IR 05000298/1985029

Insp Rept 50-298/85-29 on 851001-1130.No Violations or Deviations Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings,Annual Emergency Preparedness Exercise & Spent Fuel Shipment

ML20137H051
Person / Time
Site:	Cooper
Issue date:	01/10/1986
From:	Dubois D, Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20137H043	List: IR 05000298/1985029 ML20137H039
References
50-298-85-29, NUDOCS 8601210422
Download: ML20137H051 (15)

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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-298/85-29 License: DRP-46 Docket: 50-298 Licunsee: Nebraska Public Power District P. O. Box 499 Columbus, Nebraska 68601 Facility Name: Cooper Nuclear Station (CNS) Inspection At: Cooper Nuclear-Station, Nehama County, Nebraska Inspection Conducted: October 1 through November 30, 1985 Inspector: m h / d 8d

  [L.(OuBis,SeioYResi Inspector (SRI) Otte ~

Approved: - A .M // #d _P.MapoKI:hief,?rojectSectionA

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J Datie R cter Frojects Branch Inspection Summary Inspection Conducted October 1 through November 30, 1985 (Report 50-298/85-29) Areas Inspected: Routine, unannounced inspection of licensee action on-previous inspection' findings, annual emergency preparedness exercise, spent fuel shipment, core power distribution limits surveillance, LPRM system calibration, APRM system calibration, core thermal power evaluation, licensee event reports; followup, plant shutdown /startup, operational safety verification, and monthly surveillance and maintenance observations.: The l. '

+  inspection involved 176 inspector-hours onsite by one NRC inspecto Results: Within the 12 areas inspected, no violations or deviations were
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DETAILS

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      . Persons Contacted Principal Licensee Personnel

. *P. V. Thomason, Division Manager of Nuclear Operations

 *E. M. Mace, Plant Engineering Supervisor
 *J. M. Meacham, Technical Manager
 *R. L. Beilke, Chemistry and Health Physics Supervisor
 *H. T. Hitch, Acting Administrative Services Manager
 *R. Brungardt, Operations Manager
 *J. Sayer, Acting Technical Staff Manager NRC Personnel
 * M. McNeill, Project Engineer, Region IV
 * Indicates presence at exit meeting held November 22, 1985 Licensee Action on Previous Inspection Findings (Closed) 8326-02 (Unresolved). This item concerned inadequate licensee procedural guidance for implementing 10 CFR 50.59 requirements. Since this unresolved item was identified, CNS Engineering Procedure 1.13,
 " Station Design Changes," was replaced by CNS Engineering Procedures 3.3,
 " Station Safety Evaluations," and 3.4, " Station Design Changes." Also, Procedure NEP-10 was replaced by NEDP-09 and NEPD-10, which contain, with the exception of the cover pages, exact duplicate copies of Procedures and 3.4 respectivel The SRI's review of this item concluded that adequate procedural guidance is presently available.as a result of the following procedures revisions:
 . The licensee's 10 CFR 50.59 analysis sheet, Attachment "A" to Procedure 3.3, Revision 3, now states, "Does this proposed activity constitute a change in the facility or procedures as described in the
 , Updated Safety Analysis Report? Yes No .
    "

If the answer is

 "yes," additional guidance states that the change is reportable and-will be included in the CNS Annual Repor . CNS Procedure 3.3, Section II.A.3 requires that the CNS Annual Report
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contain a brief description of changes, tests, and experiments, including a summary of the safety evaluation of eac . . .,. _

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  .. CNS Engineering Procedure 3.5,'Revis' ion 0, "Special Test Procedures /Special Procedures," requires that a 10 CFR 50.59 Reportability Analysis Fo,rm be prepared as an attachment to any special test procedure prior' to its approval and implementatio Procedure 3.5 replaced former Procedure 1.12 with the same titl '

i This item is close (Closed) 8420-03 (0 pen Item). . This item concerned the lack of a ' licensce procedure for preventing installation of older model 120V AC HFA relays into safety-related systems. IEB 84-02 was issued'by the NRC as . result of an increased failure rat'e of General Electric (GE) type HFA

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relays. As a resulf'of IEB 84-12 and am increasing number of HFA relay

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failures at CNS, the licensee completed the replacement of all older model HFA relays in safety-related systems. The replacement relays were the GE

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  " Century Geries" model .

Licensee Maintenance Procedure 7.3.16,'" Low Voltage Relay Removal and Installation," Revision 2, was approved and implemented ou August 8, 1985.

The revised procedure included procedural steps for ensuring that all HFA relay replacements in safety-related systems were. the " Century Series" model. A checklist included as Attachment "A" to Procedure 7.3.16 contains procedure completion check points, relay specifications and data blanks, and Q.C. verification signatures for low voltage relay removal and ' installation. This item is close !

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  .(Closed). 8508-01 (Violation). This item concerned a failure of CNS

~; Procedure.0.9, " Equipment Clearance and Release Orders," Section II.A.3, to include an independent verification of switches, breakers, and other components associated with a safety-related system when that system or portions thereof, are removed from or returned to service. The FRI verified that Revision 2 of Procedure 0.9, dated May 13, 1985, requires the additional verification check This item is close i 3 .~ Annual Emergency Preparedness Exercise .

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*  The CNS annual emergency exercise occurred on October 16, 1985. Tl.e SRI ,
,  acted as an NRC observer in the Control Room and Technical Support Center
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areas. His specific comments were provided to the NRC exercise team leader and included in NRC Inspection Report 50-298/85-28. Also, the SRI !~ reviewed the minutes of the CNS Safety Operations Review Committee (SORC) meeting conducted October 17, 1985, to verify that an adequate review and - evaluation of the exercise was performed by plant management.

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, Spent Fuel Shipment'
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2 The SRI inspected the licensee's activities associated with a shipment o ' '

   . spent fuel.from CNS. ~ Included were observations and reviews of applicable
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  , . 1 procedures,l documentation, surveys, inspections, and shipping document.

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The SRI verified by review of. licensee docuraentation, through discussions i with responsible personnel, and by independent inspection that-the- '

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l, . Receiving' inspection of railcars and shipping cask ..

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   . Shipping document ;  , . Advance notification of and approval by affected state and federal
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   . Proper placarding of the transport vehicle .
   . Appropriateilabeling'of the spent fuel shipping casks.

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  * . Establishment of provisions for response by escorts and-local law enforcement agencie ~ . - Training of escort personne Testing of communications system ~
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   . Continual manning of the licensee's communications center (Movement Control).

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,   . Testing of fuel and cask handling cranes, hoists, and tool , Proper loading and sealing of the spent fuel shipping cask .- Surveillance of area radiation monitors, ventilation, systems, Land spent fuel pool water level and chemistr . Update of fuel location and accountability record . Applicable quality assurance audits and inspection i
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   . . S. Department of Energy and U. S. NRC " Nuclear Material Transaction Report," DOE /NRC Form 74 ,
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    . Bill of Lading ~.
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    . fCNSHealthPhysicsProcedure-9.5.3.7,"CaskIF-300 Shipment,"
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  .  . CNS Nuclear Performance Procedure 10.27, " Cask IF-300 Handling and       ,
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    .. CNS HP-138, " Contamination Survey     Sample Count Data Sheets."'
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    ~ CNS HP-141', " Contamination Survey - Railroad Car for IF-300,
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Irradiated Fuel Shipping Cask."

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n .' ' CNS HP-142, " Contamination Survey of IF-300 Shipping Casks."

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CNS HP-143,;"_ Radiation -

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cv .t," CNS HP-302, " Contamination Survey -:Tennelec Sample Count Dat ".

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d . CNS HP-608, " Spent Fuel Shipmant Checkoff Sheet and Certificate of tCompliance:of Number 9001' Conditions for Shipping Spent Fuel." ~

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h; " - ~ The.following independent radiation and contamination surveys were

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    . ~Co'ntasination surveys of;the shipping casks surface ,

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   , Contamination surveys of ths cask transport vehicles. '    ~
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   < . ^ The SRI reviewed:CNS Procedure 10.27,-Revision 2, dated November 1,'1985,
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   * Procedure 10.27 specific handling _ instructions _for the G.E. Type.IF-300       '

w., , Espent fuel shipping cask. JAlso included within Procedure 10.27 was- . _ .

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Attachment "A," " Handling and Loading of 'IF-300 Spent Fuel Shipping Cask Checkoff Sheet." The checkoff sheet provided'two functions: it identified .important steps used in the receipt, inspectio'n, preparation,

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c'x , movement, loading with fuel, leak testing,~ decontamination, loading'of.the

i- - +n cask onto the transport vehicle,'and final preparation for shipping; and , _ ,

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it provided a checkoff -list including spaces for-signatures and/or <

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initials of personnel who performed or witnessed the performance of key ~

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steps..of=the procedure. The SRI verified that Attachment "A" of Procedure _ 10.27 was properly' completed, signed, and date '

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 'The NRC inspector attended a special meeting conducted on-site by the licensee on November 18, 1985. Attendees included representatives from the licensee, shipper, and federal agencies. The meeting consisted of a final review of licensee preparations and documentation of activities relevant to the shipment of the spent fuel from CNS. The licensee determined that all preparations and requirements were completed. The spent fuel shipment ^1 eft the CNS at approximately 2:30 p.m. on November 18, 1985, and arrived at the G.E. Morris Operation Complex, Morris, Illinois, on November 19, 198 The shipment consisted of-2 spent fuel shipping casks, each of which contained 18 spent fuel bundles. The spent fuel cask identification numbers were:
 . IF-301
 . IF-302
 'The observations, reviews, and independent measurements were conducted to verify'that spent fuel handling and shipment operations were in conformance with the requirements established in the CNS Operating License and Technical Specification No violations or deviations were identified in this are , Core Power Distribution Limits Surveillance
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The SRI conducted discussions with CNS reactor engineers and operations department personnel, reviewed records and data applicable to core thermal limits, and verified that appropriate corrective actions were taken when core thermal data indicated an approach to limiting conditions. The 7 review and discussions included the following:

_ . Verification that the linear heat generation rate (LHGR), core maximum peaking factors (CMPF), minimum critical power. ratio (MCPR), and average planar linear heat generation rate (APLHGR)', were within prescribed Technical Specification limit *

 . Examinations of local power range monitor (LPRM) and BASE
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distribution calculations as typed out by the~00-1, "LPRM Calibration

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 -and Base Data," on-demand typewriter. Typed alarms, errors, and
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other inprocess messages were also reviewe ,

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 . Verification that traversing incore probe (TIP) machine normalization factors were properly obtaine ,
 . Examination of licensee procedures for ascertaining operation within licensed limits, should the process computer become unavailabl .

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 . Verification that average power range monitor'(APRM) channel gains were adjusted as necessary following an-LPRM calibratio . Verification that following an APRM gain adjustment, a subsequent P-1
,  .was run to assure that APRM gain adjustment factor (GAF) reflected such gain adjustment . Examination of licensee procedures which are used to correct abnormal
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core thermal. condition The following' computer printouts edited from August 20 through October 5,'1985, were reviewed: 00-1, "LPRM Calibration and BASE Data"

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 . 00-3, " Core Thermal Power and APRM Calibration" ,
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 . OD-6, " Thermal Data for Specified Fuel Bundles"
 . 00-7, " Control Rod No'tch Positions"'
 . 0D-8, LPRM Console: Readings"
 . 00-10, " Edit Specified Data Array",

, . 00-15, " Computer Shutdown and Outage Recovery Monitor" >

 . OD-16, " Target Exposure and Power Data"
 . 00-17, " Edit Periodic Core Performance Logs"
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 . P-1, " Periodic Core Performance" The following CNS procedures were reviewed:
 . 10.4, Revision 8, " Core Thermal Hydraulic Evaluation"
 . '10.-7, Revision 8, " Maximum Average' Planar and Peak Linear Heat Generation Rates and Minimum Critical Power Ratio"  ,
 . 10.8, Revision 9, " Reactivity Follow Check"
 . 10.9, Revision 13, " Control Rod Scram Time' Evaluation"

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 . 10.10, Revision '7, " Limiting Control Rod Pattern Determination"
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 . 10.11, Revision 8, " Control Rod Sequence Exchange"
 . 10.13, Revision 13, " Control Rod Sequence and Movement Control"
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being operated within licensed' power distribution limit ,.

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LPRM system which were' performed during.the period August 20 through

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Accurate determination of individual LPRM calibration currents,

"o      ~ calibration constants, and amplifier gain' adjustment factor '
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control. rod movement:did not occur during the performance;of the LPRM;

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     . AllLPRM'detectorreadingswere.obtained!prlortoandfollowingLPRM
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amplifier gain adjustmen '

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     . -5PRM channel recalibration followed the-affected APRM channel (s)
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      .' associated LPRM group adjustment .
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     . LPRM calibrations were performed at the required frequenc ,
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c . 10.15, Revision l4,f"TIP Reproducibility and Core Power Symmetry Test . .l

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   . The reviews ~were conducted.to verify that calibration of the'LPRM system
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 . . APRM System Calibration-     ,

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T* ' The SRI reviewed licensee records and. procedures ~ applicable.to calibration _.of the APRM system. This review included the following: " u w uJ

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 . Verification that APRM amplifier gains were adjusted to indicate the percent of rated power as determined by performance of CNS Procedure 1 . Verification that data was correctly transferred from the following computer printouts to Attachment "A" of Procedure 10.1: _0D-3, " Core Thermal Power and APRM Calibration"
 ~ P-1, " Periodic, Core Performance Log" ,
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The-following CNS procedure was reviewed: -

 . 10.1,7Revision 15, "APRM Calibration"-

The reviews were conducted to verify that the APRM system was properly calibrated to actual core thermal power using technically adequate and approved, procedure No violations or deviations were identified in this are . Core Thermal Power Evaluation The SRI reviewed licensee records, data sheets, and procedures applicable to reactor core thermal power evaluation and performance which were conducted during the period August 20 through October 5, 1985. The SRI verified the following:

 . Procedure prerequisites were met prior to performing the core thermal
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power evaluation . Figures and curves corresponding to specific reactor conditions were interpreted properly and recorded on data form . Calculations were correc . The' evaluation frequency met CNS Technical Specification requirement ~

 ,The following CNS procedures were reviewed:
"
 . 10.2, Revision 11, "IRM Power Calibration"
 . 10,3, Revision 6, " Core Thermal Power Evaluation"
 . 10.18, Revision 3, " Routine Core Performance Data Gathering"
 . 10.19, Revision 3, " Guide for 00-1 and 0D-2 Utilization"
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    -10-The reviews were conducted to verify that CNS is operated within the licensed core thermal power limit No violations or deviations were identified in this are . Licensee Event Reports Followup The'following Licensee Event Reports (LER) were closed on the basis of the
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SRI's inoffice review, review of licensee documentation, and discussions with licensee personnel:

 . .LER 85-003, "Setpoint Drift of Safety and Safety Relief Valves"
 . LER 85-004, " Radiation Overexposure" 1 Plant Shutdown /Startup The SRI held discussions with operations shift personnel and reviewed control room records including log entries, computer printouts, and recorder traces associated with a plant shutdown conducted on -

October 5, 1985. The shutdown was completed at 5:30 p.m. by the initiation of a manual reactor scram from 44% power (reference Scram Report 85-01).

This shutdown was unscheduled and was initiated as a result of increasin reactor vessel water conductivity accompanied by high vibration ca No. 2

 .
 - low pressure turbine. The plant was placed in cold shutdown where it
',  remained for 7 weeks to investigate and repair the cause of_ the problem ' ~
 'The CNS turbine consists of three, independent double-ended' turbines attached in series by a common shaft to one electrical generator. Th .

rotating blading'in each turbine is attached in concentric. rows'to the

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turbine shaft. The rows of rotating blading are numbered L-0 at each end 4

 , of each turbine and successive rows are sequentially numbered L-1, L-2, and etc., from that point to the center (i.e., steam input point) of the q respactive turbine <

The licensee disassembled both low pressure (LP) turbines during the outage and identified the following damage: P

• 7 . No.1 LP turbine, governor end rows of blading L-0 and L-1, had impact .

  -damag Also, the adjacent stationary blading sustained similiar e   ~ damag The damage appeared to be caused by a foreign object (s)

which were not positively identified during this inspection perio . No.2 LP turbine, . generator end rows of blading L-0 and L-1, had impact damage to both rotating and stationary blades. Also, a 5-inch piece of blading was missing from row L-1. Other blades were found to be cracked. The probable cause of damage is discussed belo ,

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The high pressure (HP) turbine outer rows of rotating blades were

- Linspected but no damage was eviden c During the inspection of the No.-2 LP turbine stationary blading, the-
 ' licensee discovered what appeared to be s -a 1-foot by 2-foot piece of sheet metal lagging laying _ across and covering an~ equivalent area of stationary blading. That blading was located adjacent to snd upstream of the L-1 row of rotating blading. The licensee has initially determined that impact

" damage to the L-0 and.L-1 rows' of rotating blading prob' ably resulted from the following sequence-of . events:

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 . The piece of sheet metal lagging. blocked steam flow in that affected area, thus causing a differential pressure.on the backside of the stationary blade . The L-1 row of blading was subjected to that change of" stress when it rotated past the blocked are . The change in stress eventually caused a cyclic fatigue failure resulting.in the 5-inch piece of blading breaking loose and traveling through both the L-1 and L-0 rotating blades prior to being exhausted'

to the main condenser. The last rows of stationary blading.was also apparently impacted by the same piece of bladin . The' piece of blading broke up into numerous small pieces during its path through the last rows of turbine stationary and rotating blading and punctured two condenser tubes. Circulating water leaked from the perforated tubes into the main condenser tubes eventually causing increasing conductivity in the reactor vessel wate Both LP turbines were disassembled and the following repairs were performed:

 . Weld fill of the numerous repairable impacted areas of the stationary and rotating blading of the No. 1 turbine and the stationary blading of the No. 2 turbin . In order to reduce repair time, two groups of L-1 blading (five blades per group) and ten individual blades in the L-0 blading were replaced in the No. 1 LP turbine rotating assembl . The entire No. 2 LP turbine rotating assembly was replaced with a spare assembly. The damaged assembly will be repaired at a future dat . The two punctured' condenser tubes were plugged and as an added precaution, ti ee additional tubes were plugged that exhibited impact damage but had not been punctured.

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.The SRI observed a reactor startup and heatup on November 22, 1985. The startup was performed following completion of the extended outage described above. He verified that plant management authorized the startup.. Also, he observed that. shift crew manning requirements were met, Technical Specification required prestartup checks were completed satisfactorily, and that operations personnel adhered to approved operating procedures. The reactor achieved criticality at 3:21 a.m. on November 22, 198 The SRI observed perfonnance of the following plant procedures during the reactor startup and heatup:
. 2.1.1, " Cold Startup Procedure"

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. 2.1.3, " Approach to Critical"
. 2.1.15, " Reactor Recirculation Pump Operation" The SRI reviewed the following procedures that were performed and completed as a result of the October 1985 shutdown and November 1985 startup:
. 2.1.1.2, " Technical Specifications PreStartup Checks"
. 2.1.4, " Normal Shutdown From Power"
. 2.1.10, " Station Power Changes"
-The discussions, reviews,~nd a observatio'ns were conducted to verify that the plant responded as designed during the shutdown, operations personnel performed appropriate followup actions, and that..no unreviewed safety questions existed. Also,'he verified that shutdown and startup operations were performed in accordance with procedures and the. requirements established in the CNS Operating License and Tachnical Specificatio No violations or deviations were identified in this are . Operational Safety Verification The SRI observed control room operations, instrumentation, controls, reviewed plant logs and records,' conducted ~ discussions with control room
. personnel, and performed system walk-downs to verify that:
.. Minimum shift mannir.g. requirements were met.

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. Technical Specification requirements were observe ~ '
. Plant ope' rations were conducted'using approved procedure . . - . -
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 . Plant logs and records were complete, accurate, and indicative of actual system conditions and configuration . System pumps, valves, control switches, and power supply breakers
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were properly aligne . Licensee systems lineup procedures / checklists, plant ~ drawings, and as-built configurations were in' agreemen . Instrumentation was accurately displaying process variables and protection system status to be within permissible operational limits for operatio . Plant equipment that was discovered to be inoperable or was removed from service for raintenance was properly identified, redundant

 / equipment was verified to be operable, and applicable limiting conditions for_ operation were identified and maintaine "
 .- Equipment safety clearance records were complete and indicated that
. affected components were removed.from and returned to service in a correct and approved manne I
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 .. Maintenance work _ requests-were initiated for equipment discovered to
 . require repair or routine preventive upkeep, appropriate priority was
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 -assigned, and work commenced in a timely manne ._ Plan't equipment conditions such as cleanliness, leakage, lubrication, and cooling water were controlled and adequately maintaine *
 , . Areas of the plant were clean, unobstructed, and free of fire

' hazards.. Fire suppression systems and emergency equipment were

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maintained'in~a condition of readines . Security measures and radiological controls were adequat The SRI performed lineup verifications of the following systems:

 . Residual Heat Removal
 . Emergency Power Distribution
 . Automatic Depressurization
 . 125V DC Distribution The tours, reviews, and observations were conducted to verify that facility operations were performed in accordance with the requirements
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established in the CNS Operating License and Technical Specificatio ,

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    -14-No violations or deviations were identified in this are . Monthly Surveillance Observations The SRI observed Technical Specification-required surveillance. test These observations verified that:
 . Tests were accomplished by qualified personnel in accordance with approved procedure . Procedures conformad to Technical Specification requirement 't
 . Test prerequisites were completed including conformance with applicable limiting conditions for operation, required administrative approval, and availability of calibrated test equipmen ,
 . Test data were reviewed for completeness, accuracy, and conformance
~_ ~ with established criteria and Technical Specification requirement J'
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 . ' Deficiencies were corrected in a timely manne . The system was returned to servic .
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The following surveillance tests were selected and observed:

 . 2.1.18, " Control Rod Drive Friction Test"
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 .4 6.2.4.1, " Daily Surveillance (Technical Specifications)"
. . 6'.3.12.1, " Diesel Generator Operability Test" L
 . 6.3.16.1, " REC Pump Operability"

.

 . 6.4.1.3, "CRD Coupling Integrity Check"

'

 .The reviews and observations were conducted to verify that facility surveillance operations were performed in accordance with the requirements established in the CNS Operating License and Technical Specificatio . Monthly Maintenance Observation The SRI observed preventive and corrective maintenance activities on portions of'the following systems / components:
 . Service Water Booster Pumps
,
 . HPCI -Auxiliary 011 Pump

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L. .:. ,

   ,
   -15- -  *
   . .

f The' observations were . conducted to verify thati -

 . Limiting conditions for operation were, met.~ 4
 . Redundant equipment was operable.

>:

 . Equipment was adequately isolated and safety tagge . Appropriate administrative approvals were obtained prior to commencement of work activitie . Work was performed by qualified personnel in accordance with approved-procedure . , Radiological controls, cleanliness practices, and appropriate fire prevention precautions were-implemented and maintaine . ~ Quality control checks and postmaintenance surveillance testing were performed as required.'
 . Equipment was properly returned to servic The reviews and observations were. conducted to verify that facility maintenance operations were performed in accordance with the requirements
,

established in the:CNS Operating License and Technical Specificatio , No violations,or deviations were identified in this are . Exit Meetings Exit meetings were conducted at the conclusion of each portion of the inspection. ' The NRC inspector summarized the scope and findings of each inspection segment at those meeting * p Y A L w p- . - }}