ML20203F681

Insp Rept 50-298/86-21 on 860601-30.Violation Noted:Failure to Have Procedure for Controlling Activities Affecting Quality

ML20203F681
Person / Time
Site:	Cooper
Issue date:	07/23/1986
From:	Dubois D, Jaudon J, Plettner E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20203F659	List: ML20203F654 ML20203F673 ML20203F681
References
50-298-86-21, NUDOCS 8607310161
Download: ML20203F681 (14)

See also: IR 05000298/1986021

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APPENDIX B

U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-298/86-21 License: DPR-46

Docket: 50-298

Licensee: Nebraska Public Power District (NPPD)

P. O. Box 499

Columbus, NE 68601

Facility Name: Cooper Nuclear Station (CNS)

Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska

Inspection Conducted: June 1-30, 1986

Inspectors: . 4. didh , 7 // !((o

E. A. Plettner, Resident Inspector, (RI). Date /

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D. L. DuBois, Senior Resident Inspector, (SRI) Date

Approved: e tuf N

CfiTef, Project Section A,.

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P. faudof,

It oject Branch

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Inspection Summary

Inspection conducted June 1-30, 1986 (Report 50-298/86-21)

Areas Inspected: Routine, unannounced inspection of previously identified

inspection findings, Licensee Event Reports, spent fuel shipments, operational

! safety verification, and monthly surveillance and maintenance activities.

Results: Withii. the areas inspected, one violation was identified (failure to

have a procedure for controlling activities affecting quality paragraph 3).

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DETAILS

1. Persons Contacted

• G. R. Horn, Division Manager of Nuclear Operations

• E. M. Mace, Plant Engineering Supervisor

• J. M. Meacham, Technical Manager

• J. Sayer, Acting Technical Staff Manager

• C. R. Goings, Regulatory Compliance Specialist

• V. L. Wolstenholm, Quality Assurance Manager

NRC Personnel

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• W. M. McNeill, Project. Inspector

• D.' L. DuBois, Senior Resident Inspector

,*E. A. Plettner, Resident Inspector

• Denotes those present during exit interview June 27, 1986.

2. Licensee Action on Previous Inspection Findings

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(Closed) Open Item (298/8110-01): " Lack of Concise and Grouped Technical

Specifications Definitiont '

This' item concerns Section 1.0 of Technical Specifications and the need to

provide more concise definitions of all plant operational conditions.

NPPD is a member of the BWR Owner's Group Technical Specification

Committee. The comaittee, which has been endorsed by the NRC, is

presently setting up a structure and long-term program to pursue various

improvement projects including:

"eview and update of the Nuclear Safety Operation Analysis (N50A) for BWRs

and check for Technical Specification completeness.

Revise and Upgrade the current Technical Specification bases.

Short term fixes allowable undar current regulations.

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NPPD commitment to this committee's proposals offers the best approach to

having more concise operational condition definitions.

This item is closed.

(Closed) Open Item (298/8504-01): "High Range Nobel Gas Effluent

Monitors"

This item involved the lack of calibration data from the vendor to verify

proper calibration of the High Range Noble Gas Effluent Monitors. The

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RI verified that vendor calibration data for detector certification of the

liigh Range Noble Gas Effluent Monitors was received and filed in the

! chemistry laboratory.

This item is closed.

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(Closed) Violation (298/8516-02): " Incomplete Test Records"

NRC Inspection Report 50-298/85-15, paragraph 3, documented the results of

an NRC inspection in the area of nondestructive examination test records

associated with the BWR pipe replacement at CNS. During that inspection,

nine generic areas were listed that exhibited various documentation

discrepancies. Eight of the discrepant areas were corrected prior to the

completion of that inspection. The ninth area dealt with numerous

failures of the licensee's inspection department to sign ofi a majority of

the listed visual and liquid penetrant inspection reports. The SRI

reviewed all of the listed work travelers and corresponding weld numbers

during this inspection period and found that the licensee's inspection

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department had subsequently signed off the associated inspection reports.

This item is closed.

(Closed) Violation (298/8519-01): " Station Procedures to Prevent or

Control the Use of Voided, Deleted or Superseded Safety-Related Drawings

and As-Built Drawings"

This item involved the licensee's failure to have procedures that provided

document control of drawings identified by as-built (status 1) from

archival (status 2) or construction (status 3). The RI reviewed

Procecure 3.8, " Drawing Control Procedure," Revision 1, dated November 14,

1985. Procedure 3.8 was changed and implemented to correct the

discrepancy.

This item is closod.

(Closed) Violation (298/8519-02): " Approved lock for Safeguards Drawings

Cabinet"

This item involved the licensee's use of a keylock instead of an approved

lock (such as a GSA con:bination padlock) on a storage cabine.t containing

safeguards information. ine licensee replaced the keylock with an

approved GSA combination padlock. NRC Inspection Report 50-298/85-19

states that replacement was accomplished on the same day the violation was

noted. The RI verified that the lock was installed on the drawings

Cabinet and that it meets GSA requirements.

This item is closed.

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(Closed) Violation (298/8524-02): " Unattended and Unlocked Security

i Records Storage Container"

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This item involved the licensee's security storage containers which were

left unlocked and unattended. The RI verified that. changes were made to

Security Procedam 1.5, " Security Administrative Procedure", Revision 0,

dated April 1, 1986. The RI verified that a document control sign-out

sheet titled " Safeguards Information Checkout Log," was developed and

implemented to ensure that the storage container would be locked at all ~

times except during the removal and replacing of safeguards documents

. normally stored in the cabinet.

This item is closed.

(Closed) Violation (298/8526-01): " Waste Gas Radioactivity Monitor

Calibrations"

This item involved the failure to calibrate gaseous radiation monitors on

a quarterly frequency. The RI reviewed data sheet 8.5.12.1.5, "SJAE Off

Gas Calibration,'? to verify that calibrations have been performed on a

quarterly frequency for all waste gas radioactivity monitors.

This item is closed.

3. Licensee Event Reports Followup

The following Licensee Event Reports (LERs) were closed on the basis of '

the SRI's inoffice reviews, reviews of licensee documentation, and

discussions with licensee personnel:

LER 85-014, "Viele .on of the APRM Flux Trip Settings and Rod Block

Monitor Trip Settings"

LER 85-020, " Invalidated ADS Surveillance Testing"

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LER 85-021, " Abnormal Handling Operation of an Irradiated Fuel Assembly"

LER 86-001, " Inadvertent RWCU Temperature Switch Setpoint Change"

LER 86-002, "High Pressure Coolant Injection Overspeed Trip Control Valve

Diaphragm Failure"

LER 86-002, Revision 1, "High Pressure Coolant Injection Overspeed Trip

Control Valve Diaphragm Failure"

i LER 86-003, "APRM Surveillence Testing"

LER 86-004, "RWCU Group III Isolation"

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LER 86-005, "High Pressure Coolant Injection Turbine Gland Seal Steam

i Condenser Failure"

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LER 86-006, " Reactor Trip"

LER 86-007, " Main Steam Line High Flow Setpoint Anomaly"

LER 86-008, " REC and DG Inoperative"

LER 86-009, " Emergency Diesel Generator 1 and 2 Switchgear Mounting

Anomaly"

In LER 85-014 listed above, the licensee described a violation of the

Average Power Range Monitor ~(APRM) flux trip and Rod Block Monitor

(RBM) trip set'.ings that occurred on September 20 and 21,1985. The

licensee at#> outed the violation to the failure to revise nuclear

performane . elated procedures 6.2.4.1, " Daily Surveillance (Technical

Specificotions)" and 10.1, "APRM Calibration." Procedure 6.2.4.1 is used

in part to determine if indicated reactor power is equal to actual power

and if not, requires that Procedure 10.1 be performed.

On September 20,1985, at 11:00 p.m. , the NRC authorized the licensee to

operate in single-loop for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> providing that correction

factors were added to the APRM flux trip and RBM trip setting

calculations. On September 24, 1985, the NRC issued Amendment 94 to the

CNS Technical Spccification. The licensee operated in single-loop from

9:17 a.m. on September 20, 1985, to 10:00 a.m. on September 21, 1985.

During that period, the licensee recognized that APRM gain adjustment

factors (GAFs) were not calculated accurately because Procedures 6.2.4.1,

and 10.1 were rot revised to meet the newly amended technical

specification requirements for single loop operation.

During the review of LER 85-014, the SRI verified that the licensee had

not exceeded the required APRM flux trip and RBM trip settings. However,

the SRI determined that the licensee did not have an approved procedure

that would have provided controls for ensuring that Procedures 6.2.4.1.

and 10.1 were revised to reflect the Technical Specification amendment

noted above. 10 CFR Part 50, Appendix B, Criterion V, requires that

activities affecting quality shall be prescribed by documented procedures.

The licensee's failure to have an approved procedure for ensuring that

requirements contained in amendments to license documents such as the

Technical Specification are incorporated as revisions into station

procedures, is an apparent violation (298/8621-01).

4. Spent Fuel Shipment

The NRC inspectors inspected the licensee's activities associated with one

shipment of spent fuel from CNS. Included in that inspection were

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observations and reviews of applicable procedures, documentation, surveys,

I inspections and shipping document preparation.

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The SRI verifiad by review of licensee documentation, through discussions

with responsible personnel, and by independent inspection that the

licensee completed the following:

o Receiving inspection of railcars and shipping casks. Shipping

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documents.

o Advance notification of and approval by affected state and federal

agencies.

j o Proper placarding of the transport vehicles.

o Appropriate labeling of the spent fuel shipping casks.

o Establishment of provisions for response by escorts and local law

enforcement agencies.

e Training of escort personnel.

o Testing of communications systems.

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c Continual manning of the licensee's communications center (Movement

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Control).

o Testing of fuel and cask handling cranes, hoists, and tools.

o Proper loading and sealing of the spent fuel shipping casks.

o Surveillance of area radiation monitors, ventilation systems, and

spent fuel pool water level and chemistry.

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o Update of fuel location and accountability records.

o Applicable quality assurance audits and inspections.

o U.S. Department of Energy and U. S. NRC " Nuclear Material Transaction

Report," DOE /NRC Form 741.

o Bill of Lading.

o CNS Health Physics Procedure 9.5.3.7, " Cask IF-300 Shipment,"

Revision 3, dated December 26, 1985.

o CNS Nuclear Performance Procedure 10.27, " Cask IF-300 Handling and

Shipping," Revision 5, dated May 8, 1986.

o CNS HP-138, " Contamination Survey - Sample Count Data Sheets."

o CNS HP-141, " Contamination Survey - Railroad Car for IF-300

Irradiated Fuel Shipping Cask."

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1, o CNS HP-142l " Contamination Survey of IF-300 Shipping Casks."

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.o- CNS HP-143, " Radiation Survey of IF-300 Shipping Cask."

, o. .CNS HP-608, " Spent Fuel Shipment Checkoff Sheet and Certificate of

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Compliance of Number 9001 Conditions for Shipping Spent Fuel."

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o CN'S'HP-14a, " Radioactive Material Shipment Record."

i The following independent radiation and contamination surveys were

performed by the RI and verified to be satisfactory: ,

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o Contact radiation surveys of the shipping casks

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o Radiation surveys at a distance of two meters from the cask transport

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o Contamination surveys of the shipping casks surfaces

l 0 Contamination surveys of the ca'sk transport vehicles

The SRI reviewed CNS Procedure 10.27, Revision 5, dated May 8, 1986. . The

licenree incorporated into Procedure 10.27 specific handling instructions

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for the G.E. Type IF-300 SPENT FUEL SHIPPING CASK. Also' included within

Procedure 10.27 was Attachment "A," " Handling and Loading of Cask IF-300

Spent Fuel Shipping Cask Checkoff Sheet." The checkoff sheet provided two

i functions: it identified important steps used in the receipt, inspection,

l preparation, movement, loading with fuel, leak testing, decontamination,

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loading of the cask onto the transport vehicle, and final preparation for

i shipping; and it provided a checkoff list including spaces for signatures-

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

I steps of the procedure. The SRI verified that Attachment "A" of

. Procedure 10.27 was properly completed, signed, and dated.

l The spent fuel shipment left the CNS on June 10, 1986. The shipment

consisted of 2 spent fuel shipping casks, each of which contained 18 spent

i fuel bundles. The shipment was transported to the G.E. Morris Operation

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Complex, Morris, Illinois. The spent fuel casks identification numbers

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o IF-301

o IF-302

The observations, reviews,and independent measurements were conducted to

verify that spent fuel handling and shipment operations were in

j conformance with the requirements established in the CNS Operating License

! and Technical Specification.

No violations or deviations were identified in this area.

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5. Operational Safety Verification

The NRC inspectors 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:

o Minimum shift manning requirements were met.

o Technical Specification requirements were observed.

o , Plant operations were conducted using approved procedures.

o Plant logs and records were complete, accurate, and indicative of

actual system conditions and configurations,

o System pumps, valves, control switches, and power supply breakers

were properly aligned.

o Licensee systems lineup procedures / checklists, plant drawings, and

as-built configurations were in agreement.

o Instrumentation was accurately displaying process variables and

protection system status to be within permissible operational limits

for operation.

o When plant equipment was found to be inoperable or when equipment was

removed from service for maintenance, it was properly identified and

redundant equipment was verified to be operable. Also, the NRC

inspectors verified that applicable limiting conditions for operation

were identified and maintained.

o Equipment safety clearance records were complete and indicated that

affected components were removed from and returned to service in a

correct and approved manner.

o Maintenance work requests were initiated for equipment discovered to

require repair or routine preventive upkeep, appropriate priority was

assigned, and work commenced in a timely manner.

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o Plant equipment conditions such as cleanliness, leakage, lubrication,

and cooling water were controlled and adequately maintained.

o Areas of the plant were clean, unobstructed, and free of fire

hazards. Fire suppression systems and emergency equipment were

maintained in a condition of readiness. It was noted that several

plastic seals were missing from fire plug indicators. Corrective

action was taken by the licensee in a timely manner to replace those

seals.

o Security measures and radiological controls were adequate.

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The NRC inspectors performed a lineup verification of the following

systems:

o Reactor Core Isolation Cooling (RCIC)

o Residual Heat Removal (RHR)

In preparation for performing the system walk-down of the RCIC system, the

RI conducted a review of and comparison between the licensee's RCIC valve

checklist and applicable as-built drawings. The following documents were

reviewed: ,

o System Operating Procedure (SOP) 2.2.67 " Reactor Core Isolation

Cooling" Revision 25, dated April 24, 1986, Appendix A " Valve

Checklist."

o As-built drawing--B&R 2043; for Reactor Core Isolation Cooling.

o As-built drawing--B&R 2049; for Reactor Core Isolation Cooling.

o As-built drawing NPPD 1.0.-19; for Reactor Core Isolation Cooling.

o As-built drawing GE 115D6015; for Reactor Core Isolation Cooling.

The above review identified the following discrepanc'es:

o S0P 2.2.67, Appendix A, listed 72 instrument related valves that were

not numbered ar labeled on applicable as-built drawings I.D.-19 or

11506015. This deficiency is similar to the violation that was

documented in NRC Report 50-298/86-14, as per EP paragraph (5).

Since the licensee will respond to this violation, it is not cited

herein. It will, however, be tracked as an cpen item pending review

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of licensee corrective action.

o S0P 2.2.67 Appendix A " Valve Checklist" Revision 17 had an error in

valve numbers. The checklist listed two valves as 1-RCIC--164 with a

l different description for each. The correct valve number for the

second description is 1-RCIC-165, not 1-RCIC-164 as is currently

listed. The licensee was notified of the error and has initiated a

procedure revision to correct the typographical error.

The SRI observed the licensee's performance of the following operating

I procedures during the period June 16 through June-19, 1986:

2.1.2, " Hot Startup Procedure," Revision 25, dated April 24, 1986.

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I o 2.1.4, " Normal Shutdown From Power," Revision 23, dated April 29,

l 1986.

o 2.2.28, "Feedwater System," Revision 38, dated April 17, 1986.

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o 2.2.60, " Primary Containment Cooling and Nitrogen Inerting System,"

Revision 30, dated January 30, 1986.

On June 16,1986, the 8:00 a.m. calculated 24-hour reactor coolant drywell

unidentified leakage rate was 0.812 gpm. Between the hours of 4:00 p.m.

and 8:00 p.m., drywell leak detection instrumentation indicated that

unidentified leakage had increased from 1.78 gpm to 3.69 gpm.

CNS Surveillance Procedure 6.2.4.1, " Daily Surveillance (Technical

Specifications), " Revision 52, dated February 6,1986, required that if

baseline unidentified leakage rate is between 1 and 2 gpm and the

unidentified leakage rate doubles in a 4-hour period, the reactor must be

shutdown. The licensee commenced shutdown at 8:55 p.m. and drywell

deinerting at 9:22 p.m.

On June 17, 1986, the 8:00 a.m. calculated leakrate indicated an increase

of 2.358 gpm within the previous 24-hour period. The CNS Technical

Specification (TS), Section 3.6.C.4, requires the plant to be in a Cold

Shutdown Condition within 24-hours if leakage increases by 2 gpm within

the previous 24-hour period and that the source of leakage can not be

identified. The licensee reduced power from 30 percent to 18 percent in

preparation for a drywell entry and plant shutdown. An initial drywell

entry was accomplished at 12:05 p.m. and the leakage source identified at

12:29 p.m. The source of unidentified leakage was observed to be an upper

packing gland leak from the reactor recirculatior, pump "B" discharge

valve RR-538. The packing gland was subsequently tightened and

unidentified leakage was reduced to 0.5 gpm. The plant resumed normal

power operation on June 18, 1986. The drywell oxygen level was reduced to

less than 4 percent by 9:00 a.m. on June 18, 1986.

The SRI observed the following licensee actions associated with the

drywell unidentified leakage rate:

o Management meetings including the Station Operations Review

Committee (SORC)

o Continual monitoring of the leak rate

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l o Evaluation of leak detection system information; planning for, and

l the initiation of, drywell entry teams; implementation of repair

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activities; and the performance of surveillance testing required by

l the power reduction and repair activities

o Drywell deinerting and reinerting operations

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l 0 Notification of the NRC

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o Implementation of procedural and Technical Specification requirements

Several differences were noted between the BWR Standardized Technical

Specification (STS) and the CNS Technical Specif; cation during the leakage

event. Those differences and additional discrepancies included:

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o The STS has separate and-distinct Limiting Conditions for

Operation (LCOs) for the' reactor coolant leak detection' system and

reactor coolunt system leakage rate during reactor power operation.

The CNS Technical Specification, Section 3.6.C.4, requires-plant-

shutdown if unidentified leakage exceeds 5 gpm and a subsystem of the

leak detection system remains inoperable for greater than 30 days.

o The STS has other separate and distinct reactor coolant system

leakage requirements such as: no pressure boundary leakage; less

than 5 gpm unidentified leakage; 25 gpm total leakage averaged over

any 24-hour period; 1 gpm leakage from specified reactor coolant

system pressure isolation valves; and 2 gpm increase in unidentified

leakage within any 4-hour period. The CNS Technical Specification

does not specifically identify no pressure boundary leakage; does

discuss a 5 gpm unidentified leakage; applies the 25 gpm criteria to

identified leakage only and not the total of identified plus

unidentified leakage; does not state criteria for leakage from

specified RCS pressure isolation valves; and establishes the 2 gpm

increase of unidentified leakage over a 24-hour period.

o The CNS Technical Specification requires that primary containment

atmosphere oxygen concentration shall be less than 4 percent by

volume within the 24-hour period subsequent to placing the reactor in

the RUN mode following a shutdown. Deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

prior to shutdown. The Technical Specification does not address when

the 24-hour inerting clock is started if a required shutdown is

aborted as a result of the initiating problem being corrected before

shutdown is achieved.

The Technical Specification differences noted above will remain an open

item pending discussions between NRC Region IV, the NRC Office of Nuclear

Reactor Regulation (NRR), and the licensee's licensing department

(298/8621-02).

A preliminary notification of an unusual event occurred on June 2 through

June 3, 1986, at LaSalle Unit 2 in Marseilles, Illinois. The cause of the

event was the failure of a reactor scram to occur at the reactor vessel

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low water level setpoint. Further investigation revealed that level

switches manufactured by Static "0" Ring were the cause of the problem.

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The RI requested CNS to determine if Static "0" Ring level switches are

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used on site and if so, in what systems are they installed. CNS responded

to the request in a timely manner. The license identified that six level

switches manufactured by Static "0" Ring are used on site. Those switches

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are used in the Radioactive Waste Treatment System. No further action was

required.

The tours, reviews, and observations were conducted to verify that

facility operations were performed in accordance with the requirements

established in the CNS Operating License and Technical Specification.

No violations or de<iations were identified in this area.

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6. Monthly Surveillance Observations

The NRC inspectors observed Technical Specification required surveillance

tests. Those observations verified that:

o Tests were accomplished by qualified personnel in accordance with

approved procedures.

o Procedures conformed to Technical Specification requirements.

o Tests prerequisites were completed including conformance with

applicable limiting conditions for operation, required administrative

. approval, and availability of calibrated test equipment.

o Test data was reviewed for completeness, accuracy, and conformance

with established criteria and Technical Specification requirements.

o Deficiencies were corrected in a timely manner.

o The system was returned to service.

The RI observed the licensee's performance of the following surveillance

tests on the indicated dates:

June 4, 1986: 6.3.4.1 " Core Spray Test Mode Surveillance Operation,"

Revision 18, dated January 30, 1986.

June 4. 1986 6.3.4.2 " Core Spray Motor Operated Valve Operability

Test," Revision 12, dated April 17, 1986.

June 4, 1986 6.2.1.4.1 "PCIS Main Steam Line High Temperature

Functional Test," Revision 11, dated July 5,1985.

5.4.1 " General Fire Procedure" Revision 20, dated

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June 10, 1986

April 10, 1986.

The SRI reviewed the following completed surveillance tests that were

performed June 18 through June 19, 1986:

o 6,3.1.1, " Primary Containment Local Leakage Tests," Revision 18,

dated March 13, 1986.

o 6.3.10.7, " Primary Containment Isolation Valve Closure Timing,"

Revision 12, dated August 8, 1985.

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 Specification.

No violations or deviations were identified in this area.

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7. Monthly Maintenance Obscrvation

The NRC inspectors observed preventive and corrective maintenance

activities. These observations verified that:

o Limiting conditions for operation were met.

o Redundant equipment was operable.

o Equipment was adequately isolated and safety tagged.

o Appropriate administrative approvals were obtained prior to

commencement of work activities,

o. Work was performed by qualified personnel in accordance with approved

procedures.

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o Radiological controls, cleanliness practices, and appropriate fire

prevention precautions were implemented and maintained.

o Quality control checks and postmaintenance surveillance testing were

performed as required.

o Equipment was properly returned to service.

Those 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 Specification.

i No violations or deviations were identified in this area.

8. Exit Meetings

l Exit meetings were conducted at the conclusion of each portion of the

j inspection. The NRC inspectors summarized the scope and findings of each

l inspection segment at those meetings.

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