IR 05000269/1977003
| ML19316A263 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/07/1977 |
| From: | Alderson C, Bradford W, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19316A249 | List: |
| References | |
| 50-269-77-03, 50-269-77-3, 50-270-77-03, 50-270-77-3, 50-287-77-03, NUDOCS 7912050837 | |
| Download: ML19316A263 (16) | |
Text
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. , IE Inspection Report Nos. 50-269/77-3, 50-270/77-3, and 50-287/77-3 ' Licensee: Duke Power Company Power Building 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 Location: Seneca, South Carolina Type of Licenses: B&W, PWR, 2560 Mut Type of Inspection: Routine, Unannounced Dates of Inspection: March 8-11, 1977 Dates of Previous Inspection: February 1-4, 16-18, and 22, 1977 Principal Inspector: C. E. Alderson, Reactor Inspector Accompanying Inspector: W. H. Bradford. Reactor Inspector Other Accompanying Personnel: J. C. Ledoux, Operations Analyst IE Study Group, IE:HQ J. Troutman, Consultant to IE:HQ Principal Inspector: [ 6.
N///27 M C. E. Alderson, Reactor Inspector Date / Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Reviewed by: , [. 7/77 R. C. Lewis,' Chief Dat'e Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch i Q -- - 0 ~1912 . j .
' . - ( ' . IE Rpt. Nos. 50-269/77-3, i fs 50-270/77-3 and 50-287/77-3-2- ' . SUMMARY OF FINDINGS 1.
Enforcement Items Infractions A.
Contrary to the requirements of Criterion V of Appendix B to 10 CFR 50 as implemented by Section 17.2.5 of the DPC QA Topical Report, maintenance performed on January 2, 1977, on containment isolation valve 3RC-7 was not in accordance with the approved work request as specified in Section 3.3.2.3 of the Administrative Policy Manual and Station Directive 3.3.5 / which resulted in post-maintenance testing not being performed.
Unit 1 only.
(Details I, Paragraph 3.m) B.
Contrary to the requirements of Technical Specification 6.4.1, the station was not operated and maintained in accordance with approved radiation control procedures, in that on March 9, 1977, radioactively contaminated trash was found in unlabeled containers and contaminated materials were found outside of a Radiation Control Zone without the required warning tags.
(Details I, Paragraph 4.a) II.
Licensee Action on Previously Identified Enforcement Matters The licensee's actions on the Infraction identified in Paragraph I.3 of the Summary of Inspection Report No. 50-269/76-12 concerning removal of fire extinguishers from their assigned locations were reviewed and this item is closed.
(Details II, Paragraph 3) III.
New Unresolved Items None IV.
Status of Previously Reported Unresolved Items 74-14/2 - Ventilation Control Between Auxiliary and Turbine Building The licensee has completed equipment modifications; however, the operating data gathered has not been analyzed to demonstrate system adequacy with open doors between the auxiliary and turbine buildings.
This item remains open.
(Details I, Paragraph 4.b) t -
- - - - -._ . . . . l tq IE Rpt. Nos. 50-269/77-3,
.y ,50-270/77-3 and 50-287/77-3-3-i ' ? ! ,
! V.
Unusual Occurrences None ! VI.
Other Significant Findings None VII. Manage 2nent Interviews i ' ' A meeting was held by C. E. Alderson and W. H. Bradford on March 11,
1977, with J. E. Smith and members of the Oconee staff to discuss
the findings presented in Details I and II of this report.
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. . - - ...-. -- - , - -. - - -. -... -. -. . -. - _ _ - . . - . (~}. IE Rpt. Nos. 50-269/77-3, , - 50-270/77-3 and 50-287/77-3 I-l ' DETAILS I Prepared by: [. d' [/7/77 h C. E. Alderson, Reactor Inspector Date V Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection: March 8-11, 1977 Reviewed by: [. 6.
kM R. C. Lewi's, Chief Date Reactor Projects Section No. 2 ' Reactor Operations and Nuclear , Support Branch l 1.
Individuals Contacted Duke Power Company (DPC) J. E. Smith - Manager, Oconee Nuclear Station J. Hampton - Director of Administrative Services L. Schmid - Superintendent of Operations 0. Bradham - Superintendent of Maintenance C. Yongue - Health Physics Supervisor , R. Bond - Technical Services Engineer M. Harris - Operating Engineer R. Nord - Associate Engineer R. Casler - Unit 1 Shift Supervisor j G. Itin - Safety Supervisor Other maintenance and operations personnel 2.
IE Bulletins and Circulars ' The inspector held discussions with Technical Services Group and Maintenance Group personnel, and reviewed maintenance requests and other documents to verify for each of the IEBs and IECs listed below that: (1) the licensee's response was prepared within the time period specified; (2) appropriate site managenent had received copies of the written response; (3) information presented in the response was accurate and supported by records or direct observation; and (4) any actions committed to in the response had been effected.
a.
IEB 76-02, Relay Coil Failures This Bulletin described failures of certain types of GE relays having nylon coil spools.
The licensee's response dated () May 5,1976, stated that 50 of the GE Type HFA relays were I - l
, _ ._~ _ -- ._._ __ _ __ _ _ _ _ _. - _-_ _ _ _ _ _ _ _ - _ _ - - - . _ _ - - - - i . "
j (-')- IE Rpt. Nos. 50-269/77-3, 50-270/77-3 and 50-287/77-3 1-2 , used in safety-related applications at Oconee and committed to have the coils on nylon spools replaced with coils on Lexan spools by September 1, 1976.
Subsequent letters from the licensee changed the commitment date to January 15, 1977.. The inspector reviewed Nuclear Station Modification ON-593 and determined that the Type HFA relays had been inspected and that replacement of the white nylon spools had been accomplished.
, This item is closed.
b.
IEB 76-07, Crane Hoist Controls / This Bulletin addressed modifications to the hoist control for ' installed cranes, especially the brake power and control circuitry.
The licensee's response dated August 17, 1976, stated that no modifications had been made to the spent fuel cranes and that none were planned.
The inspector held discus-sions with the individual who had performed the review and had no futher questions.
This item is closed.
c.
IEC 76-05, Hydraulic Shock and Sway Suppressors i ' This Circular discussed improper lockup and bleed rates of certain 1TT Grinnel shock and sway suppressors (snubbers).
The licensee's response dated November 11, 1976, stated that i ' there are no snubbers installed at Oconee from the series identified in the Circular (B-0001 through B-2000).
The inspector determined that the licensee had completed a func-tional test of all suppressors on Unit 3 and documentation from this testing confirmed that the subject suppressors were not used on Unit 3; however, similar documentation was not
available for Units 1 and 2.
The individual responsible for determining whether the subject suppressors were used stated that he had performed a spot check of suppressors in Units 1 and 2 and that functional testing of all suppressors on these units was scheduled for the next refueling outage.
The inspector stated that a spot-check was not sufficient to determine that the subject suppressors were not installed.
This item remains open pending conclusive evidence that none of the subject suppressors are installed in Units 1 and 2.
4 d.
IEC 76-06, Stress Corrosion Cracks , This Circular described through-wall cracks found in low pressure stainless piping containing boric acid solutions.
A copy of the licensee's response dated December 28, 1976, was j forwarded to IE:HQ for evaluatien of the program and schedule described therein.
The inspector determined that the testing - -
, . . { ~} IE Rpt. Nos. 50-269/77-3, ( 50-270/77-3 and 50-287/77-3 I-3 ' .,
described had not been entered into the licensee's formal system for assuring completion of commitments.
This item remains open pendn.g inclusion of the testing into the tracking system.
3.
Reportable Occurrences The inspector perf ormed an inof fice review of written event reports received from the licensee pursuant to the requirements of Technical Specification 6.6.2.1.
These reports are listed below.
The purpose of these reviews was to determine whether: (1) reporting require- 'ments were met; (2) the details provided were adequate to assess the event; (3) the stated cause appeared accurate; (4) the corrective action appeared appropriate; and (5) the licensee had considered any generic implications.
'
In addition, for all written reports submitted pursuant to Technical ' Specification 6.6.2.la and for a portion of those submitted purs2 ant to Technical Specification 6.6.2.lb, the inspector held discussions with site personnel and reviewed operating records and logs, main-tenance documents and licensee internal incident investigation reports to verify that: (1) the event was as described in the report including the reported cause; (2) the stated corrective j action had been completed or responsibilities assigned for assuring completion; (3) licensee management was notified of the event as
required by Technical Specifications 6.2.1 and 6.1.2.1; and (5) the l event did not involve operation in a manner which constituted an unreviewed safety question as defined in 10 CFR 50.59(a)(2) or an
unusual hazard to the health and safety of the public.
a.
RO 269/76-16, Inadvertent Power Level Increase Above Power Level Cutoff Limit The report stated that the apparent cause of this occurrence was a faulty level indicator on the feedwater heater "D2" flash tank.
The inspector determined that, while the failure of the level indicator contributed to the event by requiring isolation of a feedwater heater, it did not cause the event.
The power level cutoff limit established at 92.5% of full power by Technical Specification 3.5.2.3d was exceeded due to a feedwater transient created by the operator's actions to return the feedwater heater to service.
Regarding the licensee's commitment to revise procedures by December 1, 1976, the inspector determined that revision of the five affected procedures had been completed; however, it was noted that the revisions were not initiated and approved until December 13 and 31, 1976, respectively.
The inspector stated that when commitments . . __ - . .- -
. IE Rpt. Noo. 50-269/77-3, ' () 50-270/77-3 and 50-287/77-3 I-4 , ' . made in RO reports cannot be met, a revised report should be submitted to the NRC; otherwise it is expected that corrective actions will be completed by the date committed.
This item is closed.
b.
RO 269/76-17, RO 269/76-19, R0 269/77-2, R0 270/76-15 and RO 287/76-2, Steam Generator Tube Leaks The inspector determined that operating records supported the description of the events stated in these reports and that maintenance records supported the stated corrective actions.
,. The specific cause of the tube failures is not known, but is under active investigation by DPC and the steam generator manufacturer, Babcock and Wilcox.
The results of these investi-gations are being reviewed by NRR.
The inspector has no further questions on these specific reports; however, identifi-cation of the cause of the tube failures and corrective actions to be taken remain an open item.
c.
RO 269/77-1, Penetration Room Exhaust Fan Discharge Valve ' Inoperable The inspector determined that the event description, cause and . corrective actions were as stated in the report and that the j requirements of Technical Specification 3.15 were met.
This item is closed.
d.
RO 269/77-4, Primary Coolant Leak From Incore Instrument Tube The inspector determined that the event description and correc-tive actions were as described in the report.
Documentation related to the expansion of the incore instrument tube due to freeze plugging was also reviewed and the inspector had no further questions.
This item is closed.
e.
R0 270/76-13, Defective Electrical Contact Resulting in Inoperable Valve LPSW-5 The inspector determined that the event description, cause and corrective action stated in the report were supported by operating and maintenance records and that Technical Specifica-tions 3.3.5 and 3.3.7 were satisfied.
This item is closed.
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e- . j [') IE Rpt. Nos. 50-269/77-3, ' 50-270/77-3 and 50-287/77-3 I-5 , f.
RO 270/76-14, Failure of Reactor Building Electrical Penetration WMV-1 During the inoffice review the inspector noted that the event description indicated that penetration WMV-1 had failed, but that a leak test was performed on WMV-2.
The inspector reviewed completed procedure PT/0/A/150/20 which was used to perform the leak test and determined that penetration WMV-1 had been tested and that the report reference to WMV-2 was a typographical error. This item is closed.
J' g.
RO 270/76-16, Reactor Operation Exceeded the Error Adjusted Imbalance Limit The inspector determined that the event description, cause and ! corrective action were as specified in the report and that the requirements of Technical Specification 3.5.2.6 were satisfied.
This item is closed.
h.
R0 270/76-17, Quadrant Tilt Limit Exceeded The inspector determined that operating and maintenance records i supported the event description, cause and corrective actions stated in the report and this item is closed.
' 1.
RO 270/77-1 and RO 287/76-21 0ne Channel of Borated Water u Storage Tank Level Indication Inoperable The inspector determined that the event description and cause stated in these reports was accurate.
The permanent corrective action described in the reports involves installation of new heat tracing on the instrument lines which will be accomplished , by October 3, 1977 Completion of this permanent corrective ' action will be verified during future inspections. As an interim corrective action a Cold Weather Checklist must be , accomplished anytime the outside temperature is below 35 F.
The inspector noted that the only requirement to perform the checklist was on the checklist itself and further that the checklist was not classified as an operating procedure.
The ' inspector stated that the checklist should be included in PT/600/1, which addresses shif t and daily surveillance require-ments, to assure that it is performed when required.
The inspector had no further questions and these items are closed.
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RO 270/77-2, Inoperable Reactor Building Spray Valge The reported cause of this occurrence was that the valve closed too hard on the valve seat causing the valve to stick; however, the event description and coirective action both stated that the torque and closing limit switches were both adjusted but did not discuss the reason for readjustment.
The inspector reviewed work request WR 19473A and determined that the "as found" limit switch settings were incorrect.
The licensee stated that the records indicated that the switches j had been properly set during the last previous quarterly ,- maintenance on the valve positioner and the "as found" settings
could not be definitely explained.
The inspector had no further questions and this item is closed.
,
k.
R0 287/76-19, High Pressure Injection Stop-Check Valves Closed { While Reactor Above 350"F ! The inspector determined that the event description and correc-l tive action were as described in the report. As stated, the ! requirements of Technical Specification 3.3.2 were not met for a period of approximately eleven and one-half hours in that the reactor was above 350 F with the "B" HPI train inoperable.
] The licensee's diaignation of apparent cause as " personnel error in ascertaining the proper valve position" appeared appropriate as the inspector found no information to the contrary.
This item is closed.
I 1.
RO 287/76-20, Feedwater Containment Isolation Valve Inoperable The inspector determined that the event description contained ~ in the original report, and the cause and corrective action stated in the supplemental report dated February 8, 1977, were supported by operating and maintenance records.
The inspector discussed minor comments regarding Maintenance Procedure MP/0/A/1200/29 and had no further questions.
This item is j closed.
' m.
RO 287/77-1 Reactor Building Containment Isolation Valve 3RC-7 Inoperable > ! The inspector reviewed WK 19659A and determined that the work requested was to adjust the valve packing and Maintenance Procedure MP/0/A/1200/1 (Adjusting and Packing Valves) was specified as the procedure to L. used.
The specified procedure
had been changed to indicate MP/0/A/1800/22 (Controlling j Procedure for Troubleshootjng and Corrective Maintenance).
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'50-270/77-3 and 50-287/77-3 I-7 < .. , Section V of the work request indicated that the work performed was the installation of a pipe cap on a leak off line and did not indicate that the valve packing was adjusted.
The licensee stated that the maintenance craftsman assigned to do the work determined that the leak was from the leak off line rather than the packing and that the specified procedure was then changed to allow installation of the pipe cap; however, the licensee's Incident Investigation Report No. B-580 indicated that, after installing the pipe cap, the maintenance craftsman noticed that the packing appeared to be cocked and he adjusted it, an action which was no longer authorized by the modified
/ work request.
The fact that the specified procedure was changed, but the requested action was not, appeared to be the cause of the failure to stroke the valve following adjustment of the packing.
Station Directive 3.3.5, " Maintenance Work Request" assigns responsibility for specifying the procedure to be used to the Maintenance Planner or the Maintenance Group Supervisor depending on the Priority assigned.
The directive further specifies that if corrections and/or additions are made to work requests, they must be signed and dated.
This directive was not followed in that the change in specified procedure was not initialed and dated and is contrary to the requirements of Criterion V of Appendix B to 10 CFR 50 as implemented by Section 17.2.5 of the DPC QA Topical Report and Section 3.3.2.3 of the Adcinistra-tive Policy Manual.
This is an Infraction.
The inspector also noted that Station Directive 3.3.5 is inadequate in that it allows changes to be made to work requests but does not specify who must approve such char.ges nor does it specify whether such approval must be accomplished prior to starting or continuing the work.
The inspector has no further questions concerning the licensee's report; however, the licensee's corrective actions regarding the noncompliance and the inadequacy of Station Directive 3.3.5 will be reviewed during future inspections.
l 4.
Plant Operations
The inspector toured various areas of the plant to ascertain the ' general status of plant equipment and area conditions.
Specific
conditions checked included: (1) recording instruments operating properly; (2) radiation controls properly established; (3) general housekeeping; (4) existence of fluid leaks or pipe vibrations; (5) condition of pipe hangers and seismic restraints; (6) valve and
l , , - - - - - _ _. _ _ __ ._ m -. -, _- - eey
,_-. . _. _ _.. _. _ _ _ _ _ _ _ _ _ ___ ___ _ . t . ,'] ' IE Rpt. Nos. 50-269/77-3, '. '50-270/77-3 and 50-287/77-3 1-8 switch positions; (7) lockout tag information; (8) status of indica-ting lights and annunciators; and (9) control room manning.
Areas of the plant toured included: all levels of the turbine building; the Unit 1 reactor building; and the equipment rooms, cable rooms and control rooms for all three units.
In the areas toured the following adverse conditions were observed: The inspector noted several plastic bags and cardboard drums a.
of trash in the turbine building basement.
Specific articles in the trash such as paper coveralls, cloth gloves and rubber / gloves indicated that the trash might contain radioactive contamination; however, with the exception of one J um, the bags and drums were not marked or otherwise identifiea as containing radioactive waste.
The inspector also noted several pieces of lumber and a large section of hose which were wrapped in plastic, but which bore no markings to indicate radioactive contamination.
These observations were discussed with the licensee and the materials were subsequently tagged with contamination labels.
Section M of Part I of the DPC System Health Physics Manual , ,' states in part that " low level solid wastes (contaminated trash) will be collected in labeled containers." Procedural Guides II-9 and II-12 contained in the same manual state that all items removed from a Radiation Control Zone (RC2) must be
' monitored and tagged appropriately, and that all low-activity-level solid waste (contaminated polyethelene, paper, rags, disposable protective clothing, glassware, etc.) must be disposed of only in labeled radioactive waste containers.
Failure to use labeled waste containers and failure to label contaminated materials outside an RCZ are contrary to these requirements of the System HP Manual and are in noncompliance with Technical Specification 6.4.1.
This is an Infraction.
b.
On four different occasions in a period of three days the inspector found doors between the turbine building basement and the auxiliary building standing open.
While these doors are not required for security purposes, it appears necessary that they remain closed, except during passage, to assure proper differential pressures and air flows between the two buildings.
At the times the doors were found open, the inspector observed notices posted on the doors which stated this require-ment over the Station Manager's signature.
Problems with ventilation control between the auxiliary and turbine buildings was originally identified as Unresolved i _ __ _ __.__._ . ~. _
. .
f ,); IE Rpt. Nos. 50-269/77-3,
- 50-270/77-3 and 50-287/77-3 I-9 It em 74-14/2 in Inspection Report No. 50-287/74-14.
To correct the problem the licensee performed design changes and modifica-tions on the Auxiliary Building Exhaust system.
During an inspection in January 1977 (Ref. IE Report No. 50-269/77-1) the inspectors determined that the licensee was in the process of accumulating operating data on the modified systec and the item was left open at that time pending completion of a review and evaluation of the data by the licensee to determine the adequacy of the modified system.
The operating data was not reviewed during this inspection; however, the inspector noted that Section 9.8.2.1 of the FSAR ' states that "the path of ventilating air in potentially radio-activly contaminated areas of the Auxiliary Building is from areas of low activity toward areas of progressively higher activity for ultimate discharge to the unit vent" and stated that unless the operating data demonstrates that this condition is met with the doors open, corrective actions must be taken to assure the doors remain closed.
Unresolved Item 74-14/2 remains open.
It appeared that many local indicating lights on Engineered c.
Safeguards, Control Rod Drive, Vital AC and other system cabinets in the equipment and cable rooms were burned out or missing.
Several protective lenses for these lights were broken or missing.
The licensee stated that corrective action would be taken and the inspector had no further questions.
d.
The inspector noted that one hydraulic snubber on one of the Unit 1 pressurizer relief valve lines appeared to be leaking.
It also appeared that the hydraulic fluid ~, the reservoir was contaminated with a foreign substance.
This was pointed out to the licensee and the snubber was replaced prior to startup of the unit.
The inspector had no further questions.
The time lines on recorder strip charts were misaligned by as e.
much as two hours on several recorders in the Unit 1 and 2 control room.
Two such recorders, the meteorological data and unit vent flow rate recorders on Unit 1, were misaligned while a gaseous waste release was in progress.
The inspector stated that while the misaligament has no safety significance, failure to maintain proper alignment can make reconstruction of a sequence of events very dif ficult and Station Directive 3.1.14 requires that the chart paper be reset if necessary between 0001 and 0100 hours each night.
The inspector had no further question, -
.'~) IE Rpt. Nos. 50-269/77-3, ' ' 50-270/77-3 and 50-287/77-3 11-1 ' . DETAILS II Prepared by: N.
- E ? 4/ ' ' W. H. Bradfoy'd, Reactor Inspector Date Nuclear Support Section Reactor Operations and Nuclear Support Branch Dates of Inspection: March 8-11, 1977 Reviewed by: ~ I[[M < 4/77 4~ R. D. Martin, Chief ' Date Nuclear Support Section ,. Reactor Operations and Nuclear Support Branch 1.
Individuals Contacted Duke Power Company 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. T. Bond - Technical Services Engineer W. G. Itin - Plant Safety Supervisor Other Operations Personnel 2.
Fire Prevention and Protection The licensee's fire prevention and protection program was reviewed to verify that controls had been established to include: (1) work control procedures had been develvped which control welding and burning; plant operations approves and controls all work done under the work control procedure; and that a fire watch for welding and burning operations is present with capability of communication with the control room; (2) Quality Assurance Surveillance had been established for surveillance during maintenance and modification of fire stop penetration seals; (3) administrative controls require quality assurance verification of penetration seal material; (4) fire emergency procedures had been prepared; and (5) fire brigade training and fire drills had been developed and implemented.
The program was evaluated for conformance to 10 CFR 50.59; Criteria III and IV of Appendix A of 10 CFR 50; Section 3.2.3 of ANSI N45.2; and Section F of Appendix A of Regulatory Guide 1.33.
The plant was also inspected to the commitments set forth in the Duke Power i i ! - -. - ..
. . . IE Rpt. Nos. 50-269/77-3, [) 50-270/77-3 and 50-287/77-3 II-2 . *' Company Response to Appendix A to Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants.
The following documents were reviewed: a.
Administrative Policy Manual 3.0 Station Activities Section 3.4 - Modifications Section 3.11 - Housekeeping 4.0 - Administrative Instructions Section 4.4 - Modifications s-b.
Station Directives i 2.5.1 - Training 3.1.1 - Tagging and Saf ety 3.1.28 - Out of normal check sheet 3.11.1 - Storage of Combustible Material 5.1.4 - Welding and Burning Safety Procedure i 5.3.1 - Fire Brigade Orgai.ization and Training 5.3.2 - Control of Repairs to Station Fire Stops ,) 5.3.3 - Reporting of Fire Protection Impairment c.
Emergency Procedures EP/1/1500/1 - Fire - During Post Irradiation Examination of Spent Fuel Elements EP/1800/10 - Natural Disaster EP/1800/12 - Losa of Control Room (Fire in Control Room) EP/1800/16 - Loss of Power d.
Periodic Tests PT/250/10A - Weekly Fire Protection Check PT/250/10C - Semi-annual Mulsifier System Check PT/250/12 - Annual Dry Pipe Valve Check PT/250/14 - Monthly Fire Protection Procedure Check PT/250/15 - Annual Fire Protection Equipment Inspection PT/250/18 - Reactor Building Fire Protection Equipment Check PT/250/19 - Annual Fire Hydrant and Post Medicator Check Pt/250/20 - Wet Sprinkler Alarm Flow Check e.
Maintenance Procedures MP/1500/3 - Repair or Replacement of Fire Stops and Vacuum Control.
! MP/3004/1 - Periodic Inspe tion of Vault Fire Protection.
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- . . ['} IE Rpt. Nos. 50-269/77-3, 50-270/77-3 and 50-287/77-3 II-3 ' ... f.
Response to Appendix A to Branch Technical Position APCSB.9.5-1 Guidclines for Fire Protection for Nuclear Power Plants.
g.
The latest fire inspection report from the fire insurance carrier.
Implementation of the program was inspected by a representative sample review of the following: (1) completed Welding and Burning Safety Procedures; (2) completed Periodic Tests; (3) completed Maintenance Procedure 1500/03, repair or Replacement of " Fire Stops" and Vacuum control; and (4) employee training records.
The inspector was able to verify on two occasions that the Welding and Burning Work Procedure was being implemented by the licensee.
Within the above areas inspected no discrepancies were identified.
The inspectcr made a inspection tour of the major portion of the plant including the Unit #1 containment building.
There were several items of weakness which were observed.
These items were pointed out to the station safety supervisor and were discussed with plant management at the exit interview.
They are as follows: , a.
There is a conflict between figure 1 on page D-2 of the Response to the Appendix A to Branch Technical Position APCSB 9.5-1 and MP/1500/3 - Repair or Replacement of " Fire Stops" and Vacuum Control.
The maintenance procedure states that there shall be a minimum thickness of 4 inches of Mone-Kote in the penetration.
Figure 1 of the Response to Appendix A indicates the penetration is to be completely , filled.
The penetrations appear to have a minimum of 4 inches of Mono-Kote.
In penetrations which contain a large number of cables, the penetrations may be inadequately sealed.
b.
There are sprinkler heads around protected equipment inside the plant which appear to be partially obstructed by piping.
, c.
Conduit for the automatic sprinkler system fire detectors in the vicinity of the main turbine oil reservoir appears to have been bent and restraightened.
These fire detectors may have suffered damage with respect to circuit continuity.
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_ _ _ _ _ _ _ _ _.- -__ _ _ _______, ' l - . I i - ) IE Rpt. Nos. 50-269/77-3, 50-270/77-3 and 50-287/77-3 II-4 ,,,, d.
The sprinkler system fire detectors inside the building and on the transformers are supposed to actuat< the deluge system on increase in temperature, but this is not tested or verified on a periodic basis.
The insurance carrier and the licensee have discussed this item recently.
e.
There was an accumulation of combustible material (wood) in the Unit 1 penetration area. This is a contaminated zone. The inspector noted that removal of this wood had started; the wood was being wrapped in plastic and sealed and moved into a clean f area.
f.
The labels on the deluge valves on Unit 1 were missing.
g.
There is no storage provided for new resins in the turbine building.
h.
Some fire hose was found disconnected, and, on some stations, the hose appeared not to be on the clips correctly. The fire hose was reconnected.
1.
The Unit 2 control room fire deluge valve station " power indicating light" was out.
The bulb was replaced and the power indicating light burned.
There was no indication as to how long this light had been burned out.
3.
Followup on Previously Identified Enforcement Items The inspector verified that adequate corrective action has been taken by the 10:ensee with regard to Enforcement Item I.3 in IE Inspection Report 50-269/76-12.
The fire extinguishers in Unit 1 containment were examined by the inspector; the extinguishers were properly located, had no pir.s pulled, and appeared fully operational.
In addition, a plant directive regarding this matter was issued by the Plant Safety Supervisor, who verified that all appropriate plant personnel were instructed in the correct use and proper location of the plant fire extinguishers.
This item is closed.
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