IR 05000282/1989018
| ML20246N453 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/13/1989 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20246N439 | List: |
| References | |
| 50-282-89-18, 50-306-89-18, NUDOCS 8907190374 | |
| Download: ML20246N453 (9) | |
Text
{{#Wiki_filter:_ _ _ _ > I R .. d U. S. NUCLEAR REGULATORY COMMISSION REGI.0N III Reports No. 50-282/89018(DRP); 50-306/89018(DRP) Docket Nos. 50-282; 50-306 License No. DPR-42; DPR-60 Licensee: -Northern States Power Company-414 Nicollet Mall' Minneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plant Inspection At: Prairie Island. Site, Red Wing, MN Inspection Conducted: May 28 through July 1, 1989 ' Inspectors: J. E. Hard . T. J. O'Connor 7/B - Appoved By: e ef Reactor Projects Section 2A Date Inspection Summary Inspection on May 28 through July 1, 1989 (Reports No. 50-282/89018(DRP); 50-306/B13018(DRP))- Areas Inspected: Routine unannounced inspection by resident inspectors of - previous inspection findings, plant operational safety, maintenance, surveillance, ESF systems, security, quality assurance (QA) programs and followup of LERs and Generic Letters.
Results: During this inspection period, Unit 1 operated continuously at 100% . power except for a power reduction associated with the replacement and testing of a 480 volt circuit breaker for the loop B feedwater isolation valve.
Reactor coolant system radiochemistry continues to indicate the presence of a failed fuel rod. Activity levels continue to remain less than 1% of technical specification (TS) limits. At the close of the inspection period, Unit I had operated continuously for 226 days.
Unit 2 operated continuously at 100% power during this inspection period with 34 days of continuous operation at the close of the inspection period.
The plant replaced seven circuit breakers whose authenticity was questioned as a result of a vendor branch inspection, as documented in Inspection Report Nos. 50-282/88201; 50-306/88201(DRP).
Of the 7 areas inspected,1 violation of NRC requirements was identified.
h[[[k
0 _-_--
n_ . __ . j', - . . I
- j l 4-DETAILS
_ . 1.. Persons Contacte_d r ! Licensee Employees L. Eliason, General Manager, Nuclear Plants
- E..Watzl, Plant Manager R. Lindsey,. Assistant to the Plant Manager-D. Mendele, General Superintendent, Engineering and Radiation Protection (#M. Sellman, General Superintendent, Operations G. Lenertz, General Superintendent, Maintenance A. Smith, General Superintendent, Planning and Services 0. Schuelke, Superintendent, Radiation Protection G. Miller, Superintendent, Operations Engineering
- K. Beadell, Superintendent, Technical. Engineering S. Schaefer, Superintendent, Nuclear Engineering
- M. Klee, Superintendent, Quality Engineering P. Kamman,. Superintendent, Nuclear Operations QA
, R.' Conklin, Superintendent, Security and Services 'l D. Vincent, Project Manager, Nuclear Engineering and Construction ' D. Musolf, General Manager, Nuclear Projects J. Goldsmith Superintendent,-Nuclear Technical Services 'A. Hunstad,LStaff Engineer- -, T.'Amundson, Superintendent Training A. Vukmir, Site Services Representative, Westinghouse Electric Corp.
A. Vaia, Nuclear Services Division, Westinghouse Electric Corp.
L. Benson, Project Manager, Westinghouse Electric Corp.
NRC Representatives
- B. Burgess,. Chief, Reactor Projects Section 2A, DRP
. The inspectors interviewed other licensee employees, including members of j the technical and engineering staffs, shift supervisors, reactor and { auxiliary operators, QA personnel, shift technical advisors, and shift managers.
- Denotes those present at the exit interview of June 30, 1989.
-2.
Licensee Event Report Followup (93702)- (Closed)282/89007-LL: Discovery of the lack of Circuit Coordination For
the Breakers Supplying Each Unit's Loop B Feedwater Isolation Valves.
- ! As a result of inquiries made to the circuit breaker manufacturer, the-I licensee obtained the appropriate trip' characteristic curves for the
installed 600 volt G.E. THEF circuit breakers. Previously, the licensee had used the trip characteristic curves for 480 volt circuit breakers to
1 E__ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ - - _ - - - - _ - -. - - -- _ - - -. _ _ _ _ - - - _ _ - - - - _ - -
- ___-___
[, j < ', ~ . , \\ . l detennine' circuit coordination.. Review of-the 600 volt curves indicated i that the loop B _feedwater isolation valves lacked circuit coordination.
! The 600 volt curves;did not affect circuit coordination for any other-electrical ' circuits.
In response to the lack of circuit coordination, .i the licensee instituted a 24 hour fire watch to monitor the electrical i equipment associated with each unit's loop B feedwater isolation valves.
i ! The licensee was' able to locate spare circuit breakers within the plant whose trip characteristic curves would provide circuit coordination. To ensure that the replacements were acceptable, the licensee subjected the proposed replacement breakers to 100% current capacity for one hour then
measured the temperature rise,135% current capacity test, 600% overload i test, instantaneous trip test, 300% time delay overcurrent trip test, , . dielectric test and mechanical operation test. Additionally, the { breakers were visually examined for signs of tampering and remanufacture E and a records search was performed to establish traceability.
! Upon successful completion of the circuit breaker testing, the licensee { installed the breakers, partially stroked the isolation valve to verify l correct motor operation and then terminated the continuous firewatch.
(Closed) 282/89008-LL: Automatic Start of Auxiliary Building Special ! Ventilation System Due to High Radiation Spike on IR-37.
On June 18, 1989, licensed operators observed the high radiation l annunciator for train A alarming along with the auto start of train A of .the Auxiliary Building Special Ventilation System.(ABSYS). After , verifying the absence of abnormal radiation readings, radiation monitor i IR-37 was reset and the ABSVS returned to normal condition. Review of l the computer trending yielded normal operation prior to and subsequent to ! the initiating spike. Additional checks on the equipment confirmed the ! monitors operability and led to the conclusion that the high radiation l alarm was due to a random spike. The licensee's ongoing plans include i the performance of signal cable integrity testing and the eventual i replacement of the electronic component modules and signal cable.
Additionally, an equipment modification' designed to prevent random spikes from causing high radiation alarms will be installed.
! l 3.
Operational Safety Verification (71707, 93702, 92702) The inspector 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, i and to verify that maintenance work requests had been initiated for the equipment in need of maintenance.
Unit 1 operated continuously at 100% pcwer except for a power reduction l associated with the replacement and testing of a 480 volt circuit breaker for the loop B feedwater isolation valve. Reactor coolant system l i
" -
- - _ _ _ - _ _ _. - - _ _ _ _ _ _ _ _ -
_ _ _. _ _ _ - . ' q .. . radiochemistry continues to indicate the presence of a' failed fuel rod with activity levels less then 1% of TS limits.
Unit 2 operated continuously at 100% during this inspection period. As of July 1, 1989, Unit I has operated continuously fo* 226 days and Unit 2 for 342 days.
On June 6,1989, the resident inspectors observed the licensee's plant emergency exercise. The exercise was well performed and demonstrative of the licensee's ability to prevent serious consequences to public health and safety. Licensee critiques of the exercise were constructive and were accurate appraisals of performance.
(For additional information concerning the plant exercise refer to Inspection Report Nos.
50-282/89009;50-306/89009(DRP)). On June 15, 1989, the shaft of the 121 motor-driven fire pump failed.
The licensee installed a spare shaft and has forwarded the failed shaft offsite for failure analysis. This was the third failure of this pump's shaft. Additional investigation will examine pump vibrations and the pump's use under low flow conditions when used as a water supply for the intake screen wash system.
On June 18, 1989, the auxiliary building special ventilation system started as a result of a spike on IR-37.
(Reference Paragraph 2).
The 480 volt circuit breakers for the unit's loop B feedwater isolation valves, the supply for the 136 panel and the 11,12, 21, and 22 station battery chargers were determined to be of questionable origin and-replaced.
Replacement breakers came from other plant locations and were subjected to the plant's upgrade program which included an extensive testing, a visual examination and a records search to establish traceability.
(Reference Paragraph 2). The inspectors monitored variouf aspects of the electrical testing and visually examined the replacement breakers.
Installation of the 21 battery charger breaker and the loop B feedwater isolation valve breakers was also monitored.
On June 27, 1989, while performing the work authorized by work request (WR) N4669, the 480 volt breaker for the 11 Inverter Instrument Bus II was deenergized and removed from motor control center (MCC) 1AC1 instead of the 480 volt breaker for the 11 Battery Charger. Upon discovery that the incorrect breaker was removed, the breaker was reinstalled and the correct breaker removed.
10 CFR Appendix B, Criterion V, Instructions Procedures, and Drawings, requires that activities affecting quality shall be prescribed by document instructions and procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions and procedures.
Administrative Control Direction 5ACD3.2, Work Control, Rev.15, Step 6.14.3 requires work procedures to be at the job site and that the , requirements and/or precautions shall be followed and completion of ' procedural steps documented.
Removal of this breaker caused a loss of redundancy in regards to a response to a design basis accident. Although the other train was
_ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
- __- - . , ' . . available, this loss of redundancy combined with an active failure of the other train could severely affect the plants ability to respond to a design basis accident. This proposed violation in conjunction with recent operational errors, including the unexpected start of diesel generator D-1 (LER 50-282/89003) and the missed hourly logging of the axial flux ~ difference (LER 50-306/89001), prompts the inspectors to question the licensee's efforts to maintain a high threshold for attention to detail.
Of additional concern is the fact that not only did the plant operator deenergize the incorrect breaker, but that the plant electrician failed to verify that, although the breaker was deenergized, the correct breaker was removed.
In response to these incidents, the licensee has instituted an awareness program to help maintain a high threshold for attention to detail.
Failure to follow the requirements of WRN4669 is identified as Violation 50-282/89018-01; 50-306/89018-01(DRP).
Due to the need to partially stroke the feedwater isolation valve to ensure proper motor operation, Unit I reduced power on June 28, 1989, to 90% to minimize any potential feedwater transients. The inspector monitored this power reduction.
Licensed operators were briefed on the need for the power reduction and were observed using appropriate procedures. The actual breaker replacement occurred after a very thorough briefing of involved personnel. The loop B feedwater isolation valve breaker on Unit 2 was performed on June 29, 1989, without a power reduction.
Review of plant parameters revealed no feedwater perturbations as a result of the partial stroking.
Due to the lack of circuit coordination associated with the loop B feedwater isolation valves, a continuous fire watch was initiated. A number of the nuclesr plant attendants were interviewed in regards to their fire watch route, items to monitor and responses to be taken upon discovery of adverse conditions. The fire watch was terminated with the replacement of the subject breakers.
On June 28, 1989, the licensee conducted an emergency medical drill.
This drill was also well performed and demonstrative of the licensee's ability to provide proper medical attention to injured victims with appropriate regard for radiation / contamination concerns. The licensee's critique was also constructive and an accurate appraisal of the performance.
4.
Maintenance Observation (62703) Routine, preventive, a u corrective maintenance activities were observed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in 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 init;m.ing the work, activities were accomplished using approved procedures and were inspected as applicable, functional i testing and/or calibrations were performed prior to returning components l or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, radiological controls were implemented, and fire prevention control: were implemented.
>
. _ _ _ _ _ _ _ - _ -
_ _ - - _ _ ' , - q o
Portions of: the following maintenance activities were observed during the inspection period: > Repair of lube oil and fuel oil leaks on 12 diesel-driven cooling ~ water pump.
Repair of a body to bonnet leak on CV-31420, charging line~ to RCS - ~ loop isolation control. valve.
.. Seal repair on 122 CVCS monitor tank. pump.
, Replacement of the desurger on the 12 charging pump.
o Replacement.of breakers for loop B feedwater isolation vanes and
- 21 battery charger.
On June 5,;1989, the licensee removed No. 2 Emergency Diesel Generator (D2) from service for an annual preventive maintenance inspection and related modifications. The prudence of removing D2 from service at this-time of year was questioned by.the inspector in light of prior. experience with tornadoes..It should be noted that the peak tornado season 'is from-June 1 to July 15 and that tornadoes are known to occur during a large-fraction of the calendar ~ year.
Inspection Report Nos. 50-282/89012 and 50-306/87011 documents the July 17, 1987, partial loss of offsite power with D2 out of service. The partial loss of offsite power resulted from high winds, lightning and tornadoes. The licensee's event investigation report made the recommendation that planned diesel outages should not be scheduled ~during months when severe weather is likely.
Inspection Report
50-282/88012 and 50-306/88012 further documents the details of the July 27, 1987 partial loss of offsite power, noting that the licensee event report (LER) 87-015 inadequately documented the event and the-lessons learned. The licensee has not submitted a revised LER to correct the LER's- inadequacies. On June 7,1989, during the course of power maintenance testing of D2, the generator experienced a lockout of the No. 86 relay, which shut D2 down. -The relay was reset and D2 restarted. D2 then experienced another . lockout of the'86' relay. Subsequent troubleshooting involved verifying correct operation of all electrical relays and mechanical parameters which may have caused the lockout of the 86 relay. Results of the troubleshooting were reviewed and approval to restart D2 was given. Lockout of 86 relay was attributed to random fn11ure. Subsequent operability surveillance were performed on June 9 and 12, 1989, with no difficulties encountered. On June 22, 1989, a copper oil sensing line failed on the 12 diesel-driven cooling water pump. The failure mechanism was attributed to vibration induced fatigue. The licensee secured the affected pump and then proved the operability of the 22 diesel-driven cooling water pump and 02. A modification was then approved to replace the copper tubing with stainless steel tubing. Additionally, the system engineers for the emergency diesel generators and the diesel-driven cooling pumps were assigned the responsibility of walking down the engines to identify
_ _ - - - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _. ._
- - - -. - _ - _
- .
. ~ . pressurized copper lines susceptible to vibration induced fatigue and to evaluate the need to replace / alter the lines. The resident inspector will continue to monitor the licensees activities in this area. It should-be noted that, approximately two years prior, the 22 diesel-driven cooling water pump also suffered a similar failure. Through oversight, the licensee failed to identify and correct other areas susceptible to this failure mechanism. 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 were scheduled and performed within Technical _ Specification requirements, by observing that procedures were being(LCOs) were not violated, that followed by qualified operators, that Limiting Conditions for Operation system and equipment restoration was completed,- and that test results were acceptable to test and Technical Specification requirements. Portions of the following surveillance were observed / reviewed during the inspection period: SP 2032 A Sa.eguards Logic Test, Rev. 8 SP 1130 Containment Vacuum Breakers Quarterly Test, Rev. 18 SP 2186 D? Diesel Generator Operability Test, Rev. 19 SP 1106A 12 Diesel Cooling Water Pump Test, Rev. 23 SP 1544 Containment at Power Inspection, Rev. 13 SP 2544 Containment at Power Inspection, Rev. 15 The resident inspector accompanied licensee personnel into the Unit 1 and Unit 2 containment for the performance of SP 1544 and 2544. All aspects of the Unit I and Unit 2 surveillance were completed including the noting of valves which showed minute signs of leakage. The thoroughness of the Unit 1 inspection can be attributed to the presence of by an experienced senior maintenance supervisor. The plant equipment operator had not performed the surveillance before and was unfamiliar with exact equipment ! locations and ALARA considerations. An experienced person from operations ' was unavailable on the scheduled day. i Discussions with the licensee included why the surveillance was not delayed until an experienced individual was available (SP 1544 hos a l scheduled tolerance of -0, +14 days, which allows a 14 day delay) and the L appropriateness of assigning the surveillance to an inexperienced plant operator. In response to this discussion, the licensee has annotated the surveillance schedule with a note indicating that senior operations personnel should perform the at power containment inspections.
During the performance of SP 1130, the inspector noted that the containment vessel vacuum breaker isolation valve is stroke timed open after being stroke timed closed and being cycled in conjunction with the
I
l - - _ - -
,
.------- _ _ _
.
l setpoint verification of the containment to annulus differential pressure switch. Review of previous revisions showed that the stroke timing requirements were added in July 1986, in Revision 14. The order in which the testing is performed has not been changed in subsequent revisions, thereby providing a consistent stoke time data base through which degrading trends can be easily monitored. No violations or deviations were identified. 6. Licensee Action on Previous Inspection Findings (92701) (Closed)OpenItem(50-282/88010-01;50-306/88010-01(DRP)): Engineering _ Evaluation of Need for Improved Emergency Lighting and for Permanently Installed Catwalks and Access Ladders. This. engineering evaluation ha:1 been completed. It concludes that new emergency lights are to be installed at each steam generator blowdown flashtank and at the Unit 1 SI test valve locction. Modification 89Y945 was approved on May 19, 1989 for the new lights. The evaluation also concluded that dedicated portable ladders in the RHR pits and at the SI pumps were preferable to permanently installed structures which could interfere with access to other components at those locations. These new ladders are in place and appropriately labeled. (Closed) Open Item (282/85024-04; 396/85022-04): Post-Accident Emergency Ccaling Water Flow Requirement and Availability With a loss of off-site power, trip of both reactors and turbines, no immediate operator action to reduce cooling water flow, and failure of one of the two diesel-driven cooling water pump., (DDCLP) to start, the other DDCLP will be required to pump 19,800 gpm. Neither pump has been run in the plant at this required flow rate. The licensee has agreed to demonstrate this capability during the next two-unit shutdown provided such demonstration would not violate regulatory requirements in effect at the time or compromise nuclear safety. This agreement will be tracked as Open Item (282/89018-01; 306/89018-01). (Closed) Open Item (282/85024-05; 306/85022-05): Operability Requirements of the Safeguards Traveling Screens are Not Included in the Technical Specifications An amendment to the technical specifications currently being processed by NRR contains limitations on the inoperability of the traveling screens and the associated emergency cooling water line. 7. Information Notice Followup (92701) (Closed) Information Notice (282/87063-IN; 306/87063-IN): Inadequate Net Positive Suction Head in Low Pressure Safety Systems During review of this Information Notice by the licensee, a procedural inadequacy was discovered which could have, under certain post-LOCA conditions, resulted in RHR pump runout. (See also Inspection Report No. 50-282/88004(DRP); 50-306/88004(DRP)). Appropriate changes have been
_ - _ _ _ _ _ _ - _ _
- . . . - '- s t i e
- ..
.pe ' * ' , . - made in;the plant _EOPs and the operators have received training regarding
the changes. (See'alsoLER 282/88001). (Interim Report).Information Notice (282/87024-IN; 306/87024-IN): Operational Experience Involving Losses _of Electrical Inverters . Licensee review of this notice is not complete. (Closed). Information Notice (282/87034-IN; 306/87034-IN): Single Failures .in. Auxiliary Feedwater Systems Licensee review of this notice shewed that no corrective action'was - required at Prairie Island. (Closed)InformationNotice.(282/87044-IN;306/87044-IN): Thimble Tube. Thinning in Westinghouse Reactors > Eddy current testing of incore thimble tubes is performed'during each- refueling outage. 8. Exit (30703) The inspectors met with the licensee ' representatives denoted in Paragraph 1 at the conclusion.of the inspection on June 30, 1989. The inspectors discussed the purpose and scope of the inspection and the findings. lThe inspectors also discussed the likely.information content of the inspection report with regard to documents or processes reviewed by the~ 1nspector during the inspection.' The licensee did not identify any document / processes as proprietary.
l
_. _ _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ - _ }}