ML20210D501

Insp Rept 50-298/86-36 on 861201-31.Violations Noted:Failure to Have Documented Procedure Evaluation & to Follow Procedure

ML20210D501
Person / Time
Site:	Cooper
Issue date:	01/23/1987
From:	Dubois D, Jaudon J, Plettner E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20210D444	List: IR 05000298/1986036 ML20210D439 ML20210D465 ML20210D501
References
50-298-86-36, NUDOCS 8702100103
Download: ML20210D501 (17)

See also: IR 05000298/1986036

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APPENDIX B

U. S. NUCLEAR-REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-298/86-36 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: December 1-31, 1986

Inspectors: N- M //sa.//7

E. A. Plettner,- Resident inspector, (RI) Ddte'

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D. L. DuBois, Senior Resident Inspcctor, (SRI)

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J/P/ Jaud i, Chief, P 'oject Section A,

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Inspection Summary

Inspection Conducted December 1-31, 1986 (Report 50-298/86-36)

Areas Ins)ected: Routine, unannounced inspection of complex surveillance,

reactor slutdown margin,: core thermal power evaluation, core power

distribution limits surveillance, calibration of nuclear instruments,

refueling, cold weather preparation, operational safety verification, and

monthly surveillance and maintenance activities.

Results: Within the areas inspected, two violations were identified (failure

to have documented procedure evaluation, paragraph 6; and failure to follow

procedure, paragraph 7).

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DETAILS

1. Persons Contacted

Principal Licensee Employees

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

• W.' E. Crawford, Supervisor, Maintenance

• J. Sayer, Manager, Radiation Protection

• C, R. Goings, Regulatory Compliance Specialist

• D. C. Shrader, Assistant Supervisor, Operations

• H. A. Jantzen, Supervisor, Instrumentation and Control

• C. R. Moeller, Supervisor, Technical Staff

• G. Smith, Acting Manager. Quality Assurance

' The NRC inspectors also interviewed other licensee employees during the

course of the inspection.

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• Denotes those present during exit interview January 12, 1987.

2. Complex Surveillance

The purpose of this inspection was to verify that complex surveillance

tests, applicable to safety-related systems and subsystems, were

performed in accordance with the requirements established in the CNS

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Operating License'and Technical Specifications.

! The RI observed the performance of the following procedures on the

indicated dates:

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. December 30, 1986: Surveillance Procedure (SP) 6.3.4.3 "CS, RHR,

and Diesel Auto Start and Loading," Revision 21 dated August 22,

1985;

. December 31, 1986 and January 1,1987: General Operating Procedure ,

. (GOP) 2.1.14 " Reactor Vessel In-Service Leak Test," Revision 16,

dated April 3, 1986.

Reviews and observations verified that:

. Testing was performed using approved procedures, which were

consistent with regulatory requirements, industry standards, and the

Technical Specification.

. Pennanent or temporary procedure revisions were accomplished

according to administrative requirements and controls.

. Qualified personnel conducted the tests and performed the final

4 reviews and opprovals of completed test data.

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. The official test copy was available and used by test personnel.

. Procedures contained the purpose, objectives, references,

prerequisites, test equipment, precautions, limitations, and

acceptance criteria.

. Test equipment required by the procedures was calibrated and in

service.

. Procedures provided sufficient direction to accomplish necessary

evolutions.

. Systems were returned to normal lineup following completion of

testing.

The requirements for the diesel generator sequential loading are stated

in the Updated Safety Analysis Report (USAR) in Section VIII Part 5.0.

" Standby-A-C Power Source." The NRC inspectors performed a comparative

review of Procedure SP 6.3.4.3 with the requirements in the USAR.

From the review it was determined that the procedure may not meet all the

requirements as stated in the USAR in Section VIII, Part 5.0. The NRC

inspectors are waiting on information from the licensee to the following

questions:

. Are Motor Control Centers (MCC) K and S part of the load shed and

load sequencing on the IF for K or 1G for S emergency switch gear

buses?

. Are there time delays associated with the starting of the Standby

Gas Treatment fans which receive their power from MCC K for train A

or MCC S for train B?

. When is the less than or equal to 10 second timing requirement for

the diesel to start and attain rated voltage and frequency

applicable; I.E. monthly or once per cycle?

It was noted that the current SP 6.3.4.3 does not time the diesel from

start until it has attained rated voltage and frequency. Instead, the

procedure times the period from diesel start to output breaker closing.

USAR allows a three-second delay in the typical sequential loading of

diesel generator (DG) from the time the diesel attains rated voltage and

frequency until the output breaker closes. SP 6.3.12.1, performed

monthly, requires that the start time be recorded with a stop watch but

has no acceptance criteria stated in the procedure.

Pending receipt of additional information from the licensee, these

questions will be considered to be an unresolved item (298/8626-01).

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On December 29, 1986, SP 6.3.4.3 was conducted twice. The first test was

conducted by tripping breaker IFA from Board C instead of deenergizing

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undervoltage relays 27/1F2 and 27/1FA2 as specified by SP 6.3.4.3,

Step VIII.A.19. However,-the remainder of the test was conducted as

required, and DG No. I started and loaded properly. The second test was

run as specified in the procedure and demonstrated the proper response to

a simulated undervoltage condition (i.e., diesel generator automatic

start).

This is an apparent violation of failure to adhere to procedure. A

notice of violation will not be issued, since the violation was

I self-identified by the licensee, and it meets the remaining four criteria

stated in Appendix C,Section V.A of 10 CFR Part 2.

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One unresolved item was identified in this area.

) 3. Reactor Shutoown Margin

On December 21, 1986, the RI observed the performance of liuclear

Performance Procedure (NPP) 10.16. " Shutdown Margin Evaluation,"

Revision 14, dated October 16, 1986. Included were checks to ensure that

, test prerequisites were completed, testing was performed according to

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procedures, appropriate precautions and limitations were observed, and

test results were adequately reviewed for accuracy and completeness. The 1

RI also reviewed GE supplied nuclear engineering data, which was used to

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determine the sequence of rod withdrawals, expected reactivity addition

rates, and predicted shutdown margin values. The RI verified that test

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results conformed with Technical Specification requirements. -

No violations or deviations were found in this area.

4. Core Thermal Power Evaluation

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The RI reviewed licensee records, data sheets, and procedures applicable

to reactor core thermal power evaluation and performance which were

conducted during the period June 1 through October 3,1986. The RI

verified the following:

! . Procedure prerequisites were met prior to performing the core

thermal power evaluations.

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. Figures and curves corresponding to specific reactor conditions were

interpreted properly and recorded on data forms.

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. Calculations were correct.

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. The evaluation frequency met CNS Technical Specification

requirenents.

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The following CNS procedures were reviewed:

. Nuclear Performance Procedure (NPP) 10.1, "APRM Calibrations,"

Revision 16, dated October 30, 1986

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. NPP 10.2, "IRM Power Calibration," Revision 11, dated May 11, 1984

. NPP 10.3, " Core Thermal Power Evaluation," Revision 7, dated

October 31, 1986

. NPP 10.5, "LPRM Calibration," Revision 17, dated January 25, 1985

The reviews were conducted to verify that CNS is operated within the

licensed core thermal power limits.

No violations or deviations were identified in this area.

5. Core Power Distribution Limits Surveillance

The RI 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 thennel data indicated an approach to limiting

conditions. The review and discussions included the following:

. Verificationthatthelinearheatgenerationrate(LHGR), core

maximumpeakingfactors(CMPF),minimumcriticalpowerratio(MCPR),

and average planar linear heat generation rate (APLHGR), were within

prescribed Technical Specification limits.

. Examinations of local power range monitor (LPRM) and BASE

distribution calculations as typed out by the 00-1, "LPRM

Calibration and Base Data," on-demand typewriter. Typed alarms,

errors, and other inprocess messages were also reviewed.

. Verification that traversing incore probe (TIP) machine

normalization factors were properly obtained.

. Examination of licensee procedures for ascertaining operation within

licensed limits, should the process computer becorre unavailable.

. Verification that average power range monitor (APRM) channel gains

were adjusted as necessary following an LPRM calibration.

. Verification that following an APRM gain adjustment, a subsequent

P-1 was run to assure that APRM gain adjustment factor (GAF)

reflected such gain adjustments.

. Examination of licensee procedures which are used to correct

abnormal core thermal conditions.

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The following computer printouts edited from June 1 to October 3,1986,

were reviewed:

. 00-1, "LPRM Calibration and BASE Data"

. 0D-3, " Core Thermal Power and APRM Calibration"

. 00-6, " Thermal Data for Specified Fuel Bundles"

. 00-7, " Control Rod Notch Positions"

. 00-8, "LPRM Console Readings"

. 00-10. " Edit Specified Data Array"

. OD-15. " Computer Shutdown and Outage Recovery Monitor"

. 00-16. " Target Exposure and Power Data"

. 00-17 " Edit Periodic Core Perfomance Logs"

. P-1, " Periodic Core Performance"

The following CNS procedures were reviewed:

. NPP 10.4, " Core Thermal Hydraulic Evaluation " Revision 8, dated

April 24, 1986

. NPP 10.7, " Maximum dverage Planar and Peak Linear Heat Generation i

Rates and Minimum Critical Power Ratio," Revision 8, dated '

October 30, 1986

. NPP 10.8, " Reactivity follow Check," Revision 9, dated October 18,

1984

. NPP 10.9, " Control Rod Scram Time Evaluation," Revision 13, dated

November 6,1986

. NPP 10.10. " Limiting Control Rod Pattern Determination,"

Revision 7, dated January 21, 1985

. NPP 10.11. " Control Rod Sequence Exchange," Revision 8, dated

October 30, 1986

. NPP 10.13. " Control Rod Sequence and Movement Control," Revision 13,

dated October 30, 1986

The reviews and discussions were conducted to verify that the plant is

being operated within licensed power distribution limits.

No violations or deviations were identified in this area.

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6. Calibration of Nuclear Instruments

The purpose of the inspection was to determine that Source Range

Monitors, (SRM), Intermiate Range Monitors (IRM), local Power Range

Monitors (LPRM),andAveragePowerRangeMonitors(APRM)hadbeen

properly calibrated in accordance with approved procedures and met the

calibration frequency required by Technical Specification.

The RI reviewed the results of SRM, IRM, LPRM, and APRM calibrations to

verify the following:

. All precautions and prerequisites were met.

. Power supply voltages were all within tolerance.

. The instruments and calibration equipment used were traceable to the

National Bureau of Standards.

. Test equipments used and their serial numbers were recorded on the

procedure.

. Licensee's procedures ensure that the setpoint levels for alarms,

permissive and prchibitive interlocks are in compliance with the

appropriate Technical Specifications.

. Calibration results were reviewed, approved, and documented in

accordance with the licensee's administrative control procedures.

The RI reviewed the following procedures:

. SP 6.1.3, "APRM System Excluding 15% Trip Functional Test," Revision

15, dated September 18, 1966.

. SP 6.1.17, "IRM Calibration and Functional Test, (Mode Switch not In

Run)," Revision 10, dated August 8, 1985.

. SP6.1.17A,"IRMCalibrationandFunctionalTest(ModeSwitchIn

Run)," Revision 3, dated May 22, 1982

. SP 6.1.19. "LPRM Calibration Test," Revision 7, dated March 21, 1986

. SP6.1.21,"SRMCalibrationandFunctionalTest(ReactorNotIn

Run)," Revision 14 dated October 8,1984

. SP6,1,21A,"SRMCalibrationandFunctionaltest(ReactorInRun),"

Revision 6, dated October 10, 1984

. SP 6.1.22. "APRM System 15% High Flux and Inop Trip Functional Test,"

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Revision 11, dated September 11, 1986

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. SP 6.1.29, "APRM System (Flow Bias and Startup) Calibration and/or

Functional Test," Revision 14, dated January 6,1986

The RI performed an audit of calibration data, Attachments "A," of the

following procedures for the time frame of September 1985 to

December 1986:

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. Instrument and Control Procedure (I&C) 7.5.2.1, "SRM Quarterly l

Calibration Procedure," Revision 13, dated October 30, 1986. '

. I&C 7.5.2.2, "IRM Calibration Procedure," Revision ll, dated

October 30, 1986

. I&C 7.5.2.3, "LPRM Calibration Procedure," Revision 11, dated

April 10, 1986

. I&C 7.5.2.4, "APRM Calibration Procedure," Revision 12, dated

June 19, 1986 ,

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The results of the audit revealed that the " reviewed by" signature blank

and associated "date" blank had not been completed on the following:

. 4 times in I&C 7.5.2.1 Procedure, Attachment "A"

. 2 times in I&C-7.5.2.4 Procedure, Attachment "A"

Appendix B Criterion XI to of 10 CFR Part 50 and the Licensee's Quality

Assurance Plan require that test results shall be documented and

evaluated to assure that test requirements have been satisfied. The

failure to document the review of Attachments "A" in the above

procedures is an apparent violation (298/8636-02).

One violation was identified in this area.

[ 7. Refueling

The RI held discussions with fuel handling personnel, observed fuel

movement from the reactor to the spent fuel pool, verified spent fuel

pool fuel assemblies locations, accountability records, and status board

updates. The RI also reviewed the licensee's procedures and records

concerning the movement of fuel and storage of fuel assemblies.

On December 16, 1986, the RI observed the following fuel movement

activities:

STEP FUEL BUNDLE NO. FROM TO

434 LY2089 11-C-11 41-28

435 LY2034 11-B-10 43-26

438 LY2120 11-H-10 41-26

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439 LY2079 11-G-11 43-28

440 LYD808 12-M-11 41-30

441 LYD802 12-I-12 43-32

On December 12, 1986, at 2:20 p.m. refueling activities started.

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On December 16, 1986, the RI performed a review of procedures being used

on the refueling floor. The results of the review revealed that Nuclear

Performance Procedure (NPP) 10.25 " Refueling," was Revision 6, dated

. August 7, 1986. The latest issued revision of Procedure HPP 10.25 was

Revision 7, dated Decenber 10, 1986. Procedure NPP 10.25 in

Section VII.A.2 requires the licensee to verify that the procedures in the

operations refueling floor procedure book are of the latest revision and

update as necessary.

The licensee's failure to verify that the procedure NPP 10.25 in the

operations refueling floor procedure book was of the latest revision as

required by procedure NPP 10.25 is an apparent violation (298/8636-03).

The R1 reviewed Report G-HP0-6-424, dated December 3,1986, which General

Electric (GE) sent to the licensee. The report addressed the impact and

lost parts analysis for the double blade quide problem discussed in NRC

Inspec_ tion Report 50-298/86-26, paragraph 6. The GE report concluded:

. The falling blade guide has caused no structural damage to either

the fuel bundles or core top guide lissembly.

. The lost pieces of the aluminum blade guide handle in the reactor

pressure vessel would not compromise safe reactor operation.

No further action is required.

One violation was identified in this area.

8. Cold Weather Preparation

IE Bulletin 79-24, " Frozen Lines," requested the licensee to verify that

adequate protective measures had been taken to prevent safety-related

process, instrument, and sampling lines from freezing during extremely

cold weather. The RI verified the following:

. Individual plant systems operating procedures identified heating

requirements and equipment including power supplies, temperature

controls and settings, indication circuits, insulation requirements,

heat tracing, and space heaters as required.

. Backup freeze protection was provided during extended plant shutdown

in areas that are normally kept warm by heat losses from operational

systems.

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. ~ Plant procedures used during maintenance or modification of existing

systems provided reasonable assurance that cold weather protective

measures were reestablished following completion of those

activities.

. Plant preventive maintenance requirements associated with cold

weather preparation were completed on September 17, 1986. ,

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The following documents were reviewed:

. Preventive Maintenance (PM) routine 04047 and 01271

. General Operating Procedure (G0P) 2.1.11, " Station Operators Tour,"

Revision 39, dated October 30, 1985.

. Attachment "C" to G0P 2.1.11. " Station Operators Tour - R/W, A0G,

ARW Areas and Outside."

. System Operating Procedure (SOP) 2.2.30, " Fire Protection System,"

Revision 24, dated October 17, 1985.

The discussions, reviews, and walkdowns were performed to verify that the

licensee has maintained an effective program of cold weather protective

measures for safety-related components and systems.

No violations or deviations were identified.in this area.

9. 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 walkdowns to verify that:

. Minimum shift manning requirements were met.

. Technical Specification requirements were observed.

. Plant operations were conducted using approved procedures.

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

actual system conditions and configurations.

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

were properly aligned.

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

as-built configurations were in agreement.

. Instrumentation was accurately dis alaying process variables and

protection system status to be wit 11n permissible operational limits

for' operation.

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. 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 I

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inspectors verified that applicable limiting conditions for

operation were identified and maintained. O'~ -

. Equipment safety clearance records were complete and ind'i ed that

affected components were removed from and returned to. service in a

correct and approved manner. - N , [

. Maintenance work requests were initiated for equipment disco ered to'

require repair or routine preventive upkeep, appropriate priority '

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was assigned, and work coninenced in a timely manner. . Y , ,

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. Plant equipment conditions such as cleanliness, leakage, '

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lubrication, and cooling water were controlled and adequately N

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. 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. l

. Security measures and radiological controls were equate.

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The RI performed a lineup verification of the valves in'normally

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inaccessible areas in the following systems:

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. Reactor Core Isolations Cooling (RCIC) y x[ . ,

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. High Pressure Coolant Injection (HPCI) y - y ,,4

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. Standby Liquid Control (SLC) ,

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. Main Steam and Turbine Bypass (MS) ,

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. Feedwater -

In preparation for performing the system walkdown of the MS and Feedwat'er.

systems, the RI conducted a review of and comparuon between the

following licensee MS and Feedwater systems valve checklist and

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applicable as-built drawings:

System Operating Procedure (SOP) 2.2.56, "Mainsteam and Turbinh, ,

Bypass System," Revision 21, dated August 7,.1980, Appendir'A,

" Valve Checklist" .

. As-Built drawing - Burns & Roe 2002; for MS System ,,

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. As-Built drawing - Burns & Roe 2041; for MS System '

. As-Built drawing - GE 115D6014; for MS System .

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. As-Built drawing - Burns & Roe M-107; for MS System

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. As-Built drawing - Burns & Roe M-109; for MS System

.. As-Built drawing - NPPD I.D.-1; for MS System

. As-Built drawing - NPPD I.D.-3; for MS System

. As-Built drawing - NPPD I.D.-18; for MS System

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50P 2.2.28 "Feedwater System," Revision 39, dated July 17, 1986;

Appendix A " Valve Checklist"

. As-Built drawing - Burns & Roe 2002; for Feedwater

. As-Built drawing - Burns & Roe 2004; for Feedwater

. 'As-Built drawing - Burns & Roe 2005; for Feedwater

. As-Built drawing - Burns & Roe 2043; for Feedwater

. As-Built drawing - Burns & Roe 2044; for Feedwater

. As-Built drawing - Burns & Roe M-110; for Feedwater

. As-Built drawing - Burns & Roe M-111; for Feedwater

. As-Built drawing - NPPD I.D.-4; for Feedwater

. As-Built drawing - NPPD I.D.-6; for Feedwater

The review identified that S0P 2.2.56 Appendix A, listed 194 instrument

related valves that were not numbered or labeled on applicable as-built

drawings 11506014, M-107, M-109, I.D.-1, I.D.-3, or I.D.-18.

The review identified that S0P 2.2.38 Appendix A, listed 149 instrument

related valves that were not numbered or labeled on applicable as-built

drawings M-110, M-111 I.D.-4, or I.D.-6.

These deficiencies are similar to the violation that was documented in

NRC Inspection Report 50-298/86-14, paragraph 5 and similar to Open

Item 298/8626-01 identified in NRC Inspection Report 50-298/86-26,

paragraph 5. These two items will be tracked as an open item pending

review of licensee's corrective action (298/8636-04).

The RI performed an equipment safety clearance record verification for

clearance orders 86-1008 and 86-688. During the performance of the

verification the RI noted on Clearance Order 86-1008 that Tag Number 24

had been removed, but the " Return to Normal By" blank had no initials in

the blank as required. This item was brought to the licensee's attention.

Discussions with the licensee revealed that Tag No. 24 had been removed.

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The person who removed the tag had initialed the wrong blank on the

clearance order. The responsible individual corrected the discrepancy in

a timely manner.

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 deviations were identified in this area.

10. Monthly Surveillance Observations

The NRC inspectors observed Technical Specification required surveillance

tests. Those observations verified that:

. Tests were accomplished by qualified personnel in accordance with

approved procedures.

. Procedures conformed to Technical Specification requirements.

. Tests prerequisites were completed including conformance with

applicable limiting conditions for operation, required

administrative approval, and availability of calibrated test

equipment.

. Test data was reviewed for completeness, accuracy, and confonnance

with established criteria and Technical Specification requirements.

. Deficiencies were corrected in a timely manner.

. The system was returned to service.

The NRC inspectors observed the licensee's performance of the following

surveillance tests on the indicated dates:

. December 10, 1986: Surveillance Procedure (SP) 6.2.2.5.12A, "RiiR

Loops A and B Pump tnd Valve Control Logic Reactor Vessel Pressure

Less Than or Equal to 75 psig Functional Test," Revision 9, dated

November 13, 1986.

. December 12, 1986: SP 6.1.1, "SRM Functional Test (Reactor Not In

Run)," Revision 14, dated October 30, 1986.

. December 12, 1986: SP 6.1.26, " Refueling Platform Interlock and

System Functional Tests," Revision 20, dated December 4, 1986,

Attachments B and E.

. December 12, 1986: SP 6.1.27, " Refueling Platform Interlock and

System Functional Tests," Revision 20, dated December 4,1986,

Attachments E and D.

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. December 16, 1986: SP 6.2.27, " Refueling Platfonn Interlock and

System Functional Tests," Revision 20, dated December 4,1986,

Attachment B.

. December 17, 1986: SP 6.4.3.1, "HPCI Turbine Overspeed Functional

Test," Revision 10, dated May 15, 1986, with temporary procedure

change added.

. December 21, 1986: General Operating Procedure (GOP), 2.1.18

" Control Rod Drive Friction Test," Revision 10, dated March 21,

1986, with temporary change procedure added.

. December 21, 1986: SP 6.4.1.2," Withdrawn Control Rod Operability,"

Revision 16, dated December 18, 1986.

. December 21, 1986: SP 6.4.1.3, "CRD Coupling Integrity Check,"

Revision 7, dated October 30, 1986.

The reviews and observations were conducted to verify that facility

surveillance operations were perfonned in accordance with the

requirements established in the CNS Operating License and Technical

Specification.

No violations or deviations were identified in this area.

11. Monthly Maintenance Observation

The NRC inspectors observed preventive and corrective maintenance

activities. These observations verified that:

. Limiting conditions for operation were met.

. Redundant equipment was operable.

. Equipment was adequately isolated and safety tagged.

. Appropriate administrative approvals were obtained prior to

commencement of work activities.

. Work was performed by qualified personnel in accordance with

approved procedures.

. Radiological controls, cleanliness practices, and appropriate fire

prevention precautions were implemented and maintained.

. Quality control checks and postmaintenance surveillance testing were

performed as required.

. Equipment was properly returned to service.

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The NRC inspectors observed the licensee's performance of the following

naintenance activities on the indicated date.

. December 18, 1986: "A" Service Water Pump

. December 21, 1986: Design Change (DC)86-134, " Replacement of

Reactor Building Personnel Airlock Doors"

. December 22, 1986: MaintainanceonReactorEquipmentCooling(REC)

Heat Exchanger A and B Outlet Valves

. December 22, 1986: Diesel Generator No.1 Head Torque and Water

Jacket Gasket Repair

These reviews and observations were conducted to verify that facility

maintenance operations were perfonned in accordance with the requirements

established in the CNS Operating License and Technical Specification.

CNS issued a justification for Interim Operation in response to a

self-identified problem in the Standby Gas Treatment (SGT) System.

Reference Licensee Event Report 86-28, "SGT System Seismic Design

Deficiencies," dated November 20, 1986. A review of the original design

documentation revealed that no weld records existed to identify the type

of welding rod used to weld the dissimilar metals (stainless steel to

carbonsteel.) To ensure that the welding rod used to make the

dissimilar metal welds was correct the RI performed a visual inspection

of 12 different welds selected at random. The following list of hanger

welds were visually inspected for stress cracking (SC) and weld

degradation (WD) with results noted:

Hanger Designation Result Hanger Designation Result

F01 No SC or WD F02 No SC or WD

301 No SC or WD 302 No SC or WD

303 (2) No SC or WD 304 No SC or WD

306 No SC or WD 307 No SC or WD

309 No SC or WD 313 No SC or WD

317 No SC or WD

Since no SCs or WDs were found, it was concluded that the appropriate

welding rod material had been used to weld the dissimilar metals. No

further action was required.

No violations or deviations were identified in this area.

....

17

12. Unresolved Items

An unresolved item is one about which additional information is required

in order to determine if the item is acceptable, a violation, or a

deviation. The following unresolved items were identified during this

inspection.

Item Paragraph Subject

298/8636-01 2 Diesel Sequential Loading

13. Exit Interviews

Exit interviews were conducted at the conclusion of each portion of the

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

inspection segrrent at those meetings.