IR 05000315/1981018
| ML17319B131 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/01/1981 |
| From: | Dubry N, Hayes D, Swanson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17319B129 | List: |
| References | |
| TASK-1.C.6, TASK-2.E.1.2, TASK-2.K.1, TASK-TM 50-315-81-18, 50-316-81-21, IEB-78-12, IEB-78-12A, IEB-78-12B, IEB-79-06A, IEB-79-6A, NUDOCS 8201180506 | |
| Download: ML17319B131 (19) | |
Text
U.S.,REGULATORY COMMISSION
REGION III
Reports No. 50-315/81-18; 50-316/81-21 Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee:
American Electric Power Service Corporation Indiana and Michigan Power Company 2 Broadway New York~ NY 10004 Facility Name:
Donald C.
Cook Nuclear Power Plant, Units 1 and
Inspection At:
D.
C.
Cook Site Bridgman, MI Inspection Conducted:
July 1-31, 1981 Inspectors:
.E.
.
Swanson +
N. E. DuBr Approved By:
D.
W.
ayes, C
e Reactor Projects, Section 1B-Fr Ins ection Summar Ins ection on Jul 1-31 1981 (Re orts No. 50-315/81-18'0-316/81-21)
Areas Ins ected:
Routine, onsite regular and backshift inspection by the resident inspector.
Areas inspected included Operational Safety Verification, Monthly Surveillance observations, Monthly Maintenance observation, Followup on Previously Identified Findings, Licensee Event Reports, IE Bulletin 'Followup, TMI Task Action Plan, Inspection During Long Term Refueling Shutdown, Containment, Integrated Leak Rate Test and Independent Effort.
The inspection involved a total of 152 inspector-hours onsite by two NRC inspectors including 33 inspector-hours onsite during offshift hours.
Results:
Of the ten areas inspected no items of noncompliance or devia-tions were found in eight areas; three apparent items of noncompliance were identified in one area (Breach of containment integrity and failure to report Paragraph ll.d, Failure to review and approve procedure change Paragraph ll.d).
Two items of noncompliance were identified in the other area (Field design change made without proper review, approval and docu-mentation Paragraph 6 and failure to perform Type C leak test on one valve Paragraph 6).
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DETAILS Persons Contacted gf(D
"B PR J J Y(E R.
Shaller, Plant Manager Svensson, Assistant Plant Manager Townley, Assistant Plant Manager Keith, Operations Superintendent Stietzel, gA Supervisor Smarella, Technical Superintendent Dudding, Maintenance Superintendent The inspectors also conducted a number of interviews with supervisors, operators, technicians, administrative, security, and maintenance personnel during this inspection period.
+Denotes those present during exit interviews.
0 erational Safet Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the month of July 1981.
The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components.
Tours of Unit 1 Reactor Containment, the Auxiliary, and Turbine Building conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance.
The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security plan.
The inspector observed plant housekeeping/cleanliness conditions and verified implementation of radiation protection controls.
During the month of July 1981, the inspector walked down the accessible portions of the Essential Service Water, Component Cooling Water, and the Safety Injection systems to verify operability.
The inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barreling.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedures.
Monthl Surveillance Observation The inspector observed technical specifications required surveillance testing on the Protection Sets, the Seismic Monitor, and Calibration of Detector R-28 and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test
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results conformed with technical specifications and procedure require-ments and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The inspector also witnessed portions of the following test activities:
Unit 1 1 THP 4030 STP.049 1 THP 4030 STP.050 12 THP 6010 RAD 592 12 THP SP RFC 2522A
"Containment Pressure Protection Set lV"
"Steam Pressure Protection Set III"
"New Tube Calibration/Tube Requalification"
"Test of the Distributed Ignition System, both Train "A" and Train "B" (Unique to ice condenser containments)
Unit 2
OHP 4021.002.001
"Filling and Venting the Reactor Coolant System" Monthl Maintenance Observation Station maintenance activities of safety related systems and components listed below were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications.
The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qual-ified personnel; parts and materials used were properly certified; radiological controls 'were implemented; and, fire prevention controls were implemented.
Work requests were reviewed to determine status of outstanding jobs and to assure that, priority is assigned to safety related equipment maintenance which may affect, system performance, The following maintenance activities were observed/reviewed:
Unit 1 Replacement/Repair of the "N" SI Pump.
Replacement of R-28 "Liquid Monitor Detector Tube."
Unit 2 Replacement of the
"RTD Manifold Isolation Valve."
Repair of "Lower Containment Ventilation Unit."
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Following completion of maintenance on the Unit 1 "N" SIPump and Unit 2 RID Manifold Isolation Valve, the inspector verified that these systems had been returned to service properly.
5.
Followu on Previous Ins ection Findin s (Closed)
Noncompliance (50-315/80-19-01):
Failure to establish fire watch.
The inspector reviewed the licensee's corrective actions taken to prevent recurrence.
It was found that on February 15, 1981 the event recurred (Unit 1 LER 81-003/03L-0).
This was another event in which the fire watch requirement of Technical Specification 3.7.10 Action Statement was not complied with.
Corrective actions taken in response to these events included clarification of fire watch require-ments to construction personnel by memo, changes to the Plant Manager Instruction 2270 "Fire Protection and Safety Equipment" and a change to the maintenance procedure for installation of fire seals.
6.
Licensee Event Re orts Followu Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to pre-vent recurrence had been accomplished in accordance with technical specifications.
Unit 1 RO 81-019/03L-0 Undocumented, Untested Containment Penetration The above event involved the discovery of an undocumented installa-tion of a
~q inch valve and pipe section which penetrated containment.
The valve and associated piping section were installed on penetration CPN-30.
Cognizant site personnel could not establish that a design review had been performed and approved.
This constitutes a violation of the requirements of 10 CFR 50, Appendix B, Criteria III, and XVII.
(315/81-18-01)
Furthermore the valve had apparently never been tested as required by 10 CFR 50, Appendix J "Type C Tests."
(315/81-18-02)
RO 81-020/01T-0 Unit 2 Pinhole leak in NESW Cooler (IEB 80-24)
RO 81-002/03L-0 East MDAFP Spool Piece Replaced for Cracked Duplex Strainer RO 81-028/01T-0 NESW Valve Body to Bonnet Leak (IEB 80-24)
IE Bulletin Followu For the IE Bulletins listed below the inspector verified that, the Bulletin was received by licensee management.
and reviewed for its applicability to the facility. If the Bulletin was applicable the inspector verified that the written response was within the time period stated in the Bulletin, that the written response included the information required to be reported, that the written response included adequate corrective action commitments based on information presented in the Bulletin and the licensee's response, that the licensee management forwarded copies of the written response to the appropriate onsite management representatives, that information dis-cussed in the licensee's written response was accurate, and that corrective action taken by the licensee was as described in the written response.
78-12
"Atypical Meld Material in Reactor Pressure Vessel Welds."78-12A 78-12B The Safety Evaluation Report by the office of.Nuclear Reactor Regula-tion (June 4,
1981) concluded that the licensee's response to the following bulletins was acceptable.
Therefore they are being closed in this report.79-06A
"Review of Operational Errors and System Misalignments Identified during the TMI Incident."79-06A, Revision
TMI Task Action Plan Followu Item II.E.1.2 Auxilliary Feedwater Initiation and Flow:
As discussed in a June 16, 1981, Safety Evaluation Report from NRR Division of Licensing, Operating 'Reactors Branch 1, Items 1.b and 2.b had been found acceptable pending verification of the installa-tion of the enviromentally qualified auxilliary feedwater (AFW)
flow transmitters.
The inspector verified that the Barton AFW flow transmitters were replaced with qualified Foxboro units under design change RFC-DCC-12-2447 during the recent refueling outages.
The inspector verified that the change received proper review and approvals, that applicable drawings were revised, and that the trans-mitters were properly procured and qualified to IEEE 323-1974. It was noted by the inspector that the wide range steam generator level remains the control grade backup to the narrow range steam generator level and the new auxiliary feed flow transmitters.
This completes all requested actions on this item.
Item XI.K.1.3 IE Bulletins79-06A and 79-06A, Revision 1:
The Safety Evaluation Report by the Office of Nuclear Reactor Regula-tion dated June 4, 1981 concludes that the licensee has acceptably responded to these Bulletins.
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9.
Ins ection Durin I,on Term Shutdown The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the month of July.
The inspector verified surveillance tests required during the shutdown were accomplished, reviewed tagout records, and verified applicability of containment integrety.
Tours of Unit 1 Containment, the combined Auxiliary and Turbine building accessible areas, including exterior areas were made to make independent as-sessments of equipment conditions, plant conditions, radiological controls, safety, and adherence to regulatory requirements and to verify that maintenance requests had been initiated for equipment in need of maintenance.
The inspector observed plant housekeeping/
, cleanliness conditions, including potential fire hazards, and verified implementation of radiation protection controls.
The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security plan.
The inspector reviewed the licensee's jumper/ bypass controls to verify there were no conflicts with technical specifications and verified the implementation of radioactive waste system controls.
The inspector witnessed portions of the radioactive waste systems controls associated with radwaste shipments.
10.
Containment Inte rated leak Rate Test The Inspector witnessed portions of the CILRT and verified that the appropriate revision of Procedure 12 THP 4030 STP.202 was in use, test prerequisites were met, and proper plant systems were in service.
The inspector made a visual tour of the Unit 1 upper and lower con-tainment following the test.
It was noted that the lower containment pipe tunnel had a number of out of service lights and there was an oil slick on the deck.
This will be looked at for improvement during the next tour.
The inspector reviewed interim and final licensee data and forwarded raw data to the regional USNRC office for agreement computations.
(Reference:
IE Inspection Report No. 50-315/81-15.)
ll.
Inde endent Effort a
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Unit 2 Maintenance Shutdown The inspector observed the licensee's action during the main-tenance outage from July 3-11, 1981.
During this period the licens'ee identified and corrected a containment ventilation damper, misalignment, repaired/replaced the RTD manifold isola-tion valve, and investigated the excessive temperature in the Reactor coolant pump.
The inspector followed closely the effort on the above mentioned valve which included the review of clearances, procedures, parts qualification and restoration to service (including 2 MHP 5050 SPC.005,
"Hydrostatic Test Procedure" ).
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It was noted that during, this outage, GRID IV [Control Room Instrument (Power) Distribution] was lost due to the same
'roblem tha't caused the loss of GRID II. 'ad'he plant been at power this loss would have caused a trip.
(Reference:
IE Inspection Reports No. 50-315/81-13 and No. 50-316/81-16)
Licensee representatives stated that a design change has been initiated to replace the faulty component (capacitor) with one better qualified for this particular service and environment.
Personnel Contamination The inspector followed up on several minor personnel contamination incidents and verified that the licensee's actions were in accord-ance with established requirements.
I,eaka e Inside Unit 1 Containment The inspector reviewed the licensee's actions and the scenario of events which accompanied the inadvertent addition of Demineralized Water to the lower Containment volume.
The event was apparently due to lack of coordinated control when returning a newly modified system to service.
Violation of Containment Inte rit While conducting Type A Containment Leak Rate Test per
'<<12 THP 4030 STP.202 (Unit 2 in Mode 5)
a Performance Techni-cian hooked up a rotameter to a t'ee connection common to transmitter PPA-310.
During the test a Control Room Operator expressed concern that his recorder driven by transmitter PPA-311 was pegged high.
He was told it would not damage the instrument by a CSI Techrician, but that they would isolate it and bleed off the pressure anyway.
When CGI and Performance got together in the Containment Annulus, Performance realized that they were hooked up to the wrong point and moved their instrument.
In his haste he did not remove the connector and replace the plug.
After the test was over, all was forgotten and CRI put the PPA-310 back in service leaving an open path to containment (.140-inch I.D. of connector).
When the in-strument didn't track properly, Job Order 28997 (May ll, 1981)
was initiated.
.The plug was found missing on May 12 at 2230 and reinstalled.
At this time the unit was in Mode 3 having entered Mode 4 at 1045 on May 10, 1981 and Mode 3 on 0630 on May 12, 1981.
(1)
Containment integrity was not maintained while the reactor was in Modes 3 and 4 for the periods of time stated above.
This is contrary to the requirements of Technical Specifi-cation (TS) 3.6.1.1 which requires that in Modes 1, 2,
and 4>'hat containment integrity be maintained and TS 3.0.4 which requires that entry into an operational mode not be made unless the conditions of the Limiting Condition for Operation are met.
(316/81-21-01)
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Similar events involving failure to ensure operability of safety systems following surveillance testing or maintenance were dis-cussed with the licensee during an, enforcement conference on January 13, 1981.
(IE Inspection Reports No. 50-315/81-02; 50-316/81-02).
These events concerned improper positioning of valves on one containment spray system following surveillance testing and possible inoperability of one residual heat removal system for two to three months following maintenance.
"=Power operations, startup, hot standby and hot shutdown respectively.
(2)
Technical Specification 6.9.1.8 Prompt Reporting applies when:
"Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR."
CFR 50.72 notification of significant event applies when:
"Personnel error or procedural inadequacy which, during normal operations, anticipated operational occurrences, or accident conditions, pxevents or could prevent, by itself, the fulfill-ment of the safety function of those structures, systems, and components important to safety that are needed to shutdown the reactor safely and maintain it in a safe shutdown condition, or remove residual heat following reactor shutdown, or limit the release of radioactive material to acceptable levels or reduce the potential for such release."
Contrary to TS 6.9.1.8 and
CFR 50.72 a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report of the event described above was not made.to the NRC.
Instead, a 30 day report was made.
(316/81-21-02)
(3)
Technical Specification 6.5.1.6 requires that the Plant Nuclear Safety Review Committee (PNSRC)
be responsible for review of all procedures required by Technical Specifica-tion 6.8 and changes thereto.
Technical Specification 6.8 includes requirements to have surveillance test procedures.
Containment pressure transmitters PPA-310 and PPA-311 were valved out during the Intergrated leak Rate Test.
The ap-plicable procedure (12 THP 4030 STP:202, Revision 3) does not require valving these instruments out, thus the independent decision by the instrument technician and control room operator to valve these instruments out was 'improper and constituted a procedure change not authorized by plant management or reviewed by the PSNRC as required by TS 6.5.1.6.
This is an item of noncompliance.
(316/81-21-03).
The question as to whether the requirements of 10 CFR 50 Appendix J have been met when containment instrumentation is valved out during the containment ILRT has previously been referred to the Office of Nuclear Reactor Regulations for resolution.
(315/81-21-03; 316/81-21-04).
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(4)
None of the above was complied with, Further, the investiga-tion inspired several concerns:
(a)
The program for reduction of personnel and procedural errors unveiled at the April 15, 1980 Management meeting appears to be lacking effectiveness.
Technicians are routinely worked 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> days during outages, and when they are performing critical path testing they, are, pushed harder.
Corrective actions first consisted of instructions to Performance Engineers reminding them to replace plugs and caps removed for testing and a procedure change which adds precautions to "Ensure" that pipe caps and plugs are replaced.
The procedural sign off was as follows:
"5.6.15 the test connections on PPA-310 and 312, PPP-302, 303 and PPX-301 have been isolated."
(b)
The inspector's further expressed that the corrective ac-tion would not be effective given the same circumstances:
The CSI Technician could still isolate an instrument independent of the test procedure.
The procedure does not specifically call for realign-ment of these associated valves.
Double verification is not done for systems important to safety not governed by clearance permit.
(c)
TMI Task Action Plan I.C.6 "Verifying correct performance of operating activities."
Apparently the alignment of pressure transmitters aftertesting or surveillance is not addressed.
Unless the instrument requires a Clearence Permit, even ESP actuation instruments would not receive alignment verification.
(5)
Corrective actions taken subsequent to a periodic exit inter-view meeting with the inspector resulted in the following corrective actions (Supplemental LER 81-019/01X-1):
Integrated Leak Rate Test points identified with metal tags,'rocedure modified to reflect this identification and proper restoration.
Procedures to check all instrument and sampling lines to insure valves affecting containment isolation are positioned correctly prior to plant startup from cold shutdown.
IE Inspection Reports No. 50-315/80-01; 50-316/80-0 p k
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Procedure for double verification of all safety related instrumentation valve positions after repairs and calibra-tion (incorporate into applicable instrument maintenance and surveillance procedures).
These actions will be verified by the inspector at a future date as part of the routine followup of items of noncompliance.
12.
Exit Interview The inspectors met with Licensee representatives (denoted in Para-graph 1) throughout the month and summarized the scope and findings of the inspection activities.
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