ML20153G719
| ML20153G719 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 02/24/1986 |
| From: | Stratman R CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | |
| Shared Package | |
| ML20153G706 | List: |
| References | |
| OL, NUDOCS 8602280404 | |
| Download: ML20153G719 (47) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing 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 ROBERT A.
STRATMAN County of Lake
)
ss:
State of Ohio
)
Robert A. Stratman, being duly sworn, deposes and says as follows:
1.
I, Robert A.
Stratman, am General Supervisor, Opera-tions Section, Perry Plant Operations Department, The Cleveland Electric Illuminating Company.
My business address is 10 Cen-ter Road, Perry, Ohio 44081.
2.
As General Supervisor, Operations Section, I am re-sponsible for the supervision of the operation of Perry Nuclear Power Plant, Unit 1.
This responsibility includes supervision of all plant operators and all systems involved in the opera-tion of the plant.
As a part of the recovery organization es-tablished following the January 31, 1986 earthquake, I was re-sponsible for determining the plant status and whether plant structures and components had suffered damage.
A statement of 8602280404 e60225
~
PDR ADOCK 05000440 C
my professional qualifications is attached to this Affidavit as Exhibit "A".
3.
The purpose of this Affidavit is to describe the re-sults of the extensive plant walkdowns and inspections that were performed by Perry Plant personnel in response to the January 31, 1986 earthquake.
I have personal knowledge of the matters set forth in this Affidavit and believe the information set forth herein to be true and correct.
PLANT STATUS 4.
Prior to the earthquake that occurred on January 31, 1986, numerous testing, calibration, and work completion activ-ities were being conducted in preparation for fuel load of Perry Nuclear Power Plant, Unit 1.
One major activity was preparation for the Division II diesel generator response time testing.
As part of this work, all of the safety related com-ponents powered from the Division II diesel were energized and in standby readiness.
Table 1 is a list of these components.
All of this equipment behaved normally through the event; that is, there were no spurious starts or alarms.
Preparations were also underway to move the startup sources.
This work had not yet begun when the seismic event occurred.
The sources were never actually moved, and remained stored in the upper pools.
5.
In support of the ongoing test and surveillance activ-ities, a significant number of systems were in operation. In addition, numerous other systems were energized and in the standby mode.
Lists of the specific safety and non-safety ays- - -
.--. - -- ~
=
tems energized or operating prior to and during the earthquake are included as Tables 2 and 3.
All of the operating safety-related systems continued to operate through the event.
None of the. safety-related systems in the standby mode experi-enced any spurious initiations.
6.
As noted in Table 3, a large number of non-safety sys-l tems were operating or in the standby mode, and maintained their status throughout the event.
Two non-safety items tripped on protective signals as intended by their design, 2
i These were the Unit 1 instrument air compressor, which tripped on high vibration, and the auxiliary steam boiler, which tripped due to actuation of one of its protective circuits.
The instrument air compressor is a centrifugal machine that op-erates at greater than 40,000 rpm and as part of its protective devices has a very sensitive vibration switch.
The auxiliary steam boiler has several protective circuits of which one tripped during the earthquake.
The boiler was successfully 1
restarted after the event.
t 7.
The only other non-safety items of equipment that tripped during the earthquake were the Unit 1 main and auxilia-ry transformers, which tripped due to the closing of the gener-ator protection relays.
These relays although open at the time of the seismic event, were not connected to voltage.
After the event, the transformers remained in a deenergized state consis-tent with the design of the generator's protection logic.
The i
loads automatically transferred to the startup transformer per the design.
Laboratory testing of these relays since the event
! i
__.___o-.___.--__.._,-.___...._______
has confirmed that the presence of voltage on the relays sig-nificantly increases the force required to close these relays.
Had the voltage been supplied to these relays, they would not have closed during the event.
This is substantiated by the fact that other similar open relays with voltage applied did not close during the seismic event.
8.
An 1-1/2 inch increase in suppression pool level at the time of the earthquake, indicated by the water level trans-mitters, has been investigated.
A detailed investigation has concluded that the indicated increase was caused when air trapped in the two sensing lines serving the water level trans-mitters was discharged due to the earthquake.
Air apparently became entrapped in the lines when the suppression pool was refilled in early January.
The procedure for refilling the sensing lines will be revised to ensure proper filling of these
^
lines any time they are drained.
In addition, vents will be l
installed to facilitate proper filling and thus prevent air from being trapped in the future.
9.
Immediately following the event, the plant operators performed initial surveys of the plant.
Areas visually in-spected included the Transformer Yard, lower elevations of the Turbine, Auxiliary, Intermediate and Radwaste Buildings, as well as the Control Complex, Turbine Power Complex, Heater Bay, Water Treatment Building, and all levels of the Reactor Build-ing.
The reports back to the Control Room indicated that the areas were found in satisfactory condition with no major I
damage.
In addition, the Senior Operations Coordinator and I 9,
,.,,m.
-..-.m
-m s_.
made a specific survey of below grade areas.
We found no un-usual or abnormal conditions.
Further steps taken to assess' and evaluate the status of the plant included additional walkdowns by teams of plant maintenance personnel dispatched from the Operations Support Center.
PLANT IMPACT ASSESSMENT 10.
As part of CEI's response to the earthquake, a team of approximately 65 engineers and technicians was organized on the evening of January 31 to perform systematic and thorough walkdowns of all plant arear.
These walkdowns were performed using drawings of each area and checklists of components to in-spect for any abnormal conditions.
These included such items as piping, hangers, snubbers, valves, pumps, instrumentation and other components.
The results of these walkdowns were recorded and compiled into a list of approximately 480 observa-tions, many of which were later determined to be preexisting conditions.
None of the observations involved structural dam-age to the plant or equipment.
The 480 observations are typified by minor hairline cracks in concrete, burned out light bulbs and leaking valve or piping flanges, all of which are normal and expected conditions that would be identified in any i
comprehensive walkdown without the occurrence of a seismic event.
11.
In the inspections that were conducted following the earthquake, plant personnel were instructed to document all un-usual or abnormal conditions.
Those conducting the inspections l
d did not attempt to determine whether the conditions were the result of the earthquake, instead, discrepant conditions re-gardless of potential cause were documented to insure that the status of the plant following the earthquake would be fully documented for subsequent evaluation by Engineering.
Each of the observed discrepant conditions was subsequently evaluated by Engineering to determine whether the condition was caused by the earthquake or whether rework or repair was required.
The engineering evaluation of the items coac?uded that 77% were preexisting conditions, and that only two minor items could be directly attributable to the earthquake.
The remainder, ap-(
proximately 100 items, have been classified as indeterminate, i.e.,
it could not be definitively established that the condi-tion existed prior to the earthquake.
About 25% of the approx-imately 480 items will need rework or repair. (See Exhibit B).
These will be processed in accordance with a special procedure instituted in response to the earthquake.
12.
A number of other inspections were also performed to determine the earthquake's effect, if any, on specific plant structures and conditions.
A site survey was performed to as-sess any impact of the earthquake on the site environs, and in
\\
particular on the shoreline bluff.
No evidence of any earth-quake impacts could be found.
13.
A survey of settlement monitoring points was under-taken to determine if the earthquake had any effect on building settlement.
Monitoring points at various locations around the
' l
perimeter of the plant buildings are surveyed on a monthly basis to monitor building settlement.
Th'e results of the sur-veys were that the recorded movements were consistent with those measured in the past, including the amount of change from prior surveys and the absolute elevationc.
For example, a com-parison of the Reactor Building reading after the earthquake with that of February 1985, shows that the two readings were identical.
Thus, it is concluded that the earthquake had no impact on building settlement.
(See Exhibit B).
14.
A walkdown of Unit 1 Cooling Tower was performed to determine whether any damage had resulted from the earthquake.
The areas inspected included the basin walls, tower columns and footers, internal support columns, baffle system, discharge pipe and veil.
While all inspections were done from ground level, any significant cracks in the veil would have been read-ily apparent since they would have been saturated by the previ-ous day's rain.
No structural damage was found in any area of the cooling tower.
Water was observed seeping through the north and south vertical joints where the basin flume wall and pump house flume wall meet.
Seepage at this joint has been noted in the past and stopped by the application of mastic ma-terial.
(See Exhibit B).
15.
As part of the design program for the plant, seismic clearance criteria were established to assure that a seismic event would not cause any impact on a safety system either by causing swaying or by impact from a non-safety item.
Instances t m
o
~.
of these criteria not being met are termed Seismic Clearance Violations (SCV's).
SCV's are forwarded to Engineering for evaluation to determine whether repair is required.
At the time of the earthquake, there were 29 SCV's that had been dis-positioned for repair, where the repair had not yet been com-pleted.
Following the earthquake, inspectors were directed to reinspect these SCV's to determine whether the seismic event affected the SCV condition.
These inspections found neither damage nor dimensional change. (See Exhibit B).
16.
As previously noted, the plant systems, both safety related and non-safety related, operated properly during and following the seismic event.
Recognizing the sensitivity of electrical components to high frequency response, a detailed engineering study was undertaken to identify the number and types of electrical equipment that were energized during the 1
The components included motors, transformers, re-lays, switchgear breakers, switches, batteries, contacts, valve operators, chargers / inverters, meters, recorders, and transmit-ters.
A wide variety of suppliers was represented.
More than 70 separate systems were involved.
The study showed that over 47,000 electrical components were energized and experienced no adverse effects in terms of spurious system actuation.
(See Exhibit B).
17.
On February 2, 1986, the Division II diesel generator response time test that was in preparation at the time of the earthquake (see Paragraph 4 above) was performed.
The Division ;
}
II diesel operated properly and all equipment powered by that diesel operated as des.igned.
The significant number of elec-trical and mechanical components included in this test, includ-ing pumps, valves, motors, relays, switches, etc., further dem-onstrates that the January 31 earthquake had no adverse impact on safety systems.
18.
Based on all of these evaluations, inspections and tests, it is my conclusion that the plant structures and equip-ment were essentially unaffected by the January 31, 1986 earth-quake.
The large number of safety and non-safety related sys-tems which were operating or energized at the time of the seismic event responded in accordance with their design.
Ex-tensive plant walkdowns and inspections revealed no structural or equipment damage.
I therefore conclude that the plant's seismic design was adequate to handle the January 31 earth-quake.
r sht Ro'bert' A. Stratman Subscribed and sworn to before me this ffR day of February, 1986.
Oaw FAlat N
ry Public My Commission Expires:
SANE E. MOTT
_ Notary Public. Sigt* of ONa
-.... ~..
(Recorded in Lt.ke C.os.ny; 1
Exhibit "A"
Name:
Robert A. Stratman, General Supervisor, Operations Section, Perry Plant Operations Department Formal Education and Training:
Bachelor of Science Degree in Physics, Ohio State University, 1971 Master of Business Administration Degree in Finance, University of New Haven, 1981 Manter of Science in Mechanical Engeineering, Cleveland State University, 1985 Five-Week Perry Nuclear Plant Technology (GE), 1980 Nine-Week Operator Training Course, Perry Simulator (GE),
1980 (SRO Certification)
Five-Week Station Nuclear Engineering (GE), 1982 Experience:
I 1980 - Present:
The Cleveland Electric Illuminating Company Joined CEI as Operations Engineer.
Initially assigned to assist the Operations Section General Supervisor.
In 1982, assigned to develop the Plant Emergency Instructions.
In December, 1982 assumed position as General Supervisor, Nuclear Services Section with responsibility for developing and maintaining a qualified permanent plant security force and for all plant administrative and general maintenance support services.
In October, 1984 assumed position as General Supervising Engineer, Radiation Protection Section with responsibility for directing all activities of the Health Physics and Chemistry Units including the development of the Radiation Protection and Chemistry Programs for the Perry Plant.
Reported to the Technical Superintendent, Perry Plant Technical Department.
In September, 1985, was named General Supervising Engineer, Nuclear Design and Analysis Section.
As such, is responsible for the technical support on various licensing, start-up, and preoperational requirements.
The Section assumes engineering responsibility for all systems turned over to Operations.
Reported to the Manager, Nuclear Engineering Department.
___m___.,
,____,-,a-..
Robert A. Stratman In January, 1986, became General Supervisor, Operations Section, with responsibility for supervising all operations personnel.
1977 - 1980:
Northeast Utilities Engineer at Millstone Nuclear Plant.
Responsibilities included evaluation of plant systems, the design, procurement and implementation of modifications to plant systems and conformance to code and regulatory requirements.
Supervised refueling and unscheduled outages, and managed the test of the plant's reactor containment systems.
Also served as a member of the Plant Operations Review Committee.
1971 - 1976:
U.S. Navy Officer - Qualified as Engineering Officer of the Watch and qualified Engineer of a naval nuclear powered propulsion plant - duties included Electronics Material Officer, Main Propulsion Assistant, Radiation Controls Officer and Submarine Qualifications Officer j
Professional Membership:
Registered Professional Engineer - State of Ohio Page 2
TABLE 1 SAFETY REL)c3D COMPONENTS POWERED BY DIVISION II EMERGENCY DIESEL GENERATOR 1.
lE21-C002B, RHR Pump B 2.
lE12-C002C, RHR Pump C 3.
lE12-F042B, LPCI B Injection Valve f
4.
lE12-F042C, LPCI C Injection Valve 5.
M25-F250B, Control Room HVAC B Exhaust Damper 6.
M25-F255B, Control Room HVAC B Relief Damper 7.
M25-F010B, Control Room HVAC Outboard Supply Damper 8.
M25-F020B, Control Room HVAC Inboard Supply Damper 9.
M25-F263B, Control Room HVAC B Return Line Isolation Damper 10.
1C11-C001B, Control Rod Drive Pump B r
11.
C41-C002B, Standby Liquid Control Transfer Pump B 12.
lE12-F0llB, RHR B Heat Exchanger Dump Valve 4
13.
lE12-F021, RHR C Test Valve to Suppression Pool 14.
lE12-F024B, RHR B Test Valve to Suppression Pool 15.
lE12-F026B, Steam Condensing B Shutoff to RCIC 16.
lE12-F027B, RHR B to Containment Shutoff 17.
lE12-F046B, RER B Heat Exchanger Sypass Valve 18.
lE12-F051B, RHR B Steam Pressure Reducing Valve 19.
lE12-F052B, RHR B Heat Exchanger Steam Shutoff 20.
lE12-F065B, RHR B Heat Exchanger to RCIC Control Valve 21.
lE12-F087B, Steam Bypass Around Pressure Control Valve to RER B Heat Exchanger 1
2:2.
G41-F285, Fuel Pool Cooling and Cleanup Filter /Demineralizer
^
Inboard Inlet Valve 23.
G41-F290, Fuel Pool Cooling and Cleanup Filter /Demineralizer Inboard Outlet Valve
24.
1M15-C001B, Annulus Exhaust Gas Treatment Fan B 25.
1M15-D001B, Annulus Exhaust Gas Treatment-B Electric Heater Coil 26.
M23-C001B, Motor Control Center Switchgear & Battery Rooms HVAC Supply Fan B 27.
M23-C002B, Motor Control Center Switchgear & Battery Rooms HVAC Return Fan B 28.
M23-F010B, Motor Control Center Switchgear & Battery Rooms Supply Damper B 29.
M24-C001B, Battery Room Exhaust Fan B 30.
M24-F0llB, Motor Control Center Switchgear & Battery Rooms Exhaust Damper B 31.
M24-F051B, Control Room Exhaust Damper B 32.
M24-F065B, Motor Control Center Switchgear & Battery Rooms Recirculation Damper B 33.
M25-C001B, Control Room KVAC Supply Fan B 34.
M25-C002B,-Control Room HVAC Return Fan B 35.
M26-C001B, Control Room Emergency Recirculation Fan B 36.
M26-D001B, Control Room Emergency Recirculation Electric Heater B Control 37.
M26-F040B, Control Room Emergency Recirculation Damper B 38.
M28-B001B, Emergency Closed Cooling Cooling Fan B 39.
1M39-B001B, RHR B Pump Room Cooling Fan 40.
1M39-B002, RHR C Pump Room Cooling Fan 41.
1M43-C001B, Division 2 Diesel Generator Room Supply Fan 1B 42.
1M43-C002B, Division 2 Diesel Generator Room Supply Fan 2B 43.
1M43-F070B & F071B, Division 2 Diesel Generator Room Exhaust Louvers 44.
1M43-F080B & F081B, Division 2 Diesel Generator Room Exhaust Louvers 45.
1M51-F090, combustible Gas Drywell Purge Inboard Isolation Valve
46.
P41-C001B, Service Water Pump B 47.
IP42-C001B, Emergency Closed Cooling Pump B 48.
P42-F150B, Emergency Closed Cooling to Control Complex Chiller B Bypass Valve 49.
P42-F290, Nuclear Closed Cooling to Control Complex Chiller Valve 50.
P42-F295B, Nuclear Closed Cooling to Control Room Chiller B Inlet Valve 51.
P42-F300B, Emergency Closed Cooling to Control Complex Chiller B Inlet Valve 52.
P42-F325B, Nuclear Closed Cooling to Control Complex Chiller B Outlet Valve l
53.
P42-F330B, Emergency Closed Cooling to Control Complex Chiller B Outlet Valve 54.
P42-F380A, Nuclear Closed Cooling to Fuel Pool Cooling & Cleanup Heat Exchanger A Inlet Valve 55.
P42-F380B, Nuclear Closed Cooling to Fuel Pool Cooling & Cleanup Heat Exchanger B Inlet Valve 56.
P42-F390A, Nuclear Closed Cooling to Fuel Pool Cooling & Cleanup Heat Exchanger A Outlet Valve 2
57.
P42-F390B, Nuclear Closed Cooling to Fuel Pool Cooling & Cleanup Heat Exchanger B Outlet Valve 58.
P43-C001B, Nuclear Closed Cooling Pump B 59.
IP43-F215, Nuclear Closed Cooling Containment Return Inboard Isolation 60.
1P43-F400, Nuclear Closed Cooling Drywell Return Inboard Drywell Isolation 61.
1P43-C001B, Emergency Service Water Pump B 62.
1P45-F014B, RHR B Heat Exchanger Emergency Service Water Inlet Valve 63.
1P45-F068B, RHR B Heat Exchanger Emergency Service Water l
Outlet Valve 64.
1P45-F130B, Emergency Service Water Pump B Discharge Valve 65.
P47-B001B, Control Complex Chilled Water Chiller B 66.
P47-C001B, Control Complex Chilled Water Chilled Water Pump B 1
67.
P49-C002B, Emergency Service Water Screen Wash Pump (Division 2) 68.
P49-D001B, Emergency Service Water Screen Wash Screen (Division 2) 69.
P49-D003B, Emergency Service Water Screen Wash Strainer (Division 2) 70.
1R25-SO43 & SO47, Division 2 Space Heaters Distribution Panels 71.
1R25-SO45, Division 2 Space Heaters Distribution Panel 72.
EH1214, BUS EH12 Isolating Breaker 1
4 5
1 1
m
.+___ -_- _.,.,
TA3LE 2
,,_,[
I*
SAFETY RELATED SYSTEMS ENERGIZED DURING TF.E SEISMIC EVENT 0F JASUARY 31, 1986 SYSTEM DESCRIPTION -
3 C11 Control Rod Drive C41 Standby Liquid Control
.- C71
. Reactor Protection System.. _
.. c _s:.,,..... p.
a..
..w
..- a D17.~
. ; P1 ant Radiation Monitors,l.::;47;mfic.g' $.c
.... E 12.' L-i R e s i d u a 1 ' H e a t R e m o y a 1 T.. '-W.._9.M..."..#~f."I.T.r. h.t.
.. ~., - :7l
'l. *.
N,. ~ j c -
E21 Low Pressure Core Spray '..T..J. G.r~~c.
..,+, *
.. ~:
E22 H igh P r e s s u're Cor e S r ay _ ;;,'.;..;:. ;,.4... g..,3;g y..
~~
G41 Fuel Pool Cooling and, Cleanup,,, E :,.:. - - F~
~
M15 Annulus Exhaust Gas Treatnien,tyGQ:L'-jJ' y-;_ -
~
M23 MCC, Switchgear, & Mis.c. A r.e. a HV AC r-._..n
.y i.-
B a t t e r y R o o m Ex h au s t
."..2".
>'. -(_. w...- R... J - u..
a M24 M26 Control Room Emergency Recirculation, _. -.
M25 Control Room HVAC M32 ESW Pumphouse Ventilation M40 Fuel Handling Building Ventilation.
=
I M43 Diesel Building Ventilation..... r.,......
C o n d e n s a t e.T r an s f e r an d S to.. ra g e.-e
-:;., 1.
L...
J*
P11
. '. : >...P. -. -.
z P22 Mixed Bed Demineralizer 2.;,'fe.1.bp ::..
i *,.
P41 Service Water
- 9
- * 49), p..i:
?". #.9 $6 ' '. p -
P42 Emergency Closed Cooling P43 Nuclear Closed Cooling'-
P45 Emergency Service Water
-... ' Els -
P7 Control Complex Chill Water
~
Pa9 ESW Screen Wash P52 Instrument Air P54 Fire Protection-C95 Emergency Response Information System P51 Service Air R14 110 VAC Vital Inverters R22 Metalclad Switchgear R23 480 V Lead Centers R24 Motor Control Centers R25 Distribution Panels - 120, 208 & 480 volts Ra2 D. C.
System R43 Standby Diesel Generator (SDC)
Ra5 SDG Fuel Oil R46 SDG Jacket Water Coolant R47 SDG Lube 011 R61 Main Control Room Annunciator 0
TA3bE 3
NON-SAFETY RELATED SYSTEMS -
ENERGIZE: DURING THE SEISP.;C EVENT OF JANUARY 3 1986 SYSTEM dESchIPTION
?~.
F42 Fuel Transfer Equipment s
G33 Reactor' Water Cleanup M11 Containment Vessel. Cooling..~.....
--i.,.. '...
M13 Drywell Cooling
- .y.. i..;.i. k ?.:;.._.....'4.
M21 Con trolled Access HV AC ^-7.yA'sCMyr.9;jg yc'cp_Y q~
.;U..
M27 Computer Room HVAC
.--".'-N..
~ ' ~ '
~
M35 Turbine Building Cooling'&_ Ventilation..
M36 O f f - G a s "B u i l d i n g E. x. h au s t..I '... <M ~~$". ~ ~.."."... . ~.
M41 H e a t e r B ay V e n til a t i on "T --2:]".p.ip '
-D.t ry.,--- "
. b. -
M45 Circulating Water Pump Hous.e,Ve.ntilation
.z*
4 h..*
N21 Condensate N23 Condensate Filtration "P"
~
~~#
N24 Condensate De ineralizers N32 Turbine Control (EHC)
N71 Circulating Water
~~
~
P20 Water Treatment
- P21 Two Bed Demineralizer.-: c.c.d.YcX Q" eg.:..$,b.-p.....,.
P44 Turbine Building Closed Coolini.
P55 Building Heating
' g. f.0.
J
., 's j,.,
P61 Auxiliary Stea:
P62 Auxiliary Boiler Fuel 011.
r-M : ~,
P72 Plant Underdrain C91 Process Cc puter C94 Heal-h Physic Cc=; uter P56 Security R11 Sta-ion Trar.sfor:ers R15 Technical Support Center UPS R36 Heat Tracing & Anti Freeze Protection R44 SDG Starting Air RS1 Intra Plant Communications R52 Maintenance & Calibration
~
R53 Exclusic: Area Paging Syste:
R57 Radio & In-Plant Antenna System R71 Lighting S11 Power Transfer:ers
~
541 Step Up Station
EXHIBIT "B"
APPENDIX RESULTS OF SPECIFIC INSPECTIONS
^
EVALUATION OF WALKDOWN ITEMS PLANT SETTLEMENT READINGS SEISMIC CLEARANCE WALKDOWN COOLING TOWER WALKDOWN REVIEW OF ENERGIZED CIRCUITS l
i I
i
.i
. o....
THE CLEVELAND ELECTRIC ILLUMINATING CO ANY l no longe
- mish t<-
MEMORANDUMj/f
""i"****'"
.t;',,,,,,
, / 4 '"
vaF. R. Stead accM E270 raow K. R. Pech DATE 11-Feb-86 pHoNr 5246 af W220 sumaccT Evaluation of Walkdown Items As a result of the plant walkdowns conducted the evening of 1/31/86. Perry Plant Technical Department prepared a list of the observations of the inspection teams.
These observations were given the title Earthquake Inspection Team Items or EITI's.
The list of EITI's was then forwarded to Engineering for determination of whether the item was a result of the earthquake and whether or not the item needed to be repaired.
The assessment of the need for repair and the documentation of that decision whether on a Non-Conformance Report or a Work Request was done in accordance with POP-1501.
The evaluation of all 473 items was completed this afternoon, and the final summary of determinations is presented in the attached table.
Each item was placed in one of three categories with respect to its relationship to the earthquake.
1.
Caused by the earthquake 2.
' Indeterminate 3.
Not caused by the earthquake As shown in the summary. 375 items were determined not to be caused by the earthquake. 96 to be indeterminate, and 2 to be caused by the earthquake. With respect to the latter two items, one was the trip of the main transformer. noted in the walkdown of electrical bus L10.
The second was a non-safety heater exchanger drain valve that was found dripping water during the walkdown and was reported to be closed and not dripping prior to the
. earthquake.
In addition each item was categorized as to its final disposition using the procedures contained in POP-1501.
Through this process.
330 items were determined to require no repair. 119 to be repaired via a Work Request and 24 items were determined to require dispositioning via a Non-Conforuance Report.
Of the 24 NR's. 20 are anticipated to be use-as-is and the remainder constitute cosmetic repairs to concrete and drywall walls.
.--.--.--,-,-m
Page 2 Evaluation of Walkdown Items By copy of this memo, the Engineering' evaluation of the EITI's is being issued to the Perry Plant Technical Department for preparation of approiate documentation and inclusion in their Condition Report.
KRP:jg e
6 k
l
- - - - ~ -
4 i
)
I 1
I MITI 1.lST EV Al.U ATION SUMM AltY i
13:15 2/11/86 l
lilS Cil'I.IN E TOTAL.
C 1
N NR Wit NA
~
El.ECTitlCAI.
35 1
22 12 0
25 10 I
]
I&C 18 0
10 8
0 15 3
M ECil ANICAl.
83 1
29 53 2
53 28 l
)
I'll'ING 23 0
18 5
2 10 11 i
l STit UCTUlt AI.
314 0
17 297 20 16 278 i
l TOTAI.
473 2
96 375 24 119 330 1
C=
Cauwel by 1.arthquale la lu.l.-ter minat e i
N=
Not Causcal by I: art hquake i
N/A -
No Actiuu llequiscal l
l i
THE CLEVELAND ELECTRIC ILLUMINATING COMPANY b I no longer =ish to""i"***""'
c..
MEMORANDUM cav. e e n AN&
G K. Pech AccM W210 rnQw
.M.R. Krit::er oarc c-e-8 6 PHONE 6460 anow TC6 sus. scc 7 PLANT SETTLEMENT READINGS Per our discussion, plant settlement readings were taken on February 5, 1986 (Attached). No significant difference in the building elevation before and after the seismic event was observed. The maximum change occurred in the Reactor Building #1.
This change was only a minus (-)0.006 of a foot or 1/16 of an inch. The maximum growth was
+0.003 of a foot or 1/32 of an inch, which occurred in the Radwaste Building.
A review of settle =ent readings taken last February 15, 1985, revealed that the Reactor Building #1 was at the same elevation as it is today.
The minute changes in plant elevation can be expected due to structural growth as a result of weather.
cc:
E. Riley J. Eppich C. Angstadt T. Keaveney S. Dodeja 302.MRK
. m. _ m _ _
_....r_.-
f2E.7T &, A%OC,EE6.EWINEE3L% b MW MucL2Ast.pcwet.PL,any a ucreas e um, =.
.d.EYcia OF:
PLANT SETTLEMENTS DATE:
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