IR 05000282/1987012
| ML20238E168 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/04/1987 |
| From: | Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20238E151 | List: |
| References | |
| 50-282-87-12, 50-306-87-11, NUDOCS 8709140184 | |
| Download: ML20238E168 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION III
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Reports No. 50-282/87012(DRP); 50-306/87011(DRP)
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i Docket Nos. 50-282; 50-306 License No. DPR-42; DPR-60 J
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licensee:
Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401
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Facility Name:
Prairie Island Nuclear Generating Plant I
Inspection At:
Prairie Island Site, Red Wing, Minnesota l
Inspection Conducted:
July 12 through August 22, 1987 i
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Inspectors:
J. E. Hard I
M. M. Moser
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f Approved By:
R. O ay
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Reactor Projects Section 2B Date
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Inspect _ ion Summary l
Inspection on July 12 through August 22, 1987 (Reports No. 50-282/87012(DRP);
j 50-306/87011(DRP))
i Areas Inspected:
Routine unannounced inspection by resident inspectors
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l of previous inspection findings, plant operational safety, maintenance, i
surveillance, ESF systems, facility modifications, spent fuel pool j
activities, fire protection, trair.ing, followup of Licensee Event Reports,
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medical emergency drill, and region request.
l Results:
l Three violations were identified in the 12 areas inspected.
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DETAILS 1.
Persons Contacted C. Larson, Vice President, Nuclear Generation L. Eliason, General. Manager, Nuclear Plants P. Kamman, Superintendent, Nuclear Operations Quality Assurance E. Eckholt, Senior Nuclear. Safety / Technical Services Engineer
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- E. Watzl, Plant Manager l
D. Mendele, General Superintendent, Engineering and Radiation Protection
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R. Lindsey, Assistant to the Plant Manager.
M. Sellman, General Superintendent, Operations D. Schuelke, Superintendent, Radiation Protection G. Lenertz, General Superintendent, Maintenance J. Hoffman, Superintendent, Technical Engineering
- K. Beadell, Superintendent, Quality Engineering M. Klee, Superintendent, Nuclear Engineering i
R. Conklin, Supervisor, Security and Services
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D. Vincent, Project Manager, Nuclear Engineering and Construction J. Goldsmith, Superintendent, Nuclear Technical Services
- A. Hunstad, Staff Engineer j
- A. Smith, General Superintendent, Planning and Services j
A. Vukmir, Site Services Representative, Westinghouse
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The inspectors interviewed other licensee employees, including members of
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the technical and engineering staffs, shift supervisors, reactor and i
auxiliary operators, QA personnel, Shift Technical Advisors, and Shift Managers.
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- Denotes those present at the exit interview of August 24, 1987.
I 2.
Licensee Action On Previous Inspection Findings (92701, 92703)
(Closed) 282/87003-01 Violation:
Maintenance Airlock Doors Left Open.
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During Unit 1 startup preparations, the shield building maintenance airlock doors were found by the senior resident inspector to be open at a time when at least one of them should have been closed.
A procedural change to prevent recurrence has been made.
Additional corrective actions which will take longer to review and institute will be tracked under Open Item 282/87005-01.
(Closed) 282/87009-02 Unresolved:
Unit 1 NIS power range off by 10%
after refueling. This unresolved item has been determined to be a violation (282/87012-02(DRP)).
See Section 3 of this report.
3.
Operational Safety Verification (71707, 93702)
Unit 1 and Unit 2 were base loaded at 100% power except for reductions for surveillance testing.
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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, and to verify that maintenance work requests had been initiated for equipment in need of maintenance.
The NRC program RCSLK9:
Reactor Coolant System Leak Rate Determination for PWRs was used to independently verify the licensee's leakage calculations for Unit 2.
Determinations were made on two successive days with the following results:
Licensee Calculated NRC Calculated Date Leak Rate Leak Rate 7/20/87 0.1471 gpm 0.20 gpm 7/21/87-0.077 gpm-0.11 gpm i
These results are judged to represent good agreement between the licensee and NRC.
At 3:45 p.m. on 6J1y 27, 1987, with both units operating at 100% power, the licensee declared an unusual event when a tornado was sighted within two miles of the plant.
High winds and lightning strikes disabled power lines and interrupted power to the No. 1 reserve transformer, all cooling tower fans, No. 21 cooling water pump, Units 1 and 2 operating charging pumps and the No. 10 bank transformer.
Automatic closure of the alternate offsite power sources (161 KV line from Red Wing) restored power to vital buses 15 and 16.
Reactor operations continued while the licensee bought affected components back online.
The notification of an unusual event was terminated at 6:14 p.m.
LER 282/87015 is being issued for this event.
As noted in Inspection Reports (282/87009; 306/87009(DRP)) en June 19,
1987, emergency diesel generator No.1 (DG1) inadvertently started on receiving an undervoltage signal.
Investigation by the licensee has determined that the cause of the DG1 autostart was a personnel error in exceeding the scope of the work instructions while performing maintenance to replace a malfunctioning undervoltage relay.
This failure to control l
work activities which had an adverse effect on plant safety-related systems is a violation.
See Notice of Violation (282/87012-01(DRP)).
On May 28, 1987, Unit 1 was in its power escalation program following the Cycle 11-12 refueling.
Calibration of the NIS power range channels is
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done based on calorimetric data taken during the startup physics testing i
program.
At an indicated power of 34.6%, calorimetric data showed that-
-l actual power was 45.7%.
The effect of this inaccuracy is that the power i
range high flux low setpoint would have tripped the reactor at about 33%
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Technical Specification Paragraph 2.3.A.1.b specifies a setpoint l
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of 25%.
This is a> violation of Technical Specifications.
See Notice of
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.t Violation (282/87012-02(DRP)).
This item was discussed by the licensee
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in a meeting with Region III representatives injGlen Ellyn, Illinois on July 15, 1987.
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At 4:07 a.m. on July 31,.1987, with both Unit 1 aFrd Unit 2'onerating at 100% power, No.10 bank transformer was inadvertently disabled by a relay l
crew that was intending to investigate false operation of the No. 10 bank backup relays.
Due to personnel error, the re by crew inadvertently tested GH2 breaker which disabled No. 10 bank. ' Loss of No. 10 b'ank transformer caused the Number 1 reserve transformer to lockout, 'disabMng all cooling tower fans, No. 21 cooling water pump, and the Unit 1 of va-ting charging pump.
Safety systems responde'a as expected (i.e., No. I and No. 2 emergency diesel generators and No. 12 powered diesel cooling
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water pump autostarted).
This failure to follow procedures is a violation
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l of Technical Specification 6.5.
See Notice of Violstion (282/87012-03(DRP)).
4.
Maintenance Observation (62703)
Routine, preventive, and corrective maintenance activities (on safety-related systems and components) listed below were observed / reviewed 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 initiating the work, cctivities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components
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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.
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Portions of the following maintenance activities were observed / reviewed during the inspection period:
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Electrical Repair of Guardhouse Emergency Diesel Generator
No. 12 Diesel Driven Cooling Water Pump Preventive
Maintenance D2 Emergency Diesel Generator freventive Maintenance
No. 122 Spent Fuel Pool Heat Exchanger Divider Plate Repair
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 requirenants,;
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.
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Portions of the following surveillance were observed.' reviewed during the i
inspection period:
SP 2001aa Reactor Coolant System Leakage Te t
SP 2093 02 Diesel Gknerator Manual and 4KV Voltage Rejection-Restoration Scheme Test, Bus 16
SP 1661 Monthly Guardhouse Emergency Diesel Generator i
Operation.
SP 1630, 1631, 1701, 1702, 1721 Authentication of Various i
Codes Used By Nuclear Engineering.
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SP 1074 Auxili ry Building Special Ventilation Zone Test
+ ' SP 1172 Ventila: ion System Monthly Operation Test
- 9 1502 Protected Area Weekly Fence Inspection
SP'1003,
Reactor Protectim System Tests
SP 1728 Siren Test On August 1E 0937, V th Unit I and Unit 2 at 100% power, a routine monthly surveillauie for isolation testing of turbine cooling water header valves on Unit I was being performed.
This test verifies the l
operability of thd Unit I turbine building cooling water header valve and the ability of the control system to isolate the Unit 1 turbine i
building header from the cooling water system upon simultaneous high l
fiow and low pressure. signals on loop A cooling water header.
During tha surveillance it was found that the flow switch was inoperable.
Loop A cooling water header was declared out of service and the licensee commenced' the required LC0 testing specified by T.S. 3.3.D.2.C.
l Investigation determined ttat the high side sensing line was plugged and i
after clearing the line, full operation of the high flow seasing switch i
was restored. The surveillance was completed successfully.and loop A l
cooling water header was declarad operable the same day c I
i No violat. ions or deviations were identified.
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6.
ESF System Walkdown (71710)
The inspector performed a complete walkdown of the accessible, portions of Unit 1 and Unit 2 auxiliary feedwater systems.
Observations included confirmation of selected portions of the licensee's proctrdant,,' cincklists, plant drawing, verification of correct valve and power supply breaker positions to insure tnat plant eqaipment ard instrumentation are properly aligned, and local system indication to insure proper operation within prescribed limits.
No violstions or deviations were identified.
7.
Modifications (37700)
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I The post TMI Cochustion Engineering Corporation core exit thermocouple systems that were installed at Prhirie Island Units 1 and 2 in 1981-82 have been downgraded in accuracy due to a connector leakage problem discovered by Combustion Engineering during recent testing.
This leakage is expected to occur during accidents.that result in high containment
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moisture content resulting in a random " battery effect" error on the thermocouple millivolt signal.
This affects the calculation and display ofsubcoolingontheInadequateCoreCodingMonitor(ICCM)andastudyof the licensee s installation results in a maximum subcooling error of
i 40 F.
Many of the licensee's emergency operating procedures (EOP)'s use action levels considerably less than'40 (e.g., i 10, i 15 ).
Therefore, an error band of i 40 does not appear to be compatible with the E0Ps.
On July 16, 1987, the nuclear engineers issued a memo to operations personnel identifying this problem and suggesting that other methods be used to determine reactor coolant system subcooling.
The licensee is investigating possible solutions to this problem and this item will be considered open pending the licensee's resolution in this matter.
Open Items (282/87012-04; 306/87011-01(DRP)).
No violations or deviations were identified.
8.
Spent Fuel Pool Activities (86700)
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Fuel Rod Consol'idation l
Observations and commitments from the meeting of the licensee and NRC in Bethesda on August 14, 1987 regarding the demonstration program for fuel rod consolidation are as follows:
(1) Administrative control of the work will be through the NSP modification control process.
The modification package, which had not yet been prepared, will specify the procedures to be followed.
(2) Westinghouse employees will be performing the rod consolidation work under the direction of NSP Nuclear Engineers.
A Nuclear Engineer will be present onsite at all times during the work.
(3) The degree of Quality Control inspection to be applied during the work had not been decided.
(4) The plant organizations other than Nuclear Engineering are to be kept informed of the status of the consolidation work by reports during the morning meetings. Westinghouse is to provide a representative to each morning meeting.
(5) Spent Fuel Pool Emergency Procedures - The consolidation team will be required to read plant procedure D5, Reactor Refueling Operation.
(In later conversations, the Superintendent-Nuclear Engineering stated that the team will also be required to review the pertinent portions of C17, Fuel Handling System.)
(6) Boron concentration in the spent fuel pool will be maintained at or above 1800 ppm during rod consolidation operations.
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(7) The entire rod consolidation process including radiation protection aspects is being-reviewed by NSP.
A demonstration and training run cf the' equipment at Pittsburgh is to be
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vitnessed=by H5P.
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[E) IO Water temperature in the spent fuel pool reaches 120 degree's F., the rod consolidation work is to stop and fuel removed from g
tra consolidation equbment.
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On Auguit 20,1637 the resideat inspectc's attended meetings between
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the lichnt,ee anit'the Prairie Island cituens to oiscuss the fuel : rod
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conzlidation prcgram.
The Jnspectors explained the NRC role in monitoring the 6tmonstration program.
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Fuel Assembly Capeltygling t
In preparation for the fuel rod consolic'ation demonstration program _
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described above, an empty fuel assembly cage hAo a number of metal
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samples removed for analysis.
This work was performed on August 20 i
and,21, 1987 by a contractor using specialized tools with support j
from Rad Protection,;0perations, and Nuclear Engineering.
The samples will be useo to determine burial site requirements for t'he n
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empty compacted fuel assembly cages after the completion of fuel rod
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consolidation.
i No violations or deviatianstwere identified.
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9.
Training (41701)
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Rer; qualification exans were administered by the. MRC to 17 license holders durino the ;;eriod June 16-26, 1987.
Since 53% of the car.didates (9 of
',7) failed one or more portions of the examination, the Prairie lisland
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pequalification program was rated as unsatisfactory.
An KRC Confirmatory Action Letter of August 18, 1987 documents the licenses commitments for j
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corrective action which wece discussed in a meeting in Glen Ellyn,
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Illinois on August 10.
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On August 7 the licensee announced that_ effective that date, the training function at Prafrf( Island would be reporting to the General Manager,
-Nucidr Plants.
Inis places Training more directly in the line
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organizc tfon dich manages nuclear plant operation.
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During ttt ternado event on July 27, 1987 (See Section 3 above), both resident inspectors were present in the control room to observe the responses of the operating l crew.
Sptcii'ic-responses to the electrical transients included restarting many electrically operated conponents..
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,l which had be6h disabled, and realignment of powr supplies into the plant and within the plant as conditions changed on the distribution system.
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Both units reained in full power operttion during the event.
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. Performance of the operating crew and the. guidance provided by plant
' milgineerr, was judged to be excellent and reffective of pood training
,;.rdtices and a well qualified plant staff.
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Licensee' E0ent Reports' Followup - (92700)
Through directTobservations,; discussions with licerisee personnel,.and review of. records, the following event reports were. reviewed to determine that deportability requirements were fulfilled,-immediate. corrective action wasLaccomplished, and corrective action to prevent recurrence had been accomplished in accordance.with= Technical Specifications:
'(Closed) 282/87006-LL
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' Auto Start'of N'o.12 Auxiliary 282/87006-LL, Rev.'l-Feedwater Pump
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During a Unit 1 outage ~and planned n aindown of No. 12 steam-generator, No. 22 AFW pump. started because of S.G. low level.
This' was the result of personnel failure to follow plant.
rocedures which call for pump. controls to be placed in p' Manual" before draining the S.G.
All operating personnel will review the event.
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(Closed) 282/87010-LL Auto-Start of No. 1 Emergency Diesel Generator
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l-Craftsmen performing a monthly undervoltage relay test on a vital bus found a relay out of specification and decided, with their supervisor's concurrence, to replace the relay.
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replacement, diesel generator No. I experienced an unplanned start.
The replacement had not been authorized as required
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by the work procedure.
This matter was. discussed with the l
personnel involved and the report distributed to work supervisors.
(This event is also discussed in Section 3 above.)
(Closed) 282/87012-LL Three Surveillance Tests Dane Late A personnel error in the scheduling of electrical surveillance resulted in performing three surveillance tests 4-6 days. late.
A copy of the report will'be circulated among responsible supervisors to point out the importance of double-checking the surveillance dates.
The system for scheduling surveillance procedures is judged to be adequate.
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Emergency Medical Drill (82301)
q On July 28, 1987, the inspectors observed the. licensee's. annual emergency I
medical drill in order to assess.the responsiveness ~of both the plant
. staff and the pre-hospital and hospital emergancy medical services to-l handle a, contaminated and injured patientc Overall responsiveness and j
effectiveness was very good with exceptional performance being.
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demonstrated by the Plant Emergency Medical Team in rendering aid'to the j
injured and. contaminated worker.
Communications and coordination with j
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the control room were also exemplary.although the inspectors noted that
an initial PA announcement would have been helpful to alert plant-staff-l
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during the first minutes of the drill.
It should be noted that arrange-l ments with St. John's Hospital in Red Wing, Minnesota are being mad 5: that
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would permit NRC personnel with proper identification access to the
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emergency room should the need arise.
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No violations or deviations were identified.
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Fire Protection (64704)
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As noted in Inspection Report (282/87003; 306/87003(DRP)) the results of
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a special NRC insp?ction identified several areas which were in violation i
of 10 CFR Part 50 Appendix R (" Fire Protection Program for Nuclear Power
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Facilities Operating Prior to January 1,1979").
One of the violations involved a lack of circuitry coordination and resulted in the immediate implementation of compensatory measures in the form of specific fire watch tours cuery 20 minutes.
On July 19, 1987, an employee hired temporarily to perform fire watches was observed sleeping in the administration building conference room by the shift manager who had noticed on his routine building rounds that the
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5:40 a.m.' fire watch patrol had not been completed.
Normal fire watch patrols were immediately resumed.
The employee was terminated the following day and an improved fire watch log / turnover system was implemented.
No violations or deviations were identified.
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Region Request (92701)
l In response to a regional request from the Deputy Director, DRP, dated July 22, 1987, on the subject of a potential generic safety issue involving containment penetrations for process piping fabricated by
Tube Turns Corporation, the following information was provided on
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July 24,1987:
a.
Who performed the hydro testing?
(Tube Turns or licensee) The process piping with flued head welded subassembly and all pipe used
for subassemblies were hydro tested by Tube Turns. The licensee hydrotested the installed containment penetration.
b.
To what Codes were (Hydro) tests conducted? All piping was
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hydrostatically tested in accordance with ASTM A530 Paragraph 6 which requires hydrotesting to 60% of the minimum specified yield strength.
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Were hydrotests done prior to or subsequent to installation?
Process pipe and all subassemblies were shop tested and then tested
after installation.
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Did testing of the penetration assembly include a 100% visual inspection of the welds? Yes, during the system hydrostatic test.
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How many Tube Turns process' piping penetrations are used at your e.
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Unit 1 and Unit 2 have 9 Tube Turn penetrations.
facility?
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Exit (30703)
The inspectors met with the licensee representatives denoted in i
Paragraph 1 at the conclusion of the inspection on August 24, 1987.
l The inspectors discussed the purpose and scope of the inspection
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and the findings, ca The inspectors also discussed the likely information content of the m
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I inspection report with regard to documents or processes reviewed by 't I
the inspector during the inspection.
The licensee did not identify 7, any document / processes as proprietary.
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