IR 05000333/1989010
| ML19332F200 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/28/1989 |
| From: | Meyer G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19332F197 | List: |
| References | |
| 50-333-89-10, NUDOCS 8912140028 | |
| Download: ML19332F200 (16) | |
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.V.S. NVCLEAR REGULATORY COMMISSION
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Region I Report No.:
50-333/89-10 a
Docket No.:
50-333 License No.:
Licensee:
New York Power Authority Post Office Box 41 Lycoming, New York 13093 Facility:
James A. FitzPatrick Nuclear Power Plant
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Location:
Scriba, New York Dates:
September 7, 1989 through October 7, 1989 Inspectors:
W. Schmidt, Senior Resident Inspector R. Plasse, Jr., Resident Inspector R B rkley, Re cto Engineer, DRP
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Approved by:
ll'7/X-f>7 i
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g Glenn W. Meyer, Chief /' No.1B r
date Reactor Projects Sectibt Division of Reactor Projects-Inspection Summary:
This inspection report discusses routine 6,1d reactive inspections during day and backshift hours of plant activities including; plant opcrations, radiation
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protection, surveillance and maintenance, security, engineering and technical support, and quality assurance.
Results:
Two unresolved items and two violations were identified. A Table of Contents follows, i
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TABLE OF CONTENTS Page-1.
Operations-(MC. 71707,71701,93702).........................
a.
Failure to have Two Instrument Trip Systems Operable
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for.the Core Spray System during CRD' Replacement,.
F Unresolved Item' 89-10-01.............................
b.
Loss of Shutdown Cooling due to an Improper Isolation and Failure to Follow a. Surveillance Procedure, Violation 89-10-02...................................
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Full Reactor Scram While Shutdown due to transfer of RPS Busses...........................................-
1.1 - Safety Assessment......................................
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2.
Radiological Protection (MC 71707)..........................
-4 3.
Surveillance and Maintenance (MC 61726, 62703 and 92702)....
a.
-Replacement of GE Breaker Prop Springs (Resolves
. Item F-3 -IR 89-09.).................................
b.
Review of Corrective Measures to Prevent Inadvertent Removal of Tags (Resolves. Item F-8 - IR 89-03).......
c.
Replacement of-Limitorque Motor Operated Valve Torque
_ Switches Made of Me1 amine............................
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d.
Review of B and D Emergency Diesel Generator Post-i Modification and Surveillance Testing................
e.
Review of A and C Emergency Diesel Generator Post-
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Modification Testing, Concerns with.the Post-
Modification Testing Program, Unresolved
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Item 89-10-03........................................
f.
Review of Work Activities Prior to Startup.............
g.
Completion on Post-Modification Testing on the Addition of a Safety Related Air supply to the Operators for the Containment Isolation Valves for Water Supplied to the Drywell Coolers (Resolves
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Item F-4 from SSFI Inspection Report 89-80)..........
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3.1 Safety Assessment......................................
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'4.
' Emerge'ncy Preparedness.(MC 71707)...........................
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Security'(MC 71707)..........-...............................
6.
Engineering-and Technical Support (MC 92702)................-
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(Closed) Unresolved Item 89-09-05: Acceptability of Emergency Diesel Generator Testing........-.-..........
b.
Review of Program toLReplace Limitorque Motor Operated Valve Actuator Torque Switches made of Melamine......
c.'
Review of the Plant Standing Order 56, Replacement of
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Electrical Fuses (Resolves Item F-1 from SSFI Inspection Report 89-80).............................
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d.
LER 88-14',. Revision 1 - Potential Common Cause Failure of GE MagneBlast 4.76-AMH Circuit Breakers due to
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' Refurbishment Procedure Deficiencies.................
e.
Review of Canceled Modification to~ Replace the B Core Spray System Injection Piping........................
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g 6.1 Safety Assessment......................................
'7.
-Quality Assurance (71707)...................................
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Failure'to Complete Technical Specification _ Required Audits - Violation 89-10-04..........................
b.
Review of Corrective Action to Prevent Missed'ISI Inspections..........................................
7.I' Safety Assessment.......................................
8.
Exit Interview..............................................
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DETAILS 1.
Operations The plant operated at 100% power until September 16, when a normal shut-down was conducted.
The shutdown ended a FitzPatrick run of 272 days and
' began a planned maintenance outage.
The outage was completed on October 5,. when a reactor startup was commenced.
At the end of the period, the reactor was operating at 15% power.
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a.
-Based on his review of an incident identified. by a Shift Supervisor (SS),_ the -inspector questioned the - acceptability of the instrumenta-
- tion supporting the core spray initiation function during control' rod '
drive.(CRD) replacement work' on September 19.
Specifically, at
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3:15 a.m. on September 19, the SS identified :that the A core spray (CS) subsystem was inoperable based on inadequate instrumentation for
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This was determined because only.one of the two instrument trip systems had been in service (each
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trip system has two instrument channels). - A CS had been required to be operable to support. reJently completed control rod drive (CRD)
replacement work because of the potential for draining the reactor-r vessel.
Had an actual low vessel level condition occurred,,the oper-able trip system would have been able to initiate A CS if the instru-mentation functioned properly; however, any failure within the trip system would have prevented an automatic CS initiation.
The SS
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placed a ' hold on CRD replacement operations, initiated OR-168, and made the required 10 CFR 50.72 notification. The inspector concluded i
that ~ the SS's actions were acceptable and represented -good recogni-
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tion of an unsatisfactory plant condition.
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One instrument trip system had been removed from service on l
September 18 to support modification work.
A protective tagout i
request (PTR) had been implemented for modification F1-87-105, High Pressure Coolant.. Injection Analog Transmitter Trip System (HPCI ATTS)
Level 8 Instrument Replacement. The PTR deenergized ATTS panel 09-95-and thus removed from service low reactor water level trip units
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2-3-272A and 2-3-2720, which are part -of the Division I initiation
logic (trip system) for core spray. The Division II initiation logic was not deenergized and remained in service.
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i The Plant Operations Review Committee (PORC) reviewed this event on September 19, and determined that the CRD replacement work should not
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have been done while having only one trip system for core spray.
However, PORC determined that the Technical Specifications (TS)
had not been violated as a result of this incident and that the 50.72
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notification should be revoked.
PORC concluded that the plant com-plied with TS Table _3.2-2, note 1, which specifies that both trip systems be operable or the plant be placed in the cold condition, because. the plant was already in the cold condition.
Subsequently, NYPA removed the 09-95 panel isolation and restored the Division I trip system to operability prior to recommencing CRD work.
This reestablished A CS operability with two operable trip systems.
The inspector disagreed with NYPA's determination. During conditions with the potential to drain the vessel, e.g.,
CRD replacements, at least one emergency core cooling subsystem must be operable per TS 3.5.F.2.
To consider A CS operable necessitated that its instrumen-tation also meet operability requirements (both trip systems oper-able). It was. incompatible to have a condition with the potential to drain the vessel, thus requiring a ECCS cooling subsystem, and to have the cold condition of the reactor satisfy the requirements for the applicable instrumentation.
This item will remain unresolved pending further NRC staf f review.
(UNR 89-10-01)
The inspector determined that there were at least two opportunities for other operations department personnel, including licensed oper-ators and senior operators, to have identified this condition prior to the SS determining the condition was improper.
A control room annunciator alerting the operators to the loss of power to the 09-95 panel alarmed when the power was removed. This alarm was observed by the inspector to be lit on the morning of September 19.
The alarm response procedure directs the operator to consult the Technical Specifications for any Limiting Conditions for Operation (TS LCO).
The operator's daily surveillance test ST-400, required that readings on these instruments be taken in all modes of operation. The reading for the instruments that day indicated that the-instruments were deenergized because of the applicable tagout.
The inspector noted that the TS requirements were not as clear as they could have been for conditions when the potential exists to drain the reactor vessel.
However, this shortcoming had been raised to NYPA in NRC Inspection Report 88-01, and again in Inspection Report 88-09. NYPA was reviewing a TS amendment to resnive this con-cern at the time of the incident.
b.
During the maintenance outage, shutdown cooling was lost twice for short periods of time.
Both incidents were due to operator errors.
In one case, the senior reactor operator who reviewed a tagout did not fully understand what would result when the tags were hung. In the second case, a licensed operator failed to properly verify pre-requisites to a surveillance procedure.
These instances were of L
minor safety significance and resulted in no damage to equipment.
However, the incidents indicated that during conditions other then normal operation, operators must be more careful to review the effect of their actions and procedure requirement _.
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r On September 20, a Group II containment 1 solation occurred, isolating shutdown cooling and reactor building ventilation.
This happened when an operator pulled fuses while establishing an isolation for maintenance.
The fuses were replaced and shutdown cooling was restored in five minutes.
On October 5, portions of ST-2M, ECCS Trip S!/ stem Bus Power Monitors Functional Test, were performed. This test verifies the operability of the undervoltage relays in the DC ECCS logic busses. The A logic bus was deenergized during the test by pulling the power supply fuses and the associated annunciator alarmed as designed. However, the A RHR pump, which was in the shutdown cooling mode, tripped unexpec-tedly.
The inspector determined later that test prerequisite step 4.4 specified that all ECCS systems be in their standby lineup during the test.
However, the RHR system was not in its standby lineup.
The inspector concluded that the operator made an incorrect procedure signoff regarding the lineup of the RHR system and did not identify the mistake prior to pulling the fuse. The inspector found the pro-cedure to be clear with adequate guidance.
The failure of the reactor operator to properly complete the prerequisites to ST-2M is a violation.
(VIO 89-10-02)
c.
The inspector reviewed the events leading to a full reactor scram on September 20, while the plant was shutdown. This occurred when both Reactor Protection System (RPS) busses were transferred - to their alternate power supplies within an eight second period of time. The A RPS bus was initially transferred, with the expected A channel half scram occurring. The half scram was reset and with no existing scram-signals, the B RPS was transferred.
This caused the expected B channel half scram. However, before the B channel could be reset, an A channel half scram occurred, causing a full scram.
l The A channel scram was apparently caused by spiking of the main steam line and offgas system radiation monitors. This type of spik-ing had been previously observed when these channels were repowered following bus transfers or power losses.
The inspector determined that the procedure for switching power supplies was followed by the operator.
The spiking of instruments as they repower was not addressed in the procedure nor at the shif t briefing.
This example emphasized the need for operators to be attentive to potential prob-l lems and to initiate appropriate cautions in procedures.
NYPA com-pleted a procedure change to identify this possibility. The inspec-tor concluded that the corrective action was appropriate and acceptable.
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l 1.1 Safety Assessment
NYPA did not take adequate action _ to correct and provide clear guid-ance on the issue of TS clarity with regard to ECCS systems required when conducting work with the potential for draining the reactor vessel. This issue was first raised in 1988, but neither a TS amend-ment nor a TS interpretation were in place.
The loss of shutdown cooling events discussed above indicated that operators should have a questioning attitude even when the plant is not operating.
Careful review of annunciators, logs, tagouts and surveillance tests remains important during shutdowns.
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Radiological-Protection The inspector toured the plant before, during and after the outage, i
including an inspection of the drywell. The inspector did not note any unacceptable conditions.
3.
Surveillance and Maintenance a.
During the outage, the Maintenance Department replaced the prop springs _ on safety-related Magne-Blast AMH-4.76 circuit breakers with greater than 900 cycles of inservice use. These corrective actions were in response to the prop spring failure on ' the B RHR service water pump breaker on August 16, documented in Inspection Report 89-09.
The failed-prop spring for that breaker had been through 1,625 cycles. The Peach Bottom Atomic Power Plant had reported such spring failures at high cycles and recommended chcnging out the
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springs at 2,000 cycles. Inspection of the removed prop. springs dur-ing this outage did not determine any signs of fatigue cracking.
NYPA is in the process of adding the prop spring replacement to the preventive maintenance program. The inspector found these corrective actions adequate to resolve item F-3 in Inspection Report 89-09.
b.
The inspector previously observed unauthorized removal of a tagged valve handwheel by maintenance personnel.
This was discussed in Inspection Report 89-03, Section 5.f.
The valve was tagged as an isolation boundary for valve maintenance.
A mechanic ' removed the handwheel because it interfered with staging needed to support the
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valve repair.
The Maintenance Superintendent discussed this event with appropriate maintenance personnel.
In addition, a procedure change to Work Activity Control Procedure (WACP) 10.1.2, Equipment and Personnel Protective Tagging was written to ensure maintenance personnel understood that they are prohibited from repositioning or
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working on any valve or component within isolation boundaries that has not been specifically authorized to be worked. In the event that a valve or component requires repositioning to support maintenance, the Shift Supervisor has to be informed to receive appropriate authorization.
The inspector reviewed this procedure change and included. resolved F-8 from Inspection Report 89-03.
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c.
During the maintenance outage, NYPA replaced melamine torque-switches (see 5.b) on three Limitorque valves:
27 MOV 117, 23 MOV 25 and 46 MOV IDIA. The inspector observed the replacement on 27 MOV 117.
The electricians involved were knowledgeable and a Quality Control inspector was observing the work. This work was properly conducted.
When 23 MOV 25 was-stroked closed electrically after the torque switch replacement, damage was done to the valve stem.
The valve stem threads were damaged because the new torque switch had not been properly installed.
The switch leads had been installed backwards which meant that on a closing stroke the torque switch never func-tioned. This allowed the valve operator to draw locked rotor current when driving the disc into the seat. -The valve was disassembled, a
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new stem was installed and the torque switch leads were reversed.
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The valve then passed the required M0V testing.
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d.
On September 28, the inspector observed surveillance test (ST)-98, l
Monthly Operability Determinations on the B and D EDGs.
The EDGs
were being run as part of post-modification testing and to demon-
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strate operability.
The voltage adjust and. speed control switches had been rewired to correct a human engineering deficiency by pro-viding consistent switch positioning for lower and raise positions.
The operator was properly using the newly revised precedure and understood the operation of the newly rewired switches. The test was properly conducted. The EDGs were later returned to operable status after replacement of a failed fuel oil pressure sensor.
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'The inspector reviewed an instance where a tagout was still in effect l
during operation of the A EDG.
This incident identified weaknesses in the way post-modification testing is controlled. Specifically, on e
October 2, the A EDG-was started in accordance with OP-22 to support f
air start motor lubricator valve setting adjustments.
The EDG was secured by the SS when it was determined that the mechanical governor was controlling EDG electrical output at 63 HZ and there was no remote voltage or speed control. There was no damage to the equip-ment as a result of this incident.
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Upon further review, the SS determined that the fuses for EDG voltage and speed control were removed to support modification F1-87-061, Human Engineering Discrepancies for Control Room Design Review.
The EDG switch-modification was completed on September 29, and the pro-tective tagout (PTR) was still in effect. The SS cleared the PTR and authorized reinstallation of the fuses. During reinstallation it was determined that the fuse rating was 30 amps and system drawings called for 15 amp fuses. This discrepancy was documented on a WR and is discussed in section 6.c.
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The inspector determined that after completion of the modification.
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the PTR for the fuses was transferred to an SS. The PTR should have been transferred to the responsible post-modification test super-visor, but could not be since he was not a qualified tag holder. The-test supervisor then signed off the preoperational test (PT) step which required him to clear the PTR. The PT requirement to clear the
tagout was not discussed nor was it given to the SS when he received
control of the PTR. The next step in the PT was to run the EDG to verify proper speed and voltage - control.
However, before this test could be conducted, the A EDG was started as described above for i
maintenance.
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l Several weaknesses were noted:
The EDG switch installation was one of several line items on the I
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PTR for the control room modification work.
Because there was no work request for the modification, the only method for the SS
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to identify the tags outstanding was to review the PTR log.
I This did not allow a quick method for operators to determine the I
operability of the system basert on the modification work i
outstanding.
The Operations Superintendent stated that reference tags should have been placed on the switches. With a reference tag the SS would have determined the problem prior to : tarting the machine.
Reference tags are a good practice.
However, by interviewing various operators, the inspector determined that they are not
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consistently used.
An example of their lack of use was noted during previous maintenance to the 115 KV systam when a refer-
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ence tag was not used to inform the operators that one 115 KV line was unavailable.
In addition, WACP 10.1.2 equipment and personnel tagging does not require reference tags for all PTR U-
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It is standard procedure to ensure that all post-modification
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testing is complete prior to declaring a system operable. How-ever,:there is no system that tracks a modification and the required post-modification testing other than the diligence of the PT supervisor. There is no system other than tagcuts which prevents Operations from operating equipment prior to work com-pletion and preoperational testing of the modifications.
j The Operations Superintendent stated to the inspector that his
position is that part of post-modification testing is to find
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equipment that has not been properly restored, such as fuses still being tagged. The inspector disagreed with this position.
Operations should be cognizant of the status of equipment, especially prior to determining that equipment is ready to be
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operated for testing or other reasons.
The work control system
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should readily provide to the SS the necessary information to determine when a system modification is complete and that com-y ponents. have been properly restored to support final post-work i
testing. The. system should be adequate to prevent inadvertent
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testing prior to completion of system modification.
- This is a concern since, if specific items that require retest are not being properly monitored during operation prior to PT, damage to equipment can. happen.
This item is unresolved pending inspector
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review of the adequacy of the work control system. (UNR 89-10-03)
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f.
The inspector observed NYPA work activities in preparation for the i
startup from the maintenance outage. Activities in the control room were maintained in a controlled manner with a minimum of distrac-tions. At the close of the inspection, NYPA was in the early stages l"
of completing surveillance F-0P-65, Rev. 44, " Plant Startup and Shutdown".
The inspector toured the control room and walked down the control room panels.
All questions that he raised regarding the status of annunciators were resolved satisfactorily.
The shift turnover was observed and appeared to be thorough and professional, g.
NYPA perfomed a modification, F1-89-129, to provide a safety related pneumatic supply to the containment isolation valve for the water supplied to the drywell unit coolers. The pneumatic supply is from the safety related nitrogen supply for the containment atomosphere dilution (CAD) system.
This modification resolved a concern raised during the SSFI, Inspection Report 89-80, section 4.7.5, dealing with the leakage of the previously non-safety related air supply to the valves. The modification was reviewed in NRC Inspection Report 89-20
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E and found _ acceptable.
These valves are normally open and require pneumatic pressure to shut and remain shut.- The inspector ' observed the post-modification testing and noted that all the acceptance cri-
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teria for the test were met, that the operator was knowledgeable of
.the test criteria and procedural steps, and the deficiency tags on
'the system were properly cleared prior to the test.
The inspector verified that the control room. drawings reflected the modification and noted that the operations personnel on shift were knowledgeable of the. modification program cycle.
Completion of this modification resolves item F-4 from SSFI Inspection Report 89-80.
3.1 Safety Assessment Maintenance Department action to replace prop springs on safety related breakers showed good perspective and initiative to prevent further failures. Good corrective action was taken to resolve the-concern of. removing or working on tagged equipment.
Maintenance observed was properly conducted. The installation of a
torque switch with-reversed leads causing damage to a HPCI _ valve
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warrants increased management attention to ensure that personnel pay
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closer attention to detail during preventive maintenance. The method of controlling modification testing does not seem to fit into.NYPA's method of controlling normal maintenance.
Because of this, it is confusing and can lead to the operation of equipment for which post-modification testing has not been completed. This can lead to equip-ment damage -if modifications have not been completed properly or supporting equipment has not been returned to service.
The surve111anca testing observed-was adequa+e except for the one noted instance of the failure to ensure prerequisite conditions as outlined in Section 1.b.
4.
Emergency-preparedness
During-routine inspection of this area, there were no noteworthy findings.
5.
Security The inspector walked the protected area fence and observed personnel and vehicle searches.
No-deficiencies were note,
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Engineering and Technical Support a.
(Closed) Unresolved Item 89-09-05. The acceptability of NYPA's test-ing of the-emergency diesel generators (EDGs) in parallel with the main _ generator was reviewed by the NRC staff. The reason for concern l
t was that the switchgear ratings were not adequate to give proper interrupting capacity if a postulated three phase bolted short were to occur when the EDGs were operating in parallel with the main generator.
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NYPA prepared a justification for continued operation in July,1989
which allowed the EDG and main generator to be operated in parallel during EDG testing.
This allowance was based on a study of the remote probability of such a three phase short occurring during EDG surveillance testing.
During NRC staff review, the statistics used to develop the proba-l bilistic assessment were questioned, as was the advisability of using
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probability as a basis for continued operation. In a conference call
on October 2, the NRC staff requested that NYPA provide a deter-l ministic rather than probabilistic analysis.
On October 3, NYPA
presented to the staff a position that the EDGs would be only tested i
in parallel with the 115 KV offsite power supply. In this situation, the switchgear has been analyzed to have sufficient interrupting
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capacity.
NYPA will continue to - analyze the system to determine final corrective action.
This approach was acceptable to the staff and this item is closed.
b.
In November 1988, NYPA received a 10 CFR 21 notification from the l
Limitorque Corporation involving motor operated valve actuator models
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SMB-000 and SMB-00.
This notification dealt with possible failures
of valve operator torque switches made with melamine as the insula-
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ting material. The failures had been noted due to shrinkage of the
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melamine causing switch cam binding.
This binding could cause the l
torque switch to stop valve movement before a full stroke could be l
accomplished.
Limitorque. recommended actions to replace the subject torque switches.
If replacement was not possible, Limitorque specified a
test for identifying binding.
To date, NYPA has identified through
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environmental qualification documentation review that there are none of the subject switches in the drywell.
Presently, there are three
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valves outside primary containment which have these switches installed and have not been tested. These normally closed valves are 10 MOV 21A and B,
the residual heat removal (RHR) system heat exchanger drain valves to the torus, and 10 MOV 33, the RHR outside
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containment head spray valve. These valves are of little safety con-cern since they are normally closed and neither the steam ccndensing mode nor head spray functions of RHR has been used since system pre-operational testing. NYPA currently plans on replacing all melamine torque switches during the 1990 refueling outage.. The inspector found this approach and the actions taken to date to be acceptable, v
c.
The inspector observed the proper use of Plant Standing Order (PS0)-
56, Replacement of Electrical Fuses.
This procedure had been gen-
- erated based on an SSFI Inspection Report 89-80, Section 4.2.2,. con-cern dealing with fuse control. On October 2, the SS determined that fuses that had been installed in the A and C EDG voltage and speed control circuits were not as per the electrical drawing (30 amps vs 15 amps). A work request was issued to Technical Services to address the defic M cy. Technical Services determined that the 15 amp rating was correct.
PS0-56 adequately addresses the concern raised by the SSFI and resolves item F-1 for Inspection Report 89-80.
d.
NYPA conducted an in-depth study of the events leading to the November 1988 failure of a safety-related Magne-Blast AMH 4.76 cir-cuit breaker to open.
This study is documented in revised LER 88-14-01. The initial LER stated that the reason for the failure was the misalignment of the breakers prop pin when it was reinstalled following refurbishment by GE. The study showed that the misalign-ment of the prop on the breaker cam shaf t was the cause of the fail-ure. This was determined during a visit to the GE breaker refurbish-ment facility. GE has taken corrective actions including procedure changes to ensure that the prop and prop pin are properly aligned after refurbishment, e.
NYPA had planned to replace the B core spray system injection piping during tnis maintenance outage.
The replacement had been a commit-ment made by NYPA since the welds in the piping had been overlayed in 1987, due to Integranular Stress Corrosion Cracking (IGSCC), and had not received full ultrasonic testing. On September 11, a Temporary Operation Procedure (T0P)-98 was approved by the PORC for use in maintaining reactor vessel level while the CS pipe was cut.
After this procedure was approved, the Operations Department questioned whether adequate core cooling could be maintained with the lowered level and the reactor vessel internals installed. This question had not been raised before because NYPA knew that they had already replaced the A CS piping and done so with lowered level. The dif-ference with the A CS piping replacement was that this was done dur-ing a refueling outage when the vessel internal were removed, thus allowing circulation through the core.
This modification did not receive thorough evaluation for the necessary plant conditions, and had to be cancelled. NYPA discussed the delay in the replacement of B CS. piping with the NRC staff.
NYPA was directed to fully inspect the three welds that had beem overlayed and to inspect one additional CS weld in the piping that would have been replaced.
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Additional welding was necessary on the three previous overlays to
allow for surface preparation. These welds were inspected and deemed
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by - NYPA to meet the requirements of NUREG 0313, Rev. 2, for weld overlay repairs. The other inspected weld showed no signs of IGSCC.
These results are documented in a NYPA submittal to the NRC in September 1989.
6.1 Safety Assessment NYPA responded quickly and properly to NRC concerns on EDG testing.
NYPA actions -to resolve problems identified with GE circuit breakers were very thorough. This was a credit to their maintenance organiza-tion and showed that they held GE properly responsible to -take cor-rective actions.
The inability to replace the B core spray injection piping during the
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outage raises some concerns to the inspector. First, the Plant Oper-ations Review Committee and the Resident Manager approved a procedure that would not ensure core cooling while cutting and welding the piping.
Second, the planning for this modification was not ade-quately reviewed in that it did not identify that the core internals would remain installed. The Operations Department identification of this potential deficiency was commendable.
NYPA should take action to ensure' that plant conditions are properly identified and can be established for each modification to be installed.
7.
Quality Assurance a.
A Corporate Management Audit Program (CMAp) audit of the facility's Quality Assurance (QA) program uncovered several audits that were not completed when scheduled by NYPA.
Four Technical Specification (TS) 6.5.2.8 required audit areas did not get documented review within the required time periods.
The site QA department documented those missed audits via Adverse Quality Condition Report 89-107.
NYPA's site Quality Assurance Instructions (QAI) 18.3 breaks each TS required audit down, if possible or necessary, into several more manageable parts. The audits that were not completed were (listed are the scheduled completion dates):
The conformance of operations with TS and other license condi-
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tions. This requirement was broken down into nine parts, seven of which were not completed (five due by January 89, one by February 89 and one by April 89). This audit was required to be periormed at least once every twelve months.
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The performance, training and qualifications of the entire
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facility. For this audit, review of the performance (March 89)
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and training (January 89) aspects were not documented.
This.
audit was also required on a twelve month interval.
The review of actions taken to rectify deficiencies (May 89).
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This audit was required at least once every six months.
The emergency plan and procedures (May 89).
This audit was
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required at.least once every twelve months.
These audits were required to be performed by site QA under the cognizance of the offsite Safety Peview Committee (SRC).
NYPA pro-cedures for controlling these audits state that a program schedule is.
to be established by the site QA Superintendent.
This schedule should then receive-review and concurrence by the corporate QA man-ager and the SRC chairman.
Further, quarterly schedule updates were required so that the requirement could be tracked. Also, the SRC, as a matter of standard practice, should have received a presentation on completed audits, although this is not a TS requirement.
The inspector discussed the missed audits and the program with the site' QA Superintendent (QAS).
He issued a schedule showing comple-tion of the overdue audits by October 31. The QAS felt that, while the audits were not documented and as such not completed, the site QA-department had done the required reviews during the course of normal business.
He planed to gather and sort the inferution generated during these normal reviews to assess the impact of not complei N the audits when required. The inspector concluded that because NRC Inspection Report 88-10 issued a similar violation for which correc-tive actions appeared to have been ineffective, this violation should be cited. Failure to document these audits represents a violation of TS 6.5.2.8.
(VIO 89-10-04)
b.
NYPA has taken action by issuing a revision to Plant Standing Order (PS0)-31A, Inservice Inspection (ISI) Program, to address concerns raised in Inspection Report 89-07.
The concern documented in IR 89-07, Section 5.a deals with the informal method in which the required programmatic inspections are documented and verified to be complete.
This stemmed from an April 26 Occurrence Report which documented that six out of eight required reactor vessel nozzle inspections were not conducted during the 1988 refueling outage.
The revision to pS0-31A places more importance on the QA organization to ensure that the required inspections are completed.
Further, it requires that any proposed changes to outage examination schedules be sent to QA for review and approval.
To correct the cause of the April incident, every weld inspection, including multiple examina-tions on the same weld, shall be listed separately on the outage list of inspections to be completed.
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The QA organization, to. establish a baseline of what has been com-pleted,- reviewed the existing ISI documentation.
This review was completed to ensure that the program-requirements were being met. On September 16, QA discovered that two zero degree inspections were not documented as completed on two reactor vessel welds. This inspection is done as a precursor to other inspections that were completed.
Based on this and discussion with the ISI contractor and the NDE examiner who performed the inspections, NYPA determined the condition
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of these welds to be acceptable. The inspector concluded that cor-rective actions regarding the upgraded 151 program controls and the missed inspections were acceptable and appropriate.
7.1 Safety Assessment The failure to complete the TS required audits indicates that both site QA management and the corporate SRC has not been adequately aware of the status of the audit program, While normal QA activities do cover most of the missed audit areas, the failure to track the j
audits is a significant weakness.
Management actions to ensure that the ISI program is properly com-pleted are seen as positive. QA identification of the missing docu-mentation for the two forementioned reactor vessel weld examinations
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is also positive, t
8.
Exit Interview 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 they are described in this report.
The inspector decided that the tracking system for inspector safety con-cerns (the F-# system) introduced in the previous report will be applied
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to-the remaining items requiring review from previous inspection reports.
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In those cases, the item number that was listed in previous reports will i
be tracked using the F-# format. The observations listed on page four of
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the SSFI report to be followed will be tracked by the number provided in
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the report using the F-# format. These items will be identified by the F-# and the appropriate inspection report number.
l 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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