ML20210D501
| 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: |
| 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
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Facility Name: Cooper Nuclear Station (CNS)
Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska
Inspection Conducted:
December 1-31, 1986
Inspectors:
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E. A. Plettner,- Resident inspector, (RI)
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D. L. DuBois, Senior Resident Inspcctor, (SRI)
Date-
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Approved:
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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
- G. R. Horn, Division Manager of Nuclear Operations
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- 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
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course of the inspection.
- Denotes those present during exit interview January 12, 1987.
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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.
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The RI observed the performance of the following procedures on the
indicated dates:
December 30, 1986:
Surveillance Procedure (SP) 6.3.4.3 "CS, RHR,
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and Diesel Auto Start and Loading," Revision 21 dated August 22,
1985;
December 31, 1986 and January 1,1987:
General Operating Procedure
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(GOP) 2.1.14 " Reactor Vessel In-Service Leak Test," Revision 16,
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dated April 3, 1986.
Reviews and observations verified that:
Testing was performed using approved procedures, which were
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consistent with regulatory requirements, industry standards, and the
Technical Specification.
Pennanent or temporary procedure revisions were accomplished
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according to administrative requirements and controls.
Qualified personnel conducted the tests and performed the final
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reviews and opprovals of completed test data.
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The official test copy was available and used by test personnel.
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Procedures contained the purpose, objectives, references,
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prerequisites, test equipment, precautions, limitations, and
acceptance criteria.
Test equipment required by the procedures was calibrated and in
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service.
Procedures provided sufficient direction to accomplish necessary
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evolutions.
Systems were returned to normal lineup following completion of
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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
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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
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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
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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
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self-identified by the licensee, and it meets the remaining four criteria
stated in Appendix C,Section V.A of 10 CFR Part 2.
One unresolved item was identified in this area.
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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
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RI also reviewed GE supplied nuclear engineering data, which was used to
determine the sequence of rod withdrawals, expected reactivity addition
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rates, and predicted shutdown margin values. The RI verified that test
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results conformed with Technical Specification requirements.
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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:
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Procedure prerequisites were met prior to performing the core
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thermal power evaluations.
Figures and curves corresponding to specific reactor conditions were
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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
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requirenents.
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The following CNS procedures were reviewed:
Nuclear Performance Procedure (NPP) 10.1, "APRM Calibrations,"
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Revision 16, dated October 30, 1986
NPP 10.2, "IRM Power Calibration," Revision 11, dated May 11, 1984
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NPP 10.3, " Core Thermal Power Evaluation," Revision 7, dated
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October 31, 1986
NPP 10.5, "LPRM Calibration," Revision 17, dated January 25, 1985
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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
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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
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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
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normalization factors were properly obtained.
Examination of licensee procedures for ascertaining operation within
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licensed limits, should the process computer becorre unavailable.
Verification that average power range monitor (APRM) channel gains
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were adjusted as necessary following an LPRM calibration.
Verification that following an APRM gain adjustment, a subsequent
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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
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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"
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0D-3, " Core Thermal Power and APRM Calibration"
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00-6, " Thermal Data for Specified Fuel Bundles"
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00-7, " Control Rod Notch Positions"
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00-8, "LPRM Console Readings"
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00-10. " Edit Specified Data Array"
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OD-15. " Computer Shutdown and Outage Recovery Monitor"
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00-16. " Target Exposure and Power Data"
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00-17 " Edit Periodic Core Perfomance Logs"
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P-1, " Periodic Core Performance"
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The following CNS procedures were reviewed:
NPP 10.4, " Core Thermal Hydraulic Evaluation " Revision 8, dated
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April 24, 1986
NPP 10.7, " Maximum dverage Planar and Peak Linear Heat Generation
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Rates and Minimum Critical Power Ratio," Revision 8, dated
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October 30, 1986
NPP 10.8, " Reactivity follow Check," Revision 9, dated October 18,
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NPP 10.9, " Control Rod Scram Time Evaluation," Revision 13, dated
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November 6,1986
NPP 10.10. " Limiting Control Rod Pattern Determination,"
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Revision 7, dated January 21, 1985
NPP 10.11. " Control Rod Sequence Exchange," Revision 8, dated
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October 30, 1986
NPP 10.13. " Control Rod Sequence and Movement Control," Revision 13,
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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.
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Power supply voltages were all within tolerance.
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The instruments and calibration equipment used were traceable to the
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National Bureau of Standards.
Test equipments used and their serial numbers were recorded on the
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procedure.
Licensee's procedures ensure that the setpoint levels for alarms,
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permissive and prchibitive interlocks are in compliance with the
appropriate Technical Specifications.
Calibration results were reviewed, approved, and documented in
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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
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15, dated September 18, 1966.
SP 6.1.17, "IRM Calibration and Functional Test, (Mode Switch not In
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Run)," Revision 10, dated August 8, 1985.
SP6.1.17A,"IRMCalibrationandFunctionalTest(ModeSwitchIn
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Run)," Revision 3, dated May 22, 1982
SP 6.1.19. "LPRM Calibration Test," Revision 7, dated March 21, 1986
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SP6.1.21,"SRMCalibrationandFunctionalTest(ReactorNotIn
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Run)," Revision 14 dated October 8,1984
SP6,1,21A,"SRMCalibrationandFunctionaltest(ReactorInRun),"
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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
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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
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Calibration Procedure," Revision 13, dated October 30, 1986.
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I&C 7.5.2.2, "IRM Calibration Procedure," Revision ll, dated
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October 30, 1986
I&C 7.5.2.3, "LPRM Calibration Procedure," Revision 11, dated
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April 10, 1986
I&C 7.5.2.4, "APRM Calibration Procedure," Revision 12, dated
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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"
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2 times in I&C-7.5.2.4 Procedure, Attachment "A"
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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.
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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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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
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the fuel bundles or core top guide lissembly.
The lost pieces of the aluminum blade guide handle in the reactor
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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
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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
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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
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systems provided reasonable assurance that cold weather protective
measures were reestablished following completion of those
activities.
Plant preventive maintenance requirements associated with cold
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weather preparation were completed on September 17, 1986.
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The following documents were reviewed:
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Preventive Maintenance (PM) routine 04047 and 01271
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General Operating Procedure (G0P) 2.1.11, " Station Operators Tour,"
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Revision 39, dated October 30, 1985.
Attachment "C" to G0P 2.1.11. " Station Operators Tour - R/W, A0G,
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ARW Areas and Outside."
System Operating Procedure (SOP) 2.2.30, " Fire Protection System,"
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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.
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Technical Specification requirements were observed.
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Plant operations were conducted using approved procedures.
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Plant logs and records were complete, accurate, and indicative of
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actual system conditions and configurations.
System pumps, valves, control switches, and power supply breakers
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were properly aligned.
Licensee systems lineup procedures / checklists, plant drawings, and
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as-built configurations were in agreement.
Instrumentation was accurately dis alaying process variables and
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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
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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.
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Equipment safety clearance records were complete and ind'i
ed that
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affected components were removed from and returned to. service in aN
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correct and approved manner.
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Maintenance work requests were initiated for equipment disco ered to'
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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
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maintained.
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Areas of the plant were clean, unobstructed, and free of fire
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hazards. Fire suppression systems and emergency equipment were
maintained in a condition of readiness.
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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)
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Main Steam and Turbine Bypass (MS)
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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,
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Bypass System," Revision 21, dated August 7,.1980, Appendir'A,
" Valve Checklist"
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As-Built drawing - Burns & Roe 2002; for MS System
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As-Built drawing - Burns & Roe 2041; for MS System
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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
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As-Built drawing - NPPD I.D.-1; for MS System
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As-Built drawing - NPPD I.D.-3; for MS System
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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
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As-Built drawing - Burns & Roe 2004; for Feedwater
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'As-Built drawing - Burns & Roe 2005; for Feedwater
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As-Built drawing - Burns & Roe 2043; for Feedwater
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As-Built drawing - Burns & Roe 2044; for Feedwater
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As-Built drawing - Burns & Roe M-110; for Feedwater
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As-Built drawing - Burns & Roe M-111; for Feedwater
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As-Built drawing - NPPD I.D.-4; for Feedwater
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As-Built drawing - NPPD I.D.-6; for Feedwater
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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
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approved procedures.
Procedures conformed to Technical Specification requirements.
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Tests prerequisites were completed including conformance with
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applicable limiting conditions for operation, required
administrative approval, and availability of calibrated test
equipment.
Test data was reviewed for completeness, accuracy, and confonnance
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with established criteria and Technical Specification requirements.
Deficiencies were corrected in a timely manner.
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The system was returned to service.
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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
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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
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Run)," Revision 14, dated October 30, 1986.
December 12, 1986:
SP 6.1.26, " Refueling Platform Interlock and
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System Functional Tests," Revision 20, dated December 4, 1986,
Attachments B and E.
December 12, 1986: SP 6.1.27, " Refueling Platform Interlock and
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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
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System Functional Tests," Revision 20, dated December 4,1986,
Attachment B.
December 17, 1986:
SP 6.4.3.1, "HPCI Turbine Overspeed Functional
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Test," Revision 10, dated May 15, 1986, with temporary procedure
change added.
December 21, 1986:
General Operating Procedure (GOP), 2.1.18
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" 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,"
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Revision 16, dated December 18, 1986.
December 21, 1986: SP 6.4.1.3, "CRD Coupling Integrity Check,"
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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.
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Redundant equipment was operable.
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Equipment was adequately isolated and safety tagged.
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Appropriate administrative approvals were obtained prior to
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commencement of work activities.
Work was performed by qualified personnel in accordance with
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approved procedures.
Radiological controls, cleanliness practices, and appropriate fire
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prevention precautions were implemented and maintained.
Quality control checks and postmaintenance surveillance testing were
.
performed as required.
Equipment was properly returned to service.
.
I
....
16
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.