IR 05000282/1987003

Insp Repts 50-282/87-03 & 50-306/87-03 on 870215-0418. Violations Noted:Both Shield Bldg Maint Airlock Doors Found Open W/Reactor Coolant Temp Greater than 200 F & No Conditions of Tech Spec 3.6 Satisfied

ML20215L467
Person / Time
Site:	Prairie Island
Issue date:	04/30/1987
From:	Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20215L463	List: IR 05000282/1987003 ML20215L459 ML20215L465
References
50-282-87-03, 50-282-87-3, 50-306-87-03, 50-306-87-3, IEB-85-001, IEB-85-1, NUDOCS 8705120256
Download: ML20215L467 (9)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-282/87003(DRP); 50-306/87003(DRP)

Docket Nos. 50-282; 50-306 Licenses 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, Minnesota Inspection Conducted: February 15 through April 18, 1987 Inspectors: J. E. Hard

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M. M. Mose

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Approved By: a , 1 - t-Reactor rojectsSection Date Inspection Summary Inspection on February 15 through April 18, 1987 (Reports No. 50-282/87003(DRP);

50-306/87003(DRP))

Areas Inspected: Routine unannounced inspection by resident inspectors of previous inspection surveillances, ESF systems,findings,lity faci modifications, preparations forplant operational sa refueling, fire protection, plant management changes, followup of Licensee Event Reports, and TMI item Results: One violation was identified in the ten areas inspecte N$K0Ok282 PDR

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DETAILS 1. Persons Contacted

- C. - Larson, Vice President, Nuclear Generation P..Kamman, Superintendent, Nuclear Operations Quality Assurance E. Eckholt, Senior Nuclear Safety / Technical Services Engineer

• E. Watzl, Plant Manager

• D. Mendele, General Superintendent, Engineering and Radiation Protection R. Lindsey, Assistant to the Plant Manager

• Sellman, General Superintendent, Operations D. Schuelke, Superintendent, Radiation Protection G. Lenertz, General Superintendent, Maintenance b J. Hoffman, Superintendent, Technical Engineering {

• K. Beadell, Superintendent, Quality Engineering M. Klee, Superintendent, Nuclear Engineerin ,

R. Conklin, Supervisor, Security and Services D. Vincent, Project Manager, Nuclear Engineering and Construction J. Goldsmith, Superintendent, Nuclear Technical Services

• A. Hunstad, Staff Engineer A. Smith, General Superintendent, Planning and Services [

J. Proulx, NSP Battery Super. visor \

A. Vukmir, Site Services Representative, Westinghouse

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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 Advisor * Denotes those present at the exit interview of April-18, 198 . Licensee Action on Previous Inspection Findings (92701)

(Closed) Open Item (282/85004-01; 306/85004-01).

Requirement for semiannual SAC review of operational QA audit program has been added to NIACD (Closed) OpenItem(282/84008-02;306/84007-01).

Licensee has added leak rate trending requirement to SP 1070 and SP 207 (Closed) Open Item (282/84008-04; 306/84007-03).

Procedure SP 1070 (SP 2070) has been revised to make more consistent within the precedure the valve lineup verification requirement (Closed) Open Item (282/84008-06; 306/84007-05).

Procedure SP 1070 (SP 2070) has been revised to insure that actual rather than approximate leak rates are obtaine .. . .

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(Closed) Open Item (282/86002-01).

.This item was in Inspection closed Report No.previously/86006(DRSS).

50-282 (Closed) Violation (282/86011-02).

Licensed power level was exceede Corrective action instituted by licensee and future improvements in ability to monitor actual thermal power should prevent future violations of this typ (Closed) Open Item (282/86011-04; 306/86013-02).

Failure to provide adequate controls over modifications affecting systems important to safety. Significant organizational changes and comprehensive modification procedure changes have been made to improve performance in this are A review of long standing Open Items was conducted to determine if there were any items for which the expenditure of additional inspection effort was not justified. On the basis of this review, the following items are administratively closed out:

282/84014-01; 306/84016-01 - Improve performance checks of count room instrument /84016-01; 306/84017-01 - Changes in procedures for the uniform procurement progra /85004-02; 306/85004-02 - Add auditors' independence in procedure /83012-01; 306/83012-01 - Jumper bypass procedure needs revisio . Operational Safety Verification (71707)

Unit 1 was base loaded at 100% power except for (a) a forced outage from February 20, 1987 to March 3, 1987, caused by a steam generator tube leak, (b) a reactor trip on March 30, 1987, and (c) was off line on April 7 for a scheduled refueling and maintenance outage with a duration of 32 day Unit 2 was base loaded at 100% power except for reductions for surveillance testin 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 maintenanc . . . . . . _- . . - ,.. _ _ _ . _ -. . -

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On~ February 20,' 1987,Lat 8:45 p.m. the licensee began a controlled shutdown because of a primary to secondary tube leak in the No.12 steam generato The maximum measured leak rate was about 0.3 aallons per minute when the-

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shutdown began (the licensee is required to shut down when the leak. rate reaches one gallon per minute). The leak was. identified throug '

radiochemical analysis and condenser..offgas activity monitorin Eddy-current testing techniques used during the outage identified?

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a' total of 19 tubes-(including one leaking tube) having unacceptable-c - stress corrosion cracking defects in the tubesheet region on the~ hot

leg side. A11:19 tubes were plugge ,

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On March 3, 1987, during preparations for Unit 1 startup from theLforced outage described above, the senior resident inspector discovered that the 3

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shield building maintenance air lock doors were both open while the ,

reactor was no longer in cold shutdown. The Reactor. Coolant System l (RCS) temperature at the time.was about 425 degrees F., and cold shutdown

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is defined as less than 200 degrees F.

, Technical Specification 3.6, Containment System, A.1. states, " Containment

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integrity as defined-in Specification TS.1 shall not'be violate except when one of the following conditions exist: - (a) the reactor'is; in the cold shutdown condition with the reactor vessel head. installed, (b) the reactor is in the refueling shutdown condition with the' vessel -

A: head removed, or~(c) the fuel inside containment'has-not been used for '

m power operation."

In addition, Technical Specification 1.0, C. states, in part, " Containment system integrity exists when the containment vessel shield building, and

Auxiliary Building Special Ventilation Zone-(ABSVZ),are closed'and the-

. following conditions are satisfied . . 7. At least one door in each shield building airlock is closed."

None of conditions (a), (b), or (c) above were being met and thus with

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E both in violation shield ofbuilding maintenance the technical airlock doors specifications (282 op/87003-01(DRP)).

See Notice of Violation.

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On March 18, 1987, repairs to the Unit 2 instrument air line were being mad During this work, a failure of an air pressure regulator resulted.

' in the loss of air to an air operated caustic addition valve. This valve- -

! failed open and resulted in a minimal amount of caustic being introduced-

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into the line. The line was flushed, repairs made to the regulator, and ,

j the system restored.

i On March 29, 1987, while performing a routine turbine stop valve surveillance test, an AST celay coil in the Westinghouse NBFD relay failed.

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The coil was replaced and while coil testing was being performed, a second AST relay coil failed. This second relay coil was replaced, tested, and j- the stop valve test was then completed successfull *

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On March 30, 1987, at 2:19 p.m., the Unit 1 reactor tripped from 100%

power during the performance of safeguards logic testing. Cause of the trip was attributed to an.I&C technician skipping a step in the surveillance procedure. All systems responded as expected including safeguards Train B, SI pump start, RHR pump start, containment isolation, and reactor tri A small amount of concentrated boric acid (about 300

. gallons) was injected into the reactor coolant system (RCS). The injection lines were-flushed and the RCS deborated because Unit 1 was at the end of its operating cycle. .The reactor was made critical at 7:58 a.m. on March 31, and the generator placed on line at 10:02 This is the third trip attributed to safeguards logic testing in ti last 12 month On April 9,.-1987, at about 6:15 a.m. , Train A and B safeguards racks for Unit 1 were de-energized by pulling fuses to permit replacement of certain burned out control switch resistors. By so doing, power to the steam exclusion damper control circuits for Unit 1 and the operating Unit 2 were unknowingly disrupted thus rendering the steam exclusion system inoperable. This situation was noticed by an Operations Shift Supervisor _at 3:15 p.m. the same day. Power was restored and the steam exclusion system returned to normal at 4:29 p.m. A similar but unrelated event involving the steam exclusion dampers occurred on April 12. In this instance also the Operations group detected the situation within a few hours. Investigation and corrective action to prevent future such occurrences are in progress. No violations of-regulatory _ requirements occurre On April 7,1987 at 10:10 p.m. Unit 1 was taken off line for a scheduled refueling and maintenance outage with a duration of 32 day . < 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, activities were accomplished using approved procedures and were-i_nspected 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 qualif.ied personnel, radiological controls were implemented, and fire prevention controls were implemente Portions of the following maintenance activities were observed / reviewed-during the inspection period:

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02 Emergency diesel generator preventive maintenance

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Seismic monitor playback repair

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As noted in the^ previous-inspection report 50-282/87002(DRP);

.50-306/87002(DRP),-D-2 diesel generator: tripped during a routine:

. surveillance: test because of-low jacket coolant pressure. A special test-

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.was conducted'on March 5 and it was determined'that the cause of the lo .-jacket coolant pressure was air inleakage~at~the coolant pump mechanical seal. Repairs were made during the annual preventive maintenance (PM)

which began March.23. Visual inspection of the bearings during the annual

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- PM identified:the ' upper and lower crankshaft thrust bearings as having

excessive' wear. Both thrust bearings were replaced. A modification to the lubrication: system is expected to be installed by late 1987 and shouldimprovebearingilifesignificantly.'

Other activities observed that~are associated with the Unit 1 forced outage and/or! refueling outage included:

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MOVATS testing of limitorque valves for IEB 85-03-

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.Raychem splice: inspection / repair No.~ 12 battery charger preventive maintenance

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Eddy current inspection of steam generator tubes

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No. violations orideviations were. identifie . Surveillance (61726)

The inspector witnessed portions of surveillance testing'o'f' safety-related

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-systems'and components. The inspection included verifying that the tests 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 requirement ' Portions of the following surveillances'were observed / reviewed during

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.SP 2713 S.I. Pump Mini-recircline Functional Test r

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SP 2088 Safety Injection Pumps Test

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SP 1098 No, 12 Station Battery Load Test

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SP 1040A Seismic Monitor Test U

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SP 1100 No. 12 Motor Driven Aoxiliary Feedwater

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Pump Test l,

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Load testing of No. 12 Safeguard 4 ts is was performed during the report period. Capacity of this U nk . patteries has dropped to 89.6%

, (measured to 101 volts output).as compared to 95% measured in 198 According to the NSP Battery Supervisor, at this rate of degradation

,,  : No.12 Battery will only last 8-10 years as contrasted to the 20 year

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design life.' The reason for.the rapid rate of degradation is the high L ambient teinperature in the battery room which has averaged about F 95 degrees F over the past few years, and which has been measured as

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high as 107 degrees F by the resident inspection staff. The battery L . manufacturer, Exide, recommends a temperature range of 59-85 degrees F.

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/Accordingtoplantengineers,modificationstothebatteryroo l ventilation system to reduce the ambient temperatures have been

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requested. This matter will-be tracked in the future as'an Open Item (282/87003-02;.306/87003-02.)-

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No1 violations or-deviations were identifie .

- ESF System Walkdown (71710)~

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The inspector performed'a complete walkdown of the accessible portions-

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of Unit 2 Containment Spray and Caustic Addition systems. Observations included confirmation of selected portions of the licensee's procedures,

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checklists,-plant draw'ings, verification of correct valve and power suppl breaker ~ positions to insure that. plant equipment and instrumentation are-properly aligned, and local system indication to: insure proper operation within prescribed limit No: violations or' deviations were identifie . Preparations For Refueling (60705, 86700)

The Unit.1 refueling outage currently in progress is the Cycle 12 reload-and includes the following differences from previous cycles:

(a) -The first core reload which replaces more than 1/3 of the core

- to increase' cycle length from 370 effective-full day (EFD) to F 431 EFD.

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(b) The first core reload with a positive isothermal temperature i~ coefficient (ITC) and with potential derates from limiting

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total peaking factor (FQ) at the beginning of core life and

. ~end of core life. These are the result of a modeling analysis L error done by the nuclear analysis department and will require

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a technical specification change approval (this error was not

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h in the computer model for the first 11 cycles).

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-(c) Replacement of all rod cluster control assemblies (RCCAs) as described below.

Replacement of all Rod Cluster Control Assemblies (RCCAs) will be accomplished during this. outage because of-extensive stress and radiation induced cracking observed in the absorber rodlet' cladding. (See also IE Information Notice No. 87-19). The new RCCA design differs from the

, , existing design only in minor respects; via, a slightly reduced absorber diameter near the rodlet tip, tighter chemistry specifications for the cladding, plus the addition of chromium plating to the cladding. The'

, inspector reviewed the modification package and safety evaluation for

'. this change and will be observing the rod drop timing test before Unit,1 startup. On this basis it appears that the pertinent requirements of 10 CFR 50.59 regarding control of changes made in the facility without prior Commission approval is acceptable.

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-As of the end of'this-inspection reporting period, the reactor vessel-head removal and associated refueling activities is scheduled to take

place during the next inspection report period. Details of those activities will be provided in the next repor No violations or deviations were identifie . -Facility Modifications (37700)

A number of modifications associated with the Unit 1 refueling outage are in progress. The-followingisalistingofthemajormodificationsbeing

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worked:

- Redundant fusing for Appendix R; Modification 86L927

- Control Room Panel B and D Modifications

- Reactor Vessel Level Indicating System (RVLIS) Modifications-Control Rod Drive Mechanism (CRDM) Vent. Modification Replacement of RCCAs (See Section 7 Above)

The resident inspection staff.is monitoring the progress of these ongoing modifications and will provide details in the next repor , Licensee Event Reports Followup (92700)

Through direct observations, discussions with licensee personnel, and review of records, the followin'g 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 Specifications:

(Closed) 282/86011-LL Both diesel driven cooling water pumps out of service-(Closed) 282/87001-LL Diesel generator D2 tripped during surveillance testing (Closed) 282/87002-LL Tube leakage in No. 12 steam generator-(0 pen) 282/87003-LL Unit 1 shield building maintenance airlock double doors found open (0 pen) 282/87004-LL Reactor trip during safeguards logic testing 1 Closecut of Temporary Instructions (TI) (92703)

T.I. 2515/69, Inspection of Response to IE Bulletin 85-01, steam binding of auxiliary feedwater pumps. The Bulletin was closed previously in Inspection Report No. 282/867007; and No. 306/86007. All requirements of the T.I. have been me . Fire Protection (64704)

During the week of March 2, 1987, a special inspection was conducted by NRC Region based / contract personnel to ascertain the licensee's compliance with 10 CFR Part 50 Appendix R (" Fire Protection Program for Nuclear Power

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Facilities Operating Prior _to January.1, 1979"). Two Level IV violations were identified in the Inspection-Report-(50-282/87004 (DRS);_

50-306/87004(DRS)). One of the violations _ relates to lack of circuitry cocrdination and resulted in the immediate implementation of compensatory measures-in the form of fire watch tours every 20 minutes in extensive areas of the auxiliary building as well as the auxiliary feedwater pump rooms. The resident inspection staff is auditing this fire watch progra . Plant Management Change On. April 1, 1987, the plant manager announced significant organizational changes including the establishment of shift manager positions in operations and the separation of the Operations and Maintenance Department by establishing a position for general superintendent of Operations and General Superint6ndent of Maintenance independently reporting to the plant manager.. Other changes include an assistant to the plant manager, a general superintendent of planning and services, and a general

. superintendent of engineering and radiological protection. These and-other changes are part of an ongoing effort by the licensee to further improve plant performanc . Exit Interview (30703)

The inspectors met the licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on April 21, 1987. The inspectors discussed the purpose and scope of the inspection and the finding ,

The inspectors _ also discussed the likely information 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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