ML20151D527
| ML20151D527 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/30/1988 |
| From: | Dupree D, Shawn Smith TENNESSEE VALLEY AUTHORITY |
| To: | NRC |
| References | |
| NUDOCS 8807250151 | |
| Download: ML20151D527 (57) | |
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TENNESSEE VALLEY AUTHORITY OFFICE OF NUCLEAR POWER SEQUOYAH NUCLEAR PLANT MONTHLY OPERATING REPORT TO THE NUCLEAR REGULATORY COMMISSION JUNE 1988
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UNIT 1 DOCKET NUMBER 50-327 LICENSE NUMBER DPR-77 UNIT 2 DOCKET NUMBER 50-328 LICENSE NUMBER DPR-79 Submitted by:
c-1, S. J./ Smith, Plant Manager l
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TABLE OF CONTENTS Page 1.
Operational Summary.
Performance Summary 1
Significant Operational Events 1-3 t
Fuel Performance and Spent Fuel Storage Capabilities 4
i PORVs and Safety Valves Summary 4 ~~
Special Reports 4-5 Licensee Events Radwaste Summary 6-14 15 Offsite Dose Calculation Manual Changes 15 Y
II.
Operating Statistics A.
NRC Reports Unit One Statistics 16-18 Unit Two Statistics 19-21 B.
TVA Reports Nuclear Plant Operating Statistics I
22 Unit Outage and Availability 23-24 Reactor Histogram 25-27 III. Maintenance Summary Electrical Maintenance 28-29 Instrument Maintenance 30-31 Mechanical Haintenance 32-36 Modifications 37-44 b
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SUMMARY
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PERFORMANCE StiMMARY June 1988 The following summary describes the significant operational activities for the month of June.
In support of this summary, a chronological log of significant events is included in this report.
Unit 1 remained in an administrative shutdown the entire month because of design control review, configuration control updating, and resolution of significant employee concerns.
Outage-related maintenance and modifications are being performed.
Preparations are underway for restart and power operations. Unit I has been off line 1043 days.
Unit 2 entered the month of June at 99 percent po'er, producing 1124 MWe.
Br.cause of several reactor trips (3) and various secondary maintenance activities, unit 2 had a capacity and availability factor of 31.16 percent and 43.72 percent, respectively.
The unit is remaining at 70 percent poser to extend the life of the core.
SIGNIFICANT OPERATIONAL EVENTS Unit 2 Date Time Event 06/01/88 0001E Reactor at 99 percent, 1124 MWe.
1455E High steam flow bistables came in.
Reduced power.
1555E Reactor at 99 percent, 1137 MWe.
06/04/88 1155E Reduced turbine load, loop 4 AT indication was slightly above 100 percent.
1500E Reactor at 99 percent, 1130 MWe.
06/06/88 1415E Reactor tripped, low flow to No. 4 S/G.
1519E RCS at 549 degrees F, 2230 Psi, mode 3.
06/07/88 0200E Began pulling S/G banks.
0216E Banks withdrawn.
I 2053E Began withdrawing control rods, mode 2.
2217E Reactor critical.
2243E Reactor at one percent power. -
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e SIGNIFICANT OPERATIONAL EVENTS Unit 2 (Continued)
Date Time Event' 06/08/88 0028E Reactor at 2 percent.
0410E Entered mode 1, 5 percent reactor power.
0530E Holding power at 11 percent for high bearing vibration on the turbine.
1216E Maintenance continues on the turbine.
1251E On-line, reactor at 11 percent.
1315E Turbine trip, loop No. 4 high S/G level.
1318E Reactor trip, low level No. 2 S/G.
1957E Adjusting RCS boron concentration for criticality.
2231E Shutdown banks withdrawn.
t 2232E Entered mode 2, withdrawing control rods.
2330E Reactor celtical.
06/09/88 0057E One percent reactor power.
0230E Entered mode 1, 3 percent reactor pow 9r.
0330E Rolled main turbine, 8 percent reactor power.
0431E On-line, reactor at 13.7 percent.
05005 Reactor at 20 percent, 142 MWe.
0512E Reactor trip, lo-lo level No. 2 S/G.
Low pressure heaters isolated, causing a loss of flow.
0827E Mode 3. RCS at 547 degrees F.
06/19/88 1212E Diluting for criticality.
12482 Dilution completed.
1336E Shutdown banks withdrawn.
SIGNIFICANT OPERATIONAL EVENTS Unit 2 (Continued)
Date Time Event 06/19/88 13492 Entered mode 2.
(cont.)
1447E Reactor critical, 0 percent.
1925E Reactor at 2 percent, holding for maintenance on the main feed pump turbines.
06/23/88 104SE Rolling main turbine.
113SE Turbine tripped due to breaker relay operation.
Reactor at 20 percent.
1200E Reset turbine.
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1234E On-line.
1249E Power increase.
1900E Reactor at 30 percent, 310 MWe, 06/24/88 0300E Began load increase to 33 percent for_7'B" MFPT adjustment, reactor at 30 percent.
04008 Reactor at 33 percent, 312 MWe.
1000E Began load increase.
1400E Reactor at 40 percent, 410 MWe, holding for maintenance on "B" MFPT.
06/25/88 0200E Turbine load reduced to maintain T while valving avg in MSR "A-1."
Tavg began to drop and the loss of MWe occurred.
Reactor at 40 percent.
122SE Began load increase.
1342E Reactor at 50 percent, hold for maintenance on "B" MFPT, 520 MWe, 06/26/88 0427E Began load increase.
0845E Reactor at 70 percent, 770 MWo.
06/30/88 2400E Reactor at 70 percent. 770 MWo.
liolding power at 70 percent to extend the life of the core. -
FUEL PERFORMANCE Unit 1 The core average fuel exposure accumulated during June was O MWD /hTU with the total accumulated core average fuel exposure of 0 MWD /MTU.
Unit 2 The core average fuel exposure accumulated during June was 384.82 MWD /MTU with the total accumulated core average fuel exposure of 8848.59 MWD /MTU.
SPENT FUEL PIT STORAGE CAPABILITIES The total storage capability in the SFP is 1,386.
However, there are five cell locations which are incapable of storing spent fuel.
Four locations (A10, All, A24, and A25) are unavailable due to a suction strainer conflict, and one location (A16) is unavailable due to an instrumentation conf fict.
Presently, there is a total of 348 spent fuel bundles stored in the SFP.
Thus, the remaining storage capacity is 1,033.
PORVs AND SAFETY VALVES
SUMMARY
FCV-68-340 opened momentarily when the opposite PORV block valve was opened and when its block valve was opened.
No safety valves were challenged in June.
SPECIAL REPORTS The following special reports were submitted to NRC in June 1988.
1-88-12 On June 8, 1988, with unit 1 in mode 5, and unit 2 at 100 percent power, Chemistry personnel determined that the projected dose due to liquid effluent releases to unrestricted areas would exceed the 3.12 millirem (mR) limit specified by LCO 3.11.1.3 during the month o t' June. During the performance of SI-422.1, "Monthly 10 CFR 50 Appendix I Dose Calculations," the projected dose from liquid effluent releases was calculated to be 0.127 mR.
TS LCO 3.11.1.3 requires the liquid radwaste treatment system to be used t o reduce the radioactive materials in liquid effluents before dischs tge into the Tennessee River whenever the projected dose for a month (averaged over 31 days) exceeds 0.12 mR to the whole body or 0.40 mR to any organ. Hcwever, because the liquid radwaste system does not have sufficient capacity to treat all cf the low level radioactive water generated by the plant, it is not possible to comply with the requirements of LCO 3.11.1.3 when low water flow in the Tennessee River results in noncompliance with the projected dose limits.
In addition, because a significant increase in the Tennessee River water flow is not anticipated for several months, SQN may continue to have difficulty in complying with the requirements of LCO 3.11.1.3.
The root cause of exceeding t.he projected dose requirement has been attributed to the drought conditions in the southeastern United States that has resulted in low water flow in the Tennesseo River.
A contributing cause of 4
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SPECIAL REPORTS 1-88-12 this event wss a minor equipment problem with the condensato (cont.) demineralizer waste evaporator (CDWE) system that allowed two tanks containing liquid radioactive effluents to be released to the river with higher than norn.a1 activity levels. However..even with the higher activity releases, normal river flow would have provided sufficient dilution to maintain the projected dose well below ecquirements. This equipment problem has since been corrected.
2-88-11 On May 17, 1988, with unit 1 in mode 5 and unit 2 at 45 percent power, the 0300 and 0400 hourly firewatches required by TS LCO 3.7.12 were not performed because of airborne radioactivity that developed unexpectedly in the auxiliary building. A leak on elevation 669 (holdup tank valve gallery) occurred when Operations personnel attempted to let down the volume control tank to holdup tank B in order to make allowances for a unit 2 reactor coofant dilution. The operator was unable to readily open the valve, which resulted in system pressurization and leakage through the valve diaphragm in the valve gallery.
The shift operations supervisor initiated an auxiliary building evacuation at 0250E because of the leaking valves in the holdup tank valve gallery. At 0536E, access into the auxiliary building was permitted and firewatch personnel resumed firewatch activities immediately.
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LICENSEE EVENT REPORT (S)
The following licensee event reports (LERs) were transmitted to the Nuclear Regulatory Commission in June 1988.
Description of Event LER 1-88020 On May 12, 1988, it was determined that_ the cold overpressure protection system (COPS) could actuate during a postulated main steam line break (MSLBf accident.
During an MSLB, the RCS temperature in tne affected loop could decrease belov 350 degrees F, thereby autom.acically arming one crain of the COPS.
If a coincident single failuco of a second wide-range temperature channel occurred in another RCS. loop, the COPS could actuate and cause one or both power-operated relief valves (PORVs) on the pressurizer to open. Opening of the pressurizer PORVs would exacerbate the decreasing RCS pressure transient associated with the MSLB and increase the potential for departure fr.om nucleate boiling (DNB) to' occur in'the core.
This event has not been analyzed as part of the SQN design basis.
The event was caused by an inadequate design of COPS.
The automatic arming feature of this
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system was designed to ensure COPS would be functional any time the RCS temperature decreased below 350 degrees F.
However, this design did not consider the potential consequences of an inadvertent actuation as a result of a credible design basis event.
As immediate corrective action, the PORV block valves were closed on unit 2, thereby precluding an inadvertent RCS depressurization if the COPS actuated.
As an interim measure before unit 2 entered mode 1, a temporary alteration was implemented to disable the system as soon as the plant entered mode 3 and administrative controls were instituted to ensure it would be placed back in service when the RCS temperature was decreased below 350 degrees F.
As long-term corrective action, TVA is preparing a design change to install a selector switch that will delete the automatic arming.
The existing administrative controls will then be revised to require Operations personnel to manually arm the system when RCS temperature is below 350 degrees F.
1-88021 On May 23, 1988, at 1215E, while unit I was in mode 5 and the RCS was partially drained to support maintenance, a loss of the operating train of the RHR system occurred.
The "B" train of RHR was in operation when it was decided to place the "A" train RHR heat exchanger in service to enhance plant temperature control.
To place the "A" train in service, an AUO was dispatched to open two valves; however, he misunderstood his instruction and wrote down an incorrect valve number.
The incorrect valve was a manual valve (1-HCV-74-34) used to align the discharge of the RHR pumps to the RWST.
Upon opening valvo 1-HCV-74-34, the AUO heard unusual flow noise and subsequently telephoned the RO for further instructions.
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Drscription of Ev*nt LER 1-88021 (cont.) The ASOS in the CR received an RHR mlniflow alarm and he noticed RHR pump amperage oscillating, unstable flow indication, and that the indicated RCS water level was off-scale low. The ASOS subsequently stopped the "B" train RHR pump and entered the applicable action statements of TSs for a loss of RHR.
The RCS was then refilled above the top of the RCS loops by gravity. feed from the RWST via the RHR system.
Both trains of RHR were then placed in service.
The root cause of this event is attributed to poor communications between the AUO and the RO.
To prevent recurrence of this event, plant administrative instructions have been revised to require the use of "read back cards,".(cards to-write down simple instructions 6.o be read back to originator) to require a CR notification immediately before changing equipment status, and to require operators to return equipment to their "as-found" condition if unexpected occurrences are noted before contacting the MGR for instructions. Additionally, the procedures will be revised to require that a hold order tag be attached to 1-HCV-1-74-34 when the RCS is partially drained.
1-8802: On May 24, 1988, with unit 1 in mode 5 with reactor trip breakers open, all control rods fully inserted, rod control system incapable of rod withdrawal, and both source range channel (N-3_1 and N-32) detectors operable, a reactor trip signal was generated from a source range (SR) nuclear instrument channel spike. The reactor trip first out annunciator alarm for the source range channel (N-32) detector and a high flux at shutdown alarm in the MCR were received.
The NR-45 trace recorder indicated channel N-32 spiked from 1 counts per second (eps) to 6ES cps.
The RO acknowledged the alarm and informed the ASOS of this occurrence.
He. suspected welding activities as the cause of the spike since he was familiar with the source range high flux trip signal that had occurred on two previous occasions.
During investigation of the event, the welder involved with the welding activity stated that he was setting up a welding machine at approximately 0800E to weld an attachment on a pipe for a weld joint fitup.
The root cause of this event is attributed to the noise susceptibility of the existing SR channel design.
As a corrective action, hardware changes to the source range instruments will be made whe.1 the present Westinghouse nuclear instrumentation system is upgraded to meet Regulatory Guide 1.97 requirements.
As corrective action to prevent recurrence of this event, plant management ha-determined that both source range detector input signals into the reactor protection system can be bypassed when reactor trip breakers are removed from service as allowed by TS LCO 3.3.1.1 for modes 3, 4, and 5..
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Description of Ev'nt LER j
1-88023 On June 7, 1988, with unit I in mode 5, two CVIs occurred; the first occurred at 1345E and the second at 1409E. A subsequent investigation found that both resulted from work being performed on the unit 1 containment purge system radiation monitor. The first CVI was reset at 1347E, and an AUO was dispatched to investigate the source.
Before it could be determined that the work on the RM had resulted in the CVI, the second CVI occurred.
It was reset immediately, and work on the RM was terminated until further investigation could determine the associated cause and adequate measures taken to prevent recurrence of the events. At the time of the CVIs, the containment purge system was not in service.
The root cause of the CVIs has been determined to be the lack of adequate preparation to ensure that an ESFA (CVI) did not occur.
These events could have been prevented by blocking the output of the RM's analyzer module with a handswitch in the MCR before' performance of the work.
For immediate corrective action, Operations personnel reset the CVIs and terminated the work being performed to the RH o'nce it was concluded that the work initiated the CVIs. For long-term corrective action, a revision will be made to Standard practice SQM-2, "Maintenance Management System,"
Attachment C. "Preparation / Initial Planning Cuidelines for WRs " to include in the planner's checklist an evaluation to determine if worktobeperformedhasthepotentialtoinitiateag[ESFAor equipment actuation.
SQN will revise the instructions in SOI-90.1B to state RMs that een initiate an ESF actuation shall be blocked before any work / test / return-to-service is performed.
1-88SO4 On June 8, 1988, at approximately 1255E with unit 1 in mode 5 and unit 2 in mode 1 (12 percent power), an NSS officer was discovered in a less than fully alert state.
The officer was posted at the location as a compensatory measure for a degraded protection area intrusion barrier detection system.
The officer was working 8-hour shifts, 0600E to 1400E. At approximatley 1255E, a shif t lieutenant arrived at the alarm section with a security maintenance worker to evaluate the maintenance being performed on the degraded detection system. Upon arriving, the supervisor found the officer in a less than fully alert state.
This event was caused by the officer allowing himself to get in a condition that made it easy to be less than fully alert. The officer was sitting in a security vehicle with his head down. The officer was immediately relieved of the post and a new officer was placed at the location.
1-86047 This revision provides details concerning the start of the emergency 3
Rev. 1 D/Gs as a result of the opening of a 6.9 kV circuit breaker and 1
TVA's action taken in an effort to establish the cause of the spurious trip of the circuit breaker.
1-87050 This revision provides recurrence control for the event and a Rev. 2 current status of the corrective actions.
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Description of Event LER 1-87072 This revision provides additional information eclating to the Rev. 2 corrective action TVA has taken to ensure that ESF coolers have adequate cooling capacity.
1-87077 This revision describes the differences in the corrective action Rev. 1 for units 1 and 2.
1-86019 This revision provides information on corrective action to prevent Rev. 1 recurrence.
2-88022 On May 19, 1988, with unit 2 at 71 percent power, it was determined that a 10 CFR 50.49 unqualified preinsulated butt splice existed on an S/G narrow-range level transmitter. This transmitter is required in modes 1 through 3 for ESPAS and must be EQ in accordance with 10 CFR 50.49.
The unqualified butt splice could have caused the transmitter to be unavailable for mitigatio6 of a postulated feedwater line break (including AFW and S/G blowdown line breaks) while in modes 1 through 3.
However, other reactor protection systems were operational (i.e.,
(1) high pressurizer pressure, (2) overtemperature delta-T, and (3) SI) for mitigation of the feedwater line break accident.
A review of past work documents indicated that the bu.tt splice had been present since an early period in the plant life.when this type of splice was acceptable.
The splice had gone undetected through the recent effort of identifying all 10 CFR 50.49 splices, because it was located in the upper neck portion of the condulet out of normal view.
Corrective actions were to declare the channel inoperable and to replace the unqualified splice. Three additional S/G level transmitters of a similar configuration were inspected as a random sample and did not identify any further unqualified splices.
Since this condition is considered an isolated case, no further corrective actions are planned.
2-88023 On May 19, 1988, with unit 2 at 71.7 percent reactor power, a reactor trip occurred at 1413E.
At 1350E, an SRO and an IM started the process of making adjustments to the No. 3 heater drain tank (HDT) level controllers.
They proceeded to troublehsoot the problem in an attempt to reduce the level in the subject tank.
After several manipulations, the SRO noted the HDT pumps began to cavitate, and a subsequent trip of the pumps occurred.
At 1405E, the baltnce of plant (BOP) operator noted fluctuations in the No. 3 HDT discharge flow.
At 1408E, both No. 3 HDT pumps tripped.
The BOP inmediately started a reduction in turbine load.
At this time it was noted that S/G No. 1 level was dropping. The Operator took manual control of the feedwater regulator valvo and went to full open to regain icvel.
Level dropped to 21 percent before it started to ascend.
The "A" MFP backed off in speed, as it was in the automatic control.
However, "B" MEP continued in manual control, causing feedwater flows to be high.
Level continued to 9
a Dnscription of Event LER t
l 2-88023 increase to 60 percent, at which point the regulator valves (cont.) automatically closed as designed.
This resulted in a steam /feedwater flow mismatch.
The S/G loop 3 low level bistable was already tripped as a result of 2-LT-3-97 being out of service.
Therefoco, a reactor trip signal was generated'due to a steam /feedwater flow mismatch coincident with low S/G level in loop 3.
The mismatch was caused by a S/G level controller manipulation and subsequent loss of the No. 3 HDT pump. The low l
S/G level was caused by bistable 2-LS-3-97 being in the tripped condition because of environmental qualification concerns. The i
trip was reviewed with Operations personnel to ensure familiarization with the event and to detail the lessons that could be learned from the transient.
l 2-88024 On May 23, 1988, at 0028E, with unit 2 at 70 percent power, a reactor trip occurred from a low flow signal on RCS loop 4-(2 out of 3 channels tripped on any one RCS loop above 35 percent power).
At the time of the trip, SI-246, "Recalibration procedure for Reactor Coolant Flow' Channels," was in progress to recalibrate the loop 4, channel II transmitter.
The "Sequence of Events Record" showed the trip being initiated from RCS loop 4, channel III bistable.
An' investigation revealed that the transmitter being calibrated was attached to a common sense line with loop 4, channel III transmitter.
In performing SI-246, the IM did not comply with procedure when valving-out the channel II transmitter.
The SI instructs the performer to relieve RCS system pressure by cracking open the transmitter high side test tee.
Contrary to this instruction, pressure was relieved through the drain valve that routes to a closed drain system, which made it impossible to determine the amount of fill fluid (RCS water) lost when the drain valve was open.
It was theorized that when-the final step was performed to open the high side isolation valve, the void in the drain line caused a pressure drop in the common sense line and a subsequent drop in the output.
This drop caused the low flow bistable to actuate and since the bistable was already tripped, this completed the necessary 2 out of 3 logic for the reactor trip. A reenactment of the procedure steps confirmed the theory as discussed abo s and proved the event was caused by procedure deviation.
1, provide assurance that an adverse trend does not 1
exist on common interactions of equipment, a review of past reactor trips at SQN will be performed to determine a common interaction between equipment that has caused a reactor trip.
2-88025 At approximately 2330E on June 3, 1988, with unit 2 at 100 percent power, it was determined that TS 3.0.3 should have been entered at approximately 2028E. TS 3.0.3 was applicable because action statement (a) of TS LCO 3.8.1.1 was not satisfied when D/G 1A-A was taken out of service for testing.
This action statement requires the remaining D/Gs to be demonstrated operable within one hour any time one D/G has been taken out of service (or otherwise declared i
inoperable) by starting the three remaining D/Gs in accordance with SR 4.8.1.1.2.a.4.
Since the SR was not satisfied, all four D/Gs,
D$scription of Evint LER 2-88025 were technically inoperabic from 2028E until D/G 1A-A was returned (cont.)
to service at 2240E.
At this time, LCO 3.8.1.1 was exited and TS 3.0.3 was no longer applicable. The immediate cause of this event was attributed to the delays that were experienced during the performance of SI-307.1, "Degraded Voltage Relay Response Time Testing And Timer Verification." The root cause of this event was the failure of Operations personnel to adequately consider and implement the action requirements associated with LCO 3.8.1.1.
No immediate corrective action was taken following the discovery of this event, because D/G 1A-A had already been returned to operabic status and TS 3.0.3 was no longer applicabic.
To prevent recurrence of this event, TVA will issue a training letter to all SQN licensed personnel describing this event and emphasizing
- 1) the importance of TS compliance at all times, particularly during plant evolutions such as SI performances when the pat'ential for TS noncompliance la heightened; and 2) the importance of operations shift crews utilizing all available resources in the decision-making process, including other operators, the STA, management personnel, and other members of the plant staff.
2-88027 on June 6, 1988, at approximately 1415E, with unit 2 at 98 percent power, a reactor trip occurred from a loop 4 steam /feedwater flow mismatch coincident with low S/G level.
Instrument gaintenance personnel were performing SI-618, "F.ngineered Safety Features Actuation System Block Tests," to verify output continuity of the SSpS slave relays for feedwater isolation, turbine trip, and MFW pump trip (train B).
When test switch S801 in the safeguards test cabinet was depressed and released, loop 4 MFW flow control valve closed and feedwater was lost.
As a result, S/G No. A level dropped, and the reactor tripped.
The blocking circuitry of the above mentioned EFS functions incorporates test switch S801 and is designed to energize slave relays K601 and K621 while simultaneously providing a current path through diodes to block the final equipment actuation.
It was discovered, subsequent to this trip, that the diode was missing in the blocking circuit for loop 4 feedwater flow control solenoid valve.
When the slave relays were energized via the S801 switch, no current path was provided to maintain the solenoid valve in an energized state and, consequently', the associated foodwater flow control valve closed.
A review of maintenance records did not identify any work associated with the missing diode, and past testing on this circuitry was always performed in a mode in which the feedwater control valve would already have been closed.
It is believed that the diode was most likely missing when the cabinet was received on site.
The diode was replaced under WR B231541.
The performance i
requirements of SI-618 were also reviewed, and it was determined that its performance is no longer required. This procedure will be cancelled by July 6, 1988. l
D ceription of Fv*nt 3
LER 2-88028 This report described two reactor trips, one on June 8 and one on June 9, 1988. Both trips occurred immediately after startup during power ascension and were a result of S/G No. 2 Lo-Lo level.
On the June 8, 1988 event, unit 2 was at approximately 15 percent power and in the process of changing over to automatic speed control on the "A" MFP.
The MFP speed controller was placed into automatic but S/G levels were dropping, and the MFP ran back to minimum speed. The speed controller was.then placed back in manual to increase pump speed, and the S/G levels began to overfill. The main regulating valves were manually being closed to prevent overfill when a turbine trip /feedwater isolation occurred from a High-High level (75 percent) in S/G No. 4.
The MFP was reset to reestablish main feedwater (MFW) flow, but before the pump could start, the reactor tripped on Lo-Lo level in S/G No. 2.
On the June 9, 1988 event, unit 2 was at appcoximately 19.7 percent power and was experiencing gland sealing steam pressure fluctuations because of. inoperable pressure control valves.
Because of the sealing steam pressure increasing to the No. 7 heater drain tank, the low pressure heater strings began isolating on high level conditions. The cycling of the heater isolation valves caused feedwater flow and S/C level perturbations.
The UO attempted to maintain control of the S/G levels, but.the reactor tripped on S/G No. 2 Lo-Lo level.
The major causes of the events were insufficient attention to secondary side maintenance and insufficient guldelines for MFW i
control during startup. To help prevent recurrence, a work control center has been established to maintain a more comprehensive control of outstanding work orders and additional guidelines for controlling S/G level during startup have been placed in operating j
i procedures.
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i ABBREVIATIONS j
Page 1 of 2 1.
- Auxiliary Building Gas Treatment System 2.
- Auxiliary Building Secondary Containment Enclosure 3.
ABI
- Auxiliary Building Isolation 4.
- Auxiliary Feedwater 5.
AOI
- Abncemal Operating Instruction 6.
ASOS
- Assistant Shift Operation Supervisor 7.
- Assistant Unit Operator 8.
BAT
- Boric Acid Storage Tank 9.
BIT
- Boron Injection Tank 10.
CAQR
- Condition Adverse To Quality Report 11.
- Centrifugal Charging Pump 12.
- ' Component Cooling Water 13.
CRI
- Control Room Isolation 14.
- Control Room Emergency Ventilation System 15.
CSS (CS) - Containment Spray System 16.
- Containment Ventilation Isolation
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17.
D/G(s) - Diesel Generator (s) 18.
- Design Change Request 19.
- Division of Nuclear Engineering 20.
- Emergency Core Cooling System 21.
- Engineering Change Notice 22.
- Emergency Gas Treatment System 23.
- Electromagnetic Interference 24 EQ
- EnvironmentallyQualified/EnvironmentalQuali(ication 25.
- Essential Raw Cooling Water 26.
ESF(A) - Engineered Safety Feature (Actuation) 27.
FCV Flow Control Valve 28.
- Final Safety Analysis Report 29.
FWI Feedwater Isolation 30.
GOI
- General Operating Instruction 31.
GPM
- Gallons Per Minute 32.
MO
- Mold Order 33.
IM Instrument Mechanic 34.
IMI
- Instrument Maintenance Instruction i
35.
- Level Control valve i
36.
LCO
- Limiting Condition for Operation 37.
- Loss Of Coolant Accident 38.
- Maximum Allowable Stroke Time 39.
MPI
- Main Feedwater Isolation 40.
- Main Feedwater Pump 41.
- Motor Operated Valve 42.
MSI
- Main Steam Isolation 43.
- Main Steam Isolation Valve 44.
- Main Control Room 45.
- Nuclear Security Service 46.
- Nuclear Steam Supply System 47.
- Plant Operation Review Committee 48.
PRO
- Potential Reportable Occurrence 49.
- Reactor Coolant System 50.
RMR
- Residual Meat Removal,
w
ABBREVIATIONS Page 2 of 2 51.
RM Radiation Monitor (RAD Monitor / RAD MON) 52.
- Refueling Water Storage Tank 53.
- Significant Condition Report
)
54.
- Spent Fuel Pit 1
55.
S/G(s)
- Steam Generator (s) t 56.
- Surveillance Instruction /or Safety Injection 57.
SMI Special Maintenance Instruction 58.
SOI
- System Operating Instruction 59.
- Sequoyah Nuclear Plant 60.
SR Surveillance Requirement 61.
SSPS
- Solid State Protection System 62.
TACF
- Temporary Alteration Control Form 63.
TI
- Technical Instruction 64.
TS(s)
- Technical Specification (s) 65.
UO/(S)RO - Unit Operator /(Senior) Reactor Operator 66.
WP Workplan 67.
- Work Request 6
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RADWASTE
SUMMARY
June 1988 1.
Total volume of solid waste shipped offsite:
.A.
Dry active waste:
1860 ft.8
. Activity:
1.5104E1 curies B.
Spent resins, sludges, bottoms:
475.5 ft.s Activity:
12.536El curies Shipped: June 6, 23, 25, 1988 Barnwell, Inc.
2.
Radwaste onsite and awaiting shipment:
A.
Resin in storage:
158.5 ft.8 B.
Estimate resin that will be generated:
276.5 ft.8 C.
Dry active waste awaiting shipment:
993.5 ft.a OFFSITE DOSE CALCULATION MANUAL CHANGES No changes were made to the SQN Offsite Dose Calculation Manual (ODCM) in June 1988.
I e
i OPERATING STATISTICS (NRC REPORTS) l i
1 i
1 i
OPERATING DATA REPORT DOCKET NO. 50-327 DATE JULY 05, 1908 COMPLETED BY D.C.DUPREE TELEPHONE (615)870-6722 OPERATING STATUS 1.
UNIT NAME: SEQUOYAH NUCLEAR PLANT, UNIT 1
NOTES:
-2.
REPORT PERIOD: JUNE 1988 3.
LICENSED THERMAL POWER (MWT):
3411.0 4.
NAMEPLATE RATING (GROSS MWE):
1220.6 3.
DESIGN ELECTRICAL RATING (NET MWE):
1148.0 6.
MAXIMUM DEPENDABLE C APACITY (GROSS MWE):
1183.0 7.
MAXIMUM DEPENDABLE CAPACITY (NET MWE):
1148.0 8.
IF CHANGES OCCUR IN CAPACITY RATINGS (ITEMS NUMBERS 3 THROUGH 7)SINCE LAST REPORT, GIVE REASONS:
9.
POWER LEVEL TO WHICH RESTRICTED, IF ANY(NET MWE):
- 10. REASONS FOR RESTRICTIONS, IF ANY:
THIS MONTH YR.-TO-DATE CUMULATIVE
- 11. HOURS IN REPORTING PERIOD 720.00 4367.00 61368.00
- 12. NUMBER OF HOURS REACTOR WAS CRITICAL O.00 0.00 24444.91
- 13. REACTOR RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 14. HOURS GENERATOR ON-LINE O.00 0.00 23781.13
- 15. UNIT RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 0.00 0.00 77060971.91
- 17. GROSS ELECTRICAL ENERGY GEN. (MWH) 0.00 0,00 25976386.00
- 18. NET ELECTRICAL ENERGY GENERATED (MWH)
-3819.00
-30287.00 24024036.00
- 19. UNIT SERVICE FACTOR O.00 0.00 38.75
- 20. UNIT AVA~ILABILITY FACTOR O.00 0.00 38.75
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 0.00 0.00 35.24
- 22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 0.00 35.24
- 23. UNIT FORCED OUTAGE RATE 100.00 100.00 54.63
- 24. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):
- 25. IF SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
s START UP IS UNDETERMINED AT THIS TIME PENDING DESIGN CONTROL REVIEW, CONFIGURATION CONTROL UPDATING, AND RESOLUTION OF SIGNIFICANT EMPLOYEE CONCERNS.
NOTE THAT THE THE YR. -TO-DATE AND CUMULATIVE VALUES HAVE DEEN UPDATED. -
w y
9 w
p ory r
}
SEQUOYAH-NUCLEAR PLANT i
AVERAGE DAILY PCWER LEVEL l
DOCKET NO. : 50-327 UNIT : ONE DATE : JULY 05,1988-COMPLETED BY : D.C.DUPREE TELEPHONE : (615)870-6722 MONTH: JUNE 1988 l
AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)
DAY (MWe Net) 01 0
16 0
i 02 0
17 0
03 0
18 0
04 0
19 0
i 05 0
20 0
06 0
21 0
07 0
22 0
1 08 0
23 0
09 0
24 0
10 0
25 0
11 0
26 0
12 0
27 0
13 0
28 0
14 0
29 0
15 0
30 0
l i -)
UNIT SlIUTDOWNS AND POWER REDUCTIONS DOCKE*T NO. 50-327 UNIT NAME sequoyah one DATE July 5, 1988 COMPLETED BY D.
C.
Dupree REPORT !!0NTil JUNE 1988 TELEPIIONE (615) 870-6722 8
b Cause & Corrective m
q cg oy, I,icensee p,
N c
c, No.
Date g
gg y3g Event u7 gg Action to-o p
3o y
ggd Report #
go gu Prevent Recurrence 5~
EN5 A
I 880101 F
720 F
4 Design Control, Configuration Updating, and Employee Concerns.
2 3
1F: Forced Reason:
Hethod:
4Exhibit G-Instructions S: Scheduled A-Equipment Failure (Explain) 1-?!anual for Preparation of Data B-!!aintenance or Test 2-t!anual Scram.
Entry Sheets for Licensee C-Refueling 3-Automatic Scram.
Event Report (LER) File D-Regulatory Restriction-4-Cont. of Existing (NURE3-0161)
E-Operator Training & License Examination Outage F-Administrative 5-Reduction G-Operational Error (Explain) 9-Other SExhibit I-Same Source II-Other (Explain)
OPERATING DATA REPORT DOCKET NO. 50-328 DATE JULY OD, 1988 COMPLETED DY D.C.DUPREE TELEPHONE (615)B70-6722 OPERATING STATUS 1.
UNIT NAME: SEGUOYAH NUCLEAR PLANT, UNIT 2
NOTES:
2.
REPORT PERIOD: JUNE 1988 3.
LICENSED THERMAL POWER (MWT):
3411.0 4.
NAMEPLATE RATING (GROSS MWE):
1220.6 5.
DESIGN ELECTRICAL RATING (NET MWE):
1148.0 6.
MAXIMUM DEPENDABLE CAFACITY (GROSS HWE):
1133.0 7.
MAXIMUM DEPENDADLE CAPACITY (NET MWE):
1148.0 8.
IF CHANGES OCCUR IN CAPACITY RATINGS (ITEMS NUMBERS i
3 THROUGH 7)SINCE LAST REPORT, GIVE REASONS:
i 9.
POWER LEVEL TO WHICH RESTRICTED, IF ANY(NET MWE):
l
- 10. REASONS FOR RESTRICTIONS. IF ANY:
THIS MONTH YR.-TO-DATE CUMULATIVE
- 11. HOURS IN REPORTING PERIOD 720.00 4367.00 03328.00
- 12. NUMBER OF HOURS REACTOR WAS CRITICAL 404.18 785.10 22769.64
- 13. REACTOR RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 14. HOURS GENERATOR ON-LINE 314.77 680.75 22174.97
- 15. UNIT RESERVE SHUTDGWN HOURS 0.03 0.00 0.00
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 820272.14 1600976.36 70728953.58
- 17. GROSS ELECTRICAL ENERGY GEN. (MWH) 265390.00 514930.00 24051710.00
- 18. NET ELECTRICAL ENERGY GENER ATED (MWH) 265870.00 442026.00 22950171.60
- 19. UNIT SERVICE FACTOR 43.72 15.59 41.58 j
- 20. UNIT AVAILABILITY FACTOR 43.72 15.59 41.58 l
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 32.17 8.82 37.49
- 22. UNIT CAPACITY FACTOR (USING DER NET) 32.17
-0.82 37.49
- 23. UNIT FORCED OUTAGE RATE 56.28 84.41 54.21
- 24. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):
- 25. IF SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
NOTE THAT THE THE YR.-TO-DATE AND CUMULATIVE VALUES HAVE DEEN UPDATED. 4
+m n
o
--(=
4 SEQUOYAH NUCLEAR PLANT AVERAGE DAILY POWER LEVEL DOCKET NO. : 50-328 UNIT : TWO DATE : JULY 05,1988 COMPLETED BY : D.C.DUPREE TELEPHONE : (615)870-6722 MONTH: JUNE 1988 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)
DAY (MWe Net) 01 1094 16 0
02 1093 17 0
03 1088 18 0
04 1088 19 0
05 1081 20 0
06 643 21 0
07 0
22 0
08 1
23 103 09 3
24 335 10 0
25 417 11 0
26 675 12 0
27 750 13 0
28 748 14 0
29 738 15 0
30 745 f
1.
UNIT SlIUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-328 UNIT NAME Sequoyah Two DATE July 5.
1988 COMPLETED BY D.
- c. Dupree REPORT !!ONTil JUNE 1988 TELEPIIONE (615) 870-6722 n
u o
c oau Licensee gg g
Cause & Corrective Cg "c
3SJ g
ggg Event ug g]
Action to No.
Date o
N O
S"*
Report #
gu gu Prevent Recurrence 5"
E55 R
7 880606 F
46.60 A
3 Low flow No. 4 S/G while performing a test.
A diode was missing in the circuitry.
(LER:
2-08027) 2 8
880608 F
15.27 F
3 Lo-Lo No. 2 S/C. More experienced operators were needed to assist the less experienced during startup.
(LER:
2-88028) 9 880609 F
343.37 A
3 Lo-Lo No. 2 S/C.
Various valves to the feedwater heaters isolated, causing a loss
- i of flow.
(LER: 2-8802P) 1 1
2 3
i 1F: Forced Reason:
Method?
4Exhibit G-Instructions S-Scheduled A-Equipment Failure (Explain) 1-itanual for Preparation of Data B-!!aintenance or Test 2-flanual Scram.
Entry Sheets for Licensee C-Refueling 3-Automatic Scram.
Event Report.(LER) File D-Regulatory Restrict. ion 4-Cont. of Existing (NUREG-0161)
E-opera *ar Training & Li, cense Examination Outage F-Adminis_rative 5-Reduction G-Operational Error (Explain) 9-Other SExhibit I-Same Source II-Other (Explain)
I i
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OPERATING STATISTICS (TVA REPORTS) 1 l
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rve. ras: A top-a-aal NUCLEAR PLANT OPERATING STATISTtCS SEQUO"AH NUCLEAR Plant l
Period Hours 720 Month JUNE 19 88 Item No.
Unit No.
UNIT ONE UNIT WO PLANT 1
Average Hourly Gross Load, kW 0
843,132 843,132 2
Maximum Hour Net Generation. MWh 0
1,100
',100 ~
3 core Thermal Energy Gen, GWD (t)2 0
34.1780 34.1780 4
Steam Gen. Thermal Energy Gen., GWD (t)2 0
34.3155 34.3155
[
S Gross Electrical Gen., MWn 0
265,390 265,390 3
6 Station Use, MWh JE 24.520 28,339 I
7 Net Efectrical Gen., MWh
..-1 819 240.870_
237.051 O
8 Stati >n Use. Percent N/A 9.24_
10.68 9
Accum. Core Avg. Exposure. MWD / Ton!
10 CTEG This Month,10 BTU
~
0 8.849 8.849 6
7 2.799.589 2.799.589,_
11 SGTEG This Month,106 BTU 6
2.810.855 2.810.855 12 i
13 Hours Reactor Was Critrcal 0.0 404.18 4 0 't 'd 4 14 Unit Use. Hours Min.
0:00 314:46 31 IS Capacity Factor. Percent 0.0 31.16 l ').18_
16 Turbine Avail. Factor, Percent
- 0. Q__
98.87 49.44
_17 Generator Avad. Factor. Percent O_0 100.00 50.00 4
0 18 Turbonan. Avail. Factor, Percent 0.0 98.87 49.44 h
19 n.'ctor Avait. Fac'er. Percent A_O 97_94 RQD__
e 2
20 mt Ava,i. ractor. Percent 0.0 43.72 21.86 21 Turbine Startuos 0.0 3
3 22 Reactor Cold Startons 0
0 0
23 1
24 Gron Heat Rate. Btu,kWh N/A 10,550 10,550
[
25 Net Heat Rate. Rtu,kWh N/A 11 620 11,810 1
3 26 Gross Heat Rate Btu /kWh (w/o oil:
10,550 C
'7 Net Heat Rate Btu /kWh (w/o oil) 11,810 g
28 Throttle Pressure. pg N/A 875.5 875.5 g
29 Throttle Temr erature. F N/A 530.6 530.6 1
4 30 Exhaust Pr essure, inHg Abs.
N/A 2,. 7,.
2.7 E
31 Intake Water Temo.. 'r N/A 74.8 74.8 32 33 Main F eedwater, M lb/hr N/A 10.1 10.1 34
[
35 36 37 F utt Powri capacity, E FPD 404.86 363.65 768.51 sa Accum. Cycle Full Power Ocys. EFPo 0.0 230.40 230.40
{
y 39 Oil Fired fnr Generation. Gallons 1,914 2
40 Ort Heitino value. titu! Gal.
138,000 41 Qgsef Generat.on. MWh 29 19 Max. Hour Net Gen.
Max. Day Net Gen-Load MWh Time Date MWh Date Factor. %
2 43 1,100 0700E 6/1/88 26,248 6/1/88 29.93 Remarks: A For BF NP this v sue is MWD /STU and for SQNP Ond WBNP this value is MWD /MTU.
j 2(t) ind cates Thermal Energy.
a a
Dato Submitted gU L 1 1 1988 y.y,,g,aM(M)
Date Revised
_ / Plant Manager ~
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REACTOR HISTOGRAM SEQUOYAH UNIT ONE 100 90 -
80 -
g Z
70 -
T E
60 -
?
50 -
ao 40 -
G.
30 -
F-0 20 -
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REACTOR HISTOGRAM SEQUOYAH UNIT TWO 100 y
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9 13 17 21 25 29 JUNE 1988
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1.
Reactor at 99 percent power,1124 MWe, High steam-flow coming-in.
-reduced power.
2A.
Reactor trip, low flow to No. 4 S/G.
2B.
Reaactor trip, Lo-Lo No. 2 S/G.
2C.
Reactor trip, Lo-Lo No. 2 S/G, feedwater heaters isolating.
3.
Mainten&nce on "B" main feed pump turbine.
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SUMMARY
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SUMMARY
i (ELECTRICAL) l 1
I
ELECIRICAL MAINTENANCE P.ONTHLY SUMMRY 07-04-68 COMP i'
MR2.... U FUNC SYS ADDRESS. DATE.... D ESCRIPTIIDt............................. C B3370511 MV0P 067 Chi 1D 06/11/881-MV0P-047-04910 A,FLUSHOUT VRVE ON 1A LIGHT DUST DEPOSIT ON CONTACT POINTS OF STRADER WILL CLOSE ELECTRICALLY BUT THE RRAY, IN JUNCTION BOX 3243. CLEMED VILL NOT OPEN ELECTRICALLY WITH EITHER CONTACTS OF RELAY IN JUNCTION B THE IPEN PUSHPUTTON OR THROTRE GI 0FF MD VERIFIED PROKR OPERATIDH OF HMDSWITCH.
RUSHali VALVE. (WR B237051). EQUESTED THAT A FAILURE AMLYSIS BE PERFORMED BY THE MAINTENANCE ENGINEERING SECTION. (WR B237451).
B240673 2 MV0P 043 007212/05/87 2-MV0P-063-0072-A,EnNPRDu,ul0CFR50,49u,n BREMD501 IN LUBRICATDC ABRITY DUE TO P2u3,CHMGC LUBRICMT IN THE MIN GEAR A MIXTUK Of GREASES OF DIFFERENT CASE AS THEY HAVE EEEN MIXED PER CHEMICAL BASES EPLACED GREASE AND MI-10.44 PERFORED MOTOR [PERATED VALVE MALYSIS TESTINC. (WR B240473).
B274098171V0P 047 0491D 01/24/881-MUOP 067 04910-A,CnP2u,uMPRDu3,WHEN MOTOR C1RRENT WAS TOO HIGH TRYDG TO OPERATE 1-FCV-47-491E THE FUSE (APPRDXIMIELY 12 MPS) WHEN VALV 1-FC2 67-491 B-A ftEW, AFTER BEING.
OPENED MD MOTOR ROTOR WO!iD SDMETIMES REPLACED CNCE. PLEASE HAVE FCV 'ETOR' LOCK AT APPROXIMTRY 20 MPS WEN CSEEKED FCR PRC:iEM. THIS VALVE IS ON OPENING. MOTOR RATED APPROXIMTELY 20 1A-A ERCW STRAINER.
MPS AT 115V AC. REPLACG MOTOR MD ADJUSTED LIMIT SMITCES. (URt B274093).
B2765312 GEND 082 0002810/21/87 2-CEMB-082-00028-B,CxP2N,nNPRDu3,28-B THE ENGINE SPEED SWITCH WAS QUT OF D/G l'REMED TRIPPED OH DUER CURREXT $ADJUSTENT RESULTING IN OVERSPEED ON TEM TK D/G TRIPPED ON OVER SPEED ON ENGDE II, CAUSING DIESR GENERATOR 28-8 THE 01 ENG. CHECK CUT D/G - SUSMCT TO TRIP DN OVERCIRRENT. MEGGERED HEUTRAL PRCBLEM VITH THE ELECTRONIC C0VERNOR CONNECTION FOR TRMSTORMER. PERFORMED CCNTEDL. (CONTACT JIM CIEtt FCR mRE IDLE START AFTER RESETTING MOTOR IhTD.)
OPERATED PDTENTIIRIETER MD FREQUENCY MODULE. DIESR RM AT 90 RPMS. SHUT OFF REGULATOR MD ENGIE RM AT APPROXIMTRY 945 RPMS NO OVER-SPEED TRIP GCCURRED. RESET AUXILIARY RRAY TD LGCK-CUT RRAY MD ADJUSTED SPEED SWITCH FOR 2B1 ENGINE.
B279539 2 MVCP 067 ok910 05/16/88 2-MV0P-067-0491D-A ruMPRDu3,FLUSTUT THE CAPACITOR GAVE ERRATIC READINGS CN VALVE 2-PCV 67-4910 FCR 2AA STRAINER THE METER MD APRARED TO HAVE OR WRL NOT CPEN ELECT 2ICALLY USING LEAKIFG FRCM IT. WHEN VLY WAS MANUALLY j
THROITLE SWITCH CR OPEN PUSH IUTTUN.TAKEN [FF SEAT MD ATTEMPTED TO PLEASE INVESTIGATE PND REPAIR AS NEEDED. ELECTRICALLY OPEN, A FUSE BLEW IN MOTCR CONTROL CENTER. REPLtiCED THE CAPACITCH MD FUSE.
BFJ%90 2 INVD 250 GM 06/13/88 2-INVB-250-GM-D,[xW RDx3,3 WITCH W T FAN FLOW SUITCH WAS DETECTING POSSItty FUNCTICHING PRCPERLY 10THERSDME ItARM ON DUE TO NORML WEAR U-2 PNL M-1 SWITCH AND CHECKED FCR P2CPER OPERATION.
(WR R/Sil/10).
I'7597041 l'CTE 030 0078 05/26/881-i;CTB-030-0078-C.C0 MIN CONTACT NOT "C" PHASE MIN CCHTACT WAS OPEN ITLCW OPEN ENVE 02 MIN d.4 (LIS) SMI-0-317-78 MINIMUM CRITERIA 0F.02 INCH PACE 9 STATION 02 ARCKING CCNTACT NEED 10 STATICHARY MCING C,_
ELECTRICAL HAINTENANCE MONTHLY EU.1 MARY 07-04-88 CW MR2.... U fuMC SYS A DORESS. DATE.... DESCRIPTION...........................
BE REPLACED. DN-1 $nI-0 317-78 AND RETURNED BREAKER TO SEL'VICE. (WR 87517 % ).
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.s va CCnP RR2... 1 FUNC SYS ADDRE83. DATE.... DESCRIPTICN........................... CORRECTIVE ACT B228144 1 LCV 003 0164A 04/29/8814CV-003-0144A,ExP2n],BY PASS LIVEL
- CONTROLLER WILL NOT CONTROL FLOW l'2315412 FSV 003 0103014/07/88 2-FSV-003-01038-B,ErNPRDu],TROUBLESH00T A DIODE WA3 FOUND MI53ING IN THE CIRCUIT TO DETERMIE WY SAFECUARDS SAFECUARDS TEST CABINET. THE DIODE WAS BLOCK CIRCUIT DID NOT PROPERLY F1HCTION. LEFT DUT OF THE CIRCUITRY DURIN MINTEHMCE. A DIDDE WAS INSTALLED IN 2-R-53 TB859-9 AM T0840 9. PERFORRED CLOCK TEtt PORTI1M OF SI 618 FOR Pni.
B247830 1 TE 074 00400 05/10/881-TE-074-0040C,EuMPRDs],TE IS READING FOUND WIRES EVERSED AT THE ELEMENT.
DUT OF TOLERMCE PER STEP 5.1.22 03 ROLLED WIRES ON THE ELET.ENT MD
$I-87 INVESTICATE MD REPAIR CONNECTED CORRECTLY. EETURNED 10 SERVICE. (WR4 B247830).
B243C34 3 LCV 002 000105/20/88 2-LCV-002-0009,EuxPRDu3, LEVEL C(NTROL (2-LT-002-0009) THE CAUSE CF FAILURE IS VALUE 2-LCV-2-9 DID NOT RESPOND TO A UNKNOWN. THE TRANSMITTER WAS HOTWELL LCW LEVEL MD CPEN 10 REM.ENISH RECALIBRATED TO WIIHIM DESIED HDTWELL WHEN UNIT 2 TRIPPED.
TOLERMCE. VERIFIED VALVE STROKE. LCOP TROUBLE 3H00T AND RECALIERATE IF REQUIRED WORKING PROPERLY. GR4 B2dl834) ADJU PRICR TO RESTART.
CONTRILLER PROP PAND FOR FASTER.:.-
RESPONSE.
0751203 2 LT 003 0097 05/15/88 2-L1-003-0097-C,Eul0CFR50.49n,uNarDn3,CH TNE TRM!nITTER WAS DCTERMINED 10 BE ANNEL FAILED HICH REPAIR INDICATGI CME IMDPERACLE. in't CAUSE OF FAILUE IS BACK DN SCALE AT 1651 UNKNOWN. IT MY HAVE EEEN DUE TO ELECTRONICS BEAKDOWN. I'!E TRAa' nliiER WAS REM. ACED WITH A NEW DNE. ALL 0-RINGS ON SEAL WERE REPLACED, CONDUIT CONKCTICNS WERE MDE, RETCRQUED SCREWS AND BOLTS, REPLACED TRARSRITTER COVER 0-RINGS, POWERED LCDP BACK UP, CALIBRATED PER IMI-ii CC7.198, FACKFI11ED 6HD LITT IN SERVICE.
GR48731208)
B751253 211 070 173 06/14t38 2-TI-070-173,EnNFFDs),II READING 103 THE TEMPERATURE FDDIFIER WAS FCUND WITH r
DECREES F WITH PU"# NOT CPERATING MD 90 UNUS"$L DUTPUT READINGS AMD WOU CFM THRU FEATEXCHMCER. CGOLINGWATER ZERO ADJUST. RDOT CAUSE OF FAILURE TEMP IS 80 CECREES F; CALIERATE TEPP UNKNOWN. REn0VED MD INSTALLED HEW INDICATC';.
RILLIV0LIS TO CURRENT TRANSDUCER (nV/I) 1 MODIFIER (1-EM-070-0173) FRun UNIT 11 CN WRI D297600. CllECKED FCR PRCPER READINGS. REEDING OF 76 DECEEES
)
FAHREKHEIT CCERECT FCR PLMT CONDITICNs PER DPERATIONS.
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,-iewesweeswets emusu maar
.. w.24 COMP 42.... t FUNC SYS ADDREif. DATE.... DEICRIPTION....,,.
,,,,,,,,,,,,,, cy ggcityg t,cyIgg,,,,,,,,,,,,,,,,,,,,,,
B?S1338 2 Pi 003 00 % 04/11/88 2-PT-003-00 h,CmMPRDm3,1NDICAIDR HAS THE CCNDUCTOR AT THE TRAN$nIli[R W33 FAILED DOWH$CALE LOW FOUND BROKEM. TE CODUCfDR IS PART OF THE FACICRY $UPPLIED CABLE ON TE TRANSMITTER. THE CAUSE OF THE FAILED CONDUCTOR II UNKW2. IE CCDUCICR WAS KPAIRED AND THEN ETURMED TOSERVICE. NO CALIBRATION WAS PDtFORED. VERITIED INDICATDR KADING3, ( e86751338) 9 9
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MECHMICAL MINTENMCE MONTHLY $UMMRY 07-04-68 COMP MR2.... U FL'NC $YS A DDRESS. DATE.... DESCRIPTION............................. CORRECT D200%7 2 YLV 067 05520 05/02/88 2-VLV-047-0542C-A,ExWPRDu],YRVE FAILED HAD 3MLL M(IJNT Of DIRT BUILDUP DM THE SI-158.1 LEAK RATE TEST LEM RATE WA3 $ EATING $URfACE.KMOYED UALVE COKR, 507.9 $CTH.
CLEMED MD CHECKED VALUE SEATS FOR DAMGE. REINSTALLED COVER AND ETURED TO SERVICE. (WRt B200%7).
P215154 2 f!V M3 031810/22/97 2-fSV-%3-0318 B.EuMPRDu,ul0CTR50.49u],V SEAi! KE BM. SMLL AMUNT! 0F TRAtX ALVE 2-fSV-4-318 FAILED $I-158.1. LEM WEE f(I]ND IN THE VALUE. A SMLL CUT WAS RATE.1591 $CTH. RET LEAK RATE FOUND IN Tile PILUT VALVE SEAT. EMOVED
.02253CTH. RWP REQUIRED.
VALVE FROM LINE, CROUND OUT SEAL WELD MD REMYED INTEBtKS. LAPPED $ EAT MD INSTALLED E W DISC A 2 0-RING 3.
REASSERED VRVE MD TESTED. MDE SEAL WELD AND REINSTALLED VALUE. (WRt B215154).
0215163 2 f!V M3 032510/22/87 2-fSV-%3-0325 B,ExNPRDu,ul0CTR50.49u1V SEATS WEE BAD. SLIGdi UEM.
ALVE 2-TSV-43-325, TESTED TDGITER WITH DISA$$ERED VALVE, IN3fALLED NEW DISC UKUE 2-fSV-43-307 FAILED 31-158.1 LOCE MD CASKET, GROU2 OUT $EE ELD, LAPPED LEM RATE TEST. INIS WR SHOULD K WORKED 3 EAT, EASSERED YALVE, TESTED FDR IN CCNJUNCTICH WITH WRt (<215797. RWP LEAKS Am ETURNED TO SERVICE. (VR4 REQUIRED B215163).
0237600 2 VLV 072 0507 05/07/t3 2 ULV-072-0507-B,EnNPRDu1 VALVE FAILED NORML WEAR CAU!ING SLIGHT SCRATCHE! ON 3I-164.15 INVESTIGATE AND REPAIR A3 SEATING SURFACE. LAPPED VALVE DISCI HECESSARY CLEMED SEAT! AND LAPPED SEAT AND DISC.
ETURNED YRYE TO SERVICE. (WRI C237600).
0237840 2 FCV 001 0183 05/07/88 2-fCV-001-0183-A.EWRDu1 RED GPDI LIGHT DIAPHRAGM MD REGLEATOR B STAY 3 CN WHEN UALVE It CLOSED. L(EALLY MD LOC $E STEM NUT. MSGNEILM G DPERATING LIMIT SWTIEH CDE! NOT (PERATE. SIED ROTATION PRDBLEM. IN!!EL DIAPHRAGM A2 REGtLATOR. TICHTENED STEM F.' T AND STAKED T) READS. (WR1 B237840).
J C264113 2 LCV C03 0175 05/07/88 2-LCV-003-0175-A ExNPRDsLVALVE ACTUATDR DIAPHRAGM WORN OUT DIAPHRAGM LEAKS OPERATIR AND INSTALLED NEW DIAPHRAGM.
REA!!Et9 LED DPERATCH MD RETURNED TO i
SERVICE. (WRI B24k113).
IH k116 2 LCV 0%
0143A C5/07/88 2-LCV-003-0148A,ErNPRDuhDURING DIAPHRnGM CAD. INSTALLED WEW DIAPHReGM.
PERICRMHCE Of SI-75 FOLHD ATM0!!HERIC RETURNED TO SERVICE. (WRt B24 UENT LEMING ICICATE! LEAKING DIAPHRAGM. FIFAIR CR PEPLACE A!
NECESSARY 1:244119 2 LCV 003 01/5 05/07/88 2-LCV-003-0175-A,ExNPRDuh AIR LEMING DIAPHRACi! MAS IN C000 SHAPE (UT [HC
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THRDUCH ATMCS UENT ICICATING DIAPHRAGM 0-RING AND CRITACE IN PLA
).
MECHMICAL MINTENMCE MONTHLY SUMARY 07-04-88 COMP rR2... U FuM C sys A DDR Ess. DATE.... DESCRIPTION............................. CORRE CTIVE ACTIO.........
IS LEAKING AIR INSTALLED NEW DIIARAGM, REGULATOR, MD tu!LT W CMfSER OF PLATE BY WELDING.
(WR0 B2A119).
B?d72412 VLV 001 0624 05/25/88 2-VLY-M1-0424,EuMPRDs,mP2:3,VLV IS PACKING HAD BEEN ADJU$1ED 100 MUCH.
LEAKING STEM AT COTH ENDS OF VLY BONNET REPACKED VnLVE MD RETURNE WHERC STER PDETRATES.
(HRt B247241).
B290001 1 LCV 003 015dA 05/13/881-LCU-M3-015de,EnP2mM'ALVE OPulATOR DIAPHRAGM WAS LEAKING ARQUND STEM. TCP DIAPHRAGM APPEARS TO BE LEAKING - AIR OF Sitn MS ROUGH, BUT STRL ADERUATE ESCAPING FRun ATMO$PHERIC WNT IN TOP OF FOR USE. INSTELED NEW DIoPHRAG BONNET.
RETURNED TO $ERVICE (WR$ B290801).
B292910 2 VLV 072 0507 05/04/88 2-AV-072-0507-6,tnNPRDu], CHECK WLVE RUST DN $ EATING SURFACE LAPPED SEAT, FAILED 3I-164.15 RUN B C.S. PunP DN RAPPER, CLEMED EL PARf t, ERAPPED REPAIR TO FLUSH CHECK VALVE SEAT RE CAttET WITH TAPE, INSTALLED INi[RNALS, PERFDRMED 31-164.15 IF VALVE FAILS AGAIN CONNET MD RETORQUED. WORK TO BE INVE$i! GATE oND REPAIR.
COMPLETED PER UR0 B237dM. (WRt B292110).
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MEC:LVICAL MINTENANCE MONTNLY StM1ARY C H 4-88 CDir MR2.... U FUNC SYS A 00RE33. DATE... 0[$CRIPTIDN............................. CORRECTIV B26214 i PMF 070 0%d 0V30/881-PMP-070-0%4,EmP2s3,0UffilARD PUMP NORMAL KAR. PACUNC SLEEVES WORN PACKING MS EXCE33M LEAKAGE TIGHTEN OR ELOWING EXCESSM LEAKAGE. IN31ELED REPLACE EW PACKING ON TE INBOARD AND DuiBOARD ENDS. INILE EPLACING PACKING $LEEVE3 IT MS DBSERVED THAT THE EAR RING CLEMANCE MS UNACCEPTABLE AND H0 WEM i
RINGS EK AVAILABLE SEPAETELY SO TE ENTIRE ROTATING ELEMENT WAS REPLACE 0.
(WR0 P2A21's) e t
MECHANICAL MAINTENANCE MONTHLY REPORT FOR JUNE 1988 Unit 1 1.
Completed inspection of 204 valves equipped with reach rods.
2.
Installed PVC piping on RCP motors 2, 3, and 4.
3.
Completed repair on valves62-582, ~586, -583, -587, -602, and -603, 4.
Opened / closed pressurizer manways.
5.
Completed repair on valve 61-192.
6.
Completed monthly / annual inspection on lA-A and IB-B 0/G.
7.
Cleaned 18-B RHR pump room coolers.
8.
Reinstalled 18 high pressure fire pump (HPFP).
9.
Completed SI-106 on ice condenser.
- 10. Supported Combustion Engineering on S/G.
11.
Completed work on various valves during RCS and CVCS outase.
12.
Completed 51-233.2 penetration inspection.
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13.
Completed various activities on CVCS outage.
14.
Installed new pump on IB HPFP.
15.
Completed repair on FE-70-21 and -205.
16.
Completed repair on 1-FCV-62-171.
17.
Completed repair on 6 Grinnell diaphragm valves.
l 18.
Completed work on 1-VLV-61-678.
35-
MECHANICAL MAINTENANCE MONTHLY REPORT FOR MAY 1988 Unit 2 (Continued) 1.
Completed repair on 2A main bus duct cooler coils.
2.
Cunpleted repair on 2-LCV-6-105B.
3.
Completed repair on 2B penetration room cooler.
4.
Completed monthly / annual inspection on 28-B and 2A-A D/G.
5.
Completed repair on 2A-A CCP room cooler and speed increaser cooler.
6.
Applied Belzona on 2A CCW.
3 7.
Rebuilt main feed pump governor /stop valves.
B.
Changed setpoint on valves 72-512 and -513.
9.
Cleaned water boxes 2A-1, '2A-2, 28-1, and 28-2.
- 10. Completed repair on 2A-A and 28-B main feed pumps.
11.
Completed repair on annulus vacuum fan.
12.
Completed repair on 2A-A 690 penetration room cooler.
- 13. Unplugged floor drain collector tank line.
14.
Furmanited 2-FCV-3-191.
- 15. Repaired 2A2 MSR sightglass.
16.
Replaced screens on 2A traveling screens, e
Common 1.
Completed repair on precipitator agitator at now makeup 01 building.
2.
Reinstalled atmospheric drain sump pump.
3.
Completed repair on "A" glycol chiller line.
4.
Rebuilt "D" glycnl chiller.
Other Completed ice condenser inventory (measure ice baskets for acceptable weight limits).
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SUMMARY
OF WORK COMPLETED MODIFICATIONS - CURRENT STATUS JUNE 1988 l
Major capital Projects:
EN7101:
ECN 6388 - 500-kVA Switchyard Current Transformer Heaters Workplan (WP) 12223 is in progress pending safe access to current transformers.
PN7102:
ECW 5938 - Replace Feedwater Heaters 3 and 4 No work in progress at this time.
PN7105:
ECN 5009 - Essential Raw Cooling Water (ERCW) Pipins Changeout From Carbon Steel to Stainless Steel No additional pipe replacement is scheduled in the near fu turo.
PN7108:
ECN 6720 ~ Crane Consistency Program Unit 2 polar crane modification is complete.
Postmodification testing (PMT) by Electrical Maintenanco is not complete. Unit 1 polar crane work started on July 6, 1987 and is approximately 98 percent complete.
Painting of blocks and limit = ditch weight's remain to be completed.
Auxiliary Building crane WP 12596 is in the approval cycle.
Drawings have not been issued on the remaining cranes. WP 12596 is in-nonwork status. Work has surpasced $40,000 allocated toward writing workplanc.
PN7115:
ECN 6719 - Volumetric Intrusion Detection System ECN 6719 is in work.
Design is working on making the system functional. Lighting is not finished. Field Change Request (FCR) 6645 for DNE changes is approved. Work is proceeding.
PN7122:
DCR 1373 Secondary Side No work in progress at this time.
ECN 5841 - Hot Shop Fire Protection / Evacuation Alarm -
WP 12637 All fieldwork for evacuation alarm is complete.
Awaiting Work Request (WR) B240406 to be worked to restart fans to do functional test on fire protection.
Responsiblo Section -
Plant Upgrado, Allen Ashley, extension 7802, 4
_ ~
Major capital Projects (cont.):
PN7130:- DCR 1156 - Post Accident Monitoring This work is now ccheduled for unit 1 by unit 1 cycle 4 (UlC4) and unit 2 by unit 2 cycle 4 (U2C4),
e PN7132:
DCN 0026 - Sewage Treatment Facility and Civil UpReade Work has begun. WP 0026-01 has been approved and is ready to work.
PN7136:
ECN 6259 - MSR Tube Bundle Replacement ECN is complete except for PMT and inservice 'ieaketest.
Leak checks will be performed during system heatup. Unit 2 leak checks are complete.
PN7161: ECN 5855 - Replacement of Doors A56 and A57 Functional testing is complete.
PN7181:
DCR 1898 - ECNs 6832 and 6596 - Dry Active"Waste (DAW)
Building Electrical interface work is complete. Workplan closure held for Electrical Maintenance (WPs 12478, 12612), checks and update of Sols by Operatlons, and Operations (WP 12477) pending RHSI-l revision.
Significant Items:
PN7199: Miscellaneous Activities Under $100.000 This is for various work orders prepared for work under
$100,000 total sito cost. This work was done as manpower i
resources were available that did not impact unit 2 restart.
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ECN 5111 - Provide Permanent Power to Manholes 42-46 FCR 4572R2 has been written to supersede 4572R1 for conduit and cable routing for manholes 42-46.
An instruction change form is being written to revise WP 12262 to incorporate latest design information on FCR 4572R2.
ECN 5503 - Ev6c.uation Alarms O&PS/ Fire Detection O&PS WP 12482 in work July 5, 1988, i
ECN 5552 - Condensate Demineralizer Modifications and High Crud Filter Upgrade to higher range instrumentation for condensate demineralizer system neutralizati'on and nonreclaimable. waste pumps.
WP 5552 Fieldwork is complete.
ECN 5609 - Evacuation Alarm / Fire Detection Valve 26-290 WP 12387 is 90 percent complete. No work is in prEgress at this time.
ECN 5609 - Alteration to the Makeup Water Treatment Plant WP 12576 - Work is approximately 75 percent complete. Hold for Electrical Maintenance to perform breaker tests. Need to fabricate.
and install instrument tags.
i WP 12633 - Work is approximately 60 percent complete. Hold for instrument tab (47B6dl-928) set point revisions.
WP 12731 - WP is approximately 60 percent completo. Hold for instrument tab (4 7B601-928) set points.
WP 12684 - WP is approximately 70 percent complete.
Hold for instrument tab (47B601-928) set points.
WP 12665 - WP is on hold for outage and is 95 percent complete.
WP 12682 - WP is on hold for outage and is 80 percent complete.
ECN 5626 - Containment Ladders, Unit 1 Modifications need; additional design information to complete.
DNE needs to issue all drawings listed on this ECN. Work has not begun because of this holdup.
4 Other Items (cont.):
ECN 5841 - Hot Shop Firo Detection WP 12360 is complete. Workplan in closure cycle.
ECN 5935 - Correct Power 31ock Lighting Deficiencies WP 12437 is complete. WP 12275 is in work. Modifications needs DNE to provide missile protection requirements for our excavations and the USQD. Need workplan revision for FCR 7706 to add approximately 40 more lights. WP 5935-01 is in work to add security grills and grating.
ECN 6057 - Cable Tray Covers This activity is complete for the unit 2 restart. The only remaining work is the unit I annulus and this will be completed prior to July 31, 1988.
ECN 6196 - Pressurizer Hangers and Valves PHT is scheduled for unit I restart. Remaining unit 2 work is scheduled for U2C3 refueling outage.
ECN 6357 - Essential Raw Cooling Water (ERCW) Roof Access and Rails for Security Equipment WP 12238 is in work and is approximately 40 percent complete.
ECN 6388 Hydrogen Monitors in Switchyard Workplan 12223, installation of hydrogen analyrecc at 500-kV switchyard, is 75 percent complete.
ECN 6429 - Component Cooling Heat Exchanger B Replacament Plato head exchanger and framo is complete. Hangers and pipe which can be installed prior to outage is in work. Workplan to remove old heat exchanger is in review cycle.
DNE procuring the remaining needed piping material.
Estimated completion dato for nonoutage work is during the month of June 1988, i
ECN 6455 - Upgrado CU-3 Box Battery Packs WP 12295 has been issued. Modifications are completo for all CU-3 boxes.
Site Security still needs to perform SI-630 on one remaining channel before workplan is closed.
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Other Items (cont.):
ECW 6543 - Install Public Safety Access portals and Modify Entrance Road Work is being held pending the release of drawings from DME.
ECW 6601 - Removal of Unit 1 Emergency Gas Treatment System (ECTS)
Backdraft Dampers pMT remains to be completed by the Mechanical Test Section.. Fieldwork.
is complete.
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ECN 6610 - Modify Air Return Fan Supports Unit I work is in process.
ECN 6689 - Relocati'n of' Main Steam power Operated' Relief Valves (p0RV) o All work is complete for unit 2.
Work on unit 1 is being held by.
Modifications Section A (MOD A).
Mod A holding for resisters and
[
diodes on order. Work is approximately 40 perce..t complete.
l ECW 6698 - Repull 120-Volt Cables Unit 2 is complete. Unit 1 is 90 percent complete and held by ECN 6742. This ECW should be completed prior to unit I restart.
ECW 6706 14 Support Enhancement / Lost Calculations.
Repairs continue on unit 1.
This project has been combined with the calculation regeneration project for unit 1.
Unit 2 work is complete and workplans are closed. There are 25 modifications in work with 46 having been completed.
In addition, 82 maintenance items are in l
progress with 293 complete.
1 i
ECW 6739 - Alternate Analysis All unit 1 modification work has been completed. Review and closure has begun. Work is in progress on the maintenance items.
ECN 6742 - Install Fuses in Radiation Monitor power Supply Circuits The workplan is completo for unit 2.
Unit 1 is in work with Train A being complete.
Awaiting Train B tag out.
ECW 6761 - East Valve Room (EVR) Blowout panolo Implementation of unit I work is progressing as resources are availablo. Unit 2 work is completo.
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o Other Items (cont.):
ECN 6784 - Documentation to Show Pipe Class Breaks Final closure is awaiting a revision to AI-19 deleting requirements to mark shif t supervisor drawings.
ECN 6815 - Installation Power Circuit Breaker Install 500-kV power circuit breaker and associated equipment for bay 1.
Retire 161-kV PC8 and associated equipment. A total of 10 workplans will be required.
Fout.dations and conduit installation (WP 12654) are complete. WP 12740 for lighting, drain pipe, and surface ground mat is 70 percent complete. WP 12739 for the structural steel installation is 90 percent complete. WP'8815-02 is 80 percent complete. WP 6815-01 and WP 6815-03 are 80 percent complete. WP 6815-04 and WP 6815-05 for the electrical control board, main relay boards,'and the communications room are in work. WP 6815-06 for the addition of the Franklin solid state relay cabinets is field complete.
The Watts Bar No. I line was energized by Bay 1 on June 21, 1988. The estimated date to energize the Franklin line is July 22, 1988.
The estimated date to have the 161-kV AEDC line and equipment removed is by mid-August 1988.
[
ECN 6860 - Control Room Bullet Resistivity - DCR 2268 - ECN 6860 WPs 12602 and 12604 are field complete. WP 17603 is held for disposition of CAQR 880183.
WPs 12605 and 6860-01 are held for material on Contract 74014 A.
Essentially all work is complete except for pull handles, replacement of door closures, rework of one lockset, and replacement of one electric hinge.
Estimated date of completion is June 30, 1988.
WP 12605 - Functional test on door C-53 and C-60 was completed; however, door c-60 and C-37 lockset/ hinge failed approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after test. Awaiting resolution from DNE and Virgil Powoll.
Modifications Section C.
ECN 7078 - Install Hangers - Main Steam Piping Complete except for final inspections during heatup.
DCN K00006A - Remove Hydrogen Analyzers Tubing The workplan lacks PMT (31-219).
ECU 7078 - Install Hangers - Main Steam Piping Complete except for final inspections during heatup.,
)
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O Other Items (cont.):
ECN 7093 - Replacement of Feedwater Pire Work is in progress and is 90 percent complete.
DCN 200018B - Install Needle Valve for Hydrogen Analyzers
{
The workplan lacks PMT (SI-219).
DCN 70 - Hydrogen Analyzer Check Valves The workplan lacks PMT (S1-219).
DCN 192 - Pressurizer Loop Seal Work is complete. Workplan in closure cycle.
DCW 341 - Pressurizer Drain Pipe / Hangers A workplan is being prepared with fieldwork to be initiated following workplan approval.
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Instrumentation Verification Program For unit 2, there have been 953 discrepancies issued to MODS to date with 476 not required for restart.
All discrepancies required for restart have been completed; 69 noncestart discrepancies are open.
For unit 1, there have been 532 discrepancies issued to MODS to date with 218 not required for restart.
Of 314 required for restart, 10 remain open and are being worked. All unit I restart discrepancies are complete.
)
l DCN 224 - RHR Slope Rework l
Train A complete, Train B needs outage.
ECN 7318 - Capillary Tray Hanger Rework In work.
DCN 27 - Double Isolation Valves In work, DNE procuring material.
DCNs 229 and 231 - H2 Analyzer Modification i
l Work is in progress on unit 1.
i -
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s a,*,
o' Other Items (cont.):
DCN 214 - AFW Tap Rotation Complete, ' Instrument halntenance to test.
DCN 242 - Sense Line Hangers In work.
ECN 6596 - WP'12402 - Workplan is closed.
WP 12477 - Requires Instrument Maintenance calibration and checkout of area radiation monitor and air compressor.
WP 12'612 - Complete.
Closure in process.
SOIs are complete.
SI-743 is in the process of being updated b,r Elcetrical Maintenance.
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TENNESSEE VALLEY AUTHORITY Sequoyah Nuclear Plant P. O. Box 2000 Soddy-Daisy, Tennessee 37379 July 11, 1988 Nuclear Regulatory Commission Office of Management Information and Program Control Washington, DC 2C555 Gentlemen:
Entlosed
.J the June 1988 Honthly Operating Report to NRC for Sequoyah Nuclear Plant.
Very truly yours, TENNESSEE VALLEY AUTHCRITY (2
Lj'Y S, ' '. SmId Plant Manager Enclosure cc (Enclosure):
Director, Region II INPO Records Center Nuclear Regulatory Commission Suite 1500 Office of Inspection and Enforcement 1100 Circle 75 Parkway Suite 3100 Atlanta, Georgia 30339 (1 copy) 101 Marietta Street Atlanta, Georbia 30323 (1 copy)
Mr. K. H. Jenison Resident NRC Inspector Director, office of Inspection O&PS-2, Sequoyah Nuclear Plant and Enforcement Nuclear Regulatory Commission Washington, DC 20555 (12 copies)
Director Office of Special Projects 4350 East West Highway, EWW 322 Bethesda, Maryland 10814 (10 copies)
,)
Mr. T. Marston Electric Power Research Institute
/
(
P. O. Box 10412 i
Palo Alto, California 94304 (1 copy)
An Equal Opportumty Employer J
,