IR 05000289/1982002
| ML20054D263 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/01/1982 |
| From: | Fasano A, Finkel A, Haverkamp D, Richards S, Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20054D254 | List: |
| References | |
| 50-289-82-02, 50-289-82-2, NUDOCS 8204220491 | |
| Download: ML20054D263 (16) | |
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U.S. NUCLEAR REGULATORY COMMISSION 50289-810907 50289-810117 Region I 50289-811008 50289-811019 Report No.
50-289/82-02 Docket No.
5_0_-289 License No.
DPR-50 Priority Category C
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Licensee :
G_PU_ Nuclear Corporation P.O. Box 311 Middletown, Pennsylvania 17057 Facility Name: Three Mile Island Nuclear Station, Unit 1 Inspection at: Middletown, Pennsylvania inspection conducted:
February 17,1982 - March 5,1982 Inspectors:
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D. Haverkamp, Senior fpsident Inspector (TMI-1)
date signed l
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Yf l8 L A. Finkel,IEA L
ReactorInsiiectof da~te signed N/
4/ ele x-S. Richards, Reactor Ippectpr date signed
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F. Yo
, ResQent Inspector (TMI-l)
date s'i~gned-Y
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[Im /f/f.l Approved by:
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Fasano, Chief, Three Mile Island Section datedisned Projects Branch No. 2 Inspection Summary:
Inspection on February 17,1982 - March 5,1982 (Report Number 50-289/82-02)
Areas Inspected: Routine safety inspection by resident and regional based inspectors (165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br />) of licensee action on previous inspection findings; plant operations during long term shutdown including facility tours and log and record reviews; replacement of industrial cooler system valve RB-V7; steam generator tube degradation; IE bulletin and circular followup; and licensee event reports - in-office review.
Resul ts : No items of noncompliance were identified in six areas and one item of noncompliance was identified and corrected in one area (failure to follow a steam generator repair procedure, paragraph 5.c).
BE04220491 820407 PDR A90CK 05000289 G
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Detail s 1. -
Persons Contacted
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GPU Nuclear Corporation R. Barley, Lead Mechanical Engineer TMI-l
- M. Beers, Engineering Assistant Senior II, Nuclear Assurance J. Colitz, Plant Engineering Director TMI-l C. Davis, Modification Control Coordinator
- J. Fritzen, Technical Functions TMI-l Site Supervisor
~T. Hawkins, Manager TMI-l Startup and Test, Technical Functions
- W. Heysek, Supervisor Site Quality Assurance (QA) Audit, Nuclear
' Assurance H. Hukill, Vice President and Director TMI-l
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- D. Osterhout, Media Representative, Communications M. Ross, Manager Plant Operations TMI-1
- C. Smyth, Supervisor, TMI-l Licensing, Technical Functions C. Stephenson, Nuclear Licensing Engineer, Technical Functions R. Toole, Operations and Maintenance Director TMI-l E. Wallace, Manager PWR Licensing, Technical Functions
The inspector also interviewed several other licensee employees during the inspection.
They included control room operators, maintenance personnel, engineering staff personnel and general
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of fice personnel.
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2.
Licensee Action on Previous Inspection Findings (Closed) Inspector Followup Item 289/79-BU-1B:
IE Bulletin 79-01B,
" Environmental Qualification of Class IE Equipment".
The lead responsibility for evaluation of this item has been assigned to the t
Office of Nuclear Reactor Regulation, and licensee actions are being tracked under Multiplant Action Number B-60.
All plants must meet the requirements of Commission Memorandum and Order CLI 80-21 concerning 10 CFR 50, Appendix A, General Design Criterion 4.
Licensee submittals are under review by Franklin Research Center, and a proposed rule has been issued for public comment (SECY 6038).
Final safety evaluation reports must 'be issued for all plants.
The schedule for completion of licensee actions, contingent on the proposed rule, is 1985 Al though no further inspections are scheduled for IE' Bulletin 79-01B, applicable verification inspections are expected to be performed by NRC Region I subsequent to issuance of the final rule and safety evaluation report concerning this matter.
(Closed) Unresolved Item 289/79-01-01 : Complete field verification of electrical components during 1979 refueling outage.
This item was j
related to licensee actions taken in response to IE Circular 78-08,
" Environmental Qualification of Safety-Related Electrical Equipment at i
Nuclear Power Plants".
The circular requirements were superseded
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-3-by IE Bulletin 79-01B, " Environmental Qualification of Class IE Equi pment".
An installation verification of components of the r
Reactor Protection System and the Makeup and Purification System
was conducted during Region I Inspection 50-289/80-20, as described
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in Inspection Report 50-289/80-20, paragraph 7.
(See Inspector Followup Item 289/79-BU-1B for additional information concerning environmental qualification of Class IE equipment.)
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(Closed) Inspector Followup Item 289/79-20-01 : Engineering evalu-
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ation not performed for a jumper in effect for greater than 12 months.
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Administrative Procedure (AP) 1013, " Bypass of Safety Functions and
Jumper Control," requires an engineering evaluation ~ be performed on any jumper, lifted lead, or temporary mechanical modification in
effect greater than 12 months.
The inspector reviewed the jumper,
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lifted lead, and temporary mechanical modification log and noted
five lifted leads and one temporary mechanical modification which j
had been in effect greater than 12 months but did not have complete
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engineering evaluations.
This observation was discussed with licensee personnel who provided objective evidence that the discrepancies had been previously identified during licensee review and that
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engineering evaluations had been developed for the discrepancies, although the engineering evaluations had not yet been approved for entry into the log.
The inspector considered licensee implementation
of AP 1013 engineering evaluations to be acceptable and had no
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further questions concerning this item.
t (Closed) Unresolved Item 289/80-04-02: Fabrication of butt splices and review of engineering change modification (ECM), fire hazards e
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analysis (FHA) and training.
The inspector reviewed the engineering
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documentation, ECM-138, Revision 3, that defined the method of
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fabrication for butt splices as well as the quality control require-ments for inspection.
The inspector also reviewed the documentation
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dealing with the increase of flamable material associated with the
addition of butt splice material.
Licensee engineering personnel i
had determined that the addition of butt splice material would not affect the fire hazard analysis.
The quality control inspection criteria verified the inspection check points which were required
by engineering and referenced in Raychem Installation Guide 1050 and Inspection Guide 2050. Quality control inspection also verified that the training given to contracted construction workers who
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installed the butt splices included the requirements of Raychem 1050 and 2050 guide documents.
The inspector had no further questions i
concerning this matter.
(Closed) Inspector Followup Items 289/80-04-04 and 289/80-05-07 :
Certificates of compliance for butt splice task.
The licensee had
verified that material obtained by Purchase Orders (P.O.'s) 86082,
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86097, and 86522 complied with the documented engineering require-ments.
The inspector reviewed licensee engineering correspondence that stated the test data associated with these P.O.'s had been
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reviewed and approved and complied with the engineering requirements.
s The inspector had no further questions concerning these items.
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-4-(Closed) Inspector Followup Item 289/80-19-01 :
Implement Inde-pendent Onsite Safety Review Committee (IOSRC).
The licensee submitted a functional description of the IOSRC and its composition in Amendment No. 20, dated August 8,1980, to the TMI-l Restart Report.
The licensee subsequently submitted Technical Specification Change Request (TSCR) No.1008, which proposed a revised process for interdisciplinary reviews and independent safety reviews using qualified individuals / groups ~ rather than committees.
TSCR No. 1008 included a proposed Independent Onsite Safety Review Group (IOSRG),
which will satisfy the requirements for an independent, full-time,
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safety engineering staff to be located onsite.
The change request is currently under review by NRR, and a Technical Specification Amendment (TSA) is expected to be issued in April 1982, which will provide a four month implementation period.
The following aspects of the IOSRG will be reviewed during a subsequent NRC Region I inspection after TSA issuance: qualification of members; review charter and procedures; and, independence from the operating organization (289/82-02-01).
(Closed) Inspector Followup Item 289/80-19-09: Submit Technical Specification Change Request (TSCR) concerning interfacing among
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the various groups who function to review plant operations.
TSCR No.100B, dated November 25, 1981, proposed a complete revision of the administrative sections of the Technical Specifications, Appendix A and Appendix B.
(See Inspector Followup Item 289/80-19-01 for a further description of TSCR No.100B.)
The following aspects of individual / group review and safety review of activities identified in Technical Specification Sections 6.5.1 and 6.5.2 will be verified during a subsequent NRC Region I inspection after TSA issuance:
qualification of reviewers; GPU Nuclear review and approval matrix; review charter and procedures; and, independence of reviewers from the responsible technical review individual / group (289/82-02-02).
(Closed) Inspector Followup Item 289/80-19-10:
Provide more detailed resumes of plant / staff personnel to permit evaluation of specific work experience.
The licensee submitted Amendment No. 20, dated August 8,1980, to the TMI-1 Restart Report, which provided a revised Chapter 5 describing the organization of GPU Nuclear Corporation.
The revised Chapter 5 included detailed resumes which permitted NRC evaluation of specific work experience.
The inspector had no further questions concerning thi's item.
(Closed) Unresolved Item 289/80-20-01 : Licensee revising procedure to include all 10 CFR Part 21 requirements formerly found in several di fferent procedures.
The licensee has prepared and issued Licensing Procedure (LP)-007, dated December 31, 1981, entitled " Management of Potentially Reportable Items".
The inspector verified that LP-007 included all 10 CFR Part 21 reporting requirements and established within GPU Nuclear a standardized method for the identification and evaluation of potentially reportable items and the manner in which tney will be processed.
The inspector had no further questions concerning this item.
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(Closed) Inspector Followup Item 289/80-20-02: Missing, broken, and mixed screws on electrical penetration and transmitter splice boxes.
Job ticket No. 3870 was issued by the licensee on August 21, 1980, to correct the subject equipment discrepancies.
The inspector verified that the electrical equipment discrepancies have been corrected and that the equipment was installed in accordance with installation drawing requirements.
The inspector had no further questions concerning this item.
(Closed) Inspector Followup Item 289/80-20-04: Pressure Transmitter RC 3B-PT2 electronic housing not mounted firmly to sensor module.
Job ticket No. 3971 was issued by the licensee on August 21, 1980, to correct the subject equipment discrepancy. The inspector verified that the electronic housing had been secured to the sensor module, as required by the installation drawing requirements, and had no further questions concerning this item.
(Closed) Inspector F llowup Item 289/80-30-02:
Completion of o
corrective action for LER 80-006-OlT-0, " Excessive leakage past containment isolation valve RB-V7 during local leak rate testing".
The licensee replaced containment isolation valve RB-V7 with an improved type of gate valve.
Review of leak rate test data sub-sequent to the valve replacement indicated that the valve leakage is now within allowable limits.
Further review of the modification is discussed in paragraph 3 of this report.
(Closed) Unresolved Item 289/81-05-01 : Valves marked "NOT for N3 Installation" in N3 System.
Licensee letter dated April 13, 1981 (TMI-I/E 2260) stated that Purchase Order (P.O.) 72866 and Generic Procedure (GP)-1009 required the valves to be applicable for N3 use. The letter also stated that the material test reports and the certification of compliance that were supplied with the valves met the requirements of P.O. 72866.
The inspector reviewed the P.O.
data supplied with the valves and concurred with the information provided by the licensee in their letter TMI-I/E 2260.
The inspector had no further questions concerning this item.
(Closed) Unresolved Item 289/81-05-02:
Installation of differential pressure switch DPS-914 does not agree with as-built drawings.
The licensee issued design change notice DT 0168, dated April 13, 1981, to correct drawing B-308-856, Revision l A-1, by reflecting the as installed configuration.
The inspector verified that the design change notice reflects the present installation and had no further questions concerning this item.
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Replacement of Industrial Cooler System Valve RB-V7 a.
Modification Description The inspector reviewed the licensee's implementation of Task PM-1, modification of valve RB-V7.
The modification was -
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performed as corrective action for LER 80-006-01 T-0, which identified excessive leakage past the valve during containment leak rate testing.
Task PM-1 replaced the originally installed air-operated valve with a motor-operated valve.
The plant configuration changes associated with Task PM-1
included removal of the existing valve and air piston operator, installation of a new valve with electric motor operator, and the installation of conduit / conduit supports and cable to service the electric motor operator.
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Documentation Reviewed /0bservations_
The inspector reviewed the following documentation related to tisk PM-1.
Engineering Change Modification (ECM) 11 01-1-1, Revision
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(Rev.) 0-0A, " Electrical System Modification for RB-V7 Valve Operator Replacement" t:CM 1101-1-2, Rev. O, "RB-V7 Replacement"
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QC inspection reports
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Cable pull / termination sheets Work authorization notices
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Field questionaires
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Design change notices for design verification
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Weld History Records
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Firewall Breeching Notification
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Installation Records Requirements Reactor Building Air. Cooling Coils Return Isolation Valve
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RB-V7 Control Logic Test IB-E.S. MCC, Unit il D (PM-1)
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SP 1303-il.18E Data Sheet (Local leak rate test data)
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Installation drawings and "as-built" piping and instru-
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mentation drawings
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GPUNC Modification Turnover Package Nos. ECM S-1-2 and ECM S-1-1 accepted November 18, 1981.
In addition to the above documentation review, the inspector observed the installed equipment and verified the component location and installation was as described in applicable modifi-cation documentation.
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Findings (1 ) Based on the modification documents reviewed by the inspector and the observation of installed equipment and components, the inspector determined that Task PM-1 was satisfactorily completed in conformance with the above referenced commitments and requirements.
(2) There were no significant incomplete work list items related to Task PM-1.
4.
Plant Operations During Long Term Shutdown a.
Plant __ Status During Inspection Period The plant has remained in the cold shutdown condition with Reactor Coolant System (RCS) temperature less than 200 F per NRC order of August 9,1979.
At the beginning of the inspection period on February 17, 1982, the RCS was at 100 F and 0 psig with cooling to the reactor supplied by the
"A" Decay Heat Removal (DHR) Loop.
The RCS was partially drained for eddy current testing and preparation for tube plugging. Both Once Through Steam Generators were partially drained on the secondary side for tube plugging.
During the course of the inspection the following major plant changes or evolutions occurred.
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February 17, 1982 to February 21, 1982: Nineteen tubes, of which portions had been removed for metallurgical analysis, were permanently plugged
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February 20, 1980: Shi fted from " A" DHR Loop to "B" DHR Loop in order to accomplish remote oil addition modification to " A" DHR pump
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February 22, 1982: Placed both OTSG's in wet lay up on the secondary side
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Febraury 28, 1982: Placed Borated Water Storage Tank (BWST) on cleanup to remove sulfur spacies contaminants At the close of this inspection period the RCS was at 107 F and 0 psig with cooling t1 the reactor core from
"B" DHR Loop.
The RCS was partially drained for eddy current testing.
Both OTSG's were in full wet lay u.
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Plant Logs and Operating Records
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The inspector reviewed selected portions of the following plant procedures to determine the licensee established require-ments in this area in preparation for a review of selected logs and records.
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Administrative Procedure (AP) 1007, " Control-of Records,"
Revision 4
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AP 1012, " Shift Relief and Log Entries," Revision 14
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AP 1033, " Operating Memos and Standing Orders," Revision 2 AP 1037, " Control of Caution and [Do Not Operate] DN0
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Tags," Revision 2
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AP 1044, " Event Review and Reporting Requirements,"
Revision 2 The inspector reviewed the following plant logs and operating _
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Shift Foreman Log and Control Room Log Book
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Shift Turnover Checklist
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Temporary Change Notice (TCN) Log Book Active Tagging Application Book
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Locked Valve Log
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Night Order Book
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Do Not Operate and Caution Tag Log
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Plant logs and operating records were reviewed to verify the following items.
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Log keeping practices and log book reviews are conducted in accordance with established administrative controls.
Log entries involving abnormal conditions provide sufficient
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detail to communicate equipment status, lockout status, correction and restoration.
Operating orders do not conflict with Technical Specifications
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(TS) requirements.
Tagging operations are conducted in conformance with
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established administrative controls.
l No items. of noncompliance were identified.
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Facility Tours During the course of the inspection, the inspector conducted tours of the following plant areas.
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Control Room (daily)
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Auxiliary Buidling (February 23)
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Intermediate Building (February 18)
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Vital Switchgear Rooms (February 18)
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Diesel Generator Building (February 25)
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Yard Area (February 18 and March 3)
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Site Perimeter (March 3)
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Reactor Building (February 17, 19, 26 and March 5)
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The following observations / discussions / determinations were made.
Control room annunciators: Selected lighted annunciators i
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(BWST Hi/ Low Alarm, and B Diesel Trouble Alarm) were discussed with control room operators to verify that the reason for annunciation was understood and corrective action, if required, was being taken.
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Control room manning: By frequent observation during the i
inspection, the inspector verifled that control room manning requirements of 10 CFR 50.54(k) and the Technical Specifications were being met.
In addition, the inspector observed shift turnovers to verify that continuity of system status was maintained.
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Technical Specifications: Through log review and direct
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observation during tours, the inspector verified compliance with Technical Specification LCO's associated with Decay
Heat Removal System requirements during shutdown plant
operation.
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Plant housekeeping conditions:
Unsatisfactory plant housekeeping conditions observed by the inspector had been identified previously by station personnel and corrective action had been initiated as necessary,
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Monitoring instrumentation: The inspector verified that
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selected instruments (Decay Heat Removal flow, and RCS pressure) were functional and indicated that parameters were within Technical Specification limits.
Valve positions: The inspector verified that selected
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valves were in the position or condition required by Technical Specifications for the applicable plant mode.
This verification included control board indication and field observation of valv.e positions for the Industrial Cooler System and Emergency Feedwater System.
Fluid leaks:
Fluid leaks observed by the inspector had
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been identified previously by station personnel and corrective action had been initiated as necessary.
Piping snubbers / restraints: Selected pipe hangers and
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seismic restraints were observed and no adverse conditions were noted.
Equipment tagging: The inspector selected plant components
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(Condensate System) for which valid tagging requests were in effect and verified that the tags were in place and that the equipment was in the condition specified.
Security:
During the course of this inspection, observations
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relative to protected and vital area security requirements were made, including access controls, boundary integrity, search, escort, and badging. No adverse conditions were noted.
Licensee meetings:
The inspector frequently attended the
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Plan-of-the-Day (P0D) morning meetings and the daily
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status afternoon meetings held by licensee management and supervisory personnel.
The inspector observed the meetings to assess licensee evaluation of' plant conditions, status and problems and to review the licensee's plans for conducting certain major plant operations and maintenance activities. The inspector also attended bi-we-ily project status meetings to assess licensee progress and difficulties related to plant modifications required for restart.
Acceptance criteria for the above items included inspector judgement and requirements of 10 CFR 50.54(k), Regulatory Guide 1.114, Technical Specifications, and the following procedures.
AP 1002, " Rules for the Protection of Employees Working
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on Electrical and Mechanical Apparatus," Revision 22 AP 1008, " Good Housekeeping," Revision 7
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' AP 1037, " Control of Caution and DN0 Tags," Revision 2
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No items of noncompliance were identified.
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Steam Generator Tube Degradat_ ion a.
Background Repressurization of Reactor Coolant System (RCS) in November 1981 revealed degradation of tubes in both Once Through. Steam Generators (OTSG's).
In December 1981 a task force was established to coordinate and direct all actions regarding the investigation and repairs to the. steam generator tube leaks.
During the months of December 1981 and January 1982, the licensee removed sections of 19 tubes for metalurgical analysis and conducted an extensive Eddy Current test (see NRC Region I Inspection Reports 50-289/81-32 and 50-289/82-01).
Metalurgical analysis has established that circumferential cracks, all initiated from the tube inside surface (primary side), have occurred mainly in the top portion of the tubes.
Throughwall attacks and less penetrating cracks were observed with classic intergranular attack (IGA) and intergranular cracking (IGC) morphology.
The presence of sul fur on the inside surface of the tube has been clearly established by microstructural analysis conducted both at B&W and Battelle La boratories.
Furthermore, the analyses indicated that the sulfur present on the tube surface is in the reduced state but the specific form is not known at this time.
Based on the fact that the OTSG tubing is in a severely sensitized condition, it has been concluded that the failure was caused by sulfur-induced IGA, and in many cases this was enhanced and accelerated
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by stresses and manifested itself in the form of IGC.
The licensee has completed the first portion of their phase I plugging which included 19 tubes from which sections were removed for laboratory analysis.
These tubes were plugged by B&W using permanently welded plugs.
The licensee has contracted Westinghouse to complete the second portion of their phase I tube plugging using rolled plugs.
This portion of the OTSG tube repair program consists of plugging tubes that leaked during the Nitrogen (N ) bubble testing and/or tubes with
known defects.
The task was scheduled to start March 5,1982, and was expected to be completed in two to three weeks.
In addition to tube plugging, the licensee is conducting more eddy current testing. Approximately 100 tubes in each OTSG will be retested using a standard eddy current probe to determine i f there is coni' 3ing degradation of the OTSG tubes.
Eddy current testing at.1 also be performed on approximately 125 tubes per OTSG in the lower tube sheet to ensure that degradation is not occurring in that area.
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-12-In order to ascertain the extent of the degradation to the RCS and RCS boundary, the licensee is conducting a review of the entire primary system.
The review will include pulling the reactor vessel head and visually inspecting certain internal parts.
The components to be inspected are being determined by B&W at this time.
Preliminary planning has scheduled head removal co commence the week of April 5,1982 Testing and analysis of the method for repair of the tubes is continuing.
The licensee is planning to present information available and preliminary proposal for OTSG repairs to the NRC staff during the month of April.
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Review Due to the significance and severity of the OTSG tube degra-dation, the inspector has been closely monitoring the licensee's 0TSG leakage evaluation program and has initiated a review to veri fy the below listed items.
Corrective action is appropriate to correct the cause of
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the tube degradation Responsibility has been assigned to ensure proper management
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attention in correcting the problem The leakage / tube failures did not cause violation of
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Technical Specifications, license conditions or regulatory i
requirements
Information related to the event submitted to NRC is I
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accurate Stability of plant conditions, including provisions for
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Proper review and implementation of radiological controls including ALARA (As Low As Reasonably Achievable) considerations Selected sections of the following documents were reviewed.
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Radiation Work Permits (RWP's) associated with entries into the OTSG's ALARA review conducted for 0TSG tube removal and eddy
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current testing of 0TSG's Westinghouse Electric Corporation, Nuclear Service
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Division, Field Service Procedure (FSP) FP-002-82, " Roll Plug Installation - Hands On - GPU"
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and Qualification Test" GPUN Technical Specification (TSP) 1101-12-029, " Technical
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Specification for OTSG Phase I Tube Plugging" Babcock & Wilcox Operating Specification (0S) 64-1131209-00,
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"Once Through Steam Generator Temporary Tube Plugging and Removal" Babcox & Wilcox OS 64-1001546-04, "Once Through Steam
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Generator Tube Plugging and Stabilizing" GPU Maintenance Job Ticket C-7849, TMI procedure for
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hydraulically expanding tube segments in the tube sheet.
AP 1106-16 "0TSG Secondary Fill, Drain, and Layup,"
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Revision 20 In addition to conducting interviews with responsible licensee personnel, the inspector attended several licensee Steam Generator Tube Leak Task Force meetings to assess the status of actions being taken.
The inspector also conducted field observations of eddy current testing on both OTSG's on several occassions and tube plugging operations in the "B" 0TSG to independently evaluate these licensee activities.
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Findings The inspector reviewed selected portions of licensee documents and observed portions of the OTSG repair program to verify that the requirements previously stated were met by the licensee.
Observation was conducted on February 17, 1982, of B&W personnel hydraulically expanding tube segments in the "B" 0TSG.
The inspector reviewed the procedure at the job site and noticed that certain pages were marked " Draft".
To determine if the licensee was exercising adequate control in preparation and conduct of the task, the inspector proceeded to.the control room to discuss this with the Shift Supervisor. While questioning the Shift Supervisor about the procedure, it was noted that his concurrence had not been obtained for step A.4.3.2 of GPU's Job Ticket C-7849 which states, " Nitrogen purge securred (Verify through control room)".
Discussion with other control room operators and Shift Supervisors revealed that the Nitrogen purge had been secured in anticipation of the job beginning some time that day and not because of the job ticket.
During further discussion with the plant engineer (author of the procedure) and Shi ft Supervisor, the inspector determined the procedure was a P1 ant Operations Review Committee (PORC)
approved procedure.
The inspector concluded that there had been poor coordination between the licensee contractor (B&W)
and control room operators.
Failure to follow the maintenance procedure was a violation of Technical Specification 6.8 requirements for procedure implementatio.
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The inspector informed the Manager Plant Operations TMI-l of inadequate coordination of work being performed cn the OTSG.
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The Manager Plant Operations TMI-l conducted a debriefing of
all individuals that day to discuss the problem.
The licensee
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now requires that before any work is to commence on OTSG's, a briefing will be held by the supervisor of the job with the Shift Supervisor.
In addition, all OTSG work must be scheduled
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and discussed at the licensee's 4:00 PM meeting, previous to work commencing or the task can not be started.
The licensee
has instituted positive control over access to the OTSG's.
For each shift, a designated Primary Auxiliary Operator (AO)
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maintains the keys to access points to both OTSG's.
The auxiliary operator is required to inform the Shift Supervisor before work commences in either 0TSG.
If the work is not scheduled, the Shift Supervisor is not to allow the work to proceed unless otherwise directed by the Manager Plant Operations TMI-1. The inspector observed subsequent tasks performed in the OTSG and found the licensee's corrective action to be
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adequate and had no further questions in this area (289/82-02-03).
At the close of the inspection, the tube failure and repair program was still being developed.
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IE Bulletin and Circular Followup The inspector reviewed the licensee's followup action regarding the IE bulletins (IEB) and circulars (IEC) listed below.
IEB 81-03, " Flow Blockage of Cooling Water to Safety System
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Components by Corbicula' SP. (Asiatic Clam) and Mytilus SP.
(Mussel)," dated April 10, 1981 I
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IEC 81-01, " Design Problems Involving Indicating dushbutton
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Switches Manufactured by Honeywell Incorporated," dated January 23, 1981 IEC 81-02, " Performance of NRC-Licensed Individuals While on
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Duty," dated February 9,1981 IEC 81-06, " Potential Deficiency Affecting Certain Foxboro 20
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to 50 Milliampere Transmitters," dated April 4,1981 With respect to the above bulletin, the inspector verified that licensee management forwarded copies of the bulletin response to
appropriate onsite management representatives, that information and corrective action discussed in the reply was accurate and implemented as described, and that the reply was submitted within the time period described in the bulletin.
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-15-With respect to the above circulars, the inspector verified that the circular was received by appropriate licensee management, a review for applicability was perfonned, and that action taken or planned is appropriate.
Acceptance criteria for the above review included inspector judgement and requirements of applicable Technical Specifications and facility procedures.
Licensee followup to the above bulletins and circulars was acceptable.
7.
Licensee Event Reports (LER's) In-Office Review The inspector reviewed the LER's listed below, which were submitted to the flRC Region I office, to verify that the details of the event were clearly reported, including the accuracy of the description of cause and the adequacy of corrective action.
The inspectcr determined whether further information was required from the licensee, whether the event should be classified as an Abnormal Occurrence, whether the information involved with the event should be submitted to
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Licensing Boards, whether generic implications were indicated, and whether the event warranted onsite followup.
The following LER's were reviewed:
LER 81-008/03L-0, dated September 25,1981 (During the Hot
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Functional Testing Program it was discovered that the Reactor
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Building access door resiliant seals were not being tested at
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the frequency required by technical specifications)
LER 81-009/01T-0, dated October 6,1981 (Design review of the
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diesel generator breaker logic revealed that the original design did not consider certain sequences of event which could lead to possible damage to motor or generator windings.
Present logic does not allow sufficient time to elapse for residual voltage decay and for load shedding to occur)
LER 81-010/03L-1, dated January 22,1982 (Quality control
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audit determined that the Reactor Building prestressing tendon had an unacceptable lift-off force.
Technical Specifications require that two adjacent tendons be inspection, which was not performed)
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LER 81-Oll/0ll-1, dated January 22,1982 (Review of a catwalk structure for application of additional loads revealed that seismic design calculations were not performed by the original architect engineer)
The above LER's were closed based on satisfactory in-office review except LER's81-009 and 81-011.
Licensee corrective actions for those
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LER's will be reviewed during a subsequent flRC Region I inspection 289/81-LO-09 and 289/81-LO-11).
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Exit Interview
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Meetings were held with senior facility management periodically during the course of the inspection to discuss the inspection scope and findings.
The inspectors met with the licensee representatives (denoted in paragraph 1) at the conclusion of the inspection on March 5,1982, and summarized the purpose and scope of the inspection and the findings.
The licensee representatives acknowledged the findings.
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