IR 05000282/1986007
| ML20212A584 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/21/1986 |
| From: | Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20212A546 | List: |
| References | |
| 50-282-86-07, 50-282-86-7, 50-306-86-07, 50-306-86-7, GL-85-15, IEB-85-001, IEB-85-1, IEB-86-001, IEB-86-1, IEIN-85-045, IEIN-85-45, IEIN-86-003, IEIN-86-3, NUDOCS 8607290077 | |
| Download: ML20212A584 (9) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Reports No. 50-282/86007(DRP); 50-306/86007(DRP)
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Docket Nos. 50-282; 50-306 Licenses No. DPR-42; DPR-60 Licensee: Northern States Power Company 414 Nicollet Mall
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Minneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plant Inspection At:
Prairie Island Site, Red Wing, MN Inspection Conducted: April 13, 1986 through June 21, 1986 Inspectors:
J. E. Hard
M. M. Moser
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Approved By:
I. M. J ki, Chief I/-54 Reactor P jects Section 2B Date
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Inspection Summary I
Inspection on A)ril 13, 1986 through June 21, 1986 (Reports No. 50-282/86007(DRP);
50-306/86007(DR)))
Areas Inspected:
Routine, unannounced inspection by resident inspectors of previous inspection findings, plant operational safety, maintenance, surveillance, ESF systems, facility modifications, periodic and special reports, corporate management concerns, followup of Licensee Event Reports, and IE Bulletins followup.
Results: One unresolved item and three violations were identified in the ten areas inspected. The violations involved failure to maintain adequate logs, failure to adequately control modifications, and failure to follow procedures.
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DETAILS 1.
Persons Contacted
- E. Watzl, Plant Manager
"D. Mendele, Plant Superintendent, Engineering and Radiation Protection
- R. Lindsey, Plant Superintendent, Operations and Maintenance A. Hunstad, Staff Engineer A. Smith, Senior Scheduling Engineer M. Balk, Superintendent, Operations D. Schuelke, Superintendent, Radiation Protection G. Lenertz, Superintendent, Maintenance J. Hoffman, Superintendent, Technical Engineering K. Beadell, Superintendent, Quality Engineering M. Klee, Superintendent, Nuclear Engineering R. Conklin, Supervisor, Security and Services D. Vincent, Project Manager, Nuclear Engineering and Construction J. Goldsmith, Superintendent, Nuclear Technical Services A. Vukmir, Site Services Representative, Westinghouse The inspectors interviewed other licensee employees, including members of the technical and engineering staffs, shift supervisors, reactor and auxiliary operators, QA personnel, and Shift Technical Advisors.
- Denotes those present at the exit interview on June 23, 1986.
2.
Licensee Action on Previous Inspection Findings None.
3.
Operational Safety Verification (71707, 71710)
Unit 1 and Unit 2 were base loaded at 100% power except for reductions for surveillance testing, special maintenance, and weekend load following.
The inspection observed control room operations, reviewed applicable logs, conducted discussions with control room operators, and observed shift turnovers.
The inspector verified operability of selected emergency systems, reviewed equipment control records, and verified the proper return to service of affected components. Tours of the auxiliary building, turbine building and external areas of the plant were conducted to observe plant equipment conditions, including potential fire hazards, and to verify that maintenance work requests had been initiated for equipment in need of maintenance.
On March 24, 1986 with Unit 1 at cold shutdown and Unit 2 at 100% power, and while test tripping of the substation 13.8 KV ACB IH2, the relay technician inadvertently reclosed the test trip switches prior to
resetting the relay. This resulted in a loss of power to buses 25 and 26
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which started both emergency diesel generators. The diesels did not load because other power sources were available. Corrective action was immediately begun to restore substation equipment.
The cause of this event was personnel error due to a lapse of attention.
Written procedures had bean developed because of similar events involving the same test crew in the recent past and were being used at the time of this event. This failure to follow written procedures is a violation of technical specification 6.5.
See Notice of Violation (282/86007-01; 306/86007-01).
A Unit 2 reactor trip occurred at 10:57 a.m. on May 19, 1986 during safeguards logic testing. The trip was the result of a safety injection signal generated during the test.
This signal may have been caused by a technician momentarily and accidentally releasing the pressure from a test pushbutton. All systems responded as expected under these conditions. When a condensate pump was restarted following reset of the SI signal, a minor water hammer was heard in the feedwater system possibly because of collapsing steam voids in the feedwater heaters.
Subsequent walkdown of the feedwater system showed no piping or support damage from the event. The reactor was restarted at 12:55 a.m. on May 20.
A second Unit 2 reactor trip occurred at 3:19 a.m. on May 20, 1986, while at about 20% power. The operating crew had been experiencing steam generator level control and indication problems. High level in steam generator 22 caused the turbine to trip which in turn tripped the
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reactor. 'All systems responded as expected.
Steam generator level sensing lines were flushed while the reactor was shut down. The reactor was restarted at 11:25 a.m. on May 20 and the generator was placed on line at 2:27 p.m.
Steam generator level indications seemed to respond as expected during load changes.
The licensee conducted an annual emergency drill on April 22, 1986. The resident inspectors observed the drill and monitored the scenario events in the control room and technical support center. On June 17, 1986, the licensee conducted an emergency preparedness exercise with NRC Region III
- rticipating by activating the incident response center and dispatching i site team. The resident inspectors participated in the exercise by monitoring the scenario events and responses in the control room, technical support center, and emergency operations facility. NRC and Federal Emergency Management Agency personnel observed and evaluated the exercise.
Routine inspection of this radiological emergency preparedness exercise is the subject of inspectior report 50-282/86-06; 50-306/86-06(DRMSP).
No violations or deviations were identified.
4.
Myintenance Observation (62703)
Routine maintenance activities (on safety-related systems and components)
listed below were observed / reviewed to ascertain that they were conducted
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in accordance with approved conformance with Technical Specifications.
The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, radiological controls were implemented, and fire prevention controls were implemented.
Portions of the following maintenance activities were observed / reviewed during the inspection period:
Bus 26 voltage restoration / load rejection scheme special test Diesel generator D2 air cooler cleaning and replacement Diesel generator D2. preventive maintenance overhaul and bearing inspection Diesel powered cooling water pump preventive maintenance
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A major preventive maintenance overhaul of the emergency diesel generator
- 2 was accomplished during this inspection period with a region based inspector participating in a special examination of the main and crankpin
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bearings.
Indications of main bearing degradation that were found have raised concerns about the reliability of the emergency diesel generators and may result in a need for modification of the lubrication system as well as more frequent bearing inspections. This special inspection of the diesel generator bearings is the subject of Inspection Report No. 50-341/85046(DRS).
No violations or deviations were identified.
5.
Surveillance (61726)
The inspector witnessed portions of surveillance testing of safety-related systems and components. The inspection included verifying that the tests
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were scheduled and performed within Technical Specification requirements, observing that procedures were being followed by qualified operators, that Limiting Conditions for Operation (LCOs) were not violated, that system and equipment restoration was completed, and that test results were acceptable to test and Technical Specification requirements.
Portions of the following surveillances were observed / reviewed during the inspection period:
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SP 1130 containment vacuum breakers cuarterly test SP 2102 turbine driven auxiliary feedwater pump test SP 2713 21,22 SI pump test
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On June 14, 1986, while performing surveillance procedure SP 2089 (Unit 2
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i residual heat removal (RHR) test), it was found that motor operated valve i
32129 would not open. This valve must open to permit component cooling
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water flow through the #22 RHR heat exchanger. A work requ'est was issued to troubleshoot and make repairs after the valve was manually opened.
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appears that technical specifications-LCO limit of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for one RHR heat exchanger out of service may have been exceeded.
This is considered g.
an unresolved item pending further investigation by licensee and the resident inspectors (306/86007-02).
No violations or deviations were identified.
6.
ESF System Walkdown (71710)
The inspector performed a complete walkdown of the accessible portions of both turbine driven auxiliary feedwater systems. Observations included confirmation of selected portions of the Licensee's procedures, checklists, plant drawings, verification of correct valve and power supply breaker
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positions to insure that plant equipment and instrumentation are properly
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aligned, and review of control room and local system indication to insure
proper operation within prescribed limits.
No violations or deviations were identified.
7.
Facility Modifications (37700, 37701)
On May 1, 1986 at about 1:00 p.m. the limit switch compartment cover was discovered removed from the motor operator on MV-32023, the main feedwater to the Unit 1 11 steam generator containment isolation valve.
Plant management was notified of the situation and the cover was replaced
by about 3:00 p.m.
Unit 1 was operating at 100*.' power during this period
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and had been operating continuously since April 10.
The cover was likely to have been off the valve for this entire period.
I In addition to being a containment isolation valve, MV-32023 is listed in
FSAR Figure I.3-2 as required equipment for a steam line break or crack.
i Since such a break could occur in the vicinity of the valve, the licensee analyzed the safety significance of this event. Their investigative report notes that; a.
For breaks inside containment, the valve would not experience any adverse environment, as the valve is located outside containment,
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and therefore its operation could be assured. Therefore, the
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function of containment isolation would be fulfilled.
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For breaks outside containment, it could be postulated that a i
steam environment could exist in the vicinity of the valvo.
J Possible inoperability of the valve, due to a steam environment, does not pose a safety problem because 1) containment isolation
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is not required for this accident and 2) feedwater isolation is i
i accomplished by three other diverse means.
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Documentation from Limitorque regarding testing the valve with live steam in the limit switch compartment indicates that the valve probably would have operated satisfactorily even in the presence of live steam.
Conclusions from the investigative report are that there is some assurance that the valve would have operated successfully with the limit switch cover remcsid, but even if it hadn't, there would have been no
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public health and safety consequences. The inspector agrees with the conclusions.
Though the plant investigation did not verify the precise reason that the valve cover was off during plant operation, it appears likely that this error was related to modification work performed on the valve during the
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March-April refueling outage. This failure to replace a valve cover for a safety-related valve appears to be a violation.
See Notice of Violation.
(282/86007-03a)
Neither the Operations Log nor the Reactor Log for May 1, 1986 contained entries related to the valve cover matter addressed directly above. This failure to promptly log this information appears to be a violation.
See Notice of Violation.
(282/86007-03b)
On May 29, 1986 with Unit I at 100% power, control room operators noticed the step counter for the control rods in Bank C, Group I was indicating rod-in movement.
From other parameters and indicators it was clear that the rods were not actually moving.
Investigation revealed a construction worker behind control panel A in the control room digging insulation out
of an electrical penetration using a piece of sharpened conduit.
In so doing, he had shorted or grounded the control wires for the step counter.
Licensee immediately stopped this job and all other construction work on site. The next day, May 30, the plant Superintendent of Quality Engineering issued a formal Stop Work Order to the Manager of Prairie Island Projects. The licensee promptly instituted special rules over the control cf project ' work but again, on June 10, while working in the same electrical penetration containing step counter control wires, accidental step counter movement occurred.
In this case, it appears that work was being performed by construction personnel which was not yet authorized.
Throughout the balance of the report period, licensee instituted additional controls over construction work on site, released more projects for restart using the additional controls, made permanent changes in the plant modification procedure and was considering additional administrative improvements.
The matter of control of modification work at Prairie Island, first raised by the resident inspector in August 1985, remains an open issue.
This failure to control activities which had some adverse effect on plant operation appears to be a violation.
See Notice of Violation.
(282/86007-04)
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8.
Regional Requests (92705)
a.
By memo dated May 6, 1986, C. E. Norelius requested all senior resident inspectors to provide information on low level radioactive waste storage facilities. Northern States Power has no current plans to add an on-site low-level waste storage facility at Prairie Island over and above what already exists at the site, b.
By memo dated May 6, 1986, C. E. Norelius requested all senior resident inspectors to gather information in response to Temporary Instruction (TI) 2515/77 - Survey of Licensee's Response to Selected Safety Issues; specifically biofouling of cooling water heat exchangers.
The inspector reviewed the instrumentation available on safety-related equipment cooled by open-cycle service water systems, recording and logging of heat exchanger performance parameters, procedures and training, periodic inspections, and preventive maintenance.
The information obtained is being supplied by the resident inspectors to Region III management and IE.
No violations or deviations were dentified.
9.
Licensee Event Reports Followup (92700'
Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Soecifications:
(Closed) 282/86005-03 Diesel generator start on loss of 2R transformer (Closed) 282/86005-04 Diesel generator start on loss of 2R transformer (Closed) 306/86005-05 Reactor trip because of steam-feedwater flow instabilities at low power (Closed) 306/85007-0 Reactor Trip - Safeguards Logic Testing (Closed) 282/86005-01 011 pump for 11 TDAFW? failed to pick up prime (Closed) 306/86007-07 Reactor trip at low power 10. Meetings with Corporate Management (30702)
On April 17, 1986, the Region III Administrator, Division of Reactor Projects Director, and NRR BWR Licensing Director toured the Prairie Island Nuclear Power Plant and the next day met with plant and corporate management at Corporate Headquarters to discuss plant performance and implementation of the modified NRC inspection program at Prairie Island and Monticello.
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11.
IE Bulletin Followup (92703)
IEB 85-01 (282/85001-88, 282/85024-BB, 306/85001-BB, 306/85022-BB):
Steam Binding of Auxiliary Feedwater Pumps.
Responses and actions required by this Bulletin have been completed.
IEB 86-01 (282/86001-8B, 306/86001-88):
Flow Logic Problems That Could -
Disable RHR Pumps.
This Bulletin applies to BWRs.
No action is required at Prairie Island.
12.
IE Information Notice Followup (92701)
a.
" Potential seismic interaction involving the movable in-core flux mapping system used in Westinghouse designed plants."
Temporary Instruction (TI) 2500/16 issued February 28, 1986 provided guidance for inspecting PWR facilities with movable in-core flux mapping systems to determine whether PWR licensees have performed a system review as a result of IE Information Notice 85-45. The resident inspector has reviewed the analysis and assessment that was performed by the licensee in June,1985 upon receipt of the Information Notice.
For both Prairie Island units, no seismic interaction exists between the movable in-core flux mapping system and seal table as described in IE Notice 85-45 and consequently no
licensee action is required, b.
Limitorque Valve Operator EQ On February 6-7, 1986, regional specialists A. Gautam and R. Smeenge visited the Prairie Island site to perform a reactive announced inspection relative to use of non qualified wires in 10 CFR 50.49 designated environmentally qualified (EQ) limitorque valve operators identified in IE Information Notice No. 86-03.
(See Inspection Report 50-282/86003(DRS); 50-306/86003(DRS)).
I In a memo dated March 27, 1986, the Prairie Island Resident Inspector was requested to determine what action the licensee has taken as it relates to IN 86-03 and the applicability of Generic Letter 85-15 with regard to this information notice. A package of
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pertinent information was forwarded to Region III on April 10, 1986 detailing the history and current status of limitorque EQ at Prairie Island.
(Inspection Report 282/86005(DRP); 306/86005(DRP)).
In a memo dated June 13, 1986, the Senior Resident Inspector was requested to provide additional information regarding limitorque motor valve operator wiring. The response to this request was sent to J. W. Muffett, DRS, on June 27, 1986.
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It is believed that the intent of Temporary Instruction 2515/75, Inspection of Limitorque Motor Valve Operator Wiring, 3/27/86, has been met.
13. Exit Interview (30703)
The inspectors met with licensee represantatives denoted in Paragraph 1 at the conclusion of the inspection on June 23, 1986. The inspectors discussed the purpose and scope of the inspection and the findings.
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any document / processes as proprietary.
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