IR 05000336/1985026
| ML20134E576 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/07/1985 |
| From: | Mccabe E, Wenzinger E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20134E562 | List: |
| References | |
| 50-336-85-26-MM, NUDOCS 8508200334 | |
| Download: ML20134E576 (20) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
MEETING REPORT Docket / Report No.:
50-336/85-26 License No.:
DRP-36 Licensee:
Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101 Facility Name:
Millstone Nuclear Power Station, Unit 2 Meeting Date:
July 18, 1985 Reported by:
8Le b4 fr/ Cl Br E. C. McCa e, ief, RPS 3B Date d!7!"
Approved by:
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. C. We
, Chief, Projects Branch No. 3 Date Management meeting conducted on July 18, 1985 in NP.C Region I to discuss desi0n changes made during the recently completed Millstone 2 outage.
The event of in-itial concern was identification of the connection of each of the two pressurizer spray valves to the controller for the opposite spray valve (i.e., valve HIC 100E was connected to the HIC 100F controller, and vice versa).
This occurred as a re-sult of a faulty desigt. change.
The licensee presented his rationale for conclud-ing that other modifications made during the outage had no errors and that the modifications involved had been properly tested.
Review of the spray valve con-troller design change by the Haddam Neck Design Change Task Group was committed to by the licensee.
NRC Region I concurred that the licensee's evaluation provided a reasonable assurance that facility operation could be resumed without adverse affect from recent design changes.
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0508200334 850013 PDR ADOCK 05000336 G
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DETAILS 1.
Meeting Attendees Northeast Utilities R. Kacich, Licensing Supervisor J. Kelly, Millstone 2 Superintendent E. Mroczka, Vice President, Nuclear Operations NRC Region I L. Bettenhausen, Chief, Operations Branch, Division of Reactor Safety (DRS)
5. Ebneter, Director, DRS J. Gutierrez, Regional Counsel W. Kane, Deputy Director, Division of Reactor Projects (DRP)
E. McCabe, Chief, Reactor Projects Section 3B, DRP J. Robertson, Reactor Engineer, Reactor Projects Section 3B, DRP R. Starostecki, Director, DRP E. Wenzinger, Chief, Pro,' acts Branch No. 3, DRP 2.
Meeting Presentation The licensee presented his analysis of the design change which improperly connected the pressurizer spray valves to the opposite spray valve's control-1er and the results of his review of other design changes (Appendix 1 to this report with Attachments 1 and 2).
3.
Licensee Commitment The licensee committed to a review of the containment spray controller design change by the Haddam Neck Design Change Task Group.
4.
NRC Positim NRC Region I concluded that the licensee's review of this matter provided reasonable assurance that Millstone Unit 2 operation could be resumed without adverse affect from the design changes made during the recent completed re-fueling outag C
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APPENDIX 1
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NRC MANAGEMENT MEETING JULY 18, 1985 MILLSTONE UNIT NO. 2 REACTOR TRIP OF JULY 15, 1985
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-1-Millstone Unit No. 2 Reactor Trip
- July 15, 1985 Initiating Event At approximately 1525 on Monday, July 15, 1985, with the unit at 100% power, 886 MWe, pressurizer pressure began to decrease from a steady state value of 2270 psia.
Approximately two minutes later the operators were alerted to the condition when reactor protection system pretrips for thermal margin / low pressure annun-ciated at 2225 psia.
Initial actions included verifying spray
valves, power operated relief and safety valves were closed.
Turbine load was reduced in an attempt to recover pressure.
Despite the lack of abnormal indications and the reduction of turbine load, the unit tripped at 1533 on thermal margin low pressure when pressurizer pressure decreased below 2155 psia.
The unit responded routinely to the reactor and turbine trip from 100% power with the exception of pressurizer pressure and steam generator level.
The steam generators were slightly over-fed due to the actions prior to the trip but level was restored to an indicated band within ten minutes.
Pressurizer pressure decreased during the post trip transient to 1790 psia, recovered to 1860 psia ten minutes after the trip and then slowly decreased to 1725 psia.
During the decrease it was determined that a stuck spray valve was the initiating event and the valve controllers were placed in manual and given a full close signal.
This was done as insurance since the controllers for both valves indicated a full close demand.
In addition, a containment entry was
initiated to manually isolate the spray valves.
Prior to ob-taining the Radiation Work Permit for the containment entry, pressurizer pressure decreased to 1725 psia and the operators secured two reactor coolant pumps, which resulted in a signifi-cantly reduced spray flow, and allowed the pressurizer heaters to restore pressure to 2250 psia.
After the containment entry was complete, the spray valves were isolated sequentially in an attempt to identify the affected valve.
When it was determined that both valves were partially opened, the isolation valves were all closed and the two secured reactor coolant pumps restarted.
During this troubleshooting it was determined that the control room pressurizer spray controllers HIC 100E and 100F controlled the opposite valve.
This discrepancy is addressed in the attached report.
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-2-Cause The trip resulted when the pressurizer spray valve opened suffi-ciently to overcome the capability of the pressurizer heaters to maintain pressurizer pressure of 2270 psia.
While it is assumed that one or both valves were partially open during the week of operation, an unknown event caused the F valve to open further at 1525, initiating the transient.
This initiating event is assumed since the P valve was found stuck open when the operators entered the containment.
Concerning the E valve's failure to close and any initial F valve problems, these discrepancies were not noted during the week of
operation since pressurizer spray flow was desired and induced by use of the pressurizer heaters.
The power ascension test requires this spray flow to equalize boron concentration between the pres-surizer and the remainder of the reactor coolant system.
When the test was completed at noon on July 15 and the pressurizer pressure controls were returned to normal, the inability to maintain pres-surizer pressure was observed and two backup heater groups were energized.
A trouble report was issued to investigate pressurizer heater output, and the possibility of pressurizer spray valve leakage was discussed.
To obtain stable pressure control, spray was forced as it had been for the preceeding week, pending reso-lution of the problem.
There was no active investigation of the problem at the time of the trip.
Corrective Action The spray valves were functionally checked while isolated.
The E valve operated smoothly and its stroke was adjusted so that it would go fully closed.
The F valve would not stroke and was completely disassembled.
The only defects found were come wear marks on the plug and one ring of packing slightly misaligned.
The plug was cleaned and the valve repacked and assembled.
Other maintenance completed included repair of the reactor pro-o tection system matrix test relay switch which was discovered during startup testing, repair of a tear in the condenser boot seal and a balancing shot on the turbine generator.
Action to Prevent Recurrence Since there were no obvious defects in the spray valves and the reactor / turbine trip were normal, no action to prevent recurrence is planned at this time.
The spray valve preventive maintenance program and the addition of remotely controlled spray valve iso-lation valves will be evaluated.
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i-3-Reversed Wiring of Pressurizer Spray Valves Initiating Event Plant design change requests were implemented during the 1985 refuel outage to the pressurizer pressure and level instru-mentation and controls.
The modification upgraded the original GEMAC equipment to Foxboro components.
This is part of a long term and recently completed program to comply with 10 CFR 50.49.
The modification was installed during March and April, calibrated in May and tested in June.
The plant design change request was
properly followed for the installation, the calibration completed using revised unit procedures and the test performed using an approved Inservice test.
The discrepancy was discovered during the investigation following the reactor trip of July 15, 1985.
Cause As indicated on the simplified sketch, three of the drawings included in the plant design change request contained the dis-crepancy which resulted in the wiring reversal.
Sheet 50D identified wires N/P as carrying loop E signals while wires L/M carry the loop F signal.
However, sheet 50E reversed the wires such that wires N/P supply the F loop and wires L/M supply the E loop.
This reversal was repeated on sheet 50F in that ouput wires O/R and S/T switched from F to E and E to F respectively.
As a result of these wire reversals, the CO3 and C21 controllers, which were reflected on sheet SOE, were wired such that the E controller operated the F valve and the F controller operated the E valve.
This discrepancy was not discovered during testing since valve operation.was not verified.
Rather, test meter deflections were observed.
These meter deflections were observed at test points identified on the same prints which resulted in the reversal.
Corrective Action - Specific When the wiring error was discovered an addendum was initiated to the plant design change request to correct the discrepancy.
The change consisted of reversing the wires on the CO3 controllers and reversing the nameplates on the C21 controllers.
To understand why the corrective actions are different the design of the control boards must be considered.
On C21, facility separation is by a metal divider which splits the hack panel in half.
On CO3, three feet of space provides the separation.
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t Concerning the original modification to CO3, identical controllers were placed in a new section of control board for the two valves.
The labeling of these controllers was consistent with the design layout drawing which maintained the system mimic on the remainder of the panel.
However, the design layout drawing reflected the mistake made on sheet 50E and the wrong wires were connected to the controllers.
Therefore, the corrective action is to reverse the wires.
The modification on C21 consisted of identical controllers being placed in two holes in the control board.
One of the holes was on each side of the divider plate.
Since one hole was on each side of the divider plate and one cable was pulled to each side of the
plate, the connections were made without reference to the layout drawings or connection drawings which were in error.
The layout drawing was used for labeling, however, and the controllers were mislabeled.
Thus, the C21 corrective action is to reverse the
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labels.
It should be noted that no cable separation was violated since the cable routing drawings were accurate.
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To ensure the revised design change is correct, testing has been performed which verified the adequacy of the addendum.
The test will physically stroke the valves and ensure they operate to maintain the appropriate pressurizer pressure.
It will also verify the E controllers on C21 and CO3 operate the E valve and the F controllers on C21 and CO3 operate the F valve.
Corrective Action - General Since testing after plant modifications should identify design deficiencies, all plant design change requests (PDCR's), which were implemented during the 1985 outage, were reviewed for testing adequacy.
This was done in a two step process.
All PDCR'S were reviewed to determine a need for a thorough review.
Attachment 1 contains the list of design changes whose testing was not reviewed and the reason that the review was not required.
Attachment 2 contains the list of design changes whose tests were reviewed, a description of the tests and the status of completion.
The Millstone Unit 2 Plant Operations Review Committee and the Nuclear Review Board have reviewed this matter and have concluded that.it is acceptable to re-start the unit.
Action to Prevent Recurrence The plant design change request program has recently been revised to increase the controls on plant modifications.
In general, these procedure changes have resulted in more thorough reviews and more complete documentation for modifications.
They have been effective ahd confidence exits that over sixty modifications were completed and tested in an appropriate manner during the past Millstone Unit 2 outage.
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~5-The major discrepancy noted with the pressurizer spray valve change was the failure of the test to discover the error.
This is directly attributed to-inadequate testing in that the operation of the valve was not verified.
In light of other events in the NU system and the status of other related, ongoing initiatives, the following corrective actions are planned:
1.
More education (as distinguished from training) is needed to ensure that the intent of a thorough pre-operational test is conducted for all design changes.
For Millstone Unit No.
2, this education will be completed prior to implementation of any additional design changes.
For Connecticut Yankee and Millstone Unit No.
1, this will be completed by September 30, 1985.
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A memorandum from the Senior Vice President, Nuclear Engineering and Operations, will be issued to all NE&O personnel involved in the design change process which will focus on the lessons learned from this problem.
Increased emphasis will be placed on the need for a thorough pre-operational test which directly demonstrates the operability of the entire system.
This memorandum will be issued by August 21, 1985.
3.
The CY Plant Design Change Task Group (PDCTG) will include in their review the plant design change which caused this problem at Millstone Unit 2.
It is noted that the PDCTG has previously committed to evaluate the adequacy of the design change related procedures issued on November 1, 1984.
This review will be completed on the schedule required by the Order dated December 13, 1984.
4.
The PDCR's identified in Attachment 1 will be re-reviewed, to confirm the adequacy of the testing conducted, by September 30, 1985.
5.
For Connecticut Yankee and Millstone Unit No.
1, plant design changes implemented between November 1, 1984 (the date of issuance of the new procedures) and the conclusion of the next refueling outages (currently scheduled to commence January 4, 1986 and October 19, 1985 respectively) will be reviewed to confirm the adequacy of pre-operational testing.
The scope of the review will be comparable to that being completed for Millstone Unit 2.
This review will be completed prior to startup from the upcoming refueling outages.
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Attechmant 1 1985 OUTAGE PDCR'S NOT REVIEWED FOR TESTING ADEQUACY JULY 15-17, 1985 PDCR NO./ DESCRIPTION REASON REVIEW NOT REQUIRED 2-20-83/ Provide hydrolazer No effect on system operation power supply.
due to normal electrical protection.
2-57-83/ Piping improvements Pcssive modifications; normal in nitrogen system.
system usage verified proper operation.
2-71-83/ Installation of main No control functions or generator tagging compound safety-related indications analyzer.
affected.
2-113-83/ Installation of CRT Normal system usage verified for control board indication.
proper operation.
2-4-84/ Reinstallation of refuel Passive modification - no pool ladders.
operational verification required.
2-12-84/ Provide alternate power Work in progress.
to MP-3 Security.
2-15-84/ Addition of elapsed No control functions or time indicators for transformer safety-related indications oil pumps.
affected.
2-20-84/ Replace main generator No control functions or watt-hour meter.
safety-related indications affected.
2-23-84/ Addition of spare pump Normal system usage verified in CPF waste discharge.
proper operation.
2-37-84/ Replace relief valve Passive modification; normal 2-SI-466.
system usage verified proper operation.
2-42-84/ Addition of plant No effect on system operation lighting.
due to normal electrical protection.
2-52-84/ Replace feedwater Normal system usage verified heaters 2A and 2B.
system in,tegrit p
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-2-PDCR NO./ DESCRIPTION REASON REVIEW NOT REQUIRED 2-54-84/ Replace wide range Normal surveillance testing N1 power supplies.
verified system operability.
2-58-84/ Addition of plant Normal system usage verified telephone, proper operation.
2-60-84/Feedwater control Normal system usage verified-system upgrade.
proper operation.
2-62-84/ Addition of strainer Passive modification - no to refuel pool drains.
verification required.
2-63-84/ Leak repair on lube Passive modification; normal water piping.
system usage verified proper operation.
2-66-84/ Construction of shell Work in progress; passive for future radwaste reduction modification - no operational facility.
verification required.
2-3-85/ Addition of crane warning No control functions or lights in Turbine Building.
safety-related indications affected.
2-6-85/ Upgrade shutdown cooling No control functions or loop instrumentation.
safety-related indications effected.
2-10-85/ Replace RPS bistable Normal surveillance testing power supplies.
veri #ied system operability.
2-11-85/ Upgrade feedwater heater Passive modification; normal gage glasses.
system usage verified proper operation.
2-13-85/ Replace capacitors in Normal system usage verified inverters 3 and 4.
proper operation.
2-15-85/ Addition of heater drains Passive modification; normal
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tank gage glass.
system usage verified proper operation.
2-18-85/ Chemical cleaning of steam Normal system usage verified generator secondary side.
proper operation.
2-20-85/ Replace terminal block Normal surveillance testing in containment penetration.
verified system operability, i
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-3-PDCR NO./ DESCRIPTION REASON REVIEW NOT REQUIRED 2-21-85/ Modify power supply for Normal system usage verified containment temporary power.
proper operation.
2-23-85/ Relocate plant door 207.
Passive modification - no operational verification required.
2-24-85/ Core fuel reload for Normal surveillance testing Cycle 7.
and in-service tests verified system operability.
2-25-85/ Seismic qualification Passive modification - no of refuel pool drain lines.
operational verification required.
2-26-85/ Installation of chlorine Passive modification; normal system filters.
system usage verifies proper operation.
2-27-85/ Reactor cavity seal Normal system usage verified test equipment repair.
proper operation.
2-28-85/ Addition of fuel oil No effect on system operation.
test laboratory.
2-31-85/ Steam generator tube Normal system usage verified sleeving.
proper operation.
2-33-85/ Upgrade electrical No effect on system operation
facilities in snubber test room, due to normal electrical protection.
2-37-85/ Replace overcurrent Normal surveillance testing devices in 480V breakers.
verified proper operation.
2-38-85/ Replace discs on 2-AC-54 Normal surveillance testing and 2-AC-55.
verified system operability.
2-39-85/ Replace service water Passive modification; normal piping.
system usage verified proper operation.
2-40-85/ Steam generator tube Normal system usage verified plugging.
proper operation.
2-42-85/ Addition of instrument Normal system usage' verified root valves for service water to proper operation.
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-4-PDCR NO./ DESCRIPTION REASON REVIEW NOT REQUIRED 2-44-85/ Replace service water Passive modification; normal piping-in intake structure.
system usage verified proper operation.
2-45-85/ Replace valves 2-MS-12A Passive modification; normal and 2-MS-12B.
system usage verified proper operation.
2-46-85/ Control room console Passive modification - no replacement.
operational verification required.
2-47-85/ Addition of high Passive modification - no radiation area access control operational verification gates.
required.
2-48-85/ Repair and upgrade snubbers.
Passive modification - no operational verification required.
2-49-85/ Restoration of RCS RTD's Normal surveillance testing to RPS.
verified system operability.
2-50-85/ Chemical decontamination of Normal system usage verified steam generator primary heads.
proper operation.
2-51-85/ Upgrade power feed to No effect on system operation warehouse, due to normal electrical protection.
2-52-85/ Upgrade site evacuation Normal system usage verified alarm for MP-3.
proper operation.
2-53-85/ Addition of refuel pool Passive modification; normal drain filter, system usage verified proper operation.
2-54-85/ Replace cable vault drain Passive modification - no header, operational verification required.
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2-57-85/ Replace ICI guide tube Passive modification - no assembly.
operational verification required.
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-5-PDCR NO./ DESCRIPTION REASON REVIEW NOT REQUIRED 2-58-85/ Replace mechanical snubbers.
Passive modification - no operational verification required.
2-59-85/ Addition of maintenance Normal system usage verified phone jack, proper operation.
2-60-85/ Steam generator tube Normal system usage verified removal and plugging.
proper operation.
2-61-85/ Upgrade secondary sample Normal system usage verified sink instrumentation.
proper operation.
2-62-85/ Addition of sightglaus on Passive modification; normal CPF tank.
system usage verified proper operation.
2-63-85/ Upgrade steam generator Passive modification - no snubbers.
operational verification required.
2-65-85/ Addition of EPRI/UCONN Passive modification - no test materials in loop area.
operational verification required.
2-68-85/ Addition of constant Normal system usage verified temperature bath in CPF.
proper operation.
2-69-85/ Addition of sodium analyzers Normal system usage verified and constant temperature bath for proper operation.
hotwell samples.
2-72-85/ Addition of sightglass in Passive modification; normal CPF tank vent line.
system usage verified proper operation.
2-73-85/ Repair hanger base plate Passive modification - no in containment.
operational verification required.
2-74-85/ Blowdown vent line Passive modification; normal modification.
system usage verified proper operation.
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Attachment 2 19R5 OUTACE PDCR'S RFVIFWED FOR TESTING ADFQt'ACY PDCR DFSCRIPTION TESTING RFQUIRFD TFSTING STATUS & RFMARKS PDCR 2-22-85: To perform EFQ replacement of solenoids Proper valve operation to be verified by Testing coepleted satisfmetorily, under and limit switches on valves on the secondary plant testing consisting of cycling each valve listed AWO's: M2-85-05462, 05463. 05464, 12-MS-266B 2-MS-2658, 2-SW-8. l A 2-SW-8. l B 2-SW-8. lc three times using its respective control 05466, 05468, 05469, 0547I, 07146, 06781, 2-SW-3.2A, 2-SW-3.2B, 2-LRR-61.1, - Scienoids)
switch while observing each limit gives 067N3.
correct valve position.
[2-SW-8.lA. 2.SW-8.1B, 2-SW-8.lc, 2-LRR-61.1, 2-FB-88, 2-EB-89 - Limit Switches]
PDCR 2-7-85:
To improve Control Room habitability I.
Rectre air flow rate at 2500 * 2 50 CFM.
I.
Airflow Test Sat. - OP 2609F-1.
during certain incidents by giving redundant air 2.
Redundant initiation of Rectre mode.
2.
Facility I & !! Test - OP 2609A-1, rectre initiation controls, adding 3 new dampers, 3.
Radiation detectors respond to radiation 26095-1 - Sat.
adding new air supply radiation detectors and and will initiate recirc. at 2 MR/hr.
3.
Radmonitor Test and Calibration replacing chlorine detectors.
Chlorine detectors respond to chlorine IC 2410K-1 - Sat. and 24013-1.
and will initiata rectre at I ppm.
Chlorine Monitor Test and Calibration -
Sat.
PDCR 2-8-85:
To replace pressure transmitters 1.
Channel calibration of all loop 1.
Calibrations Sat. - IC 241AC.
PT-100X & PT-100Y with EFQ rated devices, and to components.
2.
Control loop operations - S3t.
T84-34 replace prensure control loop devices with new 2.
Pressurfrer pressure control loop upgraded (SPEC 200) components, operation.
RFMARKS: Additional inservice tests-T85-28 has been written tn verify trat the expected control device does actually operate its associated spray control valve.
T85-29 will verify performance of t l.,
pressure control system.
PDCR 2-34-85:
Replace the Auto / Manual controllers I.
Channel calibration of all loop I.
Channel calibration Sat. - SP 2 02D,
for Auxiliary Feedwater Flow control loops, components.
2.
AAFWIS operability per OP 2610C - Sat.
HIC-5276A6B and HIC-5279A&B with Manual loading 2.
Verification of Aux. Feedwat e r stations. Modifies the loop electronics to provide Initiation circuitry and Flow Control
- During calibration review, it was noted valve position indication for 2-FW-43A & 2-FW-438 Valve position circuitry, by visual determination in the Contrnt which is consistent with normal plant convention, Room (C05F) that 0% controller output i.e: 01 valve full closed, 100% valve full open.
corresponds to a closed valve.
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Eliminates auto flow control feature which is not utilfred.
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Page 2 of 6 PDCR DESCRIPTION TFSTING RFQUIRED TFSTING STATUS & RFMARKS PDCR 2-4 3-8 5 : Replaces non-qualified custom component 1.
Leak check of process tubing following 1.
Installation and leak test - Jat.
pressure switches PC-224X,Y,Z; PS-6319A,B C. with installation, ter AWD P2 85 6432 and H2 85 06433.
Category IE, Environmentally qualified replacements.
2.
Calibration of pressure switches per 2.
Calibration - Sat. - IC 21003.
16C 21003.
- Functi no tested by trippica charging PC-224: Charging Pump Suction Pressure pump ustng Icw vacuur device, and by PS-6119: RBCCW Pump Suction Pressure tripping RBCCW puep on decreasing suction pressure.
PDCR 2-64-85:
Installation of an ATI Bypass circuit Special test, to include verifying operation 1.
Work complete; Testing not yet for the ATI portion of the containment purge valves in the normal or unbypassed mode to check ATI performed. ATI bypass circuit not in the ESAS. The circuit will allow the operatore pattern and presence of ATI fault if called placed in service until testing is to bypass tripped purge valves for the ATI test for. Also to verify ATI Bypass does not comp **1s!.
sequence and thus eliminate a nuisance alarm for prohibit any ESAS functions, but does prevent 2.
ATI calibration IAW IC 2430A complete ATI fault.
a tripped bistable from causing an ATI fault, and sat.
3.
ATI functionally tested daily IAW OP 2619A.
PDCR 2-4R-R4:
Replace service water strainer Functional test to verify initiations and 1.
Switches replaced and tested under Differential Pressure (DP) switches and changes operation of the backwash cycle and the AWO M2-85-02371 for service water,
to pressure setpoint for start of backwash and h1 alarm in the control room.
Initial service strainers B and C and tested Sat.
DP alarm, for increased reliability.
leak test to verity pipe and tube integrity.
(A strainer not done yet).
2.
AWO's 03400 6 03401 document leak test results. as sat.
- Visually verified pressure switches sensing lines are tied in correctly to strainer high & low side.
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s Page 3 of 6 PDOR DFSCRIPTION TESTING RFQUIRED Tf5 TING STATUS & Rf' MARES PDCR 2-9-85:
To replace Pressurizer level transmitters 1.
Letdevn instrument calibration.
1.
Calibratigns: Sat.
IC2418D and LT-Il0X & LT-IICY with EEQ rated devices and to 2.
Pressurizer level instrument c a lit.r a t ion.
SP 2402E.
replace level control loop devices with new. upgraded 3.
Pressuriter level control operatien.
2.
Control loop operationg - Sat. per (SPFC 200) components.
Inservice Test T84-40.
- Copies available in NPRF.
- T85-19. Power Ascension Tear noted that level controls functioned properly.
PDCR 2-lh-85 Inst 411ation of rew instrumentation to I.
Field structural welds visual exam.
1.
N1'SCo - 908 f actory test sat. - cery sense and dis play indications of Inadequate Core Cooling 2.
Electrical checks per SP-EE-076 in unit Fngineers files.
including re;ctor vessel level (in %), Core Exit 3.
ICC factory test per NUSCo 908.
2.
HJTC test ICE-35231 sat. - copy in unit Thermocouple (CET) temperatures and, coolant subcooling 4.
HJTC vendor test per ICF-35231.
Engineers files, or superheat.
Installations included reactor vessel 5.
Functional test of CRT terminal.
3.
Functional testing of vessel level level sensor.(2) and two signal processing and control 6.
Pressure boundary inspections display (HJTC) during RCS filling, under cabinets whith received the signals from level sensors, s.
IP on seal weld.
T85-26.
Results,- Sat. - Copy in unit CFT's and RCS loop temperature and pressure for b.
Visual leak check Fngineers files, calculation of coolant subc ooling/ supe rheat.
c.
Locking mechanism
Pressure boundary inspectiors:
d.
LLRT on new electrical a.
Visual leak check performed during penetrations.
10 year ISI !!0% test with satisfactory results, b.
LIRT on new electrical penetrations performed prior to ILRT.
- 5.
Functional test of CRT still ongoing due to occasional lockup of display when memory overflows.
CET temperature display results show ICC display for CET's high by up to 12*F.
Needs final resolutions.
- A functional test of CFT/subcoolleg display functions remains to be completed.
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Page 4 of 6 PDrR DFSCRIPTION TFSTING RF?ttlRFD TESTINC STATt'S & RFMARKS PDCR 2-MS-A3:
Various process radiation monitors to 1.
Radmonitors to be calibrated and 1.
Hydrostatic test of RM 9049 and RM he modified as follows:
functionally tested.
9116 performed in accordance with 1.
RM9049 (Clean Liquid Radwaste), RM 9116 (Aerated 2.
Hydrostatic test to be perf ormed on T84-24 with sat. results.
Liquid Radwaste) and RM4262 (Steam Generator new piping and flush water systems for 2.
Calibration and function testing of Blowdown) to he relocated to a low background RM 9049 and RM 9116 RM 0049 and FM 9116 perforced by radiation area.
T84-26 with sat. tesults.
2.
RM 6038 (RBCCW) to have its shielding replaced.
3.
EM 6038 calibrated by IC 2422E 3.
Saeple pumps to be added to RM 9116 and RM9049.
with sat. results.
4.
Flust water capability to be added to RM9116 4.
RM 4262 calibrated by SP 2404P with and RM 9049.
sat. results.
RM 9049 RM 9116, RM 4262 and RM 6038 detectors 5.
Hydrostatic test by vendor sat.
and signal processing units to be replaced with 6.
AWO #M2-85-02063 installed FM 4262 units having improved shielding and electronics.
AWO #M2-85-020h4 installed RN 6038 7.
Special Procedures 84-2-7, 84-2-8 verify initial calibration.
PDCR 2-32-85:
To install manual loading (control)
1.
Inservice leak test to verify that 1.
No formal test prepared f or leak stations on the normal and high level dump valves no air leaks are present.
test or functional test. required for the heater drains tank and feedwater heaters 2.
Full operability of these loading testing was performed by AWO lA. IB. 2A, and 2B.
stations to be demonstrated by changing fM2-85-04757 with satisfactory results, f rom manual to aut omatic cont rol and automatic to manua. control.
PDCR 2-36-85:
To relocate the heater drains tank 1.
Visual inspection of welds.
1.
Calibration was performed under AWO high level drain valve controller (LC-5061) 4 inches 2.
Inservice leak test on joints.
- M2-84-03825 and completed sat.
higher than its present location.
3.
Controller calibration IAW 16C 2.
Weld inspection and leak test was procedures.
perf ormed utilizing AWO #M2-85-04559 No other formal testing was completed.
AWO inspection and leak test van sat.
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Page 5 of 6 O
PDCR DESCRIPTION TESTINC RFQUIRED TFSTINC STATUS & NFMARKS PDCR 2-19-85: To replace five existing 480 volt 1.
Complete transformer test program to 1.
The following testing was performed:
loadcenter transformers (22A. 22B, 22C, 22D and 22F)
be performed for each transformer prior
- Doble insulation test with new Brown Boveri transformers.
to connecting to bus work.
- Excitation - current test
- Windirs megger tests
- Connection Datter tests
- Tap setting verification
- Cable hypot test 2.
Functional testing of cubicle heaters, f ans and alarms was also perf ormed by
" red line" verification. No other formal resting was documented although input and output breakers were tested.
PDCR 2-41-84: Modifies the reactor trip circuit 1.
Ensure alarm sounded when switch placed Testing performed per AWO M2-85-03423. and breaker controls by adding a test switch that in test position.
matrix logic testing. found to be sat.
alarms in the test position and which allows 2.
Ensure shunt trip coil restored after resting of the undervoltage trip coils test completion.
NOTE : Procedure PT 21432 will periodically separately from the shunt trip coils.
3.
Ensure separate testing of shunt and UV test this feature. Matrix logic trip coils, testing (normal surveillance testing)
will verify restoration conditions af ter sach performance.
PDCR 2-14-R5:
Removes the seal-in circuit for Petest per OP 2322. Section 7.2 and Tested Sat. - SP 26018-1 and SP 26015-2.
the Terry Turbine Steam inlet valve motor document per SP 26015.
operator controls to allow the valve to be throttled.
PDCR 2-32-84: Modify the Existing Enclosure 1.
Calibration of PDT-8060C &
l.
Calibration Sat. AWO M2-85-04902.
Building Differential Pressure Channels.
PDT 8060J.
Functional Test ~~-Sat:
T 84-31.
(PDT-PO60C and PDT-8060J) to provide automatic 2.
Furetional test of Fan F-23 and shutdown of the Containment Purge Supply Fan Damper HV-8128.
(MF23) and closure of the Enclosure Euilding Exhaust Valve (HV8128).
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PDCR DFSCRIPTION TFSTING RFQUIRFD l t STING STATt'% 6 RFMARKS
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PDCR 2-70-85:
Modify the "B" Charging Purp 1.
Test plan to verify the "B" 3.
Testing performed rer AWO M2-8%07732 power supply crossover interlock circuitry Charging Pump would start during testing was c orpleted as sat.
by wiring in a spare contact from relay 3-1 a partial or full LNP.
2.
Testing M. OPS Form 2e>0lG-1 and OPS and 3-2 in series with the load Sequence 2.
Ferest per Operation Procedure 2f0lG Form 260lH-1.
Zero contact.
and 260lH.
3.
ALO d 's N2-8 5-07 732. 07594 per f e rred installation.