IR 05000333/1989012
| ML20012D735 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/14/1990 |
| From: | Meyer G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20012D734 | List: |
| References | |
| 50-333-89-12, NUDOCS 9003280302 | |
| Download: ML20012D735 (25) | |
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e U.S. NUCLEAR REGULATORY COMMISSION
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Region I
- Report No.:
50-333/89-12 Docket No.:
50-333
- License No.:
DPR-59 Licensee:.
New York Power Authority Post-Office Box 41 Lycoming,=New York 13093 Facility:-
James A. FitzPatrick Nuclear Power Plant Location:
Scriba, New-York Dates:
December 3, 1989 through January 24, 1990 Inspectors:
W. Schmidt, Senior Resident Inspector lasse, Jr. Re ident Inspector Approved by:
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=.JTilenn W. MeyeY,.Chie'f f
Date
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/ Reactor Projects Sectiorf No.1B
- Inspection Summary:
This inspection report discusses routine and reactive inspections of plant activities during day and backshif t-hours including; plant operations, secur-ity,_ - surveillance and maintenance, emergency preparedness, engineering and
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technical support and radiological protection. This period included deep back-shift and weekend inspection conducted on December 3, 13, 23, and 30, 1989 and January-6, 20 and 21, 1990.
- Results:
The inspector identified one non-cited violation.
An Outline of Inspection follows.
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[A[A 9003280302 900315 PDR ADOCK 05000333 o
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_0UTLINE OF INSPECTION
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Operations (MC 71707,93702)
1.a Reactor scram (January 19, 1990) due to a false reactor vessel water level signal.
SS failed to use installed panel. indication for dry-well temperature following failure of the SPDS data point.
Unre-solved Item 89-12-01
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1.b Backseating of drywell MOVs to isolate packing leakage
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(January 12,1990). NYPA committed to make changes to 0DS0-27, F-1, 1.c NYPA did not perform a temporary modification to allow closure and taping of the-latch on the RCIC room fire doors.
Unresolved Item 89-12-02 1.d Review of E0P team inspection concerns:
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(Closed) Unresolved Item 88-00-02; NYPA ensured that plant pro-cess computer and SPDS alarms were in the same units and with the same setpoints.
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(Closed) Unresolved Item 88-00-03; NYPA resolved instances of lack of equipment or information.
NYPA committed to install a permanent drain connection for emergency HCU venting by the end of 1990.
3.
(Closed) Unresloved Items 88-00-01, 88-00-04, 88-00-05, and 88-00-06; Consolidation of items that need further review dur-ing planned E0P inspection.
Unresolved Item 89-12-03 1.e (Closed) Unresolved Item 89-09-01; NYPA clarified their policy for command and control by the SS and ASS.
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1.1 Inspector Assessment 2.
Radiological Protection (MC 71707)
l 2.a NYPA appointed Mr. George Vargo, PhD. as the RES Superintendent.
L 2.b Inspector plant tours.
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Outlineof' Inspection (Continued)
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.Surve111ance'and Maintenance (MC 61726,62703,92702,92703)
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3.a The maintenance department should have promptly documented and iden-tified potentially inoperable safety related snubbers to~ the oper-ations department.
3b Evaluation of valving error that caused the January 19 scram.
3.c A ~ and C RHR pump degradation and NYPA's plans to rebuild the pumps during the 1990 outage.
3.d SBGT fan vibration corrective maintenance.
Poor identification of necessary PMT.
3.e HPCI. steam flow isolation setpoint change, based on isolations during surveillance testing; Resolved Item F-3 from Inspection Report 89-11.
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Inspector observation of EDG surveillance testing.
3.g' Compilation of deficiencies with daily shift check surveillance pro-cedure ST-400. Unresolved Item 89-12-04 3.h Inspector review. of NYPA root cause analysis following RCIC steam supply valve motor failure; Review of LER 89-021.
3.1 '(0 pen) Unresolved Item 88-11-02; Review of LER_89-22, safety related-snubbers declared inoperable.
3.j NYPA conducted testing of the UPS MG set undervoltage relay during the Fall 1989 outage; Resolves Item F-5 from Inspection Report 89-03.
'3.k' NYPA repaired the HPCI steam injection valve and turbine seals during the Fall 1989 outage; Resolves Item F-2 from Inspection Report 89-09.
'3.1 Inspector Assessment 4.
Emergency Preparedness (MC 71707,82301)
4.a. (0 pen) Unresolved Item 89-11-03 and consolidation of Emergency Plan Unusual Event classification issues,
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Outline of Inspection (Continued)
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Security (MC 71707)
'5.a -Inspector walkdown of protected area fences.
5.b NYPA switched normal access point into the protected area.
5.c (Closed) TI 2515/104 Inspection of FFD training program.
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Engineering and Technical Support (MC 37700,90712,92700,92702)
6.a NYpA committed to review the response time of RPS instruments due to spiking on sensing lines, F-2.
6.b Review of HPCI high steam line flow evaluation and calculation. Non-cited Violation 89-12-05 6,c1 (Closed) Unresolved Item 88-11-01; NYPA completed calculation and implementation of an ' acceptance criteria for determination of gross drywell leakage.
6.d (Closed) Unresolved Item 89-80-16; NYPA installed a safety related nitrogen supply to RBCCW PCIVs.
6.e (Closed) Violation 89-80-05; NYPA resolved the design control issues from the SSFI.
6.' f (Closed) Unresolved Item 89-80-07; NYPA -deve19 ped and implemented alarm response procedures for EDG local control panels.
6.g ' NYPA resolved human engineering concerns with ADS logic test switch; Resolves' Item F-1 from Inspection Report 89-03, 6.h NYPA submitted a proposed TS amendment to' address single setpoints for SRVs; Resolves Item F-8 from Inspection Report 89-02.
6.1 Inspector review of LERS; 1.
89-14-00 2.
89-15-01 and 89-15-00; The inspector planned to review correc-tive actions on RBCCW PCIVs during the 1990 outage.
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89-16-00 4.
89-17-00 iii
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s.h, L0utline of Inspection (Continued)
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89-22-00 E
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89-23-00
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89-24-00
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6.1-Inspector Assessment
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Safety Assessment / Quality Verification (MC 30703)
o 7-. a NYPA appointed Mr. George Tasick as QA Superintendent,
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7.b Inspector review of NYPA overtime. Unresolved Item 89-12-06 I
8; ExitInterview(MC30703)
Attachment A-J Acronyms
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DETAILS 1.
Operations
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From the beginning of the period until January 19, the plant operated at full power. On January 19, the reactor scrammed due to false low reactor vessel water level signals, and the unit was placed in cold shutdown.
NYPA restarted the reactor on January 22 and commenced power ascension.
At the end of the period the unit was at full power.
a.
A valving error during a surveillance test on a feedwater control system reactor vessel water level instrument (discussed in section 3.a below) caused the January 19 scram. The inspector reviewed the transient and determined that all systems functioned as designed.
The shift crew performed properly during the transients. The inspec-tor observed deficiencies with the use of drywell temperature instru-mentation as reflected in E0P-4, primary containment control.
The inspector observed operator response to the scram condition. The SS clearly controlled the situation. He gave adequate direction to and asked for the required responses from his shift crew.
The SS controlled activities from the computer bench at the rear of the con-
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trol room.
Controlling from this area allowed him to give and receive information in an orderly fashion, while maintaining a broad overview of activities.
During the transient the inspector observed the drywell average tem-perature on SPDS in the alarm. condition at 175 F.
An average drywell temperature of 135 F was an entry condition for E0P-4. The inspector observed that the SS had E0P-4 open but had not entered the proced-ure.
E0P-4 specified the instruments to be used to monitor-drywell temperature:
(1) the SPDS alarmed average drywell temperature data point; (2) the average of the reading of two drywell temperatures recorded on the main control panel (16-1-TE107 and 108); and, (3) the average of the inlet and outlet temperatures from an operating dry-well cooler.
After completion of the scram procedure the inspector reviewed the recorder strip charts from 16-1-TE-107 and 108. These charts indi-cated a drywell temperature between 135 and 140F before and shortly af ter the transients. The inspector and the SS discussed the ration-ale for not entering E0P-4 af ter the transient. The SS had author-
.ized work on the RTDs which supply signals to the SPDS average dry-well temperature, just prior to the scram.
He believed that the RTD work resulted in the SPDS alarm and indication of high temperature, since other-drywell parameters did not indicate leakage to the dry-well.
The SS asked a spare SRO to review the drywell temperature and to take action to restore the RTDs if necessary.
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The inspector discussed the use of drywell instruments with the spare
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. SR0 and the SS.
Both stated that they had not used the installed panel indication to determine drywell temperature. The spare SRO did call up another SPDS screen to monitor all the drywell RTD inputs, this indicated that several RTDs read high and caused the high aver-age temperature. The spare SRO also calculated the average tempera-ture across a drywell cooler, which showed a drywell average temper-ature less than 135F. The spare SR0 also directed I&C to restore the RTDs, and this caused the SPDS alarm to clear. The spare SR0 and the SS used this information to determine that the temperature was
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actually below the 135F entry condition.
In a technical sense the inspector found these actions acceptable, j
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The inspector discussed -the event with the Assistant Operations Superintendent and Resident Manager.
The inspector noted that there were no actual conditions which required actions by the operators and that no' adverse actions had been taken by the operators. However, it was not clear that the operators would have reacted to such a condi-tion. Accordingly, the following documents the inspector's concerns
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arid NYPA's commitments for corrective actions:
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The SS had authorized work on the drywell RTDs that had an
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effect on the SPDS average drywell temperature alarm point.
The
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SS did not review the next best. instrumentation specified by l
E0P-4 (16-1-TE-107 and 108) to ensure adequate indication.
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ther, no procedure existed to identify, to the operators, ' avail-
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able instruments that backup SPDS. The inspector considered that this information would be helpful to the operators on the loss
of a specific SPDS function due to work activities or following
loss of SPDS.
NYPA committed to reviewing the need for a pro-
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cedure to identify the instrumentation installed to backup SPDS
data.
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The installed drywell temperature recorders (16-1-TE-107 and
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108) indicated average drywell temperatures above the E0P-4 entry condition during normal full power operation.
NYPA com-l mitted to reviewing the RTO locations and acceptability of the readings of these instruments.
The inspector noticed a heavy reliance by the SS and spare SR0
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NYPA committed to provide simulator training on E0Ps i
with SPDS inoperable.
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The inspector considered these concerns to represent an unresolved
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item. Unresolved Item 89-12-01.
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On. January 12, the operations staff performed well while backseating MOVs to isolated drywell leakage.
On January 7, the control room
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received a high particulate radiation alarm on the inservice contain-f ment air monitor (CAM). In'the several days following this alarm the operations staff observed a slow increase in unidentified drywell leakage.
During this period radiation protection technicians increased the alarm setpoint on the CAM several times.
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operations and technical services personnel began preparations for F
backseating valves with the potential for packing leakage.
Operators noticed a substantial increase in the unidentified drywell leak rate during the day shif t on January 12. The SS directed deter-mination of the leak ' rate at 3:00 p.m. (one hour before the normal
determination). The calculated leakage at that time was 4.06 gpm; the leakage had been 1.97 gpm at 8:00 a.m.
The SS entered TS LC0 as 3.6,0.3, because of the increase in unidentified leakage of greater than. 2 gpm in - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This required the identification of the leakage source within four hours or a plant shutdown.
00S0-27 provided the administrative controls for backseating MOVs and a list of the valves that could be backseated.
Based on this list i
operations and technical support identified eight valves to be back-seated.
The SS coordinated the backseating of these valves, while monitoring drywell pressure.
NYPA backseated 29 MOV-102 (reactor head vent), 12 MOV-15 (RWCU inboard supply isolation), 23 MOV-15 (HPCI inboard steam supply isolation),13 MOV-15 (RCIC inboard steam supply isolation), 2 MOV-43 A and B (reactor recirculation pump suc-tion isolation), and 2 MOV-53A (reactor recirculation pump A dis-charge isolation). Backseating of -2 MOV-53A caused drywell pressure
.to decrease from 1.8 psig to approximately 1.75 psig within five minutes, This indicated isolation of the leak from the packing of-that 26 inch Anchor Darling double disc gate valve.
Leakage stab-
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ilized at 1.95 gpm by the 8:00 p.m. calculation.
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Three of the backseated valves were PCIVs (12 MOV-15, 23-MOV-15 and
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13 MOV-15).
The inspector questioned if these valves would meet their required TS closure times.
The operations department then i.
l reviewed the last closure time data on these valves and calculated a closure time for the backseated condition.
The calculated time assumed that the normal closing time started with the valve at 95%
open based on the setting of the open limit switch and linearly extrapolated the backseat condition-(100%) closure time. The calcu-lation indicted the valves would not exceed the TS closure times.
The inspector requested that NYPA determine the actual closure times during the next valve operation. On January 18, NYPA timed 23 MOV-15 closed during the monthly HPCI operability test and on January 20 they timed 13 MOV-15.
In both cases the valves closed within the TS required time and faster than the calculated 100% travel time.
MOV-15 isolated during the reactor vessel water level transient fol-lowing the January 19 reactor scram and did not get timed.
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The inspector. reviewed 0050-27 - and MP-59,20, the procedure for elec-trically backseating MOVs.
He determined that these procedures ade-quately controlled the backseatirg operation.
The inspector identi-fied two deficiencies with 0050-27. These are discussed below along with NYPA's commitment for corrective action.
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While extrapolation to the.100% valve stroke closure time for
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these valves yielded adequate results, the inspector noted that the ' procedure did not require this calculation prior to back-
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seating a valve. The inspector noted that this should be done
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first to ensure that backseating a valve does not cause a TS closure limit to be exceeded. NYPA committed to incorporating this aspect into the procedure.
The inspector determined that the procedure did not require tim-
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ing of a backseated valve during the-next valve operation. NYPA committed to incorporate the closure time determination into the procedure.
The inspector planned to review these changes in a subsequent report, F-1.
c.
NYPA did not identify or evaluate a temporary modification performed on both RCIC room fire doors, Specifically, the design of these fire doors required them to be normally open with an automatic closure
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device to actuate in case of a fire.
WR70560 removed the fusible
. link from the automatic closure device to allow replacement on December 4.
NYPA allowed the doors to close but taped the latches so-that the door would not latch in the closed position.
Each door had a special condition tag on it that stated "Do not leave door open without a fire watch present, leave latch taped".
The inspector questioned the operations superintendent about the operability of the fire door with the latch taped.
The inspector
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received a copy of a Nuclear Safety Evaluation, SE-81-018, dated July 17,1981, completed for the modification that installed the automatic closure system. The evaluation stated that the doors need to be open normally to allow blowdown of steam from a potential-RCIC steam line break. Further, the automatic closure allowed the door to shut and latch to act as a fire barrier when required. The safety evaluation also discussed an interim solution, used in 1981, of tap-ing the. door latch, but did not approve that condition.
The inspec-tor identified several concerns during his review of this condition and discussed them below along with NYPA's commitment to take correc-tive actio m
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Operators did not identify the need to use a temporary modifica-
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tion when removing the fusible links, and taping the door latches. NYPA committed to providing guidance to the operators
on such situations.
NYPA in effect performed a temporary modification to the RCIC
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doors without a-safety evaluation.
The special condition tag indicated the condition of the door, but did not require evalua-tion of the reasons for the door being open.
Based on - the inspector's concerns, NYPA perforn,ed a temporary modification to
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document and evaluate the condition of the door.
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The PTR sheet that controlled the special condition tag on the
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door did not state the specific wording on the tags.
Further, operators added the tags to a previously open PTR with no indi-cation as to the work request required to be cleared to allow clearance - of the tags.
NYPA committed to provide additional training to operators on the need to ensure that additions to PTRs are properly documented.
The inspector considered these concerns represented an unresolved item pending review of the temporary modification processed for the RCIC doors during a subsequent inspection. Unresolved Item 89-12-02.
d.
The inspector reviewed E0P Inspection. Report 88-200 concerns.
The summary of results section of this report listed seven specific con-cerns with NYPA's E0Ps. The resident inspector had assigned an unre-solved item number to each of these issues. NYPA responded to these concerns by letter dated September 30, 1988.
Of these seven items, the inspector closed' 88-00-07 previously and closed 88-00-02 and-88-00-03 in Sections 1 and 2 below.
The inspector addressed the remaining open items in Section 3. below.
1.
(Closed) Unresolved Item 88-00-02; Plant process computer set-points did not agree with emergency operating procedure (EOP)
entry conditions.
NYPA resolved this item by removing the old GEPAC process computer during 1989. The new EPIC computer sup-plies process computer and SPOS functions.
The process computer lists alarmed variables on the alarm screen.
Alarms associated with SPDS (i.e.,
E0P entry condi-tions) alarm on the process computer at the same setpoint as on the SPDS screen.
The console has a red light which indicates the. presence of an SPOS alarm.
This light informs operators that they should call up the SPDS screen, to check for other alarmed variables and E0P entry conditions.
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This item also tracked outstanding validation comments concern-
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ing' suppression pool level measurement methodology.. NYPA resolved concern by incorporating a change to E0P-4, Primary
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Containment Control.
This revision incorporated suppression pool; high/ low level entry conditions in terms' of feet of water above the torus bottom invert. This made the units of the entry.
condition consistent with the SPDS alarmed entry conditions.
The inspector closed this item.
2.
(Closed) Unresolved Item 88-00-03; The team identified several-
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instances of lack of ' equipment or information that could affect
the performance of the E0Ps.
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NYPA revised containment venting operation procedure
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F-A0P-35 to include the needed pressure curves for the different drywell pressure instruments.
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NYPA has committed during discussions with the inspector to
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install modification F1-88-116 before the end of 199u.
This~ modification allows for permanent drain connections near. the' control rod drive hydraulic control units. These connections would be used to allow venting of a control rod withdraw header to the floor drain system, in the event of a failure to scram.
NYPA made appropriate changes to prevent confusion when
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synchronizing emergency diesel generators from the remote panel.
NYPA provided. additional training to operators on the way to enter a remote shutdown panel. Near each remote shutdown panel is a shutdown equipment box, which can be opened with the operator's normal operations department key.
Inside the box is the key for the remote shutdown panel.
The inspector closed this item.
3.
(Closed) Unresolved Item-88-00-01, 88-00-04, 88-00-05, and 88-00-06; These'four issues require further NRC review. The NRC planned this review after NYPA implements EPG Revision 4.
NYPA committed - by letter dated December 20, 1989 to implement this
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revision by the end of the 1990 refueling outage. The inspector administratively closed thsse items and planned to track final resolution with a new unresolved item. Unresolved Item 89-12-03
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(Closed) Unresolved - Item 89-09-01:
NYPA clarified their policy for SR0s in the control-room.
The inspector reviewed the changes that -
NYPA made to 0050-1.
These changes included defining the SS's.com-mand __ decision responsibility and authority, which includes ensuring
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plant operation within the constraints of the license.
Further defined is control _ room command, which includes being attentive to duty and responsible for immediate actions required during a trans-ient or accident condition. The SS normally fulfills both functions.
However, the ASS may perform the control _ room command function when assigned by the SS. NYPA also addressed the issue of where the SRO with control room command should be located.
The inspector found that these actions resolved this item.
1.1 Inspector Assessment The operations staff performed well during the reactor scram and dry-well MOV backseating operations. Two problems indicated deficiencies in the area of E0Ps.
First, the failure to utilize the installed panel indication to determine average drywell temperature indicated a need-for operators to better monitor installed instrumentation.
Second,- NYPA did not identify that these instruments indicated-an average drywell temperature above the E0P entry condition at full power operation. Although these problems had not adversely affected reactor operations, the inspector concluded that the -problems could potentially hamper the operators' response to an actual escalated drywell temperature.
Operators did not have a clear understanding of what type of condi-tion required a temporary modification and what could be handled by a protective tagout special condition tag, 2.
Radiological Protection a.
On January 1, Mr. George Vargo, PhD. replaced Mr. Eric Mulcahey as the Superintendent of the Radiological and Environmental Services Department.
Mr. Vargo was previously the Radiation Engineering Supervisor.
The inspector reviewed Mr. Vargo's qualifications and-found that he exceeded the requirements of ANSI 3.1-1981.
Mr. Mulcahey has taken a position as a Senior Technical Advisor to the Resident Manager on radiological matters, b.
The inspector routinely toured the radiologically controlled areas of the plant.
Radiological housekeeping was generally goo.....,,.....
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Surveillance t d Maintenance a.
The inspector concluded that NYPA maintenance personnel should have issued an OR when they knew of safety related snubbers located in the drywell past their rebuild service life. Specifically, during review and documentation of snubber service lives (see Section 1. below) a maintenance engineer identified eight snubbers past their seven year service life.
The engineer identified this on January 12-and informed the assistant maintenance superintendent (AMS).
Upon fur-ther review the AMS noted that the list included a new snubber installed in 1985 that, therefore, had not exceeded its service life of seven years.
Based on this he requested the engineer to perform additional reviews on the other seven snubbers to determine if they were also new in 1985.
On January 17 the engineer determined the seven snubbers had exceeded their rebuild requirements and informed the AMS. The maintenance department issued OR 90-017, stating that five snubbers inside the drywell had exceeded their rebuild dates, on January 19 at 8:30 p.m. after the reactor scram.
During NYPA management discussions following the January 19 scram, the inspector observed no discussion of the need to change out or evaluate these snubbers.
When NYPA discussed the maintenance that required a drywell entry, the snubbers did not get mentioned. NYPA changed out the snubbers in question prior to restart, and all passed subsequent operability testing.
The inspector discussed this event with the AMS and questioned why an OR did not get written until January 19. The AMS responded that he did not get a chance to review the list of snubbers provide to him on January 17 until January 19.
Then the AMS review of the seven snubbers identified that five of the safety related snubbers exceeded their rebuild dates.
The inspector concluded that due to the potential for these snubbers beyond their evaluated service lives to impact the operability of safety-related systems, the OR should have been initiated more promptly. On January 17, it appeared that the maintenance department knew that snubbers in the drywell required changeout or evaluation.
_ procedures require identificatior of such conditions to the NYPA operations department via an OR.
The appropriate OR did not get generated until after the reactor scrammed.
b.
The reactor scram on January 19 occurred because of a valving error during isolation of a feedwater control system reactor vessel water level instrument 06-LT-52C per calibration procedure ISP-3-4. An I&C technician began isolating tre instrument in order to calibrate the detector, but NYPA could not clearly determine the sequence of events from this point.
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E Specifically, :,ometime after shutting the low pressure isolation valve, the technician noticed that the valve had a packing leak.
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NYPA could not determine if the packing leak occurred before or after
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the technician opened the detector equalizing valve. NYPA did deter-j mine that the technician considered some action necessary to isolate the packing.
The technician contacted the test controller in the control room and discussed the methods of stopping the leak.
The controller told the technician to continue with the isolation. NYPA speculated two possible causes for the scram.
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The low pressure detector piping depressurized, because of the
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packing leak, before the technician opened the equalizing valve.
When the technician opened the equalizing valve, the pressure t
transient in the high pressure (variable leg) caused the spike seen by the RPS transmitters.
The packing leak developed after partially opening the equal-
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i izing valve, and the technician mistakenly opened the low pres-sure isolation valve.
1&C tried to recreate the pressure spikes by performing these possi-ble valving errors witF mactor pressure at 400 psi in hot shutdown.
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A scram condition did not occur during any of the testing, but NYPA observed pressure spikes of lower magnitudes at the RPS instruments for both possible causes.
NYPA stated that the magnitude of the pressure spikes was proportional to reactor pressure at the time of the valving error and that a scram condition would have resulted at rated pressure.
In order to prevent any such occurrence until the refueling outage shutdown, NYPA performed any testing that could cause a pressure transient on reactor water level instruments prior to startup. The inspector planned to review NYPA's corrective action when the LER documenting this event is issued.
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NYPA began development of plans to disassemble and repair the A and C RHR pumps during the '1990 refueling outage.
IST review of the dif-
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ferential pressure output of both pumps showed a decreasing trend.
The IST reference differential pressures approeched the technical specification differential pressure limit.
NYPA initially believed that internal damage or wear of the orifice used to provide flow indications would cause the degradation on both pumps.
On January 9, the orifice was removed for inspection. The inspector observed this process and verified a proper tagout in effect for maintenance.
Inspection of the orifice showed no damage or wear.
NYPA performed the IST surveillance ST-2A as part of the
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retest after the orifice reinsta11ation.
The data indicated further degradation of pump performance, with the C pump differential pres-sure being one psid above the technical specification limit.
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NYf'A declared the A SBGT train inoperable on January 19 due to high t-fan vibration identified during performance testing. This placed the plant in a _ seven day LCO' AS. The subsequent reactor scram and plant cooldown on January 19 placed the-plant in a 31 day LCO AS with A
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SB3T system inoperable.
This became a restart item because TS 3.0.D states that a mode change shall not be made unless the conditions of the LCO are met without reliance on the LCO AS. The inspector mon-itored the-maintenance and troubleshooting to correct the high vibra-tion. problem.
NYPA performed the maintenance properly.
This iracluded a balance weight change that resulted in reduced fan vibra-t.ons. NYPA documented the balance weight change via a modification.
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T-) minimize impact on unit restart the operations. department per-formed PHT on the A SBGT train once the work requests were closed,
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while -technical services completed the modification paperwork.
A leak test of an access cover that had been removed did not get con-c.ucted as part of the PMT. Af ter completing the modification docu-nentation, technical' services sent a memo to operations to inform l
them of final resolution to' the A SBGT fan high vibration. In addi-i. ton, the memo noted that a leak test of the access cover was needed.
The operations staff retested the fan in order to perform-the required leak test, and identified and corrected leakage from the cover.
Then the system was declared operational.
The inspector identified the following deficiency in NYPA's PMT pro-i gram; NYPA's commitments to correct this problem are also identified.
Adequate documentation of removal of the access cover existed;
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however, the initial review did not determine the proper retest.
This error appeared as an isolated case of inattention' to detail.
NYPA committed to ensuring that SR0s who_ determine PMT
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increase their attention to detail.
e.
Between December 1 and December 11, 1989, NYPA performed varicus troubleshooting evolutions on HPCI to determine the cause of numerous system isolations on high steam flow signals.
PORC review of - the troubleshooting results coupled with review of GE SIL 475, dated November 1988 identified an incorrect high steam flow isolation set-point.
The testing showed that the steam flow, during the system I
startup transient, approached and in several cases exceeded the TS-isolation setpoint of 106 inches of water d/p.
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PORC recommended submittal of an emergency TS change to increase the
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HPCI high steam flow isolation setpoint to 160 inches of water d/p.
SIL 475 provided NYPA with a method of determining a proper setting using actual test data.
The calculation required determination of actual-instrument differential pressure with the system running at
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steady state in the test mode with discharge pressure equivalent to
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1120 psig plus head losses to the reactor vessel. At that time the -
discharge pressure used was 1160 psig based on 40 psid of head loss as determined during'startup testing.
Subsequent to this NYPA determined system startup times approaching the FSAR. limit of 25 seconds to full flow during test and the ECCS analysis limit of 30 seconds to full flow with the injection valve (MOV-19) full open. PORC reviewed and approved a 10 CFR 50.59 safety evaluation [[::JAF-SE-149|JAF-SE-149]] to change the FSAR HPCI design basis actuation
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time from 25 seconds to 30 seconds. This increase in response time
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allows lower steam flows and reduced probability of high steam flow i
isolation.
Based on this evaluation NYPA also changed the ECCS analysis opening requirement for MOV-19.
Rather than requiring the system to be discharging at rated flow and pressure with MOV-19 full open in 30 seconds, the evaluation stated that MOV-19 would pass virtually full HPCI flow when the valve stroked open at least 50%.
On December 11, the NRC granted a temporary waiver of compliance to the setpoint requirements for the high steam flow isolation prior to expiration of the seven-day LCO action statement. NYPA completed the change to the high steam flow isolation setpoint on December 11.
i Further testing and analysis demonstrated HPCI reliability and NYPA declared the system operable on December 12.
The inspector observed subsequent testing with satisfactory results. NRR approved the final'
TS amendment on December 15, 1989.
This resolved item F-3' from Inspection Report 89-11.
f.
The inspector observed performance of ST-98, monthly EDG operability on December 15, 1989 and January 17, 1990.
During the December 15 surveillance, operators secured the A and C EDG due to a. low emerg-ency service water (ESW) flow detector alarm on the C EDG.
Normal ESW pump differential pressure and no indication of abnormal EDG
- temperatures existed which indicated normal cooling flow. When oper-ators first vented the flow detector piping they observed a brown sandy sludge. Operations contacted I&C to troubleshoot.
I&C deter-mined that ' the pressure dampening orifice in the instrument line snubber partially clogged. NYPA replaced these snubbers on the four I
EDGs with a design less susceptible to clogging.
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g.
In Inspection Report 89-09 the inspector identified during review of i
Unresolved Item 88-11-01, that procedural deficiencies existed in the
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daily shift check procedure (ST-400).
The inspector identified
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several other deficiencies with ST-40D during this period.
The inspector listed these items and NYPA's commitments for correction
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below.
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During review of the situation leading to the deenergization of
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two low water level initiation instruments for the core spray
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system (Unresolved Item 89-10-02), the incpector concluded that a lack of better guidamce in this procedure may have contributed to this event. TSs require daily. channel checks on instruments that provide initiation -and isolation signals.
However, the
tabular format of the log sheets did not provide the reviewer with an indication of what instruments affect what systems.
j NYPA has committed to reviewing this procedure to determine if a t
more meaningful format is appropriate.
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Based on the review of the drywell temperature issue (discussed
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in Section 1.a. above) the inspector noted that log sheets pro-vided.for instruments that provide backup to SPDS points did not identify the E0P entry conditions. Because of this the defici-ency in the drywell temperature instruments had not been pre-viously identified. NYPA committed to review this procedure for incorporation of such information.
- The inspector considered this - item unresolved.
Unresolved Item 89-12-04.
h, The inspector concluded that NYPA performed a high quality root cause analysis of the October 31, 1989 failure of the RCIC steam supply isolation valve (13 MOV-21).
H. K. Porter (Pearless) manufactured the 125 VDC compound wound,.72 horsepower motor- (serial number AY09120). The maintenance department superintendent completed this evaluation and documented.it in a memorandum dated '.wember - 13, 1989.
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NYPA documented this failure in LER 89-21-00.
1.
(0 pen) Unresolved Item 89-11-02; The inspector reviewed LER 89-022 dated December 7,1989 which documented < the November 7 inoperability
of various snubbers.
NYPA committed to maintain a record. copy for each snubber by the plant snubber number and a history page contain-ing dates for functional test, rebuild date, visual inspection, plant installation locations and storage period up to the last rebuild date.
From discussion with the Maintenance Engineer, NYPA is still in the process of establishing the history records required by TS 4.6.9 and committed to be completed by LER 89-022.
The inspector found the LER to be adequate and the commitments made appropriate.
This item remains unresolved until completion of this program and subsequent inspector review, i
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NYPA compl_eted testing of UPS MG set under-voltage relay 71UPS-1UUR during the 1989 maintenance outage._ This action closed item F-5 from
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Inspection Report 89-03.
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NYPA repaired the HPCI steam supply valve 13 MOV 14 and worked on the turbine seal nearest to the pump during the Fall 1989 maintenance outage, This reduced the leakage past MOV 14 and the turbine seals.
.These actions close item F-2 in Inspection Report 89-09.
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3.1 Inspector Assessment
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The failure of NYpA maintenance personnel to identify potentially significant snubber information to the operations department indi-cated that NYPA management should stress the need for good communi-Cations between_ departments.
.The inspector found the identification of the A SBGT fan high vibra-tion' and the root cause analysis of the RCIC valve motor failure commendable.
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4.
a.
(0 pen)^ Unresolved Item 89-11-03; A recent resident inspection report identified several emergency plan implementation questions. Based on i
discussions with the Region I emergency preparedness staff, these issues will be tracked using this currently open item.
These con-cerns deal with implementation issues and include:
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NYPA does not currently considers that loss of the process com-
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~ puter and therefore the safety parameter display system (SPDS)
constitutes an entry condition for an Unusual Event.
The o
'p inspector reviewed NYPA's emergency plan Implementation Proced-ure 2 (IAP-2), Classification of Emergency Conditions and NUREG 0654, Revision 1 to determine the need to enter the emergency l-
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ir plan based on such a loss. The NUREG listed a loss of indica-
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tions or alarms on process parameters causing a significant loss
of assessment capabilities with examples of loss of the plant computer, SPDS or all meteorological instrumentation as possible
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reasons for entering such a condition.
The wording of NYPA's
Unusual Event Category 10, loss of' monitors, alarms and com-puters, equalled that of the NUREG, but did not specify loss of the plant process computer, SPDS or meteorological instrumenta-
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tion.
The inspector discussed this with the Resident Manager
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L who committed to reviewing this situation.
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Report 89-09,. Section 2.d. stated that their interpretation was that an Unusual Event vould be declared once a plant shutdown required by Technical Specification begins (i.e., control rod j
insertion).
NRC-had questioned whether the declaration should
be made when the shutdown action statement is entered or when (
the shutdown begins.
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restore reactor vessel water level as an entry condition for an
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Unusual Event.
5.-
Security l
a.
On January 12, the inspector walked down the protected area fence.
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In several instances he noted snow banks that would obstruct the view of secur_ity personnel as they drove around the parameter..The inspector notified the security supervisor of this finding..NYPA removed the snow banks and made security personnel aware of the need to have.this. done -in the future and the existing procedures specify-ing prompt removal.
b.
On January 8, NYPA switched the normal access to the site to the pre-vious secondary access.
This included making modifications to install a new vehicle access gate near the new primary access.
This action should allow for more efficient processing of personnel and vehicles prior to entry to the protected area.
c,
.(Closed) TI 2515/104 - Inspection of Fitness for Duty (FFD) Training Program:
The inspector attended NYPA FFD training sessions for general employees and supervisory personnel.
Both training sessions also included ' escort training.
The FFD program training presented sufficient detail: to give NYPA employees an understanding of their required responsibilities to comply with the 10 CFR Part 26 Fitness for Duty Rule. NYPA supervisory personnel answered several questions that the inspector had on the FFD Program.
It appeared to the inspector that NYPA's FFD Program complied with 10 CFR 26 by January 3, 1990.
This TI is closed.
6.
Engineering and Technical Support a.
As a result of the January 19 scram, the inspector questioned the response time-of the RPS water level instruments. Rosemount manufac-tures the 1153 detector and trip unit without any time delay in response to a short duration signal. Because of this, a spike such i
as the one that caused the scram can cause the tripping of these detectors when an actual trip condition does not exist. The inspec-tor knew of other utilities that had modified Rosemount transmitters to provide dampening on spike signals. NYPA committed to review this situation. The inspector planned to review their determination in a subsequent report, F-I (',
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The inspector reviewed the process NYPA usec to establish the new HPCI high steam flow isolation setpoint and the determination that 50% open. on MOV-19 would allow full flow (sre Section 3.e above).
The inspector identified two questions.
--~.The inspector asked how NYPA determinect the stroke time for
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MOV-19 50% open.
Engineering calculation JAF 89-046 determined
this by using motor speed and operator gear ratio and determined
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that 63% of actual valve stroke time equated to the 50% open position.
NYPA had incorporated this time into the surveillance test to calculate ~ actuation time and this appeared correct to the inspector.
The inspector asked NYPA for engineering justification for the
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statement that full flow could be achieved with MOV-19 50% open.
. NYPA then performed calculation [[::JAF-89-047|JAF-89-047]], dated December 14 that showed that with the valve 20% open the pressure drop r
across the -valve would be 6 psid.
The system design provided-for headloss of 53 psid-at full flow conditions.
The headloss measured during startup testing was 40 psid. Accordingly, NYPA concluded that with MOV-19 partially open, HPCI would achieve full flow against the increased headioss of 46 psid.
The inspector found NYPA's responses acceptable.
The inspector identified one deficiency during review of this issue.
TS 4.5 C.1 requires-testing of HPCI at rated flow with - a dis-
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charge pressure equal' to a system head with reactor pressure of 1120 psig.
Safety evaluation [[::JAF-SE-149|JAF-SE-149]] (discussed above in
Section 3.e) assumed full flow with the discharge valve at 50%
open.
The inspector. questioned the pump discharge pressure
acceptance criteria (1160 psid) for the test performed on
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December 14 in that it was less than 1120 psig plus. the increased headlo~s of 46' psid.
The inspector informed the s
Operations Superintendent and the Technical Services Mechanical Engineering Supervisor of this discrepancy.
Technical services and operations review determined that ST-4N had not been revised to take into account the increased pressure-drop through the 50% -open MOV-19. Technical services requested that the HPCI pump discharge pressure, in ST-4N, be increased to 1175 psig. This accounted for the entire 53 psid headloss from
'the initial design basis.
Operations changed the acceptance criteria for pump discharge pressure to 1180 psig, a conserva-tive value, and tested HPCI satisfactorily under the new test conditions on January 18, 1990.
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The design review did not adequately verify the affect this calcula-tion had on the technical specification surveillance test. This is a violation of 10 CFR 50, Appendix B, Criteria III, Design Control, which requires that the design basis be correctly translated into procedures.
NYPA did not receive a Notice of Violation because it took adequate corrective actions prior to the enc of the inspection report period.
NRC Enforcement Policy, 10 CFR Fart 2, Appendix C,Section V.C allows non-citing of isolated Severity Level V violations
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that are corrected before -the end of the inspecticn period. Assign-
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ment of an open item number identifies this non-cited violation solely for tracking purposes. NCV 89-12-06
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c.
(Closed) Unresolved Item 88-11-01:
The inspector reviewed NYPA's calculations and implementation of a. primary containment leakage acceptance criteria to meet TS 4.7. A.3.
NYPA performed calculation JAF 88-018 to determine the leakrate from the primary containment, The inspector. reviewed the several methods used to determine an acceptable criteria. NYPA judged the best method to be using drywell and torus pressures over periods of time when addit ons of nitrogen or venting were not taking place.
During these periods calculated leakrates varied from about 7000 SCFD to 19000 SCFD, and _ NYPA estab-
- lished a 25000 SCFD limit.
Revision 39 to -ST-40D adequately incor-porated this acceptance criteria.
The inspector agreed that this approach was acceptable.
This item is closed, d.
(Closed)
Unresolved Item 89-80-16:
NYPA performed Modification F1-89-129 to supply a safety related nitrogen supply to the reactor building closed cooling water (RBCCW) system PCIVs. The modification installation was reviewed in Section 2.2.(2) of Inspection Report 89-20. Installation of the nitrogen source to these vc1ves resolved this item..
L e.
(Closed) Violation 89-80-05:
The SSFI team identified two design l
control inadequacies. Those issues dealt with improper plugging of emergency diesel generator room floor drains and the improper design basis for safety related switchgear enclosure air conditioning units.
As discussed in Section 6.F of Inspection Report 89-09, the inspector reviewed the licensee event reports issued on these two instances and found that NYPA adequately resolved these issues.
NYPA agreed with the violation in their response dated September 21, 1989.
The inspector found the response adequate to resolve ' future concerns with design control.
The inspector closed
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this item.
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(Closed) Unresolved Item 89-80-07:
NYPA placed alarm annunciator pro:edures at the emergency diesel generator (EDG) local control panels. The inspector found these procedures adequate to address the alarm conditions.
This item is closed.
g.
During the 1989 maintenance outage NYPA installed red collars around the two ADS logic test switches on the main control room panel. Com-pleted as part of the control room upgrade, this should help prevent
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inadvertent SRV opening. when testing ADS.
This action closed item F-1 from Inspection Report 89-03.
h.
NYPA submitted a proposed Technical Specification amendment, dated December 20, 1989 to change three existing SRV performance limits.
Fi rst, NYPA requested allowance for plant operation with two, vice the currently allowed one, SRVs inoperable.
Second, NYPA requested approval of a. single setpoint for all the SRVs.
Thirdly, NYPA requested increase.of the SRV setpoint tolerance from +/- one percent to +/- three percent.
Submittal of this amendment closed item F-8 from Inspection Report 89-02, 1.
The inspector reviewed the following Licensee Event Reports and found that the information and corrective actions adequate. (The informa-tion in parenthesis includes the event date and the SALP functional area to which the report applies.)
LER 89-14-00 (August 5,1989, Maintenance); High pressure cool-
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ant injection - (HPCI) inoperable due to water in the turbine lubricating oil.
LER 89-15-01, 89-15-00 (September 18, 1989, Maintenance); Nine
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air operated RBCCW PCIVs exhibited operational deficiencies.
. LER 89-16-00 (September 18, 1989, Operations); Reactor scram
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while shut down, due to-transfer of reactor protection system (RPS) power' supplies.
LER 89-17-00 (September 20, 1989, Operations); Group II contain-
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ment isolation due to pulling fuses is part of a tagout.
LER 89-20-00 (November 5, 1989, Maintenance); Reactor scram due
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to electro-hydraulic control (EHC) system malfunction.
LER 89-23-00 (November 12, 1989, Operations); Reactor scram dur-
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ing SRV testing.
LER 89-24-00 (November 29, 1989, Surveillance); RCIC isolation
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during surveillance testing.
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6.1 Safety Assessment y-Technical services and operations needed to be more sensitive to the impact of design basis changes to ensure that changes such as the revised HPCI head loss get reflected in operating procedures and surveillance tests.
7.
Safety Assessment / Quality Verification a.
On January 1, Mr. George Tasick replaced Mr. Richard Patch as the site Quality Assurance Superintendent. Prior to this, Mr. Tasick had been the Quality Control Supervisor.
The inspector reviewed Mr. Tasick's qualifications for his new position and found them to be acceptable. Mr. Patch has assumed a technical advisory position.
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b.
The inspector identified some examples of not managing overtime dur-I ing the 1989 mini outage. Plant standing order (PS0 #26) details the administrative controls for controlling overtime within Technical Specificatien requirements. By review of time records, the inspector noted several examples where plant staff exceeded the "72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a
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seven day period" overtime restriction.
Examples were noted with
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supervisors in the maintenance, quality assurance, radiological and environmental services, and operations departments.
The overtime policy as detailed in PS0 #26 did not appear to be strictly enforced.
The Operations Superintendent appeared to have adequate review of overtime use by licensed operators and received i
verbal authorization from the Resident Manager for exceeding.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> on a case basis.
However, further discussion with the Resident Manager indicated that all departments did not request authorization prior to exceeding overtime requirements.
PS0 #26, Section 7.3 specifies that any deviation of overtime requirements be authorized via memorandum by the Resident Manager or the Superintendent of Power.
The inspector, determined that NYPA had not strictly enforced overtime guidelines, primarily regarding maintenance supervision.
However, coverage of supervisors under NYPA controls and NRC guide-lines was not completely clear.
NYPA committed to enhancing their controls on overtime during the 1990 refueling outage.
The inspector considered the problems iden-tified in the overtime policy controls and concerns about upper management's review of individual overtime usage unresolved. Unre-solved Item 89-12-06.
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8.
Exit Interview
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At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and find-ings. In addition, at the end of the period, the inspectors met with NYPA representatives and summarized the scope and findings of.the inspection as
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described in this report.
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Based on the NRC Region I review of this report and discussions' held with NYPA representatives during the exit meeting, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
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L APPENDIX A Acronyms ALARA As Low as Reasonably Achievable
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ASS
' Assistant Shift Supervisor
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EDG Emergency Diesel Generator
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E0P Emergency Operating Procedures
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EPG Emergency Procedure Guideline
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HCV Hydraulic Control Unit-
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HPCI High Pressure Coolant Injection System
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~FFD Fitness for-Duty
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IGSCC Inter Granular Stress Corrosion Cracking-
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I&C Instrumentation and Control
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ISI'
In-Service Inspection
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LIST-In-Service Testing
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LCO Limiting Condition for Operation
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LC0AS LC0 Action. Statement
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LER Licensee Event Report
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LLRT Local Leak Rate Test
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MOV Motor Operated Valve
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ODSO Operations Department Standing Order
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OR-Occurrence Report
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PCIV Primary Containment Isolation Valve
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PM Preventive Maintenance
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.PMT.
Post Maintenance Testing
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PORC'
Plant Operations Review Committee
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PTR Protective Tagout Request
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QA Quality Assurance
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QC.
Quality Control
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RCIC Reactor Core ~ Isolation Cooling System
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RHR Residual Heat' Removal System
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RPS Reactor Protection System.
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RTD Resistance Temperature Detector
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RWCU Reactor Water Cleanup System
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RWP.
Radiation Wc-k Permit
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SALP Systematic Assessment of Licensee Performance
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SBGT-Standby Gas Treatment System
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SPDS Safety Parameter Display System
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SRC.
Safety Review Committee
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SR0 Senior Reactor Operator
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SS Shift Supervisor
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SSFI Safety System Functional Inspection
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ST Surveillance Test
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STA Shif t Technical Advisor
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TAF Top of Active. Fuel
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TS Technical Specification
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