IR 05000244/1990022
| ML17262A214 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/24/1990 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17262A213 | List: |
| References | |
| 50-244-90-22, NUDOCS 9011060203 | |
| Download: ML17262A214 (18) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket/Report:
50-244/90-22 License No.
DPR-18 R. E. GINNANUCLEAR POWER PLANT INSPECTION 50-244/90-22 AUGUST 28, 1990 - OCTOBER 9, 1990 Inspectors:
T. A. Moslak, Senior Resident Inspector, Ginna N. S. Perry, Resident Inspector, Ginna Approved by:
E. C. McCabe, Chief, Reactor Projects Section 3B
<elzV(CD Date VERVIEW LANTOPERATI N: The plant operated stably at approximately full power until a September 26, 1990 reactor trip due to personnel error.
NRC review of post-trip licensee actions is documented separately in Report 50-244/90-19.
The plant resumed power operation on September 29, 1990.
Throughout this period, power operation was conducted in a safe, competent manner, with appropriate management involvement.
Plant housekeeping continued to be very good.
ADI L I AL NTR: Routine observations of radiological controls noted no significant weaknesses.
AINTENAN E/
VEILLAN E:
Licensee actions in response to a spurious actuation of the Fire Suppression System were timely and comprehensive.
A conservative approach was taken to assure safe operation.
5EQ341TY:
Observations of the implementation of site security procedures identified no weaknesses.
EMER EN Y PREPARED:
Observations of Emergency Preparedness (EP) planning drills identified good coordination, strong management support, and a pro-active EP organization.
N INEERIN /TE HNI AL P
RT: Formalprogramsidentify, track, and control stored items with limited shelf life, and assure preventive maintenance on these item ~
. t TABLE F
NTENT
~PA E
1.
1.1 1.2 1.2.1 1.2,2 1.2.3 1.2.4 1.3 Plant Operations (71707, 92701).........................................."................
Control Room Observations Follow-up on Emergency Operating Procedure Open Items Unresolved Item 50-244/89-80-02, Containment Isolation/Containment Ventilation Isolation (CUCVI) Panel (Closed)
Unresolved Item 50-244-89-80-09, Procedure Changes (Closed)
Unresolved Item 50-244/89-80-10, EOP Control (Closed)
Unresolved Item 50-244-89-80-11, Verification and Validation (V&V)
Program (Closed)
Follow-up on 4-25-90 Emergency Diesel Generator Start
2 2.
Radiological Controls (71707).......................................
13.
3.1 3.2 3.2.1 3.2.2 3.3 Maintenance/Surveillance (62703, 61726, 92701, 92702)............
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Maintenance Observations Corrective Maintenance Post-Trip Maintenance Spurious Actuation of the'Fire Suppression System Surveillance Observations
..3
4.
Security (71707)..........
5.
Emergency Preparedness (71707)..........................................................
6.
6.1.
6.2.
6.2.1 6.2.2 Engineering/Technical Support (71707, 92701).............................~...........
Plant Trip Due To Dropped Flashlight
Licensee Action on Previous Inspection Findings Shelf Life Program (88-23-01, Closed)
Maintenance Team Inspection Items (90-80-01, -03; Closed)
7.
7.1 Safety Assessment/Quality Verification (71707, 40500).
Periodic and Special Report
7 8.
8.1 8.2 TMI Action Item Follow-up (92701)...................................~.........
I.C.1.2.B and.I.C.1.3, Short Term Accident and Procedures Review II.F.2.4, Instrumentation for Detection of Inadequate Core Cooling
8
9.
9.1 9.2 9.3 Administrative (30703, Licensee Activities Inspection Hours Exit Meetings 1707)
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DETAILS
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DETAIL/
l nt ion ntr 1R m
rvai n The inspectors found the R. E. Ginna Nuclear Power Plant to be operated safely and in conformance with NRC requirements.
Control room staffing was adequate and operators controlled access to the control room.
Shift supervisors consistently maintained authority over activities and provided detailed turnover briefings to relief crews.
Operators adhered to approved procedures and understood the reasons for lighted annunciators.
The inspectors reviewed control room log books to obtain information concerning trends and activities, and observed recorder traces for abnormalities.
During normal work hours and on backshifts, accessible areas of the plant were toured and plant conditions and activities were observed with no inadequacies identified.
The inspectors verified compliance with plant technical specifications and audited selected safety-related tagouts.
Among the documents reviewed included Ginna Station Event Reports (A-25. 1) Nos. 90-84 through 90-92.
These reports were reviewed to assess whether personnel took appropriate corrective action and observed the appropriate Limiting Conditions for Operation.
No inadequacies were identified.
On September 26, 1990, the plant tripped from approximately full power when a technician performing wiring verifications in a relay cabinet accidentally dropped a flashlight, This event is described in NRC Augmented Inspection Team Report 50-244/90-19.
Immediately after the plant trip, the inspector responded to the control room.
Control room personnel were performing the required actions of the Emergency Operating Procedures.
Appropriate management and support personnel were present, assessing plant response.
The inspector noted that operator response to the event was good, and that the plant was stabilized in a timely manner.
1.2 F
11 w-n Emer enc tin Pr ur P
n Item 1.2.1 nresolved I em-244
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-02 n
inm nt I l
i n n
inmen V ntil ion I lati n I
V Pnl I
Seven isolation valve indications are not located on the CVCVIlight panel; they are on other panels and are not highlighted.
RG&E made the CUCVI valve indications more easily identifiable by using color coding to identify them as CUCVI valves.
Additionally, studies are being conducted to evaluate grouping all CI valves together.
The inspector concluded that the color coding adequately highlighted the seven valve indication.2.2 nre Iv I em-244
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Pr r
han e
Administrative Procedure (A)-601.6, "Procedure Control of Emergency and Abnormal Procedures,"
was changed to require a review to determine whether supporting procedures impact the Emergency Operating Procedures (EOPs).
An attachment to A-601.6 lists all Abnormal Procedures (APs) and the procedures referenced in the EOPs.
The inspector verified the procedure change and concluded that it adequately addressed the concern.
1.2.3 nr lv I em-244
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E P
n rol I
Initially, the governing document for the EOPs and APs (A-601.6) was not referenced in the administrative procedures controlling procedure changes (A-6.1.1 and A-601.2).
A-601.1 and A-601.2 were subsequently revised to reflect that changes to EOPs and APs are processed in accordance with A-601.6.
The inspector found these procedure changes adequate to address the concern.
1.2.4 nresolved Item-244
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-11 V rificati n nd Vali i n Pr m
This item concerned initial weaknesses with the V&Vprocess for the EOPs.
Procedure A-601.6 was revised to designate satellite procedures as requiring appropriate review.
A-601.6 was revised to require V&Vfor major changes.
Human factors training willbe conducted to familiarize individuals involved with V&V. The inspector found the above actions adequate to address the weaknesses.
1.3 F
11 w-u n 4-2 -
Emer en Di el enera or ta On April 25, 1990, the "A" Emergency Diesel Generator automatically started when the "A" Reactor Coolant Pump was started; this event was described in NRC Inspection Report 50-244/90-05.
Additional corrective actions taken by RG&E include requiring operators to routinely notify power control two hours in advance of starting a reactor coolant pump on the 751 circuit. This willenable personnel to place voltage control in manual and hold the voltage higher than normal on the feed line to the reactor coolant pump.
The inspector had no further questions.
2. 2~
2NR~
i The resident inspectors periodically confirmed that radiation work permits were effectively implemented, dosimetry was correctly worn in controlled areas and dosimeter readings were accurately recorded, access to high radiation areas was adequately controlled, and postings and labeling were in compliance with procedures and regulations.
Through observations of ongoing activities and discussions with plant personnel, the inspectors concluded that radiological controls were being conscientiously implemente ~ 4
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3.
inten nce/ urveill nce 3.1 inten n erv tion Following the September 26, 1990 reactor trip, the inspectors observed portions of various safety-related maintenance activities that supported the licensee's root cause analysis to assess whether redundant components were operable, activities violated Limiting Conditions for Operation, personnel obtained required administrative approvals and tagouts before initiating work, personnel used approved procedures or the activity was within the "skills of the trade,"
workers implemented appropriate radiological and ignition/fire prevention controls, and equipment was tested prior to its return to service.
Portions of the following activities were observed:
Reactor Trip Breaker contact inspection.
Cell By-pass Switch contact inspection.
Turbine Electro-Hydraulic Control System Filter Changeout.
No unacceptable conditions were identified, 3.2 orrec ive Mainten nce 2.2.2 After the plant trip on September 26, 1990, some apparent problems were outside the scope of the NRC Augmented Inspection Team.
Licensee resolution of the following items was reviewed by the resident inspectors.
The "B" Reheater Stop Valve indicated mid-position, but was locally verified fully closed.
Plant personnel determined that the valve's limit switch was out-of-adjustment.
The valve's position indication was readjusted, and other valves were verified as giving accurate position indication.
The out-of-adjustment condition was evaluated as an isolated case caused by a loose setscrew.
The "A" Main Feedwater Pump suction relief valve lifted and failed to properly reseat.
Maintenance personnel replaced the valve, which had been identified as leaking prior to the plant trip. It was determined that the cause of the leakage was corrosion on the valve seat.
Plant personnel plan to evaluate this corrosion to determine its origin and take further action as necessary.
Intermediate range nuclear instrument N35 initially tracked decreasing reactor power normally, but indication continued to decrease to offscale low; indication returned onscale to its normal. value several hours later.
Through discussions with Westinghouse, plant management determined that this response is normal for a fairly new detector, which needs
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activity buildup.
This response has been observed in the past and normal indication typically returns before a startup is initiated.
In addition, the range over which this offscale indication occurred is covered by the source range detectors.
After the plant trip, reactor coolant iodine activity spiked to about 0.34 microcuries/gram.
This response was not unexpected since there is a suspected failed fuel pin. The Technical Specification limitfor continuous iodine activity is 0.2 microcuries/gram; this limitwas exceeded for approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, but did not exceed the associated action statement because the plant was in shutdown.
Minor significance was given to this iodine spike since it was not unexpected, it was adequately tracked by plant personnel, the activity decayed off over a short period of time, and the plant was shut down at the time.
The inspectors had no further questions.
3.2.2 uri u Act ation of the Fire u
re si n stem On September 10, 1990, a spurious voltage spike in fire detection circuitry caused the actuation of a dedicated Halon cylinder in the Fire Suppression System for the Multiplexer Room, located beneath the Control Room.
Although ventilation dampers immediately closed and isolated this area, licensee management suspected that Halon may have communicated with the Control Room Emergency AirTreatment System (CREATS), which was operating at that time.
Since it was not be known whether Halon had been adsorbed in the CREATS charcoal filters, possibly degrading filter efficiency, the CREATS was taken out of service, the filters were replaced with fresh charcoal, and the potentially degraded charcoal was sent off-site for analysis.
Analytical results showed that the charcoal was not degraded.
To establish the root cause for the actuation, a licensee engineering evaluation was performed.
A detailed inspection of Fire Suppression System electrical, pneumatic, and mechanical components was completed.
Then, a system functional test was performed to identify system interface problems.
Finally, detector/actuator voltage was extensively monitored.
Based on this evaluation, the licensee concluded that the actuation was caused by a random voltage spike in the detection circuitry.
From a review of licensee actions in response to this incident,'he inspector concluded that the licensee acted in a conservative, safety conscious manner.
The inspector concluded that the limiting condition for operation for the CREATS was not violated, in that the system was out-of-service for about 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> for charcoal replacement and Technical Specification 3.3.5.2 allows the CREATS to be inoperable for a period not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
While the Fire Suppression System was out-of-service for testing and repair, the inspector verified that a fire watch was established for the Multiplexer Roo Through discussions with the licensee's technical staff, data review, and examination of Fire Suppression System components, the root cause analysis was assessed to be thorough and well planned.
The inspector had no further questions.
3.3 urv illan rvti n Inspectors observed portions" of surveillances to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to Limiting Conditions for Operation, and correct system restoration following testing.
Portions of the following surveillances were observed:
Periodic Test (PT)-12.2, Emergency Diesel Generator 1B, Revision 59, effective July 31, 1990, observed September 19, 1990.
PT-2.7, Service Water System, Revision 54, effective August 24, 1990, observed September 25, 1990.
No unacceptable conditions were identified.
4.
~ecurit During this inspection period, the resident inspectors verified x-ray machines and metal and explosive detectors were operable, protected area and vital area barriers were well maintained, personnel were properly badged for unescorted or escorted access, and compensatory measures were implemented when necessary.
5.
Em r nc Pre r
ne On September 21, 1990, the inspectors witnessed portions of an emergency preparedness drill involving an on-site contaminated personnel casualty and transportation to a local hospital.
The drill was found to be well coordinated, and did not adversely impact control room operations.
Site management, Wayne and Monroe County emergency management personnel, and members of the Federal Emergency Management Agency actively participated in various phases of the drill. Members of the licensee's emergency response organization appropriately and expeditiously responded.
No weaknesses were identified in the planning and execution of the drill.
6.
n in rin /7 hni l
6.1 lan Tri D e T Dr FI shli ht The plant trip which occurred on September 26, 1990 was caused by a dropped flashlight striking two MG-6 relays which caused the "A" Reactor Trip Breaker to open.
The technician who dropped the flashlight was performing wiring verifications in a relay room cabinet.
A small metal flashlight was being used to aid in reading markings on the wires.
After the reactor tripped, the technician realized he may have caused the actuation, so he immediately notified control room personnel.
After the trip, plant management stopped all
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wiring verification activities and does not plan to continue with these activities until an evaluation is performed to determine under what conditions further wiring verifications can be safely performed.
Immediately after the. trip, the resident inspector was notified by the Superintendent of Ginna Production.
The shift supervisor, with concurrence from the shift technical advisor, determined that the event was reportable to the NRC within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of its occurrence.
The trip occurred at 11:00 a.m., and the shift supervisor made the initial notification via the Emergency Notification System at 12:09 p.m..
Follow-up notification was appropriately made at 9:46 p.m. to relate further information obtained subsequent to the initial notification.
RG&E procedure A-85, Verification or Inspection of Electrical Equipment, Revision 3, effective 4-22-89, provides instructions not to use uninsulated conductive tools such as flashlights inside electrical equipment, and not to bump, jar or disturb equipment that is not a direct part of the task being performed.. On September 26, 1990, a technician, performing field verifications, used an uninsulated metallic flashlight in an energized relay cabinet and dropped it, allowing it to strike a number of relays, causing a plant trip. This potential violation of NRC requirements is unresolved pending promulgation and review of the Augmented Inspection Team Report (50-244/90-22-01).
6.2 Li n ee Actions n Previous Ins ti n Findin 6.2.2 ~26if Vi lati n-244 7-1
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-Rin N
In l d in h h lfLif Pr m
l ed n
nre lv I em-244/
-2 -
E tablish r
P Preventive Maintenance M Pr m
hed le nd Inv nt
The inspector reviewed actions taken by the RG&E Procurement Engineering Group in developing and implementing a comprehensive shelf life tracking program for items stored in the warehouse.
Included in this review was verification that computer software was developed to identify, track, and control items with limited shelf lives, e.g. 0-rings.
Due to their similar nature, the 0-ring shelf life program has been developed in parallel with the spare parts PM program.
The 0-ring program has been implemented.
Full implementation of the spare parts PM program is projected for late October 1990.
Detailed administrative guidelines have been established to carry out these programs.
Through lessons learned in the early phases of implementation, plant Administrative Procedure A-401, "Control of Procurement Documents," willbe changed to incorporate these guidelines as procedural controls.
Based on this review, the inspector concluded that the licensee has taken the actions necessary to assure that the stored material is properly controlle <
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6.2.2 ain enance T m Ins ti n I em 6.2.2.1 Vi
n-244 0- 0-1 Fail re P rf rm f t E al i nf r ntamin ted t ra eBuildin l
The inspector reviewed the safety evaluation addressing the on-site transporting and storage of contaminated materials.
The safety evaluation adequately addressed the criteria stated in 10 CFR 50.59 and concluded that an unreviewed safety question did not exist for the performance of these acivities.
This evaluation received the required review and subsequent approval by the Plant Operations Review Committee.
The inspector had no additional concerns.
6.2.2.2 Vi lati n 0-244
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Fail re o Perf rm fe Ev luation f r r
an eDeec rH uin M
ifi i n l
The inspector verified completion of the licensee's corrective actions.
Included in this review was an examination of the 10 CFR 50,59 evaluation performed in response to the Nonconformance Report (NCR 90-140) addressing the violation. The evaluation was found to be detailed and appropriately reviewed by the licensee's technical staff. It concluded that no unreviewed safety question existed as a result of making this modification.
The inspector had no additional questions.
7.
aft A
e ment/
li Vrifi in 7.1 P ri i
and i
Re Periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9.1 and 6,9.3 were reviewed.
Inspectors verified that the reports contained information required by the NRC, that test results and/or supporting information were consistent with design predictions and performance specifications, and that reported information was accurate.
The following reports were reviewed:
Monthly Operating Report for August 1990.
Semiannual Radioactive Effluent Release Report (January - June 1990).
No unacceptable conditions were identifie.
TMI A ti n Item F ll w-u 8.1 12B n I
B h
TrmA ient n
Pr ur Reiw NRC Inspection Report 50-244/85-15 review noted that operator training, final validation and implementation of the revised Emergency Operating Procedures (EOPs) remained to be completed.
NRC EOP Team Inspection Report 50-244/89-80 concluded that the Ginna EOPs are an example of a well-designed program, properly implemented.
This item is closed.
8.2 F24In m n ti nf rD i n fIn e
r lin NRC Inspection Report 50-244/85-15 review noted that the reactor vessel level monitoring system was the one remaining requirement to be completed.
The reactor vessel level indication system was installed in 1986.
An NRR safety evaluation was completed, and the Ginna Technical Specifications were amended to incorporate requirements for the system effective September 23, 1988.
This item is closed.
9.
33ii i I 9.3 ~*
At the beginning of the inspection period, the plant was operating at approximately full power.
On September 10, 1990, an inadvertent Halon discharge occurred in the Multiplexer Room located under the control room.
On September 26, 1990, a plant trip occurred when a technician, performing wiring verifications, accidentally dropped a flashlight in a relay cabinet.
This plant trip and subsequent actions taken by RG&E are described in NRC Augmented Inspection Team Report 50-244/90-19.
After. corrective actions were completed, the plant was started up on September 28, 1990.
At the close of the inspection period, the plant was operating at approximately full power.
9.2 This inspection involved 101 inspection hours, including 13 backshift hours.
9.3 At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of this inspection.
In addition, NRC exit meetings were held for the following inspections during this inspection period: 50-244/90-21 and 90-23 on September 14, 1990, 50-244/90-14 on September 21, 1990, and 50-244/90-19 on September 28, 199 NRC FORM 6 OUTSTANDING ITEMS FILE SINGLE DOCKET ENTRY FORM REPORT HOURS 1. Operations
7. Outages 2. Rad-Con 8. Training 3. Maintenance
9. Licensing 4. Surveillance
10. QA
5. Emerg. Prep.
11. Tech.Sup.
6. Sec/Sefegrde 12. Fire Prot./
Housekeeping 13. Other Docket No. 50-244 Originator: Moslak Reviewing Supervisor: McCabe Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 90-22-01 DRP 90-22 10-09-90 Originator Modifier/Closer Moslak Description: Technician dropped flashlight in relay cabinet.
Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 87-10-04 90-22 10-09-90 Originator Modifier/Closer Moslak Description:
Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 88-23-01 90-22 10-09-90 Originator Modifier/Closer Moslak Description:
Item Number Type SALP Area Area Resp Action Due'Date Updt/Close/Rpt Date 0/M/C 89-80-02 Originator Modifier/Closer Moslak 90-22 10-09-90 Description:
NRC FORM 6 Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 89-80-09 90-22 10-09-90 Originator Modifier/Closer Moslak Description:
Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 89-80-10 90-22 10-09-90 Originator Modifier/Closer Moslak Description:
Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 89-80-11 90-22 10-09-90 Originator Modifier/Closer Moslak Description:
Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date 0/M/C 90-80-01 90-22 10-09-90 Originator Modifier/Closer Moslak Description:
Item Number Type SALP Area Area Resp Action Due Date Updt/Close/Rpt Date O'/M/C 90-80-03 Originator Modifier/Closer Moslak 90-22 10-09-90 Description:
I t