ML20148B230
| ML20148B230 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/29/1988 |
| From: | Dupree D, Shawn Smith TENNESSEE VALLEY AUTHORITY |
| To: | NRC |
| References | |
| NUDOCS 8803210400 | |
| Download: ML20148B230 (71) | |
Text
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4 TENNESSEE VALLEY AUTHORITY
^0FFICE OF NUCLEAR POWER SEQUOYAH NUCLEAR PLANT MONTHLY OPERATING REPORT TO THE NUCLEAR REGULATORY COMMISSION FEBRUARY 1988 UNIT 1 DOCKET NUMBER 50-327 LICENSE NUMBER DPR-77 UNIT 2 DOCKET NUMBER 50-328 LICENSE NUMBER DPR-79 Submitted by:
5 d
S. J. Smith, Plant Manager 4
1/
8803210400 880229 I
PDR ADOCK 05000327
_ R
-DCD 1
t TABLE OF CONTENTS Page I.
Operational Summary Performance Summary 1
Significant Operational Events 1-2 Fuel Performance and Spent Fuel Storage Capabilities 3
PORVs and Safety Valves Summary 3
Special Reports 3
Licensee Events 4-12 Offsite Dose Calculation Manual Changes 13 II.
Operating Statistics A.
NRC Reports Unit One Statistics 14-16 Unit Two Statistics 17-19 B.
TVA Reports Nuclear Plant Operating Statistics 20 Reactor Histogram 21 Unit Outage and Availability 22-23 III. Maintenance Summary Electrical Maintenance 24-28 Instrument Maintenance 29-34 Mechanical Maintenance 35-39 Modifications 40-50 IV.
Offsite Dose Calculation Manual Changes 4
0288f
0 9
OPERATIONAL
SUMMARY
.h.
PERFORKiNCE
SUMMARY
~
February 1988 The following summary describes the significant operational activities for the month of February.
In support of this summary, a chronological los of significant events is included in this report.
The units remained in an administrative shutdown the entire month due to design control re/lew, configuration control updating, and resolution of significant employee concerns. Outage-related maintenance and modifications are being performed.
Unit I has been off line 921 days.
Unit 2 has been off line 922 days.
Preparations are underway for unit 2 alignment and power operation.
SIGNIFICANT OPERATIONAL EVENTS Unit 1 Date Time Event 02/01/88 0001E The reactor was in mode 5.
The administrative shutdown due to design control review, configuration control updating, and resolution of significant employee concerns continues.
02/29/88 2400E The reactor was in mode 5.
The administrative shutdown due to design control review, configuration control updating. and resolution of significant employee concerns continues.
Unit 2 Date Time Event 02/01/88 0001E The reactor was in mode 5.
The administrative shutdown due to design control review, configuration control updating, and resolution of significant employee concerns continues.
Preparations are underway for power operation.
RCS at 189 degrees.
360 psi.
Unit 2 Date Time Event 02/06/88 0904E Began heatup.
1706E Mode 4.
1740E RCS at 206 degrees.
02/07/88 2141E Received a main steam isolation while attempting to open the main steam isolation valves (MSIVs).
2355E Main feedwater isolation occurred when loop No. 4 MSIV was opened.
02/08/88 1600E RCS at 249 degrees, 470 psi.
02/10/88 0228E A main steam isolation occurred while performing maintenance on the turbine impulse pressure setpoints.
02/17/88 1616E Received a cold overpressurization mitigation actuation on pressure-operated relief valve 2-PCV-68 334 from loops No. 3 and No. 4 while attempting to perform an incore thermocouple and resistance temperature device calibration.
02/18/88 0117E RCS at 320 degrees, 520 psi.
02/21/88 0230E FCV-68-340 (PORV) opened momentarily while fuses were being replaced.
02/27/88 1310E Entered mode 3.
1540E RCS at 398 degrees, 696 psi.
2055E Began heatup to 525 degrees.
02/28/887 1708E RCS at 530 degrees, 1600 psi.
1737E Main steam isolation occurred during maintenance on loops No I and No. 3.
02/29/88 2400E RCS at 530 degrees, 1800 psi.
4 c
FUEL PERFORMANCE Unlt 1 The core average fuel exposure accumulated during February was O MWD /MTU with the total accumulated core' average fuel exposure of 0 MWD /MTU.
Unit 2 The core average fuel exposure accumulated during February was O MWD /MTU with the total accumulated core average fuel exposure of 8097.51 MVD/MTU.
SPENT FUEL FIT STORAGE CAPABILITIES The total storage capability in the spent fuel pit (SFP) is 1,386.
- However, there are five cell locations which are incapable of storing spent fuel.
Four locations (A10. All, A24, A25) are unavailable due to a suction strainer conflict, and one location (A16) is unavailable due to an instrumentation conflict. 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
A "COMS/ COPS" actuation occurred while performing maintenance on loops No. 3 and No. 4.
This action initiated PORV 2-PCV-68-334 on February 17, 1988.
FCV-68-240 opened momentarily during the replacement of fuses on February 21, 1988.
No safety valves were challenged in February 1988.
SPECIAL REPORTS The following special reports were submitted to NRC in February 1988.
88-02 On January 29, 1988, a fire barrier breaching permit for fire door A-72 axceeded the seven-day limit required by TS LCO 3.7.12.
Doct A-72 is to the unit 2 B-train CS and RHR heat exchanger room located on elevation 690 of the auxiliary building. This breach was used to facilitate maintenance being performed in the area. A i
firewatch has been established and will be maintained until the completion of all activities.
88-03 On January 22, 1988, DNE lasued a CAQR that identified that the suspended cement plaster ceiling in the control building, elevation 732, rooms C-5,
-6,
-7, and
-8, may not provide the 1-1/2 hour fire protection required by the FSAR, section 9.5.1.3.
An evaluation was performed; however, the evaluation and any action to verify the ceiling was not performed within seven days. A firewatch was established and will be maintained until the evaluation is complete.
4 This event was determined on February 4, 1988.
87-17 This revision provided the conclusions of an investigation and a Rev 1 status of corrective actions concerning liquid releases from the condensate L,mineralizer waste system directly to the cooling tower blowdown line while effluents exceeded the 31-day total body projected limit.
LICENSEE EVENT REPORT (S)
The following licensee event reports (LERs) were transmitted to the Nuclear Regulatory Commission in February 1988.
Description of Event LER l-88003 On January 4, 1988, at approximately 1425 EST, an inadvertent ABI occurred that actuated equipment common to both units during calibration of RM-90-101A (particulate channel) computer log point, which required ramping the output voltage of this channel up through the high radiation setpoint.
Since a high radiation signal will initiate an ABI, the RM block switch was pulled to block the channel signal. After the operator noted that all associated equipment actuated as designed, an immediate investigation ensued.
No high radiation signals were found on the recorder chart for the other two channels of RM-90-101, nor for RM-90-102 or -103, confirming that no high radiation condition existed. The other initiating sources of an ABI were investigated, but evidence indicated that none of these signals were the cause. Additional troubleshooting was conducted in an effort to simulate the previous occurecnce; however, an ABI was not produced until the block switch was almost completely pushed in to the unblocked position.
This testing confirmed that the event was not caused by a faulty block switch or the block switch being inadvertently bumped. All known sources of this ASI actuation were investigated; however, a cause could not be determined.
No additional corrective action is planned at this time.
1-88004 On January 11, 1988, at approximately 1500 EST, an assessment was initiated to determine the consequences of class lE cables identified as not being adequately protected from cable insulation auto-ignition temperatures.
It had previously been noted during sur~eillance testing that GE 50-amp type TED breaker had a high rate of failure. The test criteria were reviewed and this resulted in the discovery of two different time-current trip characteristic curves for GE 50-amp type TED circuit breakers. This discovery l
resulted in a review of the 10 CFR 50, Appendix R, cable and j
penetration calculations.
The calculation review discovered that all the GE 50-amp type TED breakers were assumed to have the same characteristic curve. The calculation was then revised using both curves, as applicable, resulting in the discovery of 24 breakers that did not adequately protect the No. 8 AWG and No. 10 AWG cables from auto-ignition temperatures.
The breakers allowing the cables to be susceptible to auto-ignition conditions were installed in unit 1 and unit 2.
As immediate corrective action, the breakers identified as being inadequate for auto-ignition protection and which occupied a common enclosure with cables that are part of a shutdown path were i
replaced. The surveillence test procedures were revised to ensuro use of the correct breaker characteristic curves.
To address the root cause, a review will be conducted to identify programmatic deficiencies in the transfer of vendor supplied technical information to the implementing design engineer.,
F Description of Event LER 1-88005 On January 16, 1988, at approximately 1700 EST, it was discovered that the turbine building sump release line radiation monitor was inappropriately declared operable on January 15, 1988, at 1400 EST, and the TS action statement was exited.
At the time of the discovery, units 1 and 2 were in mode 5.
The turbine building sump release line RM was declared inoperable on January 5, 1988, to clean the sample line. The applicable TS action statement was entered and grab samples were required to be taken for total gamma analysis at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
On January 11, 1988, a second job was started to repair the RM after it failed its surveillance test. The job was completed and functionally tested satisfactory on January 15.
The UO subsequently declared the RM operable and exited the TS action statement.
On January 16, an AUO was dispatched to investigate the cause of an instrument malfunction alarm on the RM.
He reported the RM valves were tagged shut due to an H0 insupport of the sample line cleaning job. The UO realized at that time that the RM was inoperable with this H0 in effect and the TS action statement was roentered.
Incorrectly declaring this RM operable resulted in a failure to take grab samples for 25-1/2 hours. The cause of this event is attributed to a cognizant operator error in not properly returning the RM to service in accordance with approved procedures. As corrective action, the procedure used to maintain cognizance of operation status will be covered in week one of the 1988 Operator Requalification Training.
1-88006 On January 27, 1988, at 1005 EST, an inadvertent CVI occurred when an Instrument Maintenance technician, working a WR to resolve a problem identified by surveillance testing on the shield building exhaust system (2-RM-90-100B), mistakenly removed the wrong monitor from service (2-RM-90-106, containment building).
The primary cause of the CVI can be attributed to inattention to detail by the IM.
A contributing factor to this event is the l
ambiguous method of labeling on the RM panel.
Personnel have been counseled on this event and the importance of correctly performing work activities and independently verifying that the equipment to be serviced is correct before removal from service.
Corrective action to be taken as a result of this roview includes relabeling i
l and demarcation of changes to more clearly identify specific I
radiation monitor modules.
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Description of Evant LER 1-88007 On January 24, 1988, at 1800 EST, a tour of the refueling floor and discussions with test personnel revealed that the ABSCE was not being maintained within the configuration set during the implementation of TS SR testing, used to determine ABGTS operability.
When TS SR testing was done to ensure the ABGTS can maintain the required negative pressure in the ABSCE, the blast doors (BDs) to unit 1 and unit 2 reactor buildings (RBs) were closed and the containment purge system (CPS) was secured. Opening the BDs will encompass the RB in the ABSCE boundary. When a unit is in mode 5 or mode 6, it is normal to have the BDs open for that unit, and it was possible that the CPS could be in operation.
Increasing the ABSCE boundary potentially causes more leakage into the ABSCE.
Considering the additional potential leakago, there was no assurance that ABGTS would be able to satisfy TS SR with the increased ABSCE boundary or when the CPS is operating.
The cause of this condition is the lack of adequate controls to ensure the ABSCE boundary is maintained within the condition set by SR testing and an inappropriate design assumption made during plant construction on how breaches would be controlled. As short-term corrective actions, the BDs were closed before unit 2 entered mode 4 on February 6, 1988, the procedures governing ABSCE breaches were revised, and the unit 1 CPS was tagged out of service.
Following subsequent leak testing of the unit 1 annulus, the unit 1 BD was reopened. As long-term corrective actions, a programmatic review of the procedures controlling ABSCE breaches will be conducted, and a design change will be implemented to isolate containment purge on an ABI.
1-88008 On January 30, 1988, at approximately 1600 EST, it was discovered that the train "B" ERCV effluent line RM was inappropriately declared operable, and the TS action statement was prematurely exited on the previous shift.
The RM was declared operable at approximately 0300 EST on January 30, 1988, without completing the required postmaintenance test. When both channels of the RM are inoperable, the applicable TS action statement requires grab samples to be taken and analyzed for radioactivity at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Compliance with this requirement had been maintained until the RM was inappropriately declared operable.
It had not been completely tested following maintenance at 1600 EST, and samples were resumed until completion of the test.
Approximately 15-1/2 hours had transpired between grab samples.
This event was caused partly by personnel declaring the RM operable without reviewing the associated work request to ensure the test was complete, and partly by not reviewing the test requirements before signing verification that the test was complete. As corrective actions, the TS action statement was reentered and personnel involved were counseled. _ _ _ _.
Description of Event LER 1-88009 On February 4, 1988, it was determined that SI-137.2, "Reactor Coolant System Water Inventory," did not adequately address the TS requirement for calculating the primary to secondary leak rate in mode 1 when the plant was operating at less than 100 percent power or in modes 2, 3, and 4.
LCO 3.4.6.2.c requires that the primary to secondary leak rate be limited to 1 gpm for all S/Gs, and 500 gallons per day for any one S/G.
Additional investigation revealed that the full power S/G water volume incorporated into the SI-137.2 mathematical model was also incorrect. A modification that occurred before original startup changed the S/G volume, but the mathematical model was not revised.
Because of the incorrect water volume, SI-137.2 underestimated the full power leak rate by approximately 28 percent.
The event was caused by an inadequate instruction for measuring the primary to secondary leak rate.
TSs have specific limitations on the primary to secondary leak rate during plant operation in modes 1 through 4 SI-137.2 should have had provisions for measuring the subject leakage during all modes of TS applicability.
The cause of the incorrect S/G water volume has been attributed to an inadequate change control process. This
[
preverted the FSAR from being revised to the correct S/G volume value.
SI-137.2 has been revised to correct the subject l
deficiency. Also, the design change control and plant modification l
procedures have been revised to prevent recurrence of this type of l
event.
1-88501 On January 5, 1988, two similar security events occurred relating to unauthorized individuals in a protected area.
Event One: At approximately 0130 EST, a QA employee on a visitors badge was observed by an NSS officer approaching the entrance to the auxiliary building on elevation 690, which is a protected area l
(escorted access). The visitor was stopped and placed under i
control.
The escort, a DNE employee, had entered the auxiliary building a few minutes earlier, leaving the visitor outside the auxiliary building unescorted.
The NSS officer located the escort and both individuals were escorted out of the protected area. NSS supervisors questioned both individuals, obtained statements, and instructed them in the proper visitor / escort procedures.
Event Two: At approximately 0340 EST, a plant laborer who I
previously had authorized escorted access to the protected area, requested her protected area visitor badge. The badge had expired l
at midnight the night before; however, neither the security clerk, who is supposed to check for expired badges at the beginning of each shift, nor the visitor noticed the date on the badge indicated that it had expired.
Upon receiving the badge, the visitor under escort key-carded into the protected area entrance card reader, l
received a red light (indicating stop), and continued into the protected area.
The NSS officer stationed at that card reader l
stated a green light was shown on his monitor.
An alarm annunciated in the Central Alarm Station (CAS)/ Secondary Alarm _ _ _ _
Description of Event LER 1-88501 Station (SAS) from the protected area entrance card reader that the (cont.)
v.lsitor had just traveled through. An NSS officer was immediately dispatched to the area hnd found the unauthorized visitor and escort in the area under the control of the NSS officer stationed at that post. The visitor and escort were removed from the protected area. The visitor was counseled by NSS supervisors on badging and key-carding procedures.
These events are applicable to both units, as the entrance area where the events took place is common to units 1 and 2.
1-87039 This revision provided details concerning design errors in the CREVS Rev. 2 that resulted in a violation of the single failure criteria.
It also updated information not included in previous LERs related to recent testing of the CREVs and corrective action.
1-87039 This report provided additional information relating to a dose Rev. 3 analysis performed to estimate the potential increase in radiological exposure to main control room personnel if a design basis accident had occurred before the corrective aations described therein were implemented.
1-87040 This revision provided details concerning the reactor shield Rev. 1 building mechanical penetration seals not qualified to ensure compliance with the design basis due to design deficiencies.
1-87041 This revision provided details concerning the reactor shield Rev. 1 building mechanical penetration seals not being qualified to ensure compliance with the design basis due to design deficiencies.
1-87061 This revision provided additional information relating to previously Rev. 1 reported 10 CFR 50 Appendix R deficiencies; to update the corrective actions taken by TVA, and to clarify the administrative and engineering controls that are in place to prevent recurrence of this event.
1-87070 This revision provided details of an investigation resulting Rev. 1 from a D/G voltage regulator that had a slow response during surveillance testing.
1-87070 This revision updates the schedule of the corrective action Rev. 2 resulting from a diesel generator voltage regulator that had a slowed voltage response during surveillance testing.
1-87074 This revision provided details concerning operating procedures Rev. I that do not adequately address ECCS requirements in mode 4, contrary to the requirements of TSs.
1-87078 This report provided additional information on completed corrective Rev. 1 actions concerning an inadequate procedure for reactor coolant system chemical addition that resulted in noncompliance with a TS action statement.
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Description of Event LER 1-88001 This revision included an additional deficiency found in the SI Rev. 1 scheduling computer program and additional corrective actions.
2-88001 On January 12, 1988, TVA received a letter from Stone & Webster Engineering Company which revealed that the lower support frame of the CS heat exchangers 2A and 2B could be overstressed during a design basis seismic event. The letter also revealed that due to close proximicy of the RHR heat exchangers to the CS heat exchangers, there existed a potential for physical interaction between the two components during an event.
TSs require both RHR loops operable during mode 5; however, the CS heat exchangers are not required operable during mode S.
The root cause of this condition was determined to be that design input control procedures did not exist during initial design to ensure vendor interface review in the generation of maximum design loads of the support frame. Further evaluations of the CS heat exchangers potential interaction with the RHR heat exchangers have concluded that there would be no physical interaction with the RHR heat exchangers.
Thus, the RHR heat exchangers would not be damaged as a result of the potential failure of the CS heat exchanger lower supports during a design basis seismic event. A CAQR was issued in March 1987, and modifications to the lower support frame of CS heat exchangers 2A and 2B were completed on January 19, 1988.
Preparations for modifications to unit 1 CS heat exchangers lower support frame are currently in process. Modifications will be complete before unit 1 enters mode 4.
2-88002 On January 18, 1988, a postperformance review of an IMI revealed that it incorrectly measured the response time of the two valves that open the CCPs suction line to the RWST.
During a subsequent search of SQN records, it was discovered that an October 1984 response time measurement was inadequate to verify compliance with the SQN TS for one of the two valves.
This event was caused by an inadequate review of a design change that was implemented to modify approximately 65 of the SQN Westinghouse Type W-2 handswitches, j
This modification connected the red and green valve status lights l
in series with the neutral (auto) switch contacts to provide plant l
operators with an immediate visual indication of electrical continuity when the handswitch was in the neutral position. Since l
the subject IMI obtained the response time of the CCP suction valves by measuring the voltage drop across the green status lights, any change to the electrical wiring configuration of the valve status lights could affect the response time measurement.
However, because this modification was not adequately reviewed for its potential effect on SQN procedures, no permanent changes were made to the IMI; thus the October 1984 performance was incorrect.
As an immediate corrective action, a permanent change to the IMI to properly measure the valves' response time has been made.
In addition, all IMIs used for response time testing of ESF-actuated equipment were reviewed for similar deficiencies.
To prevent recurrence of this event, TVA has consolidated the design change control and plant modification processes through the issuance of new nuclear engineering procedures and a revised administrative instruction.
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Description of Event LER 2-88003 This LER was submitted as a voluntary report to inform NRC of the degradation found during an inspection of the unit 2 lee condenser.
On January 19, 1988, an inspection of the unit 2 ice condenser revealed that accumulation of ice or frost in the flow passages represented a degraded condition. TS 3.6.5.1 indicates that more than one restricted flow passage per ice condenser bay (there are 24 bays in the ice condenser) is evidence of abnormal degradation of the ice condenser. An SMI was written to determine the extent of degradation in order to justify that SQN met the intent of TS 3.6.5.1.
Westinghouse had previously completed a computer analysis which concluded that with 15 percent of the ice condenser flow passages blocked, the containment pressure would be within design limits.
The SMI was written with an acceptance criteria of less than 15 percent total flow passage blockage. The SMI was performed, acceptance criteria was met, and the ice condenser was determined to meet the requirements of TS 3.6.5.1.
Thc cause of the increased ice / frost accumulation has not been deturmined.
Upon completion of an investigation, this LER will be revised to detail the results and corrective action for this condition.
2-88004 On January 27, 1988, at approximately 1743 EST, an emergency start of all four emergency D/Gs occurred during surveillance testing.
Performance of the SI required the transfer of the 2A-A 6900V shutdown board from the alternate feeder breaker to the normal feeder breaker from the control room. While attempting to perform this evolution, the normal feeder breaker failed to close while the alternate feeder breaker was open.
With both feeder breakers open, the 2A-A 6900V shutdown boards deenergized, resulting in a blackout condition. The UO recognized the blackout condition and immediately closed the alternate feeder breaker before load shedding occurred.
The immediate cause of this event was due to the normal feeder breaker failing to close.
The normal feeder breaker did not close because the applicable lockout relay had been damaged.
The damage was determined to be caused by a large rolling scaffold assembly found in the are'a, The lockout relay has been inspected and tested and has been determined to be operable until a replacement relay can be procured and installed. The scaffolding assembly has been removed from the 6900V shutdown board rooms to prevent further d amage. To prevent recurrence, rolling scaffolding that can impact installed plant equipment will be equipped with bumpers that will preclude equipment damage when it is necessary to use one in the shutdown board rooms.
2-88501 This is a safeguards event. ___. _. _..__ _
Description of Event LER 2-87008 This revision provided information relating to a CVI caused by EMI.
Rev. 2 2-87009 This revision provided additional information relating to two CVIs Rev. I resulting from EMI.
2-87010 This revised LER provided additional information relating to two Rev. 1 CVIs caused by EMI.
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f ABBREVIATIONS l
1.
- Auxiliary Building Gas Treatment System 2.
- Auxiliary Building Secondary Containment Enclosure 3.
ABI
- Auxiliary Building Isolation 4.
AOI
- Abnormal Operating Instruction 6.
AUO Assistant Unit Operator 7.
BAT
- Boric Acid Storage Tank 8.
BIT Boron Injection Tank 9.
CAQR
- Condition Adverse To Quality Report 10.
- Centrifugal Charging Pump 11.
- Component Cooling Water 12.
CRI
- Contrel Room Isolation 13.
- Control Room Emergency Ventilation System 14.
- Containment Spray System 15.
- Containment Ventilation Isolation 16.
D/G(s)
- Diesel Generator (s) 17.
- Design Change Request 18.
- Division of Nuclear Engineering l
19.
- Emergency Core Cooling System 20.
- Engineering Change Notice 21.
- Emergency Gas Treatment System 22.
- Electromagnetic Interference 23 EQ
- Environmentally Qualified / Environmental Qualification 24.
- Essential Raw Cooling Water 25.
- Engineered Safety Feature 26.
- Flow Control Valve 27.
- Final Safety Analysis Report 28.
GOI General Operating Instruction 29.
H0
- Hold Order 30.
IMI Instrument Maintenance Instruction 31.
LCO
- Limiting Condition for Operation 32.
L0t..
- Loss Of Coolaat Accident 33.
- Main Control Room 34.
- Nuclear Safety Service 35.
- Nuclear Steam Supply System 35.
- Plant Operation Review Committee 37.
PRO
- Potential Reportable Occurrence 38.
- Reactor Coolant System 39.
- Residual Heat Removal 40.
- Radiation Monitor (RAD Monitor / RAD MON) 41.
- Refueling Water Storage Tank 42.
- Significant Condition Report 43.
S/G(s)
- Steam Generator (s) 44.
- Surveillance Instruction /or Safety Injection 45.
SMI
- Special Maintenance Instruction 46.
SOI
- System Operating Instruction 47.
- Sequoyah Nuclear Plant 48.
SR
- Surveillance Requirement 49.
SSPS
- Solid State Protection System 50.
TACF
- Temporary Alteration Control Form Technical Instruction 51.
TI Technical Specification (s) 52.
TS(s) 53.
UO
- Unit Operator 54.
WP
- Workplan 55.
- Work Request o
OFFSITE DOSE CALCULATION MANUAL CHANGES No changes were made to the Sequoyah Offsite Dose Calculation Manual (03CM) in February 1988.
Correction: A change (revision 18) to the Offsite Dose Calculation Manual was made in January 1988. A copy of that revision is found in section IV of this report, l
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4 9
e OPERATING STATISTICS (NRC REPORTS)
I i
4
=
. =
DOCKET NO. 50-327 DATE MARCH 8,1988
+
COMPLETED DY D.C.DUPREE TELEPHONE (615)870-6722 OPERATING STATUS 1.
UNIT NAME: SEQUOYAH NUCLEAR PLANT, UNIT 1
NOTES:
2.
REPORT PERIOD: FEBRUARY 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 CAPACITY (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 696.00 1440.00 58441.00
- 12. NUMBER OF HOURS REi4CTOR WAS CRITICAL O.00 0.00 24444.91
- 13. REACTOR RESERVE EHOTDOWN HOURS 0.00 0.00 0.00
- 14. HOURS GENERATOR ON-LINE O.00 0.00 23781.13
- 15. UNIT RESERVE SHUTD3WN HOURS 0.00 0.00 0.00
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 0.00 0.00 77060971.91
- 17. GROSS ELECTRICAL ENE3GY GEN. (MWH) 0.00 0.00 25976386.00
- 18. NET ELECTRICAL ENERGY GENERATED (MWH)
-9853.00
-14586.00 24839737.00'
- 19. UNIT SERVICE FACTOR O.00 0.00 40.69
- 20. UNIT AVAILABILITY FACTOR O.00 0.00 40.69
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 0.00 0.00 37.02
- 22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 0.00 37.02
- 23. UNIT FORCED OUTAGE RATE 100.00 100.00 51.95
- 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:
STARTUP IS UNDETERMINED AT THIS TIME PENDING DESIGN CONTROL REVIEW, i
CONFIGURATION CONTROL UPDATING, AND RESOLUTION OF SIGNIFICANT EMPLOYEE CONCERNS.
NOTE THAT THE THE YR. -TO-DATE AND CUMULATIVE VALUES HAVE BEEN UPDATED.
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SEQUOYAH NUCLEAR PLANT AVERAGE DAILY POWER LEVEL DOCKET NO. : 50-327 UNIT : ONE DATE : MARCH 05,1988 COMPLETED BY : D.C.DUPREE TELEPHONE : (615)870-6722 MONTH: FEBRUARY 1988 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)
DAY (MWe Net) 01 0
16 0
02 0
17 0
03 0
18 0
04 0
19 0
05 0
20 0
06 0
21 0'
07 0
22 0
08 0
23 0
09 0
24 0
l l
10 0
25 0
11 0
26 0
12 0
27 0
13 0
28 0
14 0
29 0
15 0
i l
l
UNIT SIIUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-327 UNIT NAME sequoyah one
.DATE March 2, 1988 COMPLETED BY D.
C.
Dupree a
l REPORT MONTl! FEBRUARY 1988 TELEPNONE (615) 870-6722 l
nu a
0 c
gg oyu Licensee gg g n, Cause & Corrective e*,
No.
Date o
gg ggg Event ug g]
Action to o
y 3o y
5uM Report #
go gO Prevent Recurrence sv e
e.c c v
o
- c u> **,n 1
880101 F
696 F
4 Design Control, Configuration Updating, and Employee Concerns.
2 1F:
Forceo Reason:
3Hethod:
4 Exhibit G-Instructions S: Scheduled A-Equipment Failure (Explain) 1-!!anual for Preparation of Data B-!!aintenance or Test 2-!!anual Scram.
Entry Sheets for Licensee C-Refueling 3-Automatic Scram.
Event Report (LER) File D-Regulatory Restriction 4-Cont. of Existing (NUREG-0161)
E-Operator Training & Li, cense Examination Outage F-Administrative 5-Reduction G-Operationr1 Error (Explain) 9-Other SExhibit I-Same Source II-Other (Explain) r e
UM9fiNMYM trlLWFWY DOCKET NO. 50-328 DATE MARCH 8,1988 COMPLETED BY D.C.DUPREE TELEPHONE (615)S70-6722 OPERATING STATUS 1.
UNIT NAME: SEQUOYAH NUCLEAR PLANT, UNIT 2
NOTES:
2.
REPORT PERIOD: FEBRUARY 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 DEPENDADLE CAPACITY (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, CIVE 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 696.00 1440.00 50401.00
- 12. NUMBER OF HOURS REACTOR WAS CRITICAL O.00 0.00 21984.54
- 13. REACTOR RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 14. HOURS GENERATOR ON-LINE O.00 0.00 21494.22
- 15. UNIT RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 0.00 0.00 69127977.22
- 17. GROSS ELECTRICAL ENERGY GEN. (MWH) 0.00 0.00 23536780.00
- 18. NET ELECTRICAL ENERGY GENERATED (MWH)
-13176.00
-20594.00 22487551.60
- 19. UNIT SERVICE FACTOR O.00 0.00 42.e5
- 20. UNIT AVAILABILITY FACTOR O.00 0.00 42.65
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 0.00 0.00 38.87
- 22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 0.00 38.87 1
- 23. UNIT FORCED OUTAGE RATE 100.00 100.00 52.76
- 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:
THE RESTART OF UNIT-2 IS MARCH 1988.
l l
NOTE THAT THE THE YR. -TO-DATE AND CUMULATIVE VALUES HAVE BEEN UPDATED.
l l
l
s SEQUOYAH NUCLEAR PLANT AVERAGE DAILY POWER LEVEL DOCKET NO. : 50-328 UNIT : TWO DATE : MARCH 05,1988 COMPLETED BY : D.C.DUPREE TELEPHONE : (615)870-6722 MONTH: FEBRUARY 1988 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)
DAY (MWe Net) 01 0'
16 0
02 0
17 0
03 0
18 0
04 0
19 0
05 0
20 0
06 0
21 0
07 0
22 0
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
i 15 0
l.
(_.
i UNIT SlIUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-328 UNIT NAME Sequoyah Two DATE March 2, 1988 COMPLETED BY D.
C.
Dupree i
REPORT !!0 NTH FEBRUARY 1988 TELEPHONE (615) 870-6722 i
i l
a w
O c
cg
$o Licensee go,
- n, Cause & Corrective eac y
jg ytm Event ug gj Action to No.
Date o
o oo
.c u e Report #
go gu Prevent Recurrence y
g s-a:
o.c c o
o 2: m 's n
1 880101 F
696 F
4 Design Control, Configuration Updating, and Employee Concerns.
I5
)
i 1F:
Forced Reason:
3 Method:
Exhibit G-Instructions 2
4
.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 i
D-Regulatory Restriction 4-Cont. of Existing (NUREG-0161)
E-operator Training & License Examination Outage F-Administrative 5-Reduction S
G-Operational Error (Explain) 9-Ott -
Exhibit I-Same Source l
11-Other (Explain) 4 e
j
O 9
OPERATING STATISTICS (TVA REPORTS) s 4
w
l TVA M83A (DNP+84 NUCLEAR PLANT OPERATING STATISTICS SEQUOYAH NUCLEAR Plant Period Hours 696 Month FEBRUARY 19 88 l
Item No.
Unit No.
UNIT ONE UNIT TWC PLANT I
Average Hourly Gross t.oad. kW 0
0 0
2 Maximum Hour Net Generation. P4Wh 0
0 0
3 Core Thermal Eneroy Gen, GWD (t)2 0
0 0
4 Steam Gen. Thermal Energy Gen., GWD (t)2 0
0 0
g S
Gross Electrical Gen., MWh 0
0 0
j 6
Station Use MWh 9,853 13,176 23,029 E
7 Net Electrical Gen., MWh
-9,853
-13,176
-23,029 U
8 Station Use, Percent N/A N/A N/A 9
Accum. Core Avg. Exposure, MWD / Ton 1 0
0 0-10 CTEG This Month 1008TU 0
0 0
11 SGTEG This Month,106 BTU 0
0 0
12 13 Hours Reactor Was Critical 0.0 0.0 0.0 14 Unit Use, Hours Min.
0:00 0:00 0:00 15 Capacity Factor, Percent 0.0 0.0 0.0~
f 16 Turbine Avail. Factor, Percent 0.0 0.0 0.0 17 Generator Avail. Factor. Percent 0.0 0.0 0.0
.a C
18 Turbooen. Avail. Factor. Percent 0.0 0.0 0.0 19 Reactor Avait. Factor. Percent 0.0 0.0 0.0 0
20 Unit Avail. Factor. Percent 0.0 0.0 0.0 21 Turbine Startuos 0
0 0
22 Reactor Cold Startuos 0
0 0
2?
?
24 Gross Heat Rate,8tu/kWh N/A N/A N/A j
25 Net Heat Rate, Blu/kWh N/A N/A N/A
- E 26 D
27 y
28 Throttle Pressure, psig N/A N/A N/A J
d" 29 Throttle Temperature, *F N/A N/A N/A y
30 Exhaust Pressure, InHg Abs.
N/A N/A N/A E
31 Intake Water Temp., *F N/A N/A N/A F
32 33 Main Feedwater, M lb/hr N/A N/A N/A f
34 C
35 36 37 Full Power Capacity, EFPD 404.86 363.b5 706.51 38 Accum. Cycle Full Power Days, EFPD 0.0 210.8416 210.8416 j
39 Oil Fired for Generation, Gallons 2,310 2
40 Oil Heatino value. 8tu/ Gas.
138,000 41 D+tel Generation. MWh 35-Max. Hour Net Gen.
l Max. Day Net Gen.
Load M Wh Time Date l MWh Date Factor. %
2 43 N/A N/A N/A i
N/A N/A N/A S
_ R emark s: IFor BFNP this value is MWD /STU and for SQNP and WBNP this value is MWD /MTU.
j 2(t) indicates Thermal Energy.
2 MAR 141988 C/
Date Submitted Oate Revised
/.
? Plant Superintendent
SEQUOYAH ONE REACTOR HISTOGRAM 110 100 -
90 -
80 -
70 -
eA 60 -
3 50 -
2 40 -
0 30 -
20 -
10 -
0 0
0 l
-10 i
i i
i i
i i
i i
0 4
8 12 16 20 24 28 FEBRUARY 1988 SEQUOYAH TWO REACTOR HISTOGRAM 100 90 -
80 -
70 -
O 60 -
t f
50 -
2 40 -
8 30 -
g 5
a 20 -
10 -
2 0
-10 i
i i
i i
i i
i i
i 0
4 8
12 16 20 24 28 FEBRUARY 1988 iVQ 6%0e' (PP 3 76)
UNIT OUTAGE AND AVAILABILITY h
SEQUOYAH Nuclear Plant Lue v.. <! Itca4 tew Power 3.4II MW(th) gr,,t No.. ONE Genas it, a Rat ng,1220.6 MW(c)
Month / Year FEBRUARY 1988
- 0...g.. i.tou ( la te ical Ratmg _l l81 MW i
Period Houss 696 T.rne on.t Awa.14tWe Time Not Available EH F
UNIT CORRECTIVE ACTION t.m.
T.m.
SHU T T ING STATUS OUTAGE CAUSE TAKEN TO PREVENT rote cen.
Not Uwit T ue h.nc Gen.
Rea< tor Unit Out en DOWN DURING RE PE T ITION oav ms l e.o ms ' u.n.ms un 3
aslun as ' u.o a s ' u.n ms ' u.n. g u.n si,5 I u.,
REACTOR OUTAGE I
00 eon 00,00 l
24 800 24 '00 24.00 24a00 Design Control, Configuration N/A Mode 5 g-~
2]p 2 { 00 2Q 24 '00 24 8 00 00 00 iOO t
t]pdating,, and Employee 3
nq 99_ 00A qq_._, e _.24 10_0 2._4_100 24 300 24 00 i
_1_
Concerns
- 3 "Q_l00_ OQ_100_
24 i00 2 Q Q0_
2 00 24 8 00 e
8
_g__.
5 00 lDQ_..QQ_!QQ_.__t._.
. 24 ' 00.. Z.4.400 24 i00 g 00 8
T4], UO '
6 00 800 00 100 8
3 24~ OO 24 +00 24 '00 24,00 QQ_fQQ_
00 00 C
7 Mk 24 l00 24800 s
8 M
00 00 0[00_.
l.
.' b 0Q_ QIL 1QQ_
's __ 24 s00 24 r00 24 10 0 24l00 e
4 MQO 24 8 00 24 '00 24 s O_0_
i i
y 10 00200_ 00 800 1
24 LOO _MOO 24__j00 24 9 0 r
[-
- M00 11 00 iOO 00 oo 24 '00 24 iOO 24 '00 24 i 00 i
12 00 800 00 Ioo 1-24 '00 24 [00 24400 e
i 24 iOO' 2M00 24
!3 no ' 00 00 t00 iO 24!00 O
8 s
14, -q_.j QQ_ QQ_ Q[~__
24dQ0 24j QQ_
QQ_._24 OQ
!), 00200 00 00 24 s00 24 '00_ 24.00 24 1 00 i
i
. BdO_. 2M00 24 00 24100 e
i 16 30 e00 00 00 e
8 l
I7, 00 800 00 00_
24go...24 000, 2d00_ 24!00 g
3 18 00' 00 00 100 24 *00
] 00~ N.]t {
- ]E 24 OO 24 10 0 24l00 1
i l ') h(00_ 00 600
,24j00 24 i 00
?O, 00 00 00 IOO
._g_
24 8
24 '00
- 5. 700_ 24 '00 241 00 4
00{00._{_ 24 [00 24 l00 2 LOO 241 00 j]_. 00g00 00 100 00 100 24 t00 24 TOO 24 n00 24!00 22 ~' OP 0000 100 t
24-IIiO' ~2NOT 2P00 24 ; 00 e
3 i
?)
24 O_ii 00 06~~I,00 7 4 70O' '24 800 24 IOO 24g00 3
I 3
25 00 00 00 00 t
24 800 2 '400 24 50 0 24 3 00 g
3 2(,
00 00 C00
, 24 ' 00 24_200 24 '00 244 00 00 t 8
2 OO l
24 100 24 e00 24 iOO 241 00
.. _ _ i 27_. e0 e 00 e
i 28 OG 100 00 t00 i
24 ' 00 24 '00 24 ;00 24 ; 00 8
29 00]I( 0 00_
24 e00 MOO 24 '00 24 g l
8 3h l
H l
T ot al 00,00 00 {00 696s00 h96 '00 h96 iOO 696' 00
,%,X NM
' l--
J_
TX 32<
21 N JC 8
8 6
Y va 060C (0174 UNIT OUTAGE AND AVAILABILITY S EQUOYAlf Nuclear Plant Lxenwd ikxtor Power 3.411 MW(th)
Unit No.
WO Genceato. Ratir3gl220.6 Mw(c)
Mon t h/Ye.ir FEBRUARY 1988 Deugn G ou Electrical Rating. I.183 MW Per.od Hours 696 T.me Unit A.a.satise T.me Not Avadable H
F UNIT CORRECTIVE ACTIOt4 T.m.
T m.
SHU " ' N b^
S OUTAGE CAUSE TAKEN TO PREVENT f ut.si Gen.
Not Uwd T w ts.ne Gen.
Heu tor Unit Out in DOWN DURING R EPE TITION 04v eus ' u.a 6 % ' u.n e m ' u.n ti.s ' u.a
- s ' u.n tus ' u.n tm ' u.n 6*s f u.n ei.s ' u.n REACTOR OUTAGE g
1,DD e6 F0{0_C.
24 eDO-7 0~2DF7 de o
i Design control Configuration N/A Mode 5 2
2 00_ 00_ 00_t0Q_
i
. 24jo0_. 24 l00 24 '00 24 ' 00 8
Updating. and Employee i..00a 00_ 00 '0R_
4 00a 00_ 00__iOQ__ _!_._..
24_t00_ 24_100 24 300 24 ; 00 1_
concerns 14_j00 24_J 00_. 24_'00 14_
24'00_
i i
.J4d, OQ.
'a. 00_'00_ 00._t00_._L__
2.4 '90_ 24_4. QQ_
a 8
10 0 _
6
.0 QQ_ QQ 1
24_ 00 24 00 24 '00 8
.0Q_100 24 00_
Mode 4 7
Q0_. M_ 24m2q_.2dD0_. 24 l00_
24'00 00-,00 8
00_400 00 00
_24 '00 24 s00 24 10 0 24l00 i
3 i
9 24_1Q2_.24 '00 24 'QO 24 00 I
[
00_400_ 0Q_100 10 y
0020Q_ 00 '00 _ {__
24_1_00_ 24 l 00_ 2_4__iOO 24 3 00 e
11 00 100 00 00 24 800 24 iOO 24 t00
_lL 0Q) 09_ QO 00_ ___1
.[2_4)0] 24 ' 0('24
'00~
24100 i
4 24 i00
.L_
i 2 ( 00 _
13 90 00 00 100__._j_
24 '00 24 iOO 24I00 8
i
_lLo0400_ 00 '00_
24 Ino_.24_100_ 2q00_ l G_00 00_,je0_
e J 1..90_100_
00_
i 24 *00 24 '00 24 00 24i00 i
i
_24 l 00_ 24I00 24 '00 24t00 e
16 20_ Qq_ qQ 17 20 800__ 00 00 1
24200_
24 800_. 24_[00 24 ! 00 3
IM OQ 00...Q_QaQ0_ __1 24j 00 24]OO 24 l00 24 d0 a
_k_
19 90y 0q_..p0qiqq_._j_
J4a 00 24_i00__ 24 iOO 24 t 00 0_,00_ ___' _ 24lpo 24joo
?o. no 4 00_
24 '00 24t00 i
21 90 !00_
QO._ j o0_..
a__
24_100 24.00 24 00 24100 1
24]OO 22 00 100 00 100 i
24 800 24 s00 24 00 i
'2 { 00 R. 0Q 00 00 100 24300 24 {00 24,00 _ e 8
24.S2400 00 00__. __}_
24 e00 24_]800 24 100 24 i 00 4
25 00 00 00_ 100 a
24 '00 24 OO 24 00 24i 00 1
i
?6 iUo0_
24[00 M n00-~ I '00 2E00 r00 8
.00_100 2
g0_.
24 '00 24 *00 24 00 2 D 00.
i 27 sq_e pL t
i Mode 3 28 00 800 00 sc0 e
24 ' 00 24 '00 24 ;00 24 ; 00 6
8 j_
29 00 '00_.ogo0_ __l__.Es 00_,24,00~24 00 24i00'
-~
_ _J
]
l
_ _j_ q_
_4_
.__t__.
intai 00 ! OO 00 '00 5
696 00 696 '00 s96,00 696 l 00 x i ><..xfx rm
. c z:
xz
_X
SUMMAgy 0F MAINTENANCE ACTIVITIES
e D
MAINTENANCE
SUMMARY
(ELECTRICAL) l l
l l
G.EGRICAL MAINTENANCE MONill.Y 8CMARY COMP MR2.... U FCMC SY8 ADORESS. DATE.... DEsCRIPT?CN.......................... CO RR 021543d 2 ;$V 047 003511/08/87 2-TSV-047-0035-A.DNPRDm], VALVE WILL HOT REPLACED THE CLOSE LIMIT SWITCH fuMCTIDH PRC?ERLY. INVESTICATE MD (2-23-047-0085-1A) AHD ADJUSTED THE REPAIR LIMIT SWITCH ARM FOR PRUPEP LIGHT INDICATION WR B215436 C2742311 GEND 082 000107/27/871 GENS 082-0001-A,1NVEST THE ROOT CAUSE "C" PHASE SATURAftE REACTOR TRMSFORER Of TFE FULLCWDA RELAYS OPERATICH THAT (T52) HA0 A GROUND."A" PHASE SA TRIFFED TEE D/G 59N 50 MD 87 TRAN3f0RMER (T51) AND "C" PHA!E SATURACLE TRMSflRMER (T53) FAILED TE MNT. INSTRUCTION 10.39 LEAKAGE CURRENT TEST. (T51 AT 13 KILOVOLTS AND T53 AT 11 KILOVOLTS). REPLACED 4ATilRt.8tE REACTOR TRMSFORMERS T51, T52, MD T53 RETESTED TRAN3 FIRER $ MD RETERMINATED CABLES.
WRt B-276251.
027dM 1 GEND 082 00018 09/01/871-GERB-082-00010 0,E8HFRDu],THIS eLAKM ADJUSTED TMDEM EXCITER PDTENTIUMETER TO CCNES IN WHEN D/G IS I: LING.
EDUALIZE YOLTAGE. MNUALLY LITTED THE ACTUATOR LEVER UP TO CLEAR DIFTERENCE ALARM. RETURNED LEVER TO IDLE POSITION PROPER INDICATION WR B27dM B276599 2 INVD 250 GP 12/05/87 2-INVB-25047-E,CEFRDMJ,PER SI-3 THE CALIBRATED FREQUENCY TO 40H2 AND LIMIT SHDL'LD LE BETWEEN 40.6 HZ $ 59.3 REINSTALLED METER WR C274599 HZ; INVESTIGATE & RPAIR.
B279474 2 IS 001 0181 101/31/88 2-Z3-001-0181 f,CTUATOR WAS DUT Uf ADJUSTMENT DUE TO 1-A,Ex10CTRS0.498,xNPRDn],VLV SH3 LING LODSE STEM NUT. STEM ROTATICH PROBLEM RED MD GEEEN LIGHT WHEN"0PENED MD GENERIC TO THIS TYPE MSONEILAN VALVE.
CLDSED.
TURNED ACTUATOR CH YALVE TO ALIGN WITH LIMIT SWITCH DPERATOR ARM. WR B292232 WAS WRITTEN TO TIGHTEN SIEM NUT DN VALVE TO PREVENT STEM FROM TURN 0261311 0 003 250 KH 11/24/87 0-0CC-25bKH-C,ExFMDx],THE MHtHCIATOR GRCUND INDICATOR METER LAMP WAS BAD DUE 2-XA-55-1C24 IS IHLISTED. INVESTICATE TO NDRMAL WEAR. REPLACED I2ICATIN AND TMPJTi FPIM THE DATIERY LDARD.
IN GRCUND METER MD CALIBRATED THE DETERMTHED MD I!CLATE CAUSE Of ILARM GRCUND INDICATING SWITCH. REINSTALLED THE METER MD RELAY MD CHECKED FOR PRCPER INDICATION. WR B281311.
0204641 f/:C 032 10101/22/881-EEG-032-101,ADJJST THE FRICTION CRAKE ADJUSTED THE MOTOR BRAKE SETTD% TO 45 TO SLCW THE ENGINE SIART 10 H0 SECCND!. LETT THE SET POINT AT RFNS/3CJJST CRAKE TC2 0.S. AMP MOTGR APPRDXIMTELY 9.45 3ECONDS 10 SECONDS 0FERATI'B CURRENT DRAWING LEf3 THAN OR EGUAL TO 2 MPS Ema91 D ? N2 (001M 01/04/E8 *.-ENS-0324001Al,LxWP204]JDJJST THE ERRATIC START TIME DUE TO SPEED SETTING FRICTIDH (2ffE TO SIDW Ilit D:CIFE STf.RT h0TCR ORAKE LEING DUT Cf A0JUST TD00!bs eDJUSTED CRAKE CN THE SFEED SETTING MOTCH TO AU.UW 0-%5 RPM i;AhCE IN 10.3d
!ECONDS. WR C28 % 9.
l 02$.% 2 E 202 Kt' /X1 02/12/88 2-fRC-202-rE / 01-A,(.0F.Ua),C f HA!E THE 84-318 RELAY 1HRU 00LT! (TH0$E FRCH
+
DVEECt9Edi FELAY PICKED UP:84-718 RELAY FRONI TO BACK) WEFE L0utE AND FOSS!!tY, i
l i
i,_-,--,.
,p
ELGTRICAL MA1HTENANCE MNill.Y SUMARY COMP
- MR2.... U TUNC SYS A DDRE83. D ATE.... DESCRIPTION.........
................ CO RK CTIVE ACTION.....................
DID NOT TRIP-HANDLE IS EENT. INVESTICATE THE CCHTACTS WEK DIR RELAY AND DETERMIME WHY RELAY DID NOT CAUSE COULD BE DETERMIED. PERf0 RED i
OPERATE.
MAINTENMCE ON 84-818 RELAY. TICNTENED 00LTS, CECKED CETACT RESISTAHOE AND CLEMED CONTACTS. RETURED E.KAKER TO SERVICE MD TRANSfERED FROM ALTERNATE TO NORMAL FEEDER. WR B-285494.
'B285840 0 C)GD'250 GK 11/02/87 0-CHGB-25MK-3,CNNPRDd], BATTERY CHARGE THE YOLTAC" DROPPED PELOW 125 FAILED SI 303 AT 150 MPS VOLTAGE DROP THE CURRENT ENT FROM 147 M BELCW 125VDC PR0f1Erl IN CURRENT LIMITING MP3. THE CURENT LIM DEVICE DUT Of ADJUSTENT. CLEMED AND i
LUMICATED TK FIIIAT PUTENTIOMETER RESISTER 93 DN TE AMPLIFIG BOARD.
ADJUSTED LIMIT Ale BODSTED TRANSfDRMER POTENTILMETER. YERITIED PROPER OPGATION Cf CHARCER. WR B285860.
B2885971 GEND 082 0001A 11/12/071-CENB-082 0001A-A,CuMPR0x], min DURIEC SPEED SETTING PDTENTIDMETER WAS PERfDRPMIE Of SI 26.2A, THE ELECTRIC DEFECTIVE, POSSIIt.Y DUE TO BURNED / PITTED GOVERH02 031A-A D/G FAILED.
CONTACT 6KPLACED POTEXTIDMETER WITN UNE TAKEN.fRDM, TE 5fH DIESEL GENERATOR ON WR B285304. SET BASE SPEED AND VERIFIED 0 TO FULL LOAD TIE WAS LESS THM 40 SECONDS. WR B288597.
B2899191 MVCP 047 0491010/19/871-MV0P-067-%910-A,C1NFR0n3 min (LU DGES THE CCNTROL FUSES WERE PLOWN POSSIBLY NOT EPEN kHEN "THR0lILE" FUSH-PUTTON IS DUE TD EXCESSIVE DPERATION O DEPRESSED.
THE PUSHEUTTON. M OPERATOR KPLACED THE FUSES MD CHECKED TUR PROPER OPERATICN.
REFEREHCE WR B289921 FCR SME TYPE Cf FAILURE ON 1-MUCP-6M91A-A. WR B289919.
B269921 1 MV0P 047
%91A 10/19/871-MWP-067-%91A-A.Ed?R0m],NIN THE VLU THE CONTROL FUSES WERE CLOWN POSSIBLY DCES NOT CPEN LHEN "TMIDTTLE" PC 13 D'JE TO EXCESSIVE CPERATIDH DR FCRCE Of DEPRES!ET THE PUSHEUTTCW. M OPERATCH REFLACED THE FUSES MD CHECKED FOR PROPER OFERATIDH Of VALVE. REFERENCE WR i:289919 FOR SAME TYTE Uf FAILURE CH 1-NDP-47-4910 A. WR B269921.
0292234 2 NV0P 072 0%101/21/88 2 N'.'0P-072-0%1-0,E:NP7DN,W10CfR;0.4943, VALUE FAILED TO DPERalf tuRING PEUTCRMMCE Of SI-1(4.6.
F293:A2 0 ECIO 201 CM /W 12/04/87 0 0CTB-201-DM /40,LMPPDs],IRf/IER FJEAKER MPTECTCR FAILED THE PRIMARY FAILED SI 275.1 Sil? 73.2.1 PPIMRY INJECTICN TEST TOR LCHG DELAY TIME.
EJECTIDH IEST FOR LUNG OELAY TIhE (0Fil) REPLACED AMPTECTOR AhD PERftaED REPL!.CE 'AMPTECTOR SI-275.1 DN THE tREAKER. WRi B-293%2.
02mN 2 it') 041 01ri.' $1M/88 2-fCM61-0192,EnW1,110CfR30 Ala],V3 LOCAL MD REM]TE STRCIE II.NES DIFFERED LUE tCCAL/REMDIE Sim IIMES DIFFERED 1.Y f.Y MORE THM 2 SECONDS. THE LIMIT SW MEEE TH4N 2 !EC. RW'8 EQ'D.
ACTUATCR WAS OUT CF ADf)SIMENT DUE TO
~
FORcAL GrERATIUN. I.DJUSTED LIMIT ! WITCH.
ELECTRICAL HAIXIERANCE MUWTHLf Sll MARY COMP M2.... U FL'HC $Y! ACCRESS. DATE.... DESCRIPTIDH........
.................. CO RREC TIVE ACTIGH.....................
ACTUATOR AND STROKE TESTED YALVE. WR B-2H320.
s m
9 m
o
- G O
e g
J D
9 1
.c
ELECTRICAL nh1NTENANCE MONTHLY SunHARY CONP MR3.... U FUNC SYS ADDRE38. DATE.... DESCRIPTIEN...................... CORREC C2515001 CEhD 032 0031810/29/871-CEXB 062-0N18-A,DNPkD13,EEPLt(E REPLACED ENGINE 1A2 40VERNDR ACTUATCR.
ACTUATCR DN EXCIEE 1 AND 2 EEINSTALLED LINKAGE AND OIL LIKES AM REPLACED DIL IN THE ACTUATOR. PERFORED SPECIAL MAINTENANCE IN STRUCTION (SHI) 0-82-d WR B251500 0235427 0 LS 018 UARIDUS 11/19/87 0-LS-018-VARICUS,E*NPRDs),HI LEWL 0-LS-018-006B/1-A. THE LEVEL SWI1CH HAD ALAEM (0-LS-18-43011) IS OPERATIHC INITIALLY DEEN INSTALLED AT AN INCORRECT BEr0RE THE LEVEL SWITCH (0-LS-18-62B11) ELEVATICN. REPLACED LEVEL S WHICH SHUTS DOWN THE PRIMRY PUMP.. REPERfDRMED SHI+18-1. WR B-285427.
THERETURE WE ARE RECEIVING NUISANCE ALARn! AD DOES NOT MEET INSTRU TIB SET POIT REQUIREENTS. THIS PRUBLEM UERITIED BY PERFDWMACi 0F SMI-0-18-1.
B292053 2 GEND 082 0002A 01/23/88 2-CENB-082-0002A-A.EuRPRDm],TP1 ELAY RELAY PICKUP TIME WAS 5.4 SECORDS.
FAILED TO HEET SI-102 E/SA TIPE SPECIFIED TOLERANCE SHOULD HAVE BEEN 4.5 ACCEPTf.NCE STEP 4.25 REFAIR AS TO 5.5 SECONDS. CALIFRATED TIME DELAY HECES$6RY. 'DN 11 FOR SI 102 E/SA EELAY. LEFT TIME DELAY SET AT 4.8 SECONDS. VERITIED READINGS WERE WITHIN TOLERANCE.' WR B-292053.
'8293023 2 r"TJP 067 04910 01/21/88 2-HVU?-067-04910-A,ENNPEDu,nP2x],2A-A THE STARTING CAPACITOR WAS FAD POSSIDLY STRAINER FLUSHOUT ULY WILL NOT CEN WHEN CUE TO ACE REPLACED THE START ITS IN 'CN' CR OPEN POSITICN. PLEASE CAPACITUR AND CYCLED THE FLUSHCUT AND INVESTIGATE $ PEPAIR.
BACKWASH UALUES. WR B293023.
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SUMMARY
(INSTRUMENTATION) l i
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INSIFUMGT EAINTEMMCE h0HTHLY SUMARY COMP MR2.... U FUNC SYS A DDRE83. DATE.... DESCRIPTION....................
RECALI!*ATED TO K! IRED TOLERANCE AND RETURNED TO SERVICE. WR8 B279684 B279953 i FI 067 0136 '02/05/881-FI-067-0136,EKNFRDu,m10CFR50.49u3,ERC W FLOW IKDICATING 3000 CPM. - WITH BOTH TK INLET, 8 OUTLIT VALVE ISOLATED.
e B285697 2 FT 072 0034 02/01/88 2 FT-072-0034-A.EueRDs n10CFR50.49a],2-CAUSE OF FAIlliRE: THE IXDICATOR WAS FI-72-34 INDICATES 1000 CPM UITH NO CF TOLERMCE LOW ON BOTH THE LOW AND FLOW. CONTAIMENT $ PRAY REQUIRED HIGH ENDS OF THE RMGE. THIS MAY HAVE OPERABLE FUR MDDE 4.
EEEN DIE TO AGING OR NORME CYCLING.
CORRECTIVE ACTION: THE INDICATOR WAS RECEIFAATED TO MSIED TulIRANCE AND RETURHED TO SCRUICE. WR8 B285697 B290949 2 LT 063 017712/28/87 2-LT-043 0177-E EnNPRDN,m10CFR50.49u3,WH THE TRMSMITTER WAS FOUND BELO ILE PERr0RMING PM-1653-063 FOUND TOLERANCE THROUCEUT THE ENTIRE RMGE.
TRM!MITTER TEST POINT YALVE TO tt DUT THIS MY HAVE BEG DE TO EQ OF TOLERAECE. RECE DRIFT. THE TRANSMITTER MS ECEIERATED.
PERFORfED PRESSURE TEST AND RESPONSE TIME TESTING. VERIFIED CAPILLARY FILL f.CCEPTABLE' AND ETU9ED TO SERVICE. WR8 B-290949...,
I:293906 2 fi 068 VARICU! 01/12/88 2-FT-048-VARICUS-F,EdN?RD*],BOTH LOOP 3 2-FT-068-00710-F MD 00718-E/iHE FAILED PMT REQUIRED OF MI-li.1.11.
TRANSMITTER WAS FOUND TO BE DUT OF TRM1MITTER! E30 NEED HEEKE UtJ.YE IN TOLERANCE LOW THROUGHOUT T BLEED PARTS REPLACID 2-FT-68 719 8 71D RANCE. THIS MY MVE CECM DL CYCLING. THE TRANSMITTER WAS RECALTCRATED MD RETURED TO SERVICE.
WRt B-293906.
P298595 2 li 043 UARICUS 01/22/88 2-LT-043-YARICUS,E5P283,RWST LT-63-46.49 00 HOT READ TEE SAME.
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4 INSTRUN0ff MINTEMMCE MONTHLY 81WLiRY CCRP
- 02.... U FUNC SYS A DDR ESS. DATE.... DE6CRIPTIUM.......................
^
REPLACED, MD CALIDRATED PER SI-75.
VERIFIED OPERATIDI AND RETURED TO SERVICE. WR4 B247928 B249232 2 LT 043 0179 09/23/87 2 LT-063-0179-C,[810CFR50.4%uMPRDu1RE THE TRMSMITTER AND PELLOWS BOTH WERE PLiCE THE OR IN THE CAPRLMY FILLED FOUND TO BE DEFECTIVE, TNIS MY HAVE SYSTEM USE CFLY WESTIXCHDUSE PROCESSED DEEN DUE TO ERUIPENT ACE OR.
WEAR. THE TRMSMITTER MD BELLOWS WEE BOTH REPLACED, 01 ADDD, MD FRL CAPS WELDED. THE TRANSMITTER WAS CALIBRATD MD QUMTIFIED MINTENMCE PERFORMED.
THE TRMSMITTER WAS RETURED TO SERVICE MD POST MAINTENANCE TESTIMC PERFORnD.
WRt B-249232.
025 N 1 2 FM 068 0334D 01/14/88 2 Pn-068 0334B-E,[uMPRDu1 READS Sant OF THE MUDITIER TRMSISTORS kTRE "10.30nA" DUT PUT WITH "10.00 MA"INPUT FOUND TD BE DEFECTIVE. THIS WAS PR (KOTE: LCW LIMIT LDCKUP IS ADJU$ FED DUT DUE TO HCRML WEAR. THE TRMS OF THE WAY)
REPLACED IN THE lEDIFIER AND CALIERATD.
THE MDDITIER WAS RETLRED TO SERVICE MD A POST MINTDfANCE TESTING PERFORE9 WRt B 257461.
C264020 2 FT 072 0034 01/21/88 2-TT-072-003'<-A,[uP2u,uRPRDs,u107R50.49 THE MODIFIER WAS FCUND TO K OUT CF x12-FI-72-34 SHCWS 800 CPn WITH NO FLCW TOLERANCE HICH IN THE LOW RMCE PRESENT.
WOULD CouSE THE INDICATCR TU READ INCCRRECTLY. THIS MY HAVE BCEN DUE TO ACIN: CR NCRML CYCLINC. THE TRMSnITTER WAS EQUALIZED AND THE MDDITIER WAS RECALIBRATED MD WAS RETURHED TO SERVICE. WRt B-2M20.
C267209 2 FR 003 0090 02/18/83 2-FR-003-0090,THE ItUE PEN DOES NOT RIDE DUUN ON THE PAPER AS IT SH0!1.0 E.279455 2 LT 063 0040 01/29/89 2-LT-063-0060,[*NP2Dx12 LI-43-82 READING (THICHER THEN 2-LI 43-60 AMD ALARM DH HICH LEVEL CANE IN AT 922 (ALMM STP + 98.%) IETSTICATE LEVEL
027x68'4 2 P! 063 62 02/01/38 2-PI-063-62,[xNFP0nLCALIFPATE I4 ACC, CAUSC DF FATLURE: THE TRARSnITTER WAS PEESS IND.
DUT OF TOLERMCE HICH THROUGHOUT THE ENTIRE RANCE. THIS MAY MVE FEEN DUE TO ACING DR HCRML CYCLINC.
CORRECTIVE ACTIOR: THE TR M$nITTER WAS i
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INSTRUNDIT MINTEHANCE MCNTHLY SINIARY COMP
' MR2.... U FUNC SYS ADDRE88. DATE.... DESCRIPTIGN............
B226395 2 LT 063 017810/01/87 2-LT-063 0178-F,EWPRDW,ul0CFR50.49u3,RE THE TRMSMITTER SCH30R BELLCWS PLACE THE FILL FLUID IN THE CAPILIARY TO BE DEFECTIVE. THI! MY HAVE CEEN DUE SYSTEM USE THE WESIINGHLTSE APPRWEDTO NORMAL MECMNICE EAR. A NEW SENSOR PROCESS REF. SMI-2 63-5 tELLOWS Wh! INSTALLED, FILLED WITH t
APPROYED DIL MD PECALIBRATED. WELCD FILL CAPS MD PERFORED CALIBRATICM DF TRANSMITTER. PERFORMED ESPUNSE TIME TEST. ETURNED BISTABLE TEST SWITCH TO HORML. WR4 B-224385.
B2451312 LCV 003 01544 01/04/88 2-LCV-003-0154A,EmP2s,nNPRDn],LCV-3-154 THE UALVE POSITIDHER MLTUNCTIONED AN A WILL SUDDENLY CLOSE ' 1/4"-38" FROM WAS DETERMING TMT IT EDG TO EE FULL OPEN CAUSING THE FLOW TO DRIP FRDM REPLACED. THIS MY HAYE
NORML DR CYCLINC EAR. THE POSITIDXER WAS REPLACD MD CALIERATD. THE UALUE WAS DBSERVED WITH PunP 2A-4 RUNNINC. THE INSTRUMENTATION SECTION CHECKED WITH CPERATICWS MD F111ND Tmi TEY HAD ConPLETED SURYEILLMCE INSTRUCTION
~ SI-144.6 $UCCESSFULLY MD SIGNED THE WORK RE90EST OFF AS,,CC5PLETED. WRt B-245131. -
g247200 2 in 048 0002C 01/28/88 2*In~468'0002E'0'[ #PN0*)'0URIEC PERFORMRCE CF SI-90.12 DVERPCHER D!l.TA T CISTABLE WAS FCUND OUT OF TOLERMCE B247355 2 !E 068 0056B 02/01/88 2-TE 068-005fB-T,EnMPkDs,W10CFR50.49u3,5 IT WAS DETERMINED THAT THE RTD CASLE HAD HIELD IS SHC2 TID TO CALINET GROUS CEEN CRUSHED CAUSING SHIELD TO CE SHCETED TO CROUND.-THE RTD SHIELD LEAD WAS LIFTED AND MEASURED FROM RfD INNER SHIELD TO CROUND. IN$tALLED RAYCEEM CN l
SHIELD WIRE TERMINAL. RTD WAS TESTED TO UERIFY PROPLR OPERATICN.
j 0247927 2 LCY 003 017101/21/88 2-LCU-00F0171-B ExP24,sMPRDn],DLRING t
PERFCRMRCE CF RT 411D FOUND UALE TIME RESPONSE 100 LCNC.
0247928 2 LCV 003 0172 01/25/28 2-LCU-003-0172-A,E*FPRDs),DURIWC CAUSE OF FAILURE: THE CCNTRCLLER WAS PERFORMNCE CF RT6110, UALVE WOULD NOT FOUND TO HAVE A DEFECTIVE SERVQ K3tCR RESPOND TO TEST SICML. PO!!ICLE LIMIT MD ALCORITHM BOARD. THIS MY HAV SWITCH PRDBLEM.
DUE TO ACING CW HORNAL l
CYCLINC. CORRECTIVE ACTICN: THE CONTRULLER'S SERVD MOTC2, ALCORIttin
[0ED, AND CABLE AS!ESLY WERE l
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-"3 INS?nMT P.AINTENANCE MilNilu.Y $latARY CGn*
C2.... U fdic $YS A0MES8. SATE.... DEEC2IPiiCS......................... CORRECTIt'E A B247184 2 L' N3 s'li112/13/87 2-ti-063 0179.C,EuXP20s,u1CCT230.49u3,00 THE TRtMITTER EAS TOUNO C'1T OF RING PERTE2n*FCE OF P511453-043, TEST TDLERANCE LOW AND POR0ER LINE LOW POINT READIEC MAS C' T Cf TOLER. N THROUCllCUT THE ENTIRE RANGE. THIS WAs J
REQ 'D.
r% ABLY DUE TO HORMAL kEAR. THE TRANSnTTTER NAS RECALIERATED AND Reil:RNED TO SEWYICE. RRt B-247134.-
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INSTRUMENT MAINTENANCE MONTHLY REPORT FOR FEBRUARY 1988 COMMON 1.
Completed corrective actions for CAQR SQP871408 concerning the preparation of calibration procedures for monitor tank pressure gauges used as compliance instruments.
2.
Completed corrective actions for CAQR SQP880080 concerning the failure to initiate a PRO when compliance instrument 0-TS-30-156 was found out of calibration.
3.
Completed corrective actions for CAQR SQQ871603 concerning the failure to perform a quarterly calibration of portal monitors.
UNIT 2 1.
Completed corrective actions for CAQR SQP880121 concerning the replacement of missing covers fro 6a the millivolt-to-current modules associated with the RCS hot leg temperature instrumentation.
2.
Completed corrective actions for CAQR SQP880091 concerning administrative errors associated with the performance of Special Test Instruction (STI) 108, 3.
Completed WP 0066A-01 to revise the diesel generator temperature recorders to provide alarms in both the main and backup control rooms.
This work was authorized by DCN X00066A and allowed the closure of SAL 1031, 4.
Prepared and completed WP 00150-01 to extend the range of the reactor coolant pump thermal barrier differential pressure transmitters. This work was authorized by DCN X00150C.
5.
Provided technical support to the Reactor Engineering Group for performance of SI-488, RCS RID cross-calibration. During performance of the SI, abnormal lead resistances were measured on several RTDs.
One wide range RTD is currently out of service.
The problem also occurred on two narrow range RTDs, but was corrected by either switching to the associated spare RTD or by switching to spare lead wires.
It is suspected that the problems are caused by bad electrical penetration splices.
Further investigation will continue.
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During the recent heatup from mode 5 to mode 4 on unit 2, the pressurizer loop seal low temperature alarm did not clear as expected.
Instrument Maintenance and DNE entered the pressurizer enclosure and determined that the alarms should be in due to the loop seal temperature and because two of the temperature switches were defective. The defective switches were replaced and the magnitude of the current to the heat trace verified to be correct. A Thermon representative verified that the heat trace rating was adequate to obtain the needed heat transfer to the loop seals.
It was then determined to trend the hestup of the loop seals as the unit heats up to determine if the alarms will clear.
If the alarms do not clear prior to mode 2, an evaluation will have to be made to determine the action required. To date the loops have heated up as the unit heated up, but not enough to clear the alarms.
DNE and the Systems Engineering Section are continuing the investigation.
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INSTRUMENT MAINTENANCE MONTHLY REPORT FOR FEBRUARY 1988 UNIT 2 7.
While Operations was attempting to maintain a constant temperature with the main steam cooldown valves, it was identified that the valves were not fully open when they should have been receiving a full open (25 percent full steam dump) signal from the controller. WR-B257426 was performed to check the calibration of the output indicators of the flow controllers, the stroke of a sample of the eight steam dump valves, and the outputs of the regulators that supply air to the various instruments that control the steam dump /cooldown valves.
WR-B263562 was then performed to verify the stroke of the cooldown valves and the operation of the pneumatic controllers that control the four banks of steam dump /cooldown valves.
The WR revealed three items of concern.
A.
2-FCV-1-103D was inoperable since it had no output while the valves were being stroked. The controller was repaired on a WR.
B.
The stroke for 2-FCV-1-103 was shorter than the 1-3/4 inch desired for full stroke. The stroke was checked on a WR and no problem was found. The problem was then found to be that 2-FCV-1-303 was leaking, thereby not allowing the cooldown valve to receive its full pressure. WR-8256432 was written to replace the controller and correct the stroke problem on the cooldown valve.
C.
The stroke of the other two cooldown valves (2-FCV-1-107, and -111) was found to be within tolerance for the 25 percent full steam dump signal. However, when the signal was increased (which would normally start opening the second bank of valves), the valves stroked another 7/16 inches.
The extra stroke on the valves is the reason for the valves being identified as not stroking fully open at a 25 percent full dump signal.
To prevent confusion in the future Operations was notified of the overstroke on February 26, 1988, such that a training letter could be issued. Also WR-8257439 was submitted to fabricate tags / scales for the valves that show the calibrated stroke of the valves (1-3/4 inch) and give an explanation of the extra stroke movement.
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MAINTENANCE
SUMMARY
(MECHANICAL) 4
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MECH /.N1 CAL hAINTENANCE MONTHLY Stat 1ARY COMP MR2.... U TUNC SYS A DCR ES8. D ATE.... D ESCRIPTIDN.......................... C B215799 2 FSV 043 0250 01/29/88 2-fSY-043-0250-A,ExNPRDu,n10CfR50.490,Y VALVE HDT SEATIHC PROPERLY CUE TO BEIN ALVE FAILED SI-153.1 LOCAL LEAK RATE DIRTY. REMOVED URVE FROM LINE, TEST. REFERENCE LEAK RATE IS 0.0225 SCfH INSTALLED TUBE CAPS ON BOT AND YALVE LEAKED 0.1057 SCfH. RWP TUBING, REMOVED EODIET SEAL WELD REQUIRED.
MACHINED URUE DISC, LAPPED SEAT MD
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REINSTALLID VALVE. WR8 B-215798.
D234133 2 LCV 003 0154 02/04/89 2-LCU-003 0154 A,CnXPRDO,UALVE DID ROT UALVE STEM NEEDED LUBRICATING CAUSED BY MEET STFCKE TIME PCR SI-166 ICING DIRTY.-DISASSEMBLED YALUE, CLEMED STEM, CAE AND STUFFINC BOX. REINSTALLED STEll A@ LUCRICATED, CAGE MD REPACKO.
REINSTALLED BONNET AND TOROUED.
REINSTALLED AIW LINES AND LIMIT SWITCHES.
02341% 2 FCU 030 0014 01/29/89 2-fCV-030-0014-A,EuP20, VALVE FAILED !! UALVED SEEMED TO l.E STUCK ON SEAT, DIE 164.1 FOR VfLVE STRCKE TIME.
TO ACTUATOR SETTINGS. REMOVED ACTUATOR FRGM VALVE, VERITIED ACTUATOR OPERAID SMOUTHLY THRDUCH ENTIRE STROKE.
REINSTALLED ACTUATOR DN YALVE AND 1DJUSTD GPERATOR STOPS. RETURNED TO SERVICE. WR 2341%
B2394241 VLY 067 0582C 01/20/S81-VLU-057-0592C-A,[xNPCC,RELIET VLV RUST MD SEALE CH SEATIMC SLTFACE CAUSD LEAXING THRD'JCH liY EXPOSLTE TO RAW WATER. DISASSEMBLED UALUE, CLEMED SEATING SURTACES, LAPPED SEAT, REA!!EMSLED MD REINSTALLED VALVE MD CHECKED RELIf7 PRESSURE. WRt B-239424.
024745] 2 PMP 062 0108 02/16/88 2-PN?-062-0103-6,Et10CTR50.49s,4FRDO,2 THE DIL PuhP PACKING CLMD itLTS HAD A-A CH/;RCING FUMP SPEED CHt.NCES DJTBCARD IACKED DUT Of CLMD ALLCWING THE PUMP TO SICE CIL LEt.K MD DIL SM04:E MD m0M LODSE SUCTION AND THE SPEED INCREASER TULL Dr DIL Ut:POR SUPPORT LAD SEtt AND BEARING TO CO CAD.- REPLACED SPEED OR LEA 211:C CH SPEED CERCER.
INCREASER, REMOVED SPEED INCREASER LLTE
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DIL COOLER HEAD AND INSTALLED 0H THE NEW SPEED INCREASER LLTE DIL COULER. PAEKID WITH CREASE.
02717951 ULV 032 0543-102 02/22/831-VLV-OS2-0543-102-B,[dP2x,nKPRDC,Yt.LVE VALVE WDEN DUT.-INSTALLED NEW VALVE.
0CDY IS LEAKINC AIR CAUSING EXCESSIVE STI.RIS Of AIR COMMESSCR.
02717961 '.iY 082 0521-10102/22/881 VLV-082-0521-101-D,[2P?u,GEDO,UALVE UALUE WORM DUT.-INST ALLED NEW VALVE.
00DY IS LEAKING AIR.
f:2852351 i;V 062 0009 02/22/88 1 fCV-062 0009,[ 4?RD C,ThTFE IW REARS DIAPHRACM IAD.-INSTALLED KEW DIAPHRAGM.
TO LE M AIR LEAK FROM 1-f CV 62-9. THE 4 -
MECMNICAL MINTENMCE NONTHLY SUMMRY
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COnP MR2.... U FUNC SYS ADDREW. DATE.... DESCRIPTION......................
YKVC MD RECULATOR ARE EXCLOSED IN A BOX-UNABLE TO DETERMINE EXACT LOCATICH W/0 ENTRY IN TO ENCLOSURE PLEASE INVESTIChiE MD REPAIR B285722 2 TCV 041 0122 02/01/88 2-fCV-041-0122,EuMPRDx2, VALVE LEARING OIAPHRAGM WQM DUT. INSTALLED FEW AIR ARCUND CFERAIDR DIAPHRACM.
DIAPHRACM MD RETURNED YALVE BACK TO SEEVICE. WR6 C-285722.
02922112 FSV %3 0307 02/01/88 2-FSV-%3-0307-8,REh0VE COMPLETE VRVE INTERNAL VALVE CLEARMCE ARE NOT FROM LINE FCR TRCintESHCOTING RE'RK AS SUFFICIENT TO ALLCW FOR R NECESSARY. REINSTALL VALVE. RWP PIQ'D.
OPERATIDN. VALVE WOULD NOT STRGKE.-CSTAIKED NEW VALVE TMT WAS REMOVED WILL BE REWORKED OH WR48-208150 INSTALLED FDL'R REED SWITCNES.FROM WATTS FAR AND INSTALLED. THE VALVE THAT WAS REF.UVED WILL EE REWORKED ON WR68208150.
INSTALLED FOUR REED SWITCHES.
B292782 2 VLV 068 0538 02/%/88 2-VLV 048-0538-S,ERHPEDm],UALVE 2-68-538 PACKING HAD BEEN ADJUSTED UNTIL THERE LEAKS W/NO PEE!!URE WAS NO MCRE ADJUSTMENT LEFF.-REPACKED vAtvE.
B295149 0 MP 047
%60 01/28/88 0-PEP-067-0460,EnNPRDM), REPLACE LCWER CAD PACKING 00X DUE. TO EXPOSU HEAD PUnP SHAFT. LEAKIEC CETWECH SLEEVE WATER / REPLACED LOWER HEA 8 SHATT.
PACKING BOX. RETUGED PUMP TO SERVICE.
(WR0295169) 0295922 2 VLV 070 067(D 11/04/87 2-YLV-070-06740-A,EDNFRDu],REF SIR DUE TO EXCESSIVE ELEARMCE IN THE HMCER 84-003 REMOVE 00ENET TER INSPECTIUN.
PND DISK, THE DISK RUCBED THE SIDE Of THE VALVE BODY IN TWO LOCATIENS.
REPLACED INTERNALS AND REINSTALLED CONNET. VERIFIED NO EXTERNAL LEAKA3E.
WRi B-298922.
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ElCCTRTCAL HAINTENAKE MONTil.Y SimnARY COMP MR2.... U TUNC SYS A DORESS. DATE.... D ESCRIPTION............................ COR B2025871 IN 077 Cold 03/18/871-FSU-077-0016 B.EGFOn],COING FRDn THE SEAL-IN RELAY WAS SHORTED CAUSING THE CLOSE TO CPEN POSITION (10WS THE NCRMAL NORML CONTROL POWER FUSE CONTROL POWER FUSES BATT BD.II CKT. A24 UALVE WENT FRCn THE CLOSE PO!!TIUN. REPLACED THE SEAL-IN RELAY WR B202587.
8209043 2 f!V M1 0029A 09/18/87 2-FSV-001-0029A-A,ExNPRDu,ul0CTR50.49u], THE 'SHAP-LOCK' LIn!T SWITCKS (UNDER FOUND 80 RC VOLT CPOUND ON VLV THE JUNCTICH BOX HOUSING THE SCLENOID VALVES) WERE GROUNDED DUE TO WATER IN THE JUNCTION BOX (2-JE0X-911-0401).
REPLACED LIMIT SMITCHES 2-23-001-M29/1 AND 2-ZS401-M2W2. REPAIRED 2-23-M1-0029A/3 AND 2-ZS-M1-M29A/5 BY REPLACING THE BOOT AND RETAININC RINGS, CAR SPRINS WASHER, CAM RETAINING RING AND TOP AND ICTT!Il CASKET KITS. REPAIRED 2-ZS-M1-002W4 BY REPLACING THE BOOT AND RETAINING RING AND THE TOP AND E.0TT05 CASKET KITS. ALSO REPLACED THE
'100T, CONTACT' BLOCK KIT AND TOP AND 00TT0n CASKET,KIIS ON 2-ZS-M1-002W4.
WR $209043.
0209825 2 NLP 074 M24 %/02/87 2-MUDP-074-0024-0,Cm10CFR50.Mu,mHPRCz], REPLACED TRIPPER ASSEniLY HYPOID CEAR, CEEASE ORCKEN DOWN.
TRIPPER CAn, LIMIT SWITCHES TORCUE SWITCH AND CA!KET AND SEAL SET. REPLACED CREASE AND PERTORMED M0f0R OPERATOR VALVE ANALYSIS TEST. WR B209825 D22809*l 2 IN C47 00% 10/11/87 2-TSV-OdM094-A,CeDa], VALVE }MS AIR THE SDLEWOID VALE COIL WAS OPEN.
LEAK IN THE CLD!ED PO*1 TION. LEAK STOPS REPLACED THE SOLENGID UALUE. INST WHEN PLACED IN DPEN POSITIGN.
C-TITTING AND SPLICED WIRES COING TO DISCOVERED DURIHC TUNCTION TEST VALVE. WR B228091.
02490931 (Cif 070 0%f 12/22/871-TCTB470-0%4-A,E<NFliDm],EREAKER IREAKER FAILED LDFC DELAY Tint FAILED L DELAY TINE SIEP ?.3.1 GF SI253, SURVEILLANCE CRITERIA. f.nPIECTOR WAS ALSD FAILED (2) EEff!TS Cf SI258.
CEFECTIVE. CAUSE UXKWN. REPLACED REPL6CE MPIECTOR MPTCTGR AND PERf0RMED SI-258. WR B-249098 F251W2 0 CHGli 250 Di 11/10/87 0-CHGB-250-cH-F,E*P7Dd3,CH'tRGER WOULD 150 OHN LOAD RESISTORS, ACQOSS NOT MEET (LDAT U Pa Il 100 TP d.10 2 CT CAPACITORS, WERC 111RNED OPEN. REPLACED 415) r03 s EES ACra$$ EAPACITORS RESISTORS 1, 3 8 5. TERMINATED RESISTCRS INCICATED 1 AND 3 f.HD ADJUSTED V0 Lit.CE DUTPUI CT CHf.RCER TO DEET SI-100 ACCEPTANCE CRITERIA. WR B-251W2.
0249 32 2 '
M1 MC.'<101/17/23 2-2340MW/1-C.LWF.itd,410CTR50.4913, THE LIMIT SWITCH ACit!ATOR WAS IENT.
A01031 UffM" LIMIT ! WITCH CR ACilfalCR CAUSE tMNCWN. CCJ' LD PCSSIt 50 TPAf /.CT!'ATCP SlHKES R3tLER CN LIMIT CENT DURING PREUICUS MAINTENAN SWITrH /,In PDJUSTED THE LIMIT SWITCH ACTUATOR AND CHECKED TOR FROPER LIGHT INDICATION Uti 0-264032.
MECMMICAL MIXTEXARCE f.ONTHLY StanARY CCMP C2.... U FUN C SYS A DDRES8 DATE.... DE3tRIPTIDH.........................
B215798 2 ftU 043 0250 01/29/8') 2-FtV-0434250-A.EuMPRDu,ul0CTR50.49m],V VALVE HOT SEATING PROPERLY DL'E T ALVE TAILED $1-158.1 LOCAL LEAK RATE DIRTY. REn0VED VALVE F7GM LINE, TEST. REFEREXCE LEAK RATE 13 0.0225 SCFH INSTALLED TUBE CAPS OH AND UALVE LEAKED 0.1057 3CTH. RWP TUBIHC, REn0VED [CNNET SEAL EELD REQUIRED.
MACHIREh VALVE DISC, LAPPED SEAT MD REIWS'ALLED VALVE. WR6 B-215798.
0234133 2 LCV 003 0154 02/04/88 2 LCV4034154-A.EnWRtu3, VALVE DID NDI V3fft. STEM HEEDED LUBRICATIFC CAUSED MEET STR3E TIr.E FLs31-164 t!IhC DIRTY.-DISA!!EnCLED UALVE, CLEt+ED
$!',1, CAGE MD STUFFING BOX. REINSTALLED STEM MD LUCRICATED, CASE AND REPACKED.
REIRSN1 LED BONNET AND TORQUED.
REIN 3htLED AIR LIKS MD LIMIT SW1TCn D2341% 2 FCU 030 0014 01/21/88 2-fCV-030 0014 A,ExP2n],t. 4 TAILED 'T
'iALVED JtiMED TO BE STUCK OH 3 EAT, DUE 146.1 FDR VtLVE STPCKE Tint.
13 ACTUATOR SETTINC3. REMOVED ACTUATCR FRGM VoltC. VERIFIED ACTUATOR OPERaiED inCDihU T,RDUCH ENTIRE STROKE.
REIW3TellE0 ACTUATOR DN VAltt AND
'3DJutilD GPERATOR STOP3. 4 TURNED TO ffRVICE. WR 234L%
B23%241 Vty 06?
0582C 01/20/881-VLV-Of7 0532C-A UNT4J. RELIEF %s Etn MD SCALE GX SEATIE SUffhCE CAUSED LEAKIWC TFFCUCH CY cPDtuRE TO RM WATER. DI!As3EMBLED VALVD M.!MED SEATING $URfACES, LAPPED IEAT. e3SEfSLED MD REINSTALLED VALVE M CHECKG RELIIT PRESSURE. WR8 D"23'1424.
0247453 2 Pl* 042 0108 02/16/88 2-!??-052-0103 A,D10CFR5). Wen)?R@3,2 ?3E OTL PunP PttKIKC CLAND FCLT3 NAD A-A CllfRCINC PP"! $Pito CHMO %CtMD FACKED CUT OF CL)ND ALLCHING THE Fun SKE OIL LDT f ND all I",0KE rWD eI:05 LODSE SUCTION A0 THE SPEED INCREA3ER ff. Of Ott. VAFOR SUPPC2f LAD SEfs. AMD (14RIT.10 CD Bt).- REPLACED ! FEED 01 EfARIC.3 $ PEED CISCtR.
1[fiEr,ttR, RitalVED STEED INCREA3ER LLEE
~
0[L Lui1r.R HEAD AND INSTALLED CN THE NEW LfEr3 IEREATER LISE DIL CCELER. PACKED
?IT4 'sFIA!L.
02717951 VW 082 0'43-102 02/22/881-VLV-0824%3112-Detw,*kPRCn],ViiVr. vAttE isiuti.IN$rALLED NEW VALVE, IC0Y 13 LEAKING AIR CAUSING EXCESSIVE Sif5TS Cf AIR CONNE!!CR.
02717H 1 VLV G82 0521-1!102/22/881 VLV492 0521-101-0,E P28,wh?RCn],Vt10E V'L','E ut? DUI. 'N!it.LLED REW utVE.
FC0Y I LEAKIEC AI2.
P255235 1 fCV 052 COM C2/22/891-fCV-02-00M,ErNPni),1HERE ArTREf.23 DItfHRA61 CAD.-IASTALLED PEW DIAPHRACn.
10 IE N AIR LEAK FRDn 1-fCV-62 9.
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MECHakl[AL MINTEMMCE h04THLY ttMMARY COMP MR2.... U f t'HC 8Y8 A DDR CES. DATE.... DESCRIP TION.........................
VALVE MD REGULATCR ARE FMCLOSED IN A BOX-UHADLE TO DETERn1XE tXACT LOCATIDH W/0 EMTRY IX TO EMCtD$tRE PLEASE INVESTICATE AND REPAIR B285722 2 FCU 061 0122 02/01/88 2-TCV-061-0122,[mMPRDn), VALVE LEAKING DIAPHRAGM WORN OUT. INSTALLED EEW.
AIR MOUN0 CPERATOR DIAPERAGM.
DIAPdRAGM AND RETURNED YALVE BACK TO SERVICE. WR6 B-285722.
B2922112 FSU 043 0307 02/01/88 2-FSU-043-0307-A REn0VE CunPLETE VALVE INTERML VALVE CLEAR FROM LINE FCR TRCUCLESH00 TING REW.RK AS SUFFICIENT TO ALLDW FOR HECESSARY. REINSTALL VALVE. RWP RE0'D.
OPERATION. VALUE WUULD NOT STROKE.-0BTAINED MEN VALUE THAT MS REMOVED WILL BE REWDRKED ON WR68-206150.
INSTALLID TOUR REED SWITCHES.FROM WATTS EAR AND INSTALLED. THE VALVE THAT WAS REMOVED WILL BE REWORKED ON WR68208150.
INSTALLED FER REED SWITCHES.
B292782 2 VLV 068 0533 02/04/89 2-VLV-068-0523-S,ExNPRDu],UALVE 2-68-538 PACKTNG HAD FEEN ADJUSTED UNTIL THER LEF.K! W/hD FPISSURE WAS NO MDRE ADJUSTMENT LIFT.-REPACKED VALVE. ~
B295169 0 pnp 067 0460 01/23/89 0-pnp-067-0460,[xXPRDn], REPLACE LOVER CAD PACKING FOX DUE TO EXPO!URE TO RM HEAD PUNP SHATT. LEAKING CETWEEN SLEEVE MTER/ REPLACED LOWER HEAD
& SH.V T.
PACKING BOX. RETURNED PUMP TO SERVICE.
(WRB295169) 0293922 2 VLV 070 0 6. 11/06/87 k-VLV-070 06768-A,E*WFRM ], PEF $[0i DUE TO EXCESSIVE CLEARMCC IN THE HMCER 84-003 RCn0VE EDHNET T02 INSPECTIDN.
AND DISK, THE DISK RLTSED THE SIDE CT THE VALVE 00DY IN TWO LOCATIDMS.
REPLACED INTERNALS AND REINSTALLED I4HNET. VERIFIED KO EXTERHAL LEAKAGE.
WRt B-298922.
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MECHANICAL MAINTENANCE MONTHLY REPORT FOR FEBRUARY 1988 COMMON 1.
Completed repair on valve 12-713 in auxiliary boiler system.
2.
Completed repair on link on CO2 supply system.
3.
Completed work on boric acid transfer system.
4.
Completed repair on "A" auxiliary boiler.
5.
Sepported Operations in installation and removal of test flanges.
6.
Supported Operations in completion of various sis.
UNIT 2 1.
Compitted repair on AFW level control valve (LCV-3-173).
2.
Completed insulation repair on S/G No. 4.
3.
Complet3d replacement of HEPA filters in postaccident sampling system.
4.
Completed SI-106, -107, and -108 on ice condenser.
5.
Completed element replacement on terry turbine.
6.
Completed repair on terry turbine governor valve.
7.
Supported restart and test group in shim removal, STI-62.
8.
Supported Operations in installation and removal of test flanges.
9.
Supported Operations in completion of various sis.
10.
Completed modification on RHR manways.
OTHER 1.
Continued closure of various CAQRs, CARS, DRs, etc.
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MAINTENANCE
SUMMARY
(MODIFICATIONS) l I
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SUMMARY
OF WORK COMPLETED MODIFICATIONS - CURRENT STATUS FEBRUARY 1988 l
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Major Capital Projects:
I PN7101:
ECN 6388 - 500-kVA Switchyard Current Transformer Heaters i
Workplan (WP) 12223 is in progress pending safe access to current transformers.
PN7102:
ECN 5938 - Replace Feedwater Heaters 3 and 4 i
No work in progress at this time.
PN7105:
ECN 5009 - Essential Raw Cooling Water (ERCW) Piping Changeout From Carbon Steel to Stainless Steel Insulation of new stainless steel piping has been completed.
No additional pipe replacement is scheduled in the near future.
PN7108:
ECN 6720 - Crane Consistency Program Unit 2 polar crane modification is complete.
Postmodification testing (PMT) by Electrical Maintenance is not complete.
Unit 1 polar crane work started on July 6, 1987 and is approximately 98 percent complete.
Painting of blocks and limit switch weights remain. 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 is stopped pending resolution of budgetary problems.
PN7115:
ECN 6719 - Volumetric Intrusion Detection System ECN 6719 is still on hold.
Security is dissatisfied with system operation.
Security and DNE have not concluded discussions on resolution.
Design is working on making the system functional.
Lighting is not finished. Workplan change for drainage is being written. Field Change Request (FCR) 6645 for DNE changes is approved. Work is stopped pending resolution of budgetary problems.
PN7122:
DCR 1373 Secondary Side Steam Generator Preservation ECN 5657 - Installation of Moisture Separator Reheater (MSR)
Drain Valves Work is essentially complete on all the ECNs. Some minor insulation installation activities are in progress as resources permit..-
Major Capital Projects (cont.):
ECN 5712 - Evacuation Alarms - Auxiliary Building - WP 12663 Complete.
ECN 5841 - Hot Shop Fire Protection / Evacuation Alarm -
WP 12637 All fieldwork for evacuation alarm complete. Awaiting Work Request (WR) B240406 to be worked to restart fans to do functional test on fire protection.
PN7123:
ECNs 5938. 6305, 6571 - Replace Feedwater Heaters 1 and 2.
Units 1 and 2 and Eroded Pipe Modifications Group B is supporting PMT on heaters 1 and 2 on both units as requested.
Insulation work on unit 1 continues.
PN7130: DCR 1156 - Post Accident Monitoring This work has slipped to allow work to continue on priority unit 2 restart efforts. NRC commitment on NUREG 1.97 called for unit 1 completion by September 1987 (this has been moved to 1988) and unit 2 by U2C4. A request has been sent to the NRC to move the unit 2 completion date to U2C4.
PN7132: DCR 2348 - Sewage Treatment Facility and Civil Upgrade New DCR for onsite sewage treatment facility upgrade.
City of Soddy-Daisy will be responsible for interf ace work on site, lift. station, and pipeline between Sequoyah Nuclear Plant and Highway 27.
PN7136:
ECN 6259 - MSR Tube Bundle Replacement Complete except for PMT and inservice leak test.
Leak checks will be performed during system heatup.
PN7136:
ECN 6808 External Conduit Sealing Complete.
PN7161:
ECN 5855 - Replacement of Doors A56 and A57 Installation of door A56 is complete.
Security of A56 to be worked later. Work on door A57 and interlocks will begin later. Workplan 11654 is in work. \\
Major Capital Projects (cont.):
PN7181: DCR 1898 - ECNs 6832 and 6596 - Dry Active Waste (DAW)
Building Electrical interface work is complete. Workplan closure held for Instrument Maintenance and Electrical Maintenance checks and update of SOIs by Operations.
PN7199: Miscellaneous Activities Under $100.000 This is for various work orders prepared for work under
$100,000 total site cost. This work is done as manpower resources are available that will not impact unit 2 restart effort.
Significant Items:
The following major significant items are under Operation and Maintenance Expense (0&M) and will continue to be until restart of unit 2.
1.
Fuse Program Work ECN 5880 - Work complete for class IE unit 1 and 2 and non-IE fuses associated with unit 2 restart.
Remaining work is non-class IE unit I and non-class IE unit 2.
WP is being written for non-IE checklist for unit 1.
ECN 6854 - Complete.
2.
Cable Ampacity Program Unit 1 is in progress and nearing completion.
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Other Items:
ECN 5111 - Provide Permanent Power to Manholes 42-46 Electrical conduit and wiring installation for manholes 44-46 is complete according to WP 12262. Manholes 42 and 43 are being held ~for information from DNE on power cable routing and terminations.
FCR 4572R1 was sent to DNE and was logged in, but it was not sent to the responsible engineer. Manholes 42-46 will be worked by FCR 4577RI.
This item is not considered unit 2 restart by DNE, and they will not authorize the engineer to do the necessary work to obtain pull cards or drawings for manholes 42-46.
FCR 4572R2 will be written to supersede 4572R1 for conduit and calle routing for manholes 42-46.
WP 12276 is 100 percent complete. Workplan closure in process.
ECN 5412 - ERCW Nuisance Alarms WP 5412 Fieldwork complete. Workplan in closure cycle.
ECN 5435 - Fire Doors The installation of weather stripping continues as resources permit.
ECN 5503 - Evacuation Alarms O&PS/ Fire Detection O&PS WP 12482 is held for manpower (0&PS fire detection).
WP 12664 - Work is in progress and is approximately 95 percent complete.
Remainder restrained by CAQR SQP 871616 (evacuation alarm O&PS).
ECN 5552 - Condensate Demineralizer Modifications and High Crud Filter l
Upgrade to higher range instrumentation for condensate demineralizer system neutralization and nonreclaimable waste pumps. A mode 2 restart item.
WP 5552 In work.
Functional test in progress.
l ECN 5599 - Conduit Supports for Fifth Vital Battery and Waste Disposal Area Complete.
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Other Items (cont.):
SCN 5609 - Air Compressors Makeup Water Treatment Plant Installation and checkout is complete. Work is complete. WP 11987 is field complete.
Obtainin; all documentation for closure of workplan.
ECN 5609 - Evacuation Alarm / Fire Detection Valve 26-290 WP 12387 is in work and is 90 percent complete. Workplan change is needed to allow local testing of alarm horn.
ECN 5609 - Alteration to the Makeup Water Treatment Plant WP 12576 - Work is in process and approximately 90 percent complete.
WP 12633 - Work is in process and approximately 50 percent conflete.
WP is being held for material on WP 12665 (Mechanical).
WP 12731 - WP is approved and being held for material procured on purchase requests 74262A and 74321A.
WP 12684 - WP in work, approximately 60 percent complete.
WP 12665 - WP is being worked, and is 20 percent complete.
WP 12682 - WP is being worked, and is 20 percent complete.
DCN 113 - WP 113 Workplan in work. WP 113 Workplan in work.
DCN 66A - WP 0066A Complete.
ECN 5626 - Containment Ladders, Unit 1 A workplan is being written to incorporate this ECN.
Modifications needs additional design information to complete. DNE needs to issue all drawings listed on this ECN. Work has not begun because of this holdup.
STI - 105 - Complete.
ECN 5726 - CAQR SQP870478 - Reroute Sense Line Piping for Train Separation to Panels 2-L-58 and 2-L-68 Fieldwork is complete.
In-service leak check will be performed during heatup.
Other Items (cont.):
ECN 5754 - ERCW Nuisance Alarms WP 5754 Fieldwork complete. Workplan in closure cycle.
ECN 5841 - Hot Shop Fire Detection WP 12360 is 95 percent complete. Material ordered for PNL 0-L-633.
ECN 5935 - Correct Power Block Lighting Deficiencies WP 12437 has been approved and is in work. This is a noncestart item. WP 12275 is being written.
Modifications needs DNE to provide light mounting detail for raw water storage tanks and estimate design date after unit 2 restart. WP 5935-01 has been written to install security grills and gratings.
ECN 6005 - Replacement or AFW Steam Ejector The workplan is complete except for the leak check to be done by SI-130.1 in mode 3.
ECN 6057 - Cable Tray Covers This activity is in progress for the unit 2 restart.
ECN 6082 - Vent for Boric Acid Tank (BAT)
Installation of vent header and supports, vacuum relief valves for three boric acid tanks, and the new HEPA filter is in progress. Work is scheduled to be completed by mode 2 for unit 2.
ECN 6185 - Offsite Paging System Functional tests of the remote mobile units remain to be completed.
ECN 6196 - Pressurizer Hangers and Valves PMT is scheduled for unit i restart. Remaining unit 2 work is scheduled for U2C3 refueling outage.
I ECN 6205 - Replacement of Instrument Loop Power Supply Fuses Unit 2 and unit 0 work is complete except for spares.
Fuses are not on order. United Engineering has not notified materials group about fuse type (100 milliampere). WP 12447 has been turned in to Document Control for partial completion.
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Other Items (cont.):
ECN 6357 - Essential Raw Cooling Water (ERCW) Roof Access and Rails for Security Equipment WP 12238 is in work.
ECN 6380'- Replacement of Barton Pressure Transmitters Unit 2 fieldwork is complete. Final closure awaiting instruction revision by Instrument Maintenance.
ECN 6388 - Hydrogen Monitors in Switchyard Workplan 12223 - Craft support as needed for installation of hydrogen analyzers at 500-kV switchyard (50 percent work complete).
ECNs 6402 (Unit 1) and 6439 (Units 1 and 2) - Pressurizer Instrumentation Relocation Inservice tests will be performed when the system is brought up to temperature and pressure.
ECN 6429 - Component Cooling Heat Exchanger B Replacement DNE to procure piping, hanger, and PHE frame material. DNE to issue hanger drawings. Workplans are in work to fabricate and weld piping and fabricate and install new plate heat exchanger.
ECN 6437 - Replace FE-2-256 and 257 Complete.
j ECN 6455 - Upgrade CU-3 Box Battery Packs WP 12295 has been issued. Modifications are complete for all CU-3 boxes.
Site Security still needs to perform some PMT before the workplan can be closed out.
ECNs 6491 and 6534 - Essential Raw Cooling Water (ERCW) Supports All restart work is complete. A workplan to remove previously j
installed piping was approved and will be worked as esources permit.
il ECN 6543 - Install Public Safety Access Portals and Modify Entrance Road Work is being held pending the release of drawings from DNE.
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Other Items (cont.):
ECN 6557 - Installation of CO2 Abort Switches Complete.
ECN 6601 - Removal of Unit 1 Emergency Gas Treatment System (EGTS)
Backdraft Dampers PMT remains to be completed by the Mechanical Test Section. Fieldwork is complete.
ECN 6610 - Modify Air Return Fan Supports i
Unit I work is incomplete.
ECN 6631 - Modify Snubbers Unit 2 work is complete.
ECN 6689 - Relocation of Main Steam Power Operated Relief Valves (PORV)
All work is complete for unit 2.
Work on unit I has started.
ECN 6698 - Repull 120-Volt Cables Unit 2 is complete. Unit 1 is 90 percent complete and held by ECN 6742. This ECN should be completed prior to unit I restart.
ECN 6706 14 Support Enhancement / Lost Calculations Repairs continue on unit 1 as resources allow. This project has been combined with the calculation regeneration project for unit 1 and work has started with one workplan being prepared and placed in the approval cycle. Unit 2 work is complete and workplans are being closed.
ECN 6739 - Alternate Analysis Unit 1 work has started and 50 of 83 hangers are complete. Work has not started on the maintenance items.
ECN 6742 - Install Fuses in Radiation Monitor Power Supply Circuits The workplan is complete for unit 2.
Unit 1 is 50 percent complete.
Instrument Maintenance cannot support until March 14, 1988. This ECN should be ecmpleted prior to unit I restart.
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Other Items (cont.):
ECN 6761 - East Valve Room (EVR) Blowout Panels Implementation of unit I work is progressing as resources are available.
Unit 2 work is complete.
ECN 6784 - Documentation to Show Pipe Class Breaks Final closure is awaiting a revision to AI-19 deleting requirements to mark shift supervisor drawings.
ECN 6808 - Conduit External Sealing Complete.
ECN 6815 - Installation Power Circuit Breaker Install 500-kV power circuit breaker and associated equipment for bay 1.
Retire 161-kV PCB and associated equipment. A total of eight workplans will be required.
Foundations and conduit are presently being installed by WP 12654. WP 12740 for lighting, drain pipe, and surface ground mat has completed the review cycle.
WP 12739 for the structural steel installation has been written and approved.
Completion of WP 12654 will probably be delayed somewhat by late delivery of anchor bolts and bad weather.
An attempt is being made to make up for lost time. Workplan 6815-02 has been written and is in the review cycle. Workplan 6815-01 is now being written.
ECN 6860 - Control Room Bullet Resistivity - DCR 2268 - ECN 6860 WPs 12602, 12603, 12604, and 12605 are in closure cycle.
ECN is complete except for miscellaneous touchup painting. The lockset for doo.- C-37 malfunctioned and a CAQR has been initiated. Work is estimated to be completed by March 14, 1988.
ECN 6866 - Modification of Valve Room Doors Complete.
ECN 6871 - External Conduit Sealing DGB/CB/ERCW Complete.
ECN 7000 - Addition of Support to Main Steam Dump Header Fieldwork is complete except for final inspec~ ions during mode 2 heatup when steam dump header reaches 5000F+. ---
Other Items (cont.):
ECN 7078 - Install Hangers - Main Steam Piping Complete except for final inspections during heatup.
ECN 7190 - Replace Duct Detectors and Install Thermal Detectors WP 7190 Installation and checkout is complete. Workplan closure is in process.
DCN X00006A - Remove Hydrogen Analyzers Tubing The workplan lacks PMT (SI-219).
DCN 2000188 - Install Needle Valve for Hydrogen Analyzers The workplan lacks PMT (SI-219).
DCN 49 - Computer Room Air Conditioner Awaiting instruction revision to close workplan.
DCN 70 - Hydrogen Analyzer Check Valves The workplan lacks PMT (SI-219).
DCN 89 - Crane Wall Penetration Additional penetrations were added by DNE and completed by Modifications.
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; 94 nonrestart discrepancies are open.
For unit 1, there have been 193 discrepancies issued to MODS to date with 30 not required for restart. Of 163 required for restart. 93 remain open and are being worked.
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ECN 6596 - WP 12402 - Complete with the exception of five percent of j
paying to be completed spring 1988.
}
Other Items (cont.):
WP 12477 - Requires Instrument Maintenance calibration and checkout of area radiation monitor and air compressor.
WP 12612 - Complete.
Closure in process. SOIs are complete,. SI-743 is in the process of being updated by Electrical Maintenance.
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6 E WE CALCULATION MANUAL CHANGES i
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0FFSITE DOSE CALCULATION MANUAL CHANGES REVISION NO. 18 4
SQN ODCM Chango Description Form Et!EEj B.iSD.of Chad 12:
Table 3.1 and Figure 3.6 need to be roviced to reflect the environmental radiological, monitoring requirements for SQN. Specifically, Rev. 17 of cho SQN ODCM contains the environmental monitoring requirements for BFN not SQN. Further, monitoring locations 00 and 23 need to be deleted frem-Figure 3.6. Those locations were, in a previous ODCM revision, doloted from Table 3.2 but inadvertently retained on Figuro 3.6 Affected pagec: 69.through 72, 85.
Iqq;1f_isation f or chanag:
The SQN ODCM must reflect the onvironmental radiological program for SQN.
Ef fects on setpolnis_ and deso calculations:
This change will have no effect on effluent setpoints or dose calculations.
s bb RARC Review:
_ ate-RARC Chairman i
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e TABLE 3.1 (Sheet I cf 4)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
~
Exposure Pathway S q ling and Type and Frequency and/or Sanole Sample Locations
- Collection Frequency of Analysis
- 1. AIRBORNE
- a. Particulates 4 samples from locations Continuous sanpler Analyze for gross beta (in different sectors) at operation with sample radioactivity > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or near the site bnundary collection once per following filter change.
(LM 2, 3, 4, and 5) 7 days imore frequently Perform gansna isotopic if required by dust analysis on each sanple loading) if gross beta > 10 times yearly mean of control sample.
Conposite at least once per 92 days (by location for gamma scan) l 4 samples from coninunities approximately 6-10 miles I
distance from the plant (PM 2, 3, 8, and 9) 3 sanples from control locations greater than 10 miles from the plant (RM 1, 3, and 4)
- b. Radioiodine Sanples frG i same locations Continuous sanpler l'1 1 at least once per as local (LM) and Remote (RM) operation with filter 7 days air particulates collection once per 7 days
- c. Soil Samples from same locations Once per 3 years Gamma scan,
- 5r, Mr as air particulates once each 3 years
- 2. DIRECT RADIATION 2 or more dosimeters placed Once per 92 days Gamma dose at least at 10 of the air particulate once per 92 days sampling stations (LM-3, LM-4, LM-5, PM-2, PM-3, PM-8, PM-9, RM-1,k RM-3, and RM-4) 2 or more dosimeters placed at each of at least 30 other Iocations.
- o o us (Figures 3.2 and 3.5) 1S~3
- *f m
o o@E 7 5anple locations are shown on Figures 3.1, 3.2, 3.3, 3.4, 3.5, and 3.6.
p,a Entire page changed.
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TABLE 3.1 (Sheet 2 cf 4)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sanpling and Type and Freguency and/or Sanple Sanole locations
- Collection Frequency of Analysis
- 3. WATERBORNE
- a. Surface TRM 497.0 Collected by automatic Ganma scan of each (Figure 3.4)
TRM 483.4 sequential-type sanpler**
conposite sample.
1RM 473.2 with conposite sanples Composite for tritiun collected over a period analysis at least of $ 32 days once per 92 days
- b. Ground 1 sanple adjacent to At least once per 92 Gross beta, ganma scan (Figure 3.2) plant (location W-6) days and tritium analysis at least once per 92 days I sample from ground water source upgradient
- c. Drinking 1 sanple at the first Collected by automatic Gross beta and ganma (Table 3.3) potable surface water sequential-type sanpler**
scan of each composite (Figure 3.4) supp1v downstream from with conposite sangle sar,pJ e.
Comoosite for the plant (TRM 473.0) collected over a period trittun, * *Sr, *oSr at least of $ 31 days once per 92 days 1 sapple at the next 2 G ab sanple once per 31 downstream potable surface days water suppliers (greater than 10 miles dc<nstream)
(TRM 470.5 and 466.3) 2 sanples at control
!amples collected by locations (TRM 497.0 automatic sequential-and TRM 503.8) type sanpler with composite sarple collected over a period of $ 31 days
- d. Sediment TRM 496.,
At Mast once per 184 Ganma scan of each TRM 483.4 days sangle TRM 480.8 TRM 472.6
- o m 6 en
- Sanple locations are shown on Figures 3.1 3.2 3.3, 3.4, 3.5, and 3.6.
- Sanples shall be collected by collecting a,n al,quot at intervals not exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
$ $ *3 mo i
ra o
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TABLE 3.1 (5heet 3 cf 4) a RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sangling and Type and Frequency and/or Sancle Sanole locations
- Collection Frequency of Analysis
- e. Shoreline TRM 485 At least once per 184 Gamna scan of each Sediment IRM 478 days sanple (Figure 3.4) 1RM 477
- 4. INGESTION
- a. Milk I sample from milk producing At least once per 15 days Ganma isotopic and (Figure 3.6) animals in each of 1-3 areas 1311 analysis of each indicated by the cow census sanple.
- Sr, 'OSr once where doses are calculated per quarter to be highest. If samples are not available from a milk animal location, doses to that area will be estimated by projecting the doses from concentrations detected in milk from other sectors,or by sangling, vegetation where milk is not available (Table 3.1, 4.d)
At least I sangle from a control location.
- b. Fish 1 sample each for Nickajack. At least once per 184 Gamma scan on edible Chickamauga, and Watts Bar days. One sanple of portion Reservoirs each of the following species:
Channel Catfish White Crappie Smallmouth Buffalo
- 5 apple locations are shown on Figures 3.1, 3.2, 3.3, 3.4, 3.5, and 3.6.
Entire page changed.
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TABLE 3.1 (Sheet 4 cf 4) e RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sangling and Type and Frequency and/or Sample Sanole Locations
- Collection Frequency of Analysis
- c. Invertebrates TRM 496.5 At least once per Gamna scan on edible (Asiatic Clams)
TRM 483.4 184 days.
portion TRM 480.8
- d. Food Products 1 sample each of principal At least once per 365 Ganna scan on edible food products grown at days at time of portion private gardens and/or harvest. The types farms in the immediate of foods available vicinity of the plant.
for sangling will vary.
Following is a list of typical foods which may be available:
Cabbage anc/or Lettuce Corn 1
Green Beans Potatoes Tomatoes
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- e. Vegetation 1 sample from up to three At least once per 31 days Ganma scan at 1 cast (Figure 3.6) locations of milk-producing once per 31 days.
animals where a sanple of
- *Sr and 'osr milk is not available and analysis and least at each air particulate once per 92 days station
- Sanple locations are shown on Figures 3.1, 3.2, 3.3, 3.4, 3.5, and 3.6.
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Figure 3.6 MILK AND VEGETABLE SAMPLING LOCATIONS T
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l 5 Milk and vegetation collected O Vegetation only collected O
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~~ f Note: Vegetation is also collected ca h r rnoni toring s tat ion.
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-- I TENNESSEE VALLEY AUTHORITY Sequoyah Nuclear Plant P..O.
Box 2000 Soddy-Daisy, Tennessee 37379 March 14,.1988 Nuclear Regulatory Commission Office of Management Information and Program Control Washington, DC 20555 Gentlemen:
SEQUOYAH NUCLEAR PLANT - MONTHLY OPERATING REPORT - FEBRUARY 1988 Enclosed is the February 1988 Monthly Operating Report to NRC for Sequoyah Nuclear Plant.
Very truly yours, TENNESSEE VALLEY AUTHORITY c/ /
)
S./J.) Smith J
)nu(
Plant Manager Enclosure cc (Enclosure):
Director, Region II Nuclear Regulatory Commission Office of Inspection and Enforcement Suite 3100 101 Marietta Street Atlanta, Georgia 30323 (1 copy)
. Director, Office of Inspection
\\
and Enforcement Nuclear Regulatory Commission I
Washington, DC 20555 (10 copies)
)
Mr. T. Marston l
Electric Power Research Institute P. O. Box 10412 Palo Alto, California 94304 (1 copy)
INPO Records Center Suite 1500 1100 Circle 75 Parkway I
Atlanta, Georgia 30339 (1 copy) y)
Mr. K. M. Jenison, Resident NRC Inspector l
0&PS-2, Sequoyah Nuclear Plant l
An Equal Opportunity Employer
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