemetrics proved that the brackets are rigid enough to render the data valid. Partial preliminary time history data obtained from the Kinemetrics instruments located at the reactor building foundation (575') and on the containment vessel annulus (682') and the associated response spectra generated by Kinemetrics are shown in enclosure 5. Note that the grade level o' the plant is at 620'. Enclosures 4 and 5 indicate that there is a strong possibility that both OBE and SSE design response spectra may have been exceeded in the frequency range of approximately 16 Hz and above. The design basis for SSE is a Regulatory Guide 1.60 design response spectra anchored at 0.159 , with the OBE spectra
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I t anchored at 0.0759 Preliminary data indicated a Zero Period Acceleration (ZPA) value of 0.139 at the reactor base mat. These recorded datc are being further studied,by the NRC and CEI consultants and instrument suppliers. ,
- 2. Results of Preliminary Plant Inspection .
Damage to the Perry Plant was essessed by CEI and found tb te limited to hair line cracks in rad waste and auxiliary building walls and a pipe flange leak in a hot water heater in the rad waste building. In most cases CEI was not able to determine whether or not these cracks were there before the earthquake. The hot' water heater supplies clean noneradioactive steem to an , evaporator, and the leakage rate was estimated to be about 20 drops per minute. In addition, a rotating disc type protective relay for the turbine ' generators tripped, which in turn caused an auxiliary boiler to trtp. Since the turbine generators were not in operation at the time of the earthquake, there was ao voltage across the relay .and, therefore, the cotating disc was hanging loose there. It tripped due to the earthquale motion which would be expected. An instrumEat air compressor tripped; however, one of the three other instrument / service air compressors in the plant automatically started. There was also an indication that suppression pool water level indicators ; might have malfunctioned during and after the earthquake. CEI is still , looking into this matter. Stepage of Ground water through the junction of wall and floor and through cracks in the wall was also observed in the rad waste building after the earthquake. The seepage was not considered sevue. The NPC staff participated in a n ik-down of the plant to inspect for damage ' and anomalies. The NRC staff walk-down observations are recorded in enclosure 6. CONCLUSIONS
- 1. No damage of any significance was observed at the Perry plant by the NRC staff during the trip.
- 2. Based on the preliminary records of seismic monitoring instruments at the plant, the staff concluded that there is a strong possibility that the OBE and SSE design response spectra (FSAR values) may have been exceeded in the frequency range above approximately 16 Hz. It has been pointed out that the,high frequency range 6f Regulatory Guide 1.60 spectra is not as conservative as the lower frequency range,. However, .
the records obtained at Perry are still puzzling since the seismic ' instrumentation on the containment wall at 682 (about 107' above base - mat) still shows exceedance of FSAR floor spectra in the frequency range , higher than approximately 16 Hz. One would normally expect that the responses in the frequency ranges higher than the building fundamental frequency would be filtered out by the building. Further analyses and study of the seismic data by CEI staff engineers and consultants may yield new infomation on this matter. It should be kept in mind that response spectra do not present all the information
6 bout an e6rthquake. The energy centent as a function af frequency is i' better represented by the Power Spectral Density (PSD), which ik an fir,portant indicater ef damage potentisl. .Another damage potential indicator is the daretion of the earthquake. On b9tn accounts, the r effects cf tfie eart'nquake on the Ferry plant were cbserved to be minimal. Norelly responses in the high frequency range do t'ot have much snergy content. Furthermore, the peak tithe history reading et t'he base rr,at lasted for cr.ly about one half'second. , NRt Ag SQUG F0tt0W-UP ACTIONS l
- 1. The locaticru af tershocks and travel mechanism of the e6rthquake sequence w611 have to be examined end co g ared with previous seismicity ;
.so as to determine if there is any fropact upch the adequacy of assarnptions nude in the Safety Ev510atien Report. _
L DSP.0/NRR had a meeting with representatives of tne Division of SWR - Licensing en 2/3/86 to brief them of the findings at Perry. It was ; decided that since the irWestigation of the effect$ of th'is e4rthquake on Per;ry is plant-spetific, the BWR 1.1 censing Divsion should take the ; lead. DSRO will perform on oversight function and provide technical assis?ance if needed. R. Kemann and t.. Reiter were appointed as tean " leaders in the stru?tural/ mechanical and geascience archt respectf vely. ! 4 perrr.anent review team is being formed to review the 6ffects of this earthquake on tne . structural, equilpment and geoscience aspects of the Perry plant. 3- A BWR site inv6stigation tc6m was sent to Perry vt 2/6/86. This team consists 6f A.. l.ee (Engir.2cring 6' ranch /cWR l.icensing Division /NRR) and J. Singh (Il4Et contractor), and will ' perform an independent review of ; safety related equipment. ~ l
- 4. Sciso>ic Qeslificaticn Utility Group (SQU6) will send 6 team to Perry on R/10/UG to collect seismic experience data. This informatien could be t uteful for the on-going Onresolved Safety Issue (USL) A-46, " Seismic 09alification of Equipmunt in Operating risnts."
- 5. As the analysis and review of t'en earthquale, and recorded data proceed, additional recommeridat,tosis te, specific gehgric issues may be forthcoming, a,nong th.ese presently being cor: sider.ed include increased use of parameters other than peak arcelerations and respons'e spectra to ,
t'etter FtfleC!t the d@m&gG potential of Earthquake grQuDd motion and a ' clearer definition of 0BE ar.d/or SSE er.ceedance and their implications.
/L ~ -psy D e D, T. Y. Chstg Leon.Reiter Task Manager Sr. Reliability and Risk ,
Analyst I cc: See Page 5. i
- - - - , ~- g -e g . ey-
I W i cc: 21. Denton S. Brocoum D. Eisenhut L. Heller i R. Bernero N. Chokshi . W. Butler L. Shao, RES ! G. Lainas G. Bagchi , J. Stefano G. Giese Koch S. Stern P. T. Kuo T. Speis R. Rothman B. Sharon A. Lee - N, An.iorson P. Sobel Lic. Div. Directors R. Hermann Lic. Div. A/Ds C. P. Tan G. riotto, RES ' J Richardson, RES C. Faperiello, Reg. III ; I t [ I b
- e i
r F i [ t 4 4
ENC &osaqE / RECOVERY DABANIZATI M MEDWERY : IWHER :
- M. R. EDELMAN :
- ===================i M. D. LYSTER :
A. KAPLAN : l l l l
. ns :::::: ns. . ::::u = ======. .==========3 .== -=. .a=======---==. .===- - ---
- EEt3ENCY : : CPERATICNS : MAINTENANCE : : EN61NEERIN6 : : LICENSIMi : PUBLIC
- PLAN : : : AND : : : : : RELATIONS :
ACTICNS : : : : WCRK ORIERS : : : : : RND :
- : : : : : : : : : : EDIA :
S. F. KENSIGI : : R. A. STRATMAN : : D. J. TMACS : : F. R. STEAD : : E. M. BUZZELLI : : W. E. COLEMN :
-anumes nsui u=====u====== ----===i - i -
- =======------*
M. W. GMYREK M. CthEN L. D. BECK R. L. FARRELL
- 6. R. LEIDICH e
- i I
__, .__ _ __, . . ., ~ ,
l enc cosygg z I MEETING ATTENDANCE February 1, 1966 - TEC I NAME- COMPANY _ POSITION N. J. Lehman CEI Staff Analyst J. J . Waldron CEI Manager, Technica; Dept. , K. A. Connaughton NRC Fesident Inspe: tor , i J. W. Muffett NRC, Region 1.II Plant Sy. stem .Section Section Chief - , J. J. Harrison NRC-, Region III Chief, Er.g BrEnch, DRS i
- C. J. Paperiello (Team lender) NRC, Region 217 Director, Div., Peactor Safet ;
i C. R. Angstadt CEI Leaf Structural Eagireer E. M. Buzzelli CEI Sr. Licensing Engineer K. R. Pech CEI Gi.n. Sup. Ecg. , Nuc1 Cor.str. M. N. Gmyrek CE3 Sr. Operatiens Coordinater
- l R. A. Stratman CEI Gen. Supsevisor, Operations H. L. Williams CEI Executi te Vice Presidect A. Kaplan CEI C.P. Nuclear Operations Civ.
?
M. D. Lyster CEI Manager, PP00 ' F. R. Stead CEI Mana6er, NED , B. A. Stiles CEI Coccmaity 3 elations Asst. 4 John J. Stefano NRC/NPR Per *y Project Manager . Leon Reiter NP.C/KRR Sr. Reliability & Pipe AnalyS:. l T. Y. Chang N E"N3R Task Menager ; W. E. Coleman CEI CSE, Concunity Relations l G. R. Leidich CEI CSE, OutaEe Mana.gener.% D. J. Takace CEI Gen. Sup. , Maintenance J. A. Grobe NEC Sr. Besident Ir.spector M. J. Hayner CEI 1.icensing :Engireer 5
- -..s, .. .. w ~ ' _. . .. _
-2 Attendance - Februa'y r 1, 1986 - TEC KME C0t!PANY POSITION M. H. Minne CEI Licensing Engineer R, G. Schuerger CEI Chairman, NSRC E., N. Levine Weston Geophysical Seismologist, Consul-tant to CEI
- Gabriel Leblanc Weston Geophysical Seismologist, Consul-tant to CEI Steve Kennicki CEI Technical Superintendent Bill Kanda CEI GSE, Technical Dan Hulbert CEI Emergency Planning Sup.
Gecrge Siegel Kinnet.etrics Field Operations Sup. Vince Concel CEI FPTD System Engr. Lead i j Paul Engdahl Engdahl Enterprises President t ! Roger Alley Gilbert, Inc. Project Structural Eng. Franic Kocsis CEI Independent Safety Eng. i Cyril Sh9 ster .CEI Manager, OA Eonald L. Farrell CE Manager, FFSD l Nurray R. Edelnar. CEI V,P., Nuclear Croup i 4 I i 4 4 )
ENCLOJupf .3 g TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION .
= .
. MINIMUM f MEASUREMENT
- INSTRUMENTS
- RANGE '_ OPERABLE.
ENTS AND SEMSOR LOCATIONS _ l . Triaxial Time-History Accelerographs 9,e4 4
@ J 51-M101 Reactor Building e Foundation 0.01 - 1.0g i [ . t-Q '
0 .- 0.01 - 1.0g Ig h g. D51-M111 Containment Vessel 0.005 - 0.02g 1
@ D51-M100 Reactor Building N
- c. .
T d.
Foundation D51-N110 Reactor Building 0.005 - 0.02g
~ N)
'1 b h(
Foundation Triaxial Peak Accelerographs 051-R120 Reactor Recirculation 0.05 - 1.0g_ _ _ , 1 ._ % 4 @ Pump . 1 ak@
. 0.05 - 1.0g '
D51-R130 HPCS Piping in Reactor
@ Butiding 0.05'- 1.0g 1
@ D51-R140 HPCS Pump Base Nat 5b Triaxial Seismic Switches D51-N150 Reactor Building # , 0.025 - 0.25g I I")
a. f 9 " Foundation ay, qm N Triaxial Response-Spectrum Recorders 44 D51-R160 Reactor Building 2 - 25.4 Hz 1(*)%' ~
/
# @ Foundaticn 1 ( 'p ~W,
' @ 051-R170 Reactor Recirculation \ 2 - 25.4 Hz " ,
Piping Support 1
' D51-R180 HPCS Pump Base Mat fo}f 2 - 25.4 Hz 1 D51-R190 RCIC Pump Base Mat i.,2 - 25.4 Hz
; @J Q[ . \ .
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4 U.S. NUCLEAR REGULATORY COMMISSION AUGMENTED INCIDENT RESPONSE TEAM Report No: 50-440/86005(DRS) Docket No: 50-440 License No. CPPR-148 Licensee: Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, Ohio 49101
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Facility Name: Perry Nuclear Power Plant, Unit 1 Inspection At: Perry Site, Perry, Ohio
- Inspection Conducted
- February 1, thru 27, 1986 Team Leader: C. J. Paperiello, Director Division of Reactor Safety Region III Team Members: J. J. Harrison, Chief J. M. Jacobson Engineering Branch Reactor Inspector, I Region III Region III J. W. Muffett, Chief J. A. Gavula Plant Systems Section Reactor Inspector, Region III Region III D. H. Danielson, Chief J. H. Neisler Materials & Processes Section Reactor Inspector, Region III Region III J. T. Stefano J. A. Grobe, Senior BWR Project Directorate #4 Resident Inspector NRR Perry Plant, Region III L. Reiter, Sr. Reliability K. A. Connaughton and Risk Analyist Resident Inspector Reliability and Risk Perry Plant, Region III Assessment Branch, NRR T. Y. Chang, Task Manager D. E. Jones Engineering Issues Branch Reactor Inspector NRR Region III Approved By: A. Bert Davis Deputy Regional Administrator Region III a i2mt 2 > <~ E7~
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TABLE OF CONTENTS
- 1. S u nna ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Pa g e 1
- 2. Initial Perry Nuclear Power Plant Response. . . . . . . . . . . .Page 4
- 3. Plant Conditions Subsequent to Earthquake. . . . . . . . . . . . .Page 6
- a. Visual Inspection by Perry Plant Staff. . . . . . . . . . .Page 6
- b. Initial NRC Inspection...........................Page 7
- c. NRC System Wal kdown - Febura ry 5-7. . . . . . . . . . . . . . .Page 9
- 4. Sei smic Instrument Response to Earthquake. . . . . . . . . . . . .Page 11
- a. Kinemetric Instruments...........................Page 11
- b. Engdahl Instruments..............................Page 12
- c. Conclusion.......................................Page 13
- 5. Confi rmatory Action Letters . . . . . . . . . . . . . . . . . . . . . . . . . . .Page 14
- 6. Pe rs on s Co n ta c ted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Pa ge 16
- 7. Op e n I t em s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Pa g e 17
- 1. Summary The Perry Nuclear Power Plant (PNPP) operated by the Cleveland Electric Illuminating Company (CEI) is a two unit boiling water reactor plant located on Lake Erie in Lake County, Ohio, about 7 miles northeast of ,
Painesville, Ohio. Each reactor is rated at about 3579 megawatts thermal, i Unit 1 is essentially complete. Fuel for this unit is on site in the l upper and lower level storage pools. On January 31, 1986, a number of testing, calibration, and work completion activities were being conducted in preparation for fuel load. Unit 2 is 40-45% complete. Essentially no i construction work has been done on Unit 2 in the past year and the owner has not decided at this time when work will resume. At about 11:47 am (EST) on January 31, 1986, theplantexperiencedan-earthquake which was computed to be about 5.0 on the Richter scale by the U.S. Geological Survey at Golden, Colorado. The PNPP staff responded to the event in accordance with the Perry Emergency Plan in order to expedite notification of the NRC and off-site officials, to ensure personnel safety and to manage response personnel. The counties of Astabula, Geauga and Lake were notified by 12:24 pm and the NRC Operations Center was notified by 12:29 pm (EST). The NRC Resident Inspector was on-site at the time and , went to the Control Room. The NRC's Region III office activated its l Incident Response Center at 1:15 (EST) and established contact with the NRC Operations Center, the Perry resident inspectors, and the Perry Plant operators. By 2:25 (EST) based on preliminary walkdowns by plant personnel which showed little or no damage and in consultation with the NRC, the Perry : Plant terminated the emergency declaration. ' The NRC dispatched an Augmented Investigation Team (AIT) to the plant the f following day. In the interim Cleveland Electric Illuminating Company agreed to conduct a review of the event to determine if the earthquake i was within the Design Basis of the plant, to identify any damage, to ! ensure all equipment had returned to normal operating conditions, and to . maintain all affected equipment in the "as found" condition until the NRC l AIT could retrieve any evidence which would be needed to investigate this i event. This agreement was documented in a Confirmatory Action Letter j (CAL-RIII-86-01; Attachment 1). > The AIT made up of NRC Region III and Headquarters (NRR) representatives i arrived at the Perry Plant at about 11:00 am (EST) on February 1,1986. ! Based upon the initial meeting with Cleveland Electric representatives ! and their consultants, the team decided that the items of initial concern j were the confirmation of CEI's findings based on plant walkdowns during i the night that there was no significant damage to the plant and the ! review and the validation of the data captured by the plant's seismic instruments. The initial review indicated that the emergency response aspects of the event were adequate and a detailed review could be delayed j especially since the plant, having no license, was not yet bound by NRC' emergency response requirements, f The NRC team after an initial quick plant tour split into two groups to ! investigate the above two items of concern. One group reviewed the seismic l data and over the next two days the other group reviewed the data gathered ; by the plant staff on their walkdowns, conducted independent walkdowns in +
i a number of selected areas, and re-inspected off-normal conditions
- identified during the plant staff's walkdowns to detemine the validity of the plant staff's data. The NRC team concluded that.the plant staff had i
been very conservative in the identification of off-normal conditions and that there was no significant damage to the plant. No damage to safety- ! related systems could be identified. In completely non-seismic structures the only clear indication of the earthquake was the shifting in position
- of piping associated with the fire suppression system. Other events I reported by the plant staff indicate that a few objects fell over on desks, dust from ceiling tiles shook loose and a light fixture fell a few inches. New fuel and other radioactive materials on site were in storage
!~ and were not affected. Most of the off-normal conditions identified by
! the plant staff could not be distinguished from nomal wear routinely i associated with ordinary plant construction and operating activities. l Many of these conditions were identified as existing prior to the event. The fine cracks in concrete observed in several areas of the plant had no structural significance. Some were known by the plant staff to have existed prior to the event and some, as observed by the~NRC team, could be shown to have existed prior to the event based upon the cracks being covered in part by uncracked paint. Since fine hairline cracks with no structural significance normally occur when concrete cures, the NRC team found that there were no clear indications that the cracks had been caused by the earthquake. j At the time of the event CEI had a large number of safety and non-safety
- systems in operation as part of its testing program. Of these, two
- non-safety items tripped on protective signals. These included the Unit 1 i
instrument air compressor and the auxiliary steam boiler. Neither unit f was damaged. Relays on the Unit 1 main and auxiliary transformers tripped due to vibration. A leak (20-30 drops per minute) occurred on a flange in i a potable hot water heater in the radwaste treatment system. This system ) is not safety related and does not contain radioactive water. Instruments indicating suppression pool level showed a 1.5 inch increase in suppression pool level. The water level transmitters were found to be i slightly out of calibration but not enough to account for the 1.5 inch increase. No real increase in wate'r level was identified. The total nominal height of the water in the suppression pool is about 17 feet. l The dislodging of trapped air in sensing lines is now believed to have been the cause of this change but this item remain open for further NRC review. Similar instruments elsewhere in the plant did not show similar i behavior. I ! In order to increase the NRC's level of confidence, on February 5-7, four
! reactor engineer inspectors and their supervisor from the Region III
! office walked down all or part of 21 safety-related systems that were
- judged to be the most vulnerable to seismic acceleration due to location j and system dimensions. No damage or significant movement that could be
! attributed to seismic activity was identified. l The plant had three different types of seismic monitoring instrumentation i provided by two different vendors, Engdahl Enterprises and Kinemetrics, Inc. A number of instruments were in the process of being calibrated by one of the vendors at the time of the event. For these instruments, 2 i .
manufactured by Engdahl, some had been calibrated the day prior to the earthquake, some had been calibrated as much as a year earlier and one was out-of-service due to calibration work in progress. The Kinemetrics instruments were calibrated in December,1985. Due to the importance of the data and the difference in the instruments, part of the NRC team followed the evaluation of the seismic instrument data by the plant staff and the instrument vendors. Preliminary data, available by February 2 and later submitted formally to the NRC on February 12 (Attachment 2), indicated the accelerations measured over a narrow range of frequencies (approximately 16-20 Hz and higher) at a few locations exceeded the acceleration expected at these locations for a design basis earthquake (Safe Shutdown Earthquake). Cleveland Electric Illuminating Company and its consultants believe that due to the short duration and low energy of the earthquake these exceedences were not significant from an engineering point of view and are not inconsistent with the seismic design basis of the plant. This issue has been assigned the Office of Nuclear Reactor Regulation for resolution. However, recognizing at an early stage the importance of this matter, the team carefully scrutinized all activities involving the seismic instruments to ensure the accuracy of the data. Subsequently in an amended CAL (Attachment 3) dated February 4, Cleveland Electric was permitted to resume all routine work activities but consnitted to notify the NRC of any activity involving the seismic monitoring instru-mentation. Essential portions of the testing and recalibration work on these instruments was witnessed by NRC inspectors. At this time the AIT believes the instruments were accurate within their design although some of the Engdahl data may be suspect because some instruments appear to have been distrubed by construction activities. However, even without these data the conclusion stated above is not changed. Sufficient validated seismic data exists to analyze the event. In a letter dated March 12, 1986, all procedural restraints ~on equipment and instruments by the NRC were removed. The NRC conluded that essentially all data that could be obtained concerning this event was available. 3
- 2. Initial Perry Nuclear Power Plant Response Title 10, Code of Federal Regulations, Part 50.54(q) requires a licensee authorized to possess and/or operate a nuclear power reactor to follow and maintain in effect an emergency plan. Since the Perry Nuclear Power Plant is not yet licensed and there is no fuel in the reactor, the above requirement does not apply. However, in preparation for licensing, emergency plans, emergency implementing procedures and off-normal instructions were in place. The Shift Supervisor responded to the event by following Off-Normal Instructions ONI-051 " Earthquake (Unit 1)" and the PNPP Emergency Plan as directed by this procedure. The following summarizes the emergency actions in chronological sequence:
Time (EST) Event ; 11:47 Earthquake occurs. 12:01 Plant Emergency Alarm sounds. 12:06 Shift Supervisor declares precautionary Site Area Emergency, makes evacuation announcement. 12:16 Begin notification of CEI emergency personnel. 12:25 Astabula County notified. Geauga County notified. Lake County notified. 12:29 NRC Operations Center notified. 12:32 Perry Operational Support Center activated. 12:35 Technical Support Center activated. - 12:40 State of Ohio notified. 13:02 Site Area Emergency downgraded to alert. 13:03-13:15 Notification of downgrading to alert made to Counties State, Coast Guard and NRC. 13:41 INP0 contacted. 14:20 NRC-Bethesda and Region III concur on termination of emergency. 14:25 Emergency Event Tenninated. 14:31-14:42 Notification of Emergency Event termination made to Counties, State, Coast Guard and NRC. 15:31 Deactivated Technical Support Center. 16:30 Set up recovery organization. The team's initial impression, based on the participation of some members in the activation of the NRC incident response organization during the event, was that the plant staff's response was prudent and overall satis-factory. A decision was made to have the emergency plan implementation examined more closely by a Region III Emergency Preparedness Inspector. This was done on February 5 and 6, and reported in IE Inspection Report Number 50-440/86004; 50-441/86002. This report identifies a number of problems many of which were caused by the peculiar preoperational status of the plant. The overall response was judged to be adequate. With respect to the off-normal procedures, the control room staff responded as required. The Earthquake (OBE) y receivedlimits acceleration initialhad indication that the by been exceeded Operating Basis use of the Triaxial Response Spectrum Recorder Instrument No. D51-R160, that has an indicator in the control room. The instrument itself is located on the reactor building foundation mat in the intennediate building at 4
1 elevation 574'-11". Initial walkdowns were conducted as required. The j recording devices on the seismic measurement instruments were recovered a as required. The appropriate surveillances were conducted on the accelero-l graphs as required. All activities conducted by the plant staff at this l stage appeared appropriate. - 4 1 i 'l 1 P i l : t i s' l. h , i k b i l ! s ( ) i 5 , t
- 3. Plant Conditions Subsequent to Earthquake The information presented in this section is based on a review of the Perry Plant walkdown checklists and results; the summary presented in CEI's " Seismic Event Evaluation Report" transmitted to Mr. Harold Denton in a letter dated February 12,1986; statements made to the NRC at the February 11, 1986, public meeting at the Perry Plant; interviews with the Perry Plant engineering and operations staff, control room staff, and plant management; review of plant staff evaluations of condition reports; and visual observation by the A1T staff during their walkdowns.
- a. Visual Inspection by Perry Plant Staff In the initial several hours after the earthquake visual inspections were performed by plant personnel. Areas inspected included Turbine Building, Auxiliary Building, Intermediate Building, transformer yard, Control Complex, Service Water and Emergency Service Water pump house, suppression pool and the cooling towers and basins. Later in the evening a team of approximately 65 engineers and technicians performed systematic walkdowns of all plant areas. These walkdowns identified approximately 480 potential off-normal conditions none of which involved significant structural damage or major equipment malfunctions.
Cleveland Electric's engineering evaluation concluded that 77% were preexisting conditions, two minor items were attributable to the earthquake and the remainder were indeterminate, i.e., it could not be definitely established that the conditions existed prior to the earthquake. These items included hairline cracks in concrete, burned out light bulbs, dents in sheet metal surfaces, damaged ceiling tile, , putty coming out of cracks, and non-safety piping that had shifted in place. The items concluded to have been earthquake caused were the trip of the main transfonner breaker and a non-safety water heater flange which was found to be dripping water (about 20-30 drops per minute). This is a potable water system and would never contain radioactive steam or water even during operation of the plant. Tests conducted by Cleveland Electric indicated that the trip of the generator protection relays on the main transformer system would not have occurred if they had been energized. During this event they had been deenergized due to testing. Two non-safety pieces of equipment tripped when their protective systems actuated. These were the Unit 1 instrument air compressor which tripped on high vibration. When instrument air pressure dropped to 100 pounds, a back-up air compressor started. The Unit I air compressor was later restarted with no problems identified. The auxiliary steam boiler tripped apparently on an abnormal change in water level caused by splashing or waves induced by the earthquake. It was also restarted with no apparent problems. l Immediately after the earthquake a step increase of approximately one and one half inches was observed on both the Suppression Pool . wide and narrow range level ' recorders. The plant staff investigated l this event and documented their analysis in a Condition Report Event l Investigation CR-86-0106 (Attachment 4). They eliminated a number ' of potential causes: namely, potential seismic damage to the Rosemount differential pressure transmitters, installation of the 6 l 1
level transmitters, calibration procedure or technique and recorder pen sticking. They concluded that the most likely cause was air entrapment in the variable sensing lines due to inadequate filling and venting after the recent suppression pool outage. The plant staff has proposed corrective actions to ensure adequate sensing line filling and venting. This item remains open pending completion of corrective actions and further review by an NRC instrument specialist (440/86005-01). A civil engineering survey of settlement monitor points was conducted on February 5 to determine if the position of buildings had been affected by the earthquake. The maximum change was observed in Reactor Building Number 1 which was a change of 1/16 of an inch in elevation since January 15. However structural settlement changes occur due to a number of causes and the Reactor Building on February 5, 1986, was at the exact same elevation as on February 15, 1985.
- b. Initial AIT Damage Assessment The NRC AIT, in order to assess the possible damage to the Perry Plant caused by the earthquake, reviewed the PhPP's instruction for responding to this off-nomal event " Earthquake"; the checklists used by the licensee for their walkdowns; CEI walkdown results; and performed general walkdowns to determine if any damage had occurred.
The team reviewed CEI's instruction ONI-D51, " Earthquakes", for responding to this off-normal seismic event. This procedure although detemined to be adequate, did not formally contain a generic damage assessment checklist or a controlled record form to record the walkdown results. The PNPP used an informal checklist and radio-l logical survey sheets (maps of plant areas) to evaluate the plant areas and to record the results. This method and control was accep-table to the NRC. The PNPP subsequently developed and issued a temporary instruction GTI-0003, Identifying and Evaluating January 31, 1986, Suspected Earthquake Damage, to control all work activities as a result of the walkdowns and for any ongoing maintenance activities. PNPP plans to incorporate a generic walkdown checklist and a walkdown record form into instruction ONI-051 which will enhance this instruction for future application. The team reviewed the plant staff's walkdown chechlists and all walk-down results; both were found to be adequate. The team reinspected a number of areas walked down by the plant staff. The team verified PNPP's findings to conservative. Almost all of the identified items were attributed to normal plant operating problems and construction damage, i The team upon arrival on site perfomed a general plant tour and did not observe any damage attributable to the earthquake. The team subsequently toured the following plant areas and buildings and per-formed general walkdowns in order to make a preliminary assessment of damage to the plant: i 7 l
Rad Waste Building Auxiliary Building Containment Building and Drywell Turbine Power Complex Control Complex Turbine Power Complex Off-Gas Building Turbine Building Feedwater Heater Bays Water Treatment Building Emergency Service Water Pump House Circulating Water Pump House During these tours /walkc' owns the team inspected for items such as damage, distortion, impact, deformation, looseness and cracking. The team generally inspected piping, HVAC and electrical systems including components, equipment foundations, attachments, walls and floors. The inspection revealed no plant damage conclusively attributable to the earthquake. The team did not find any piping systems in safety related buildings where the earthquake had caused any notable movement. The team did observe that the fire protection systems in the Water Treatment Building and in the Circulating Water Pump House (both non-safety related) had moved as was evidenced by misaligned pipe supports and chipped paint. These systems are very flexibly supported and therefore the noted conditions may have been caused by plant workers not the earthquake. In any case there was no damage to these two systems, s Loose grout was observed by inspectors around a pedestal supporting one end of a feedwater heater. The inspectors believed this was caused by thermal expansion of the heater pressing against grout that had been applied around the edge of the steel base plate. A formal analysis by CEI reached the same conclusion and stated the grout had been applied too high. Other feedwater heaters in the area did not show the same phenomena because the grout was properly applied. These feedwater heaters are not part of a safety system. The grout that had loosened did not provide structural support. The plant conditions observed by the team were typical of what one would expect to find at any power plant, nuclear or fossil, that is in operation or near operation. That is, minor leaks, an occasional broken gauge glass, scrapes, dents and chipped paint would be expected. Most of the identified damage was attributable to construction activities. The cracks in the concrete structures were typical of any massive concrete structure. No unusual or large cracks were identi-fied, all cracks could have existed prior to the earthquake, no cracks were identified as being earthquake related. None of the cracks identified and observed were structurally significant. The initial 1 AIT inspection supported the findings of the inspection conducted by the PNPP staff. { I I i 8
- c. NRC Systems Walkdown - February 5-7 In order to increase the NRC's confidence level in direct NRC observations a detailed inspection of selected safety-related piping and associated mechanical and electrical support systems was conducted. Systems were selected based on safety significance, location in areas of highest acceleration, and dimensions judged to be most vulnerable to vibration.
(1) The scope of the walkdown and visual inspection included: (a) Assessment of the general condition of the systems and components selected for inspection to determine whether there was visible evidence of damage or significant movement as a result of seismic activity. (b) Examination for bent or deformed pipe support structures ' or components. (c) Inspection for loose anchor bolts or cracked concrete associated with anchor bolts and embed plates. (d) Inspection for signs of significant movement such as damaged pipe insulation and scraped or cracked paint at support location. (e) Examination of pipe snubbers and spring cans for changes in initial settings. (f) Examination of the exterior and interior of electrical and control panels for cracks in frames, termination integrity, instrument damage, and glass breakage. (g) Inspection of components for misalignment, foundation cracks, and fluid leakages. (h) Inspection for movement and cracks in battery racks, and batteries, leaking cell jars. (2) The inspection team selected all or portions of the following safety-related systems / components for the walkdown and visual inspection: (a) High Pressure Core Spray system (b) Low Pressure Core Spray system (c) Automatic Depressurization System (d) MSIV Leakoff system (e) Main Steam line piping (f) Residual Heat Removal systems 9
i i (g) Shutdown Cooling system (h) Standby Liquid Control system * (1) Control Rod Drive Hydraulic system (j) Hydrogen Recombiners (k) Division 1, 2, 3 batteries, battery racks and battery ~ chargers (1) 125 vdc distribution panels ' (m) Remote shutdown panels (n) Inverter cabinets
- (o) RCIC suction and discharge piping i
(p) Feedwater piping a j (q) Safety-related HVAC duct, hangers and air handlers (r) ERIS/ PAS panels , (s) 480 vac load centers (t) Drywall in equipment /switchgear rooms (for cracks) , (u) Cable trays, conduits and their support structures ;
- In addition to the safety-related systems the insp'ectors j examined nonsafety-related systems that shared connon areas l with the safety-related systems above .and in the intermediate and turbine buildings.
1 No damage or significant movement that could be attributed to I seismic activity was identified during the walkdown or the detailed visual inspections. i i a t i l ! 10 i
- 4. Seismic Instrument Response to Earthquake Cleveland Electric Illuminating Company in its " Seismic Event Evaluation Report" (hereinaf ter called CEI Sejsmic Report) presents a detailed description of the earthquake event, an analysis of the plant seismic design and its conclusions. The appendices to the report presents a description of the seismic instruments, their location in the plant and the data obtained from these instruments. The responsibility for the NRC evaluation of the seismic consequences of this earthquake in regard to the design basis is being performed by the Office of Nuclear Reactor Regulation and is not part of this report. The AIT concentrated on ensuring that the data from these instruments was preserved and validated to the maximum extent possible. .
The Perry Plant has three different types of seismic monitoring instruments. These include the Kinemetrics Model SMA-3 strong motion triaxial time history accelerograph of which there are two; the Engdahl PSR 1200-H/V response spectrum recorder of which there are four and the Engdahl PAR , 400 peak accelerograph of which there are three. One of the Engdahl PAR ' 400's was out of service at the time of the earthquake because it was being recalibrated. The location and detailed description of these instruments are accurately described in the CEI Seismic Report.
- a. Kinemetric Instruments The Kinemetrics Model SMA-3 strong motion triaxial time-history accelerograph detects and records three mutually perpendicular components of acceleration over the entire duration of the earthquake onto cassette magnetic tape. Power to the unit is supplied by internal rechargeable batteries which are kept in a charged state by 120 V AC line power. Two instruments of this type were used and were located on the Reactor Building Foundation Mat at an elevation of approximately 575 feet and on the Reactor Building Containment Vessel at an elevation of approximately 686 feet.
Both instruments were last calibrated on December 1,1985. Imediately after the event and in accordance with the Earthquake off-normal instruction, a technician conducted a surveillance test of the instruments and replaced the magnetic cassette tapes. A representative from the vendor of this instrument was on-site on February 1. The vendor's representive made paper trace transcrip-tions of the magnetic tapes for the NRC representatives. Copics of these are in NRR and Region III files. A copy was also provided in CEI's seismic report. In response to questions raised concerning the natural frequency of vibration of the metal supports on which the instruments were mounted, the vendor's representative conducted a " ring" test of the support at the 575 foot level. The support was tapped lightly with a screw driver and with an individual's ring. A damped wave was observed with a frequency too high to measure (above 100Hz). This eliminated any question that resonance of the mounting plate could have influenced the readings on this instrument. The amplification was also turned up on the instrument to look for noise sources that might have influenced the readings. None were observed. 11
These tests were witnessed by the hRC. Both Kinemetrics instruments were recalibrated between Fcbruary 4 and 7 by CEI personnel under the supervision of the vendcr's representative. An NRC inspector observed essential steps in the calibration process. The recalibra-tion showed both instruments to be accurate within the vendor's tolerances. A preliminary report on the measurements of the seismic motion during the earthquake was sent directly to the NRC by the vendor (Attachment 4). The data presented in the CEI Seismic Report matches that sent directly to the NRC. The AIT believes the data for the Kinemetrics instruments presented in the CEI Seismic Report to be within the accuracy of the instruments,
- b. Engdahl Instruments The Perry Plant had two types of Engdahl instruments. Four Engdahl PSR 1200 - H/V response spectrum recorders were used. They are mechanical passive devices and require no other power source. Two were located on the Auxiliary Building Foundation Mat at an elevation of approximately 568 feet, one was located on the Reactor Building Foundation Mat at an elevation of approximately 575 feet and one on the Reactor Building Inside Drywell Platform at an elevation of approximately 630 feet. This last instrument was last calibrated on January 30, 1986, the day before the event. The other three were last calibrated during January 1985.
Another type of instrument used was the Engdahl PAR 400 peak . accelerograph. There are three of these instruments installed. This instrument is also a totally passive mechanical device. One instrument was located on the Auxiliary Building Foundation Mat at an elevation of approximately 568 feet and a second on the Reactor Recirculation Pump at the elevation of approximately 605 feet. The first instrument was calibrated on January 30, 1986. The second instrument was last calibrated on December 5, 1985. A third instrument was out of service and was being calibrated at the time of the event. The Engdahl response spectrum recorder (D51-R160-VR) on the Reactor Building Foundation Mat at elevation 575 contains in addition to the
. mechanical recording devices of the normal instr'u ment, limit switches that actuate indicators and alarms in the control room. Amber indicators . indicate accelerations are at or above 0.7 OBE and red indicators indicate OBE. There are three banks of indicators, one for each axis and each bank has 12 pairs of indicator lights each corresponding'to a portion of the vibration frequency spectrum.
During this event indicators at the high frequency end above 16 Hz were illuminated. Red and amber lights corresponding to OBE were illuminated at 20.2 Hz in the North-South and East-West horizontal direction and at 25.4 Hz in the North-South horizontal direction. This was the first indication the plant had that OBE might have been exceeded. In addition an amber light was lit at 16 Hz in the North-South horizontal direction. 12
A Although the Perry Plant staff r ad the recording plates from these instruments, the vendor was also requested to read the plates. Some 1 plates chowed signs that some of the instruments might have been struck while construction activities were in' progress. In addition, some instruments were used by technicians during training for practice in replacing the recording plates. Cleveland Electric was in the process of replacing the plates and recalibrating the instruments when the earthquake occurred. 4 Mr. Paul Engdahl, President of Engdahl Enterprises, was at the site r on February 1-3 to read the recording plates. Attachment 5, a report from the NRC seismic experts on the AIT, includes data obtained by . the AIT during this period for a number of these instruments. A ;
- corrparison of this data with the data presented in Appendix B of the CEI Seismic Report appears to be in agreement. As can be seen the floor acceleration spectra derived from the data, the OBE and SSE appear to have been exceeded in some cases at high frequencies (above 16Hz).
- The recalibration of the Engdahl instruments, which was performed between February 11 and 19, was inspected by the NRC. The as-found calibration of two of the four response spectrum recorders was completely within tolerances. These were instrument number D51-R170 on the Reactor Building 630' platform and D51-R190. on the Auxiliary 4
Building Foundation Mat in the RCIC Pump Room. Instrument number D51-R180 in the Auxiliary Building Foundation Mat in the HPCS Pump Room was found to have vertical reed #5 out of specification with respect to sensitivity. This has little significance since no ~ vertical motion portion of this instrument was out of service. ' Instrument number 051-R160 located on the Reactor Btilding Foundation Mat had 4 of its 36 reeds out of specification with respect to damping. These include Vertical Reed #6, Horizontal Reeds (E-W), #2 and 9 and Horizental Reed (N-S) #9. , In spite of these out
- i. of specification conditions, most' of the dataf from the instrument is acceptable. Furthermore instrument number D51-N101, a Kinemetrics/
, SMA-3, was located close by and provided off setting data for this l instrument. Of the three peak' shock records, D51-R120, located on ! the Reactor Recirculation Pump was found to have sensitivities that were out of specification. Overall data recovery from the Engdahl instruments was good. i ,
- c. Conclusion l Based on their observations the AIT believes that the data presented i
! in the CEI Seismic Report accurately represent the measurements made , by the seismic instruments in place during the' earthquake on 2 January 31, 1986. The Cleveland Electric Illuminating Company took reasonable steps to preserve and validate the. data. Interpretation i of the data will be perfonned by the NRC's Office of Nuclear Reactor
- Regulation.
- i 1
P ! 13
- 5. Confirmatory Action Letters 1
A Confirmatory Action Letter from James G. Keppler, Regional Administrator, Region III, to Murry R. Edelman, Vice President, Nuclear, Cleveland Electric Illuminating Company, was issued January 31, 1986 as noted in Paragraph I. The line items of the Confirmatory Action Letter and status of CEI's implementation are as follows:
- a. Conduct a thorough review to determine if the earthiluake was within 1
the Design Basis of the plant (FSAR). 1 (Closed for purpose of Region III) CEI in a report, dated February 12, } 1986, presented its argument that the earthquake does not call into
- question the design basis of the plant. Review of the report and the presentation of a Safety Evaluation Report on this matter has been
- assigned to NRR. Any further questions and resolutions of this issue 4
will be from NRR. Therefore in respect to the Region III CAL this , matter is closed. I b. Identify any damage as a result of this seismic event; determine if i that level of damage was as expected. , (Closed) Identified d rage was minor. With the exception of an open item in this report we h6ve no further concerns in this area. This open item is considered to be minor issue,
- c. Determine that all equipment, including snubbers, that actuated during this event returned to normal operating conditions / positions; identify any abnormalities. ;
(Closed) Two non-safety items, the Unit 1 instrument air compressor and auxiliary steam boiler, tripped on protective signals as designed. They were restarted successfully. The generator protective relays that tripped the main and auxiliary transformers remained in their
- tripped state as designed, Loads were transferred to the startup transformer. No other changes of state due to the earthquake could be identified,
- d. Identify any actions required to complete licensing of the plant -
related to this event. ' (Closed) Any further action needed to license the plant as a result of this event will be determined by NRR after they review CEI's Seismic Report. For the purposes of Region III this item is closed.
- e. Maintaine all affected equipment in the "as found" condition. .
Therefore, take no action such as removing, repairing or replacing 1 equipment which would destroy or cause to be lost, any evidence ' which would be needed to investigate this evtnt. Routine main- ) tenance may be performed provided that no infonnation related to l
- the event is altered or destroyed. I j 14
(Closed) Based on the initial observations of the AIT and review of CEI's walkdown reports this item was modified by issuance of CAL-RIII-86-01A.
- f. Submit a fonnal report of your findings and conclusions to the NRC Region III Of fice within 30 days.
(Closed) CEI completed and submitted this formal report which was transmitted by letter dated February 12, 1986, to Mr. Harold R. Denton, Director, NRR. Region III received from CEI a copy'of the report. A second Confirmatory Action Letter from James G. Keopler to Murry R. Edelman was issued on February 4, 1986. The line items of the Confirmatory Action Letter and status are as follows: ,
- a. Maintain all seismic instrumentation in the "as found" condition. Take no action such as removing, repairing, calibrating;or replacing instru-mentation which would destroy or cause to be lost, any evidence which w'ould be needed to investigate the event. Maintenance, surveillance and calibration will be performed only with the prior concurrence 'of NRC Region III management. ,
(Closed) All data retrieval including post event tests and calibraticn is complete. There is no further need to restrict: the use of these instruments.
- b. Resume all other activities including work, such as, but not limited to, maintenance, training, surveillance, operations and calibrations under the following conditions:
- 1. All off normal conditions identified during these activities will be documented in accordance with CEI's programs drd procedures.
- 2. Off normal conditions will be evaluated to determine if they ,
were potentially earthquake related.
- 3. Equipment identified in an off normal condition will be maintained "as found" until evaluated. Equipment determined to be potentially earthquake affected will be maintained in..
an "as found" condition until released by the NRC.
- 4. The NR'C will be notified of all off normal potentially ,
earthqu'3ke related conditions within 24 hours. (0 pen) This item will be addressed in the Resident'. Inspector's _ Report (50-440/86006). ,
- c. Develop special procedures to implement Item 2.
(Closed) CEI developed procedure OM19A:GTI-0003 which met this requirement. _ i 15 ' l 1
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- 6. Persons Contacted Major Cleveland Electric Company representatives contacted were as follows:
M. R. Edelman, Vice President, Nuclear Group A. Kaplan, Vice Presdient, Nuclear Operations Division C. M. Shuster, Manager, Nuclear Quality Assurance Department F. R. Stead, Manager, Nuclear Engineering Department R. L. Farrell, Manager, Perry Project Services Department J. J. Waldron, Manager, Perry Plant Technical Department M. D. Lyster, Manager, Perry Plant Operations Department S. F. Kensicki, Technical Superintendent, Perry Plant Technical Department R. A. Stratman, General Supervisor, Operations Section. In addition to the above a large number of CEI and Perry Plant supervisors and employees were contacted and provided information used in this report. A number of CEI contractor and vendor representatives were also contacted as follows: P. E. Engdahl, President, Engdahl Enterprises G. Siegel, Field Operations Supervisor, Kinnemetrics R. Alley, Project Structural Engineer, Gilbert Commonwealth, Inc. C. Chen, Gilbert Commonwealth, Inc. E. N. Levine, Seismologist, Weston Geophysical G. Leblanc, Seismologist, Weston Geophysical J 16 i I
1
- 7. Open Items Open items are matters which have been discussed with the applicant, which will be reviewed further by the NRC, and which may involved some action on the part of the NRC or applicant or both. An open tem disclosed during this inspection is disclosed in Paragraph 3a.
6 1 i i ? s I 17
CONFIRMATORY ACTION LET.T,E_R UNiff3 STATES [ pus 49[*g NUCLEAR REGULATORY COMMISSION
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jgg 3 g gggg CAL-RIII-86-01 Docket No. 50-440 Docket No. 50-441 The Cleveland Electric Illuminating Cortpany ATTN: Mr. Murray R. Edelman Vice President Nuclear Group Post Office Box 5000 - Cleveland, OH 44101 Gentlemen: This letter confirms the telephone conversation between Dr. C. J. Paperiello of this office and yoursulf en January 31, 1986., related to the seismic event that occurred on January 31, 1986, in' the vicinity of the Perry Nuclear Plant. It is our understanding, With regard to this matter, that you wilh
- 1. Conduct a thorough review to detsrmine if the earthquake was within the DesignBasisoftheplant(FSAR).
- 2. Identify any darnage as a result of this seismic event; detemine if that level of demage was as expected,
, 3. Determine that all equipment, including snubbers, that actuated during this event returned to nomal operating conditions / positions; identify any anomalies.
- 4. Identify any acticns required to complete licensing of the plant related to this event.
- 5. Maintain all affected equipment in the "as found" condition. Therefore, take no action such as removings repairing or replacing equipment which w would destroy or cause to be lost, any evidence which would be needed to 1 investigate this event. Routine mairitenance er.ay be perfomed provided that no information related to the event is altered or destroyed.
- 6. Submit a fomal report of your findings and conclusions to the hRC Region III Office within 30 days.
CONFIRMATORY ACTION.LETIE_R_
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C0hFIRMAT6AY ACTI0ft LETTER _ l The Cleveland Electric Illu:r.inating 2 JAN 311966 Coinpany i
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Please let us know icrediately if ycur unde'rstanding differs from that set out , above. _ Sincerely. MSk%' Jarnes G. Keppler; Regional Administrator cc w/ enclosure: , J. J. Waldron, Manager, Perry Plant Technical Department M. D. Lyster, Manager, Perry Plant Operations Department L. O. Beck, General Supervising Engineer, Nuclear Licensing and
- Fuel Managerent Section -
DCS/R5B(RIOS) Licensing fee Management Branch Resident Inspector, RITI Harold W. Kohn, Ohio EPA Terry J. Lodge, Esq. James W. Harris, State of Ohio Robert H. Quillin, Ohio Department of Health F 4 9 h P
DRAFT ATTACHMENT I CONFIRMATORY ACTION LETTER CAL-RIII-86-01 35
4 i n THE CLEVELAND ELECTRIC ILLUMIN ATING COMPANY P O. BO X 5000 - CLEVELANO. OHIO 44101 - TELEPHONE (216) 622 9800 = ILLUMINATING BLOG. - 55 PUBLIC SOUARE Serving The Best Location in the Nation MURRAY R. EDELMAN VICE PRESIDENT NUCLE A R February 12, 1986 PY-CEI/NRR-0437 L Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 - Seismic Event Evaluation Report
Dear Mr. Denton:
My letter to you dated February 5, 1986, committed to provide a report on our response and activities related to the earthquake which occurred in the vicinity of the Perry Nuclear Power Plant. Enclosed is the Cleveland Electric Illuminating Company . , (CEI) report titled " January 31, 1986 Earthquake-Seismic Event l Evaluation" for the Perry Nuclear Power Plant. This document has been prepared by CEI and our consultants following a thor-ough and detailed assessment of the plant response to the January 31, 1986 earthquake. This report demonstrates the appropriateness of the seismic de-sign for the Perry Nuclear Power Plant. Although this recent event provides an additional " data point" for historical seismic event activity, it will not alter any of the design criteria or licensing basis. ; i
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Mr. Harold R. Denton >
- February 12, 1986 Page 2 ,
We believe that this report prcvides the information necessary to support the staff's review and we are available to steet with , your staff as necessary. Should you or your staff have any questions please feel free to call. . Very truly yours, ,
@ W Murray R. Edelman Vice President Nuclear Group MRE:L cc: Jay Silberg, Esquire John Stefano -
J. Grebe D. Eisenhut R. Bernero . W. Butler G. Lainas : - J. Keppler ) C. Morelius C. Paperiello R. Knop 9 8 6
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DRAFT ATTACHMENT III CONFIRMATORY ACTION LETTER CAL-RIII-86-01A I l 1 l 37
DRAFT ATTACHMENT II LETTER To hR. HAROLD R. DENTON February 12, 1986 =
-1 36 l
I
i CONFIRMATORY ACTION LETTER CAL-RIII-86-01A uh Qtp' 4 2.s FEB 0 41985 Docket No. 50-440 Docket No. 50-441 The Cleveland Electric Illuminating Company ATTN: Mr. Murray R. Edelman Vice President . Nuclear Group Post Office Box 5000 Cleveland, OH 44101 Gentlemen: This letter refers to our earlier Confirmatory Action Letter, CAL-RIII-86-01, dated January 31, 1986. Item 5 of that letter states that all affected equipment will be maintained in the "as found" condition but allowed routine maintenance to be performed. As a result of the inspections by our Augmented Investigation Team (AIT) and the subsequent conversations between the AIT leader, Dr. C. J. Paperiello, and Mr. M. D. Lyster, of your staff, we are amending and clarifying Item 5 as follows. It is our understanding, with regard to this matter, that you will:
- 1. Maintain all seismic monitoring instrumentation in the "as found" e
condition. Take no action such as removing, repairing, calibrating or replacing instrumentation which would destroy or cause to be lost, any evidence which would be needed to investigate the event. Maintenance, surveillance, and calibration vill be performed only with the prior concurrence of NRC Region III management.
- 2. Resume all other activities including work such as, but not limited to, maintenance, training, surveillance,' operations and calibrations under the following conditions:
A. All off normal conditions identified during these activities will be documented in accordance with CEI's programs and procedures. B. Off normal conditions will be evaluated to determine if they were potentially' earthquake related. ggn,j/ % CONFIRMATORY ACTION LETTER
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CONFIRMATORY ACTION LETTER The Cleveland Electric Illuminating 2 Company C. Equipment identified in an off normal condition will be maintained "as found" until evaluated. Equipment determined to be potentially earthquake affected will be maintained in en "as found" condition until released by the NRC. The NRC will be notified of all off normal potentially earthquake D. related conditions within 24 hours. -
- 3. Develop special procedures to implement Item 2.
Please let us know immediately if your understanding differs .from that set out above. Sincerely, Original s.it ad by J.rms G. Keppler James G. Keppler Regional Administrator cc: J. J. Waldron, Manager, Perry Plant Technical Department M. D. Lyster, Manager, Perry Plant Operations Department L. O. Beck, General Supervising Fuel Management Section DCS/RSB (RIDS) Licensing Fee Management Branch . - Resident Inspector, RIII Harold W. Kohn, Ohio EPA Terry J. Lodge, Esq. James W. Harris, State of Ohio Robert H. Quillin, Ohio Department of Health
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RIII M RIII,,i RIII RJII RIIp gd- ppg) A % Knop /pd Wa'rnick Pi iello N {s Davis K ler 2/ 4 /86 '@/'16 73 4 FG y (, y/g6-CONFIEMATORY ACTION LETTER
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DRAFT ATTACHMENT IV CONDITION REPORT EVENT INVESTIGTATION CR-86-0106 4 l l 38}}