IR 05000269/1976013
| ML19329A466 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/27/1976 |
| From: | Epps T, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19329A464 | List: |
| References | |
| 50-269-76-13, 50-270-76-13, 50-287-76-13, NUDOCS 8001031033 | |
| Download: ML19329A466 (9) | |
Text
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IE Inspection Report Nos. 50-269/76-13, 50-270/76-13-and 50-287/76-13 Licensee:
Duke Power Company Power Building
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422 South Church Street Charlotte, North Carolina 28201 Facility Name:
Oconee Units 1, 2 and 3 Docket Nos.:
50-269, 50-270 and 50-287 License Nos.:
DPR-38, DPR-47 and DPR-55 Category:
C, C and C Location:
Seneca, South Carolina Type of Inspection:
Routine, Unannounced Dates of Inspection:
November 16-19 and November 30 - Decercher 3,1976 Dates of Previous Inspection:
October 13-15, 1976
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Principal Inspector:
T. N. Epps, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Accompanying Inspectors:
None Other Accompanying Personnel:
None Principal Inspector:
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T. N! Epps, Reactor Inspectogr
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Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch
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Reviewed By:
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R. C. LAwis, Chit gT '
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Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch e
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SUMMARY OF FINDINGS L
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I.
Enforcement Items None
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I II.
Licensee Action on Previously Identified Enforcement Matters Licensee corrective action on item I.A.1 of IE inspection
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report 50-269, 270, 287/76-6 remains open.
(Details I, paragraph 4)
III.
New Unresolved Items 76-13/1 Verification of Redundant Equipment Operability f
l Redundant. component operability verification prior to maintenance is not always required.
(Details I, l
paragraph 6.e.)
IV.
Status of Previously Reported Unresolved Items
Not' inspected.
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V.
Unusual Occurrences l
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See Details I, paragraph 6.
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Other Significant Findings
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None VII.
Management Interviews
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. Meetings were held on November 19 and December 3, 1976, by T. N. Epps with J. E. Smith and J. W. Hampton respectively k
to discuss the findings of this inspection presented in the
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Details of this report.
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/M.7-7 7 DETAILS I Prepared by:
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T. N. Epps, Reactor Insp(ctor Date'
Reactor Proj ects Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection:
November 16-19 and November 30 -
December 3, 1976
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Reviewed by:
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Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch 1.
Individuals Contacted Duke Power Company (DPC)
Oconee Personnel J. E. Smith - Manager, Oconee Nuclear Station J. W. Hampton - Manager, Administrative Services L. E. Schmid - Superintendent of Operations 0. S. Bradham - Superintendent of Maintenance R. M. Koehler - Superintendent of Technical Services R. T. Bond - Technical Services Engineer J. N. Pope - Operating Engineer G. A. Ridgeway - Assistant Operating Engineer W. R. Campbell - Reactor Engineer W. M. Harris - Operating Engineer D. Hunter - I&E Supervisor L. Knight - I&E Technician
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Other Operations Personnel Corporate Office M. Tuckman - Staff Engineer (Licensing)
Nuclear Regulatory Commission D. Neighbors - Nuclear Reactor Regulation (Oconee Project Manager)
F. Clemenson - Nuclear Reactor Regulation P. Atherton - Nuclear Reactor Regulation
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' Plant Operations Reviews conducted, in this part of the inspection, are su=marized below.
All 3 units were operating during this inspection, a.
Records The inspector reviewed shift supervisor's and control room log books for the two weeks ending on November 18, 1976, on Units 1 and 2 and for a one month period ending on November 17, 1976, on Unit 3.
Recent control room log sheets, work orders and the out of normal log were also reviewed on all 3 units.
The above records appeared to conform to sections 6.5, 3.0 and 4.0 of the Technical Specifications.
b.
Control Room Observations The inspector observed control room data, switch positions and monitoring instrumentation to verify the following instrumenta-tion to be operational on each unit in accordance with Technical Specifications (TS):
BWST Level - T.S. 3.3.1.e CF Tank Pressure and Boron Concentration - T.S. 3.3.3 Reactor Power Range.nannels - T.S. 3.5.1 RCS Pressure Instrusents
'T.S.
3.5.1 The inspector also verified that operating control rod group overlap was within Technical Specification 3.5.2.5 require =ents and that control rod overlap requirements of T.S. 3.5.2.2.a were met.
Discussions were held with reactor operators concerning control room alarm indications.
Heat tracing alarms were on in the Unit 1 and 3 control rooms.
This wac due to several local heat-tra2ing alarms being on and acknowledged in the auxiliary building.
The licensee took action to assure that the alarms were functioning properly.
Heat tracing was not observed to be inadequate on any safety systems.
The control rod drive fault alarm was al'so on in the Unit 3 control room.
The licensee stated that this was due to a problem with the alarm systen and was being investigated.
The inspector also verified that control rooms were staffed to meet the requirements of Technical Specificatier. Table 6.1-1.
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k RadiationmonitoringinstrumentationwasobservegontheUnit 1 air ejegtor to be reading approximately 6.5x10 cpm, as the licensee had stated, du2 to a primary to secondary leak in the 1B steam generator.
The licensee stated that this reading would be cauged by a.05 gpm leak and that if the air ejector reached 6x10 cpm and if air ejector grab samples reached 0.01 micro-curies per millileter indicating a 1 gpm leak, the unit would be shut down for repair.
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c.
Plant Tour The inspector toured various portions of the facility on 3 different occasions.
These tours included the turbine building and auxiliary building and the Lake Keowee condenser circulating water system discharge structure.
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The licensee showed the inspector a Unit 3 hotwell pump suction header that was deflected such that one of 3 expansion joints was nisaligned several inches.
The licensee dye pene-trant inspected the appropriate welded joints on the header and on an 8-inch diameter emergency feedwater line connecting to the header.
The licensee stated that no adverse indications resulted from the dye penetrant testing.
During the tour the inspector observed approximately 10 drums full of used oil and other debris stored on the basement floor of the turbine building.
This was identified as a possible fire hazard to the licensee.
The licensee stated that the material was to be removed.
3.
Turbine Building Flooding Three Nuclear Regulatory Commission personnel from the Office of
Nuclear Reactor Regulation visited the site on Friday, November 19, 1976, to observe equipment and gather information on conditions
.related to the turbine building flooding incident (RO-287/76-18) of October 9, 1976.
The personnel are involved in reviewing the safety implications associated with the turbine building flooding and proposed corrective actions.
4.
Previous Items of Noncompliance
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The inspector reviewed licensee corrective actions on noncompliance item I.A.1 of IE Inspection Report 50-269, 270, 287/76-6.
This item remains open pending further review.
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5.
Organization and Administration Reviews were conducted, during this inspection, to verify that the
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station organization meets the requirements of Technical Specifica-tion 6.1.
This review included personnel conducting reviews onsite and serving on the Nuclear Safety Review Board (NSRB) at the licensee's corporate office and discussions and observations by the inspector.
The inspector questioned how the NSRB meets Technical Specification
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6.1.3.4 which defines items included in NSRB audits.
The NSRB Chairman stated that the NSRB takes credit for QA audits conducted by the site QA organization.
The inspect'r reviewed QA records of audits of corrective actions on incident; affecting nuclear safety (Technical Specification 6.1.3.4.C).
It was determined that QA audits meet the frequency requirements ot Technical Specifica-tion 6.1.3.4.C.
The inspector had no further q'restions on this subj ect.
6.
Reportable Occurrences The following reportable occurrences were reviewed at the site during this inspection:
a.
R0-269/76-15 involving isolation of the 230 kv yellow buss during electrical breaker testing, on September 30, 1976, was reviewed.
This incident resulted from a breaker failure relay malfunction.
Corrective actions were verified and there were no further questions.
b.
R0-269/76-14 involved exceeding control rod insertion limits.
Review of station records and discussions with licensee personnel showed the cause to be rapidly increasing power (257. per hour)
during recovery from a reactor trip with transient xenon conditions.
Reactor power was increased from 20 percent to 70 percent in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and then to 90 percent approximately two hours later.
Power imbalance became a problem so demineralized water was added to the reactor coolant system to drive control rods in but xenon was being reduced at the same time and the control rods were inserted past the control rod insertion limit.
This was contrary to Technical Specification 3.4.2.5.c and Figure 3.5.2-1A1.
Licensee corrective actions stated in the licensee's-event report were verified and the inspector had no further questions.
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c.
'RO-270/76-12 involved an inoperable containment isolation
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valve.
This was reviewed during a previous inspection (IE report 50-269, 270, 287/76-12) and there were no further questions.
d.
R0-270/76-ll involved isolation of the 230 Kv switchyard, Red Buss and startup transformer CT2 on September 24, 1976.
This was contrary to Technical Specification 3.7.2.a which allows only one startup transformer to be out of service, and was caused by an error in the implementation of a station modification.
The affected equipment was returned to service within 18 minutes of the incident.
The inspector had no further questions.
e.
R0-270/76-9 resulted from taking LPI train "A" out of service without verifying Train "B" operability.
- Technical Specification 3.3.7 requires that prior to initiating maintenance on any HPI, LPI, LPSW, RB spray or RB cooling component, the duplicate (redundant) component shall be tested to assure operability.
The Bases of the subject Technical
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Specification states that the allowable maintenance period is acceptable if the operability of equipment redundant to that removed from service is demonstrated immediately prior to removal.
The specification is unclear on whether to test a redundant pump when a pump is taken out of service by performing maintenance on a valve in the same train.
This is a new unresolved item.
f.
R0-270/76-10 involved increased activity in the component cooling system resulting from letdown cooler leakage.
The leaking cooler "B" was isolated and the "A" cooler is presently being used.
The licensee plans to replace the "B" cooler.
The inspector had no further questions.
g.
R0-287/76-17 resulted from a defective motor which caused a reactor building cooling unit to be inoperable.
Conditions of Technical Specification 3.3.6.d were met and corrective actions were verified.
There are no further questions on'this item.
h.
R0-287/76-16 involved loss of power to some ES equipment due to failure of a static inverter.
The inverter was repaired
'and. returned to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> which is within the
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There are
no further ' questions on this item.
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R0-287/76-13 involved an inoperable containment isolation valve in the containment sampling system. Action was taken by the licensee to meet Technical Specification 3.6.4.b.2.
The inspector verified that the valve was repaired and there were no further questions.
j.
R0-287/76-15 involved an erroneous data input to the plant computer which caused an error in the plant heat balance from which core thermal power is determined.
This input error resulted in a 9 percent error in thermal power such that the indicated power level was 44 percent when actual power was 53 percent.
The reactor was operated at the above power level for 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> before the error was discovered and corrected.
The licensee is reviewing proposed administrative controls for computer software.
This item will receive further review to determine chat the administrative controls are appropriate to prevent recurrence.
This incident was contrary to Technical Specification 2.3 which defines reactor power trip setpoints.
k.
R0-287/76-12 involved loss of a motor control center feeding Reactor Building (RB) Cooling Unit 3B and three ES RB isolation valves.
Redundant equipment was operable and the
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motor control center in question was out of service for nine 9 minutes. A station modification is to be completed by January 1, 1976, which will reroute the power supply to RBCU's and a study of the breaker coordination for 600 volt ES motor control.
centers is to be completed by the same date.
This item will receive future review.
1.
R0-287/76-14 involved failure to sample core flood tanks, after makeup, for boron concentration, six times during June, 1976.
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This was contrary to Technical Specification 4.1.2.
Corrective action was reviewed and the inspector had no further questions.
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LPI Motor Bearings The in'spector discussed LPI motor locking devices used on Westinghouse motors that could be damaged if they used the type of locking nuts with threaded nylon inserts.
A licensee representative stated that
- the Unit 3 LPI pumps 'were inspected during the recent refueling outage and were verified to not have the nylon type inserts.
Units 1 and 2 LPI pump motors were not of the type that could have used the nylon insert locking devices.
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Caustic Mix Tank-
- Review of the caustic mix' tank' location and NaOH concentration showed'the tank to'contain a 50 percent concentration of NaOH:and
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the ' tank is located on the second floor of the auxiliary building.
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Weekly visual inspections of the contents of the tank are conducted by chemistry personnel. This system would be manually operated af ter a loss of coolant ' accident for the purpose of adding NaOH to
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the LPI' System to neutralize'the boric acid effect on equipment.
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9.
IE Bulletins and Circulars The licensee's response dated May. 5,1976, y
IEB-76-02 stated that 50 GE type HFA relays as described in IE Bulletin a.
The coils of 76-02 were used in safety-related applications.
these :50 relays will be replaced with the recommended Lexan F
This item remains open.
type ' coils by January'15,1977.
The licensee's letter dated May 5,1976, stated IEB-76-03 that no GE type STD relays as described in IE Bulletin 76-03 b.
This item is closed.
were used at Oconee.
These Bulletins are closed based on s-IEB-76-05 and 76-06_~
-licensee letters dated April 30, 1976 and July 30, 1976, (
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stating 'that the subject equipment is not used at Oconee.
This circular is closed based on the licensee's
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IEC-76-02_
letter dated October 11, 1976, stating that Westinghouse BF and BFD relays are not used in safety-related systems at Oconee.
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