IR 05000213/1984014

Insp Rept 50-213/84-14 on 840730-0802,14-20 & 0829-1031. Violations Noted:Penetrations to Cable Vault & Auxiliary Feed Pump Room Unsealed & No Fire Watch Established.Qa Program Requirements Not Implemented

ML20140D356
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	12/03/1984
From:	Johnson T, Mccabe E, Shedlosky J, Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20140D293	List: IR 05000213/1984014 ML20140D288 ML20140D313
References
50-213-84-14, NUDOCS 8412180443
Download: ML20140D356 (12)

Text

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DCS NO 50-213/84-06-11 50-213/84-06-12 50-213/84-07-21 50-213/84-08-01 U.S. NUCLEAR REGULATORY COMMISSION 50-213/84-08-17 50-213/84-08-19 Region I 50-213/84-08-21 50-213/84-08-24 Report No. 50-213/84-14 Docket No. 50-213 License No. DPR-61 Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06101 Facility Name: Haddam Neck Plant Inspection at: Haddam, Connecticut Inspection conducted: July 30 - August 2, August 14-20, and August 29 -

October 31, 1984 Inspectors: [ // / T4 Paul D. Swetland, Senior Resident Inspector Date S'igned Jh Cr J6hn T. Shedlosky, Senior Resident Inspector

// /St Y4 Date Signed

%wd)h. ) 1m>& ll29f8 T. Johnson,yeactorInspector Date Signed Approved by

b Obb )t E. C. McCabe, Chief, Reactor Projects ik3 {ST/

Date Signed Section 3B, Division of Project and Resident Programs i

Inspection Summary: Routineresidentinspection(352 hours0.00407 days <br />0.0978 hours <br />5.820106e-4 weeks <br />1.33936e-4 months <br />)ofplartopera-tions, maintenance, radiation protection, preparation for refueling, refueling operations, licensee events and followup on licensee recovery actions for a refueling cavity seal failure.

l Inspector witnessing of cavity seal failure corrective actions identified l satisfactory completion of the approved recovery progra One violation related to the operability of plant fire barriers was

! identified. (Detail 2.2)

One violation related to the process for review and approval of field changes to design modification packages was identified. (Detail 6.3)

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DETAILS-l

' Followup on Previous Inspection Findings

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1.1 (0 pen) Unresolved Item Ut13/79-20-07), Torrect:the humidity control prob-les in the records stoi. age vault. "After r? work.and adjustment, worst ,

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case humidity remained about 40%. Documer ts which would be_ endangered j- by long tem exposure to.such conditions were; transferred to the corpor-4 ate office storage facility. This item is open pending-detemination of  ;

i adequacy of procedural controls over storage of such records.

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1.2 (Closed) Followup Item (213/83-13-04) The licensee was to restore Xray equipment to operation and improve fire protection of the -

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ammunition storage location. A new Xray device was put into service

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on October 29, 1984. The inspector verified the satisfactory operation

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of this equipment. A fire detection system had been previously

installed in the a m unition storage area. The inspector had no j further questions in this are . Review of Plant Operations 2.1 The inspector observed plant operation during regular plant tours i throughout the reporting period. The following plant areas were

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-- Control Room -- Security Building

-- Primary Auxiliary Building --FenceLine(ProtectedArea)

i -- Vital Switchgear Room -- Yard Areas i -- Diesel Generator Rooms -- Turbine Building

-- Control Point -- Intake Structure and Pump Building l -- Containment

Control room process instruments were observed for correlation be-

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tween channels and for confomance with Technical Specification 1 requirements. The inspector observed various alarm condidions which r had been received and acknowledged. Operator awareness and proper

response to these conditions were reviewed. Control room and shift

! manning were observed to be in confomance with regulatory require-ments. Proper posting and control of radiation and high radiation j areas was inspecte Compliance with Radiation Work Permits and use '

! of appropriate personnel monitoring devices was verified. Plant

! housekeeping controls were observed, including control and storage

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of flammable material and other potential safety hazards. The in-4 spector also examined the condition of various fire protection systems.

l During plant tours, logs and records were reviewed to verify that entries were properly made and comunicated equipment status /deficien-

cies. These records included operating logs, turnover sheets, j tagout and jumper logs, process computer printouts, and Plant Infoma-

tion Reports. The inspector observed selected aspects of the licensee's ,

j security organization including access control, physical barriers, '

and personnel monitorin Except as noted below, no unacceptable conditions were identifie ,

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. 2.2 During a tour of the auxiliary feedpump room on October 30, 1984, the in-spector observed two electrical penetrations in the floor of the north end-

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of the building. Penetration No. 801 was temporarily sealed with an approved

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- -fiber material. The temporary sealant material in penetration No. 800 had -

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been removed to pull a test instrumentation cable through the penetration and had not been replaced. These penetrations provide access to the safety-

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related cable vault and are required to be sealed. Upon notification, the . -

licensee.promptly replaced the temporary penetration seal. The inspector-

verified that no fire watch had been established to monitor these penetra-F

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tions.

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i ~Further review of this discepancy revealed that these penetrations had not j been previously sealed. The licensee identified the open penetrations while initially running the temporary test cable in early August 1984. The pene-trations were temporarily sealed; however, during a subsequent running o the test cable, penetration No. 800 was left unsealed. Technical Specifi- y

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l cation 3.22F requires all penetration fire barriers to safety-related areas 1 to be functional or be monitored by an established fire watch within one 1 hour. Pursuant to the above, maintenance procedure 8.5-143. Installation.

l Repair, and Inspection of Fire Barrier Penetration Seals, Revision 0 re-l quires a fire watch to be established and maintained when a penetration is

opened. - The open penetrations had apparently not been identified previously, .

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l because this location was not specifically covered by the above inspection

- and maintenance procedure. The existence of two unsealed, unmonitored pene-trations to the cable vault for more than one hour prior to August 1984,

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constitutes a licensee identified violation. However, the licensee's

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corrective actions in response to this condition did not prevent one pene-,

j tration from again being made inoperable when the temporary sealant was  !

! removed for more than an hour and no fire watch was established. These

! items constitute a violation. (213/84-14'-01) ,

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2.3 During this refueling outage, the steam door at the entrance to the con-trol room was kept open to facilitate access. This door is not a security j access door. It provides blast protection from a main steam line rupture

! in the turbine building adjacent to the control room. This protection is

' not required during cold shutdown conditions. Since this steam door is .

also a fire barrier, control room personnel were assigned fire watch duties.

l The licensee normally assigns a dedicated fire watch to compromised fire l barriers in accordance with a maintenance work order. In this case, these femal controls were not implemented. Control room personnel were aware of their obligation to shut the door in case of fire, and their fire watch duty was noted in.the control room log. This deviation from standard prac-tice for fire watches was found acceptable due to the availability'and'

fire brigade qualification of control room personnel. The licensee com-mitted to fomalize guidance for the control of open fire doors to insure consistency in implementation of compensatory action. The ins  !

review this guidance in a subsequent inspection (213/84-14-02)pect .

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In addition, the inspector noted that the control room and switchgear room blast doors are not self-closing doors. Fire doors are generally

, required to be self-closing and latching by National Fire Protection Association Code 80. Inspector review of the licensee's fire hazards analysis did not identify that this had been presented or addressed by the NRC's fire protection review. The licensee stated that both these doors are scheduled for replacement with code approved doors. Since the control room is continuously ranned and the switch-gear room door is verified closed once per shift, the inspector determined that there would be adequate control of these fire barriers until suitable replacements are obtained. This item will be closed *

upon completion of the door replacement project. (IFI 213/84-14-03)

3. Preparations for Refueling j 3.1 New Fuel Receipt Inspection '

The inspector reviewed the receipt inspection of 54 new fuel assemblie These inspections were conducted in accordance with procedure 1.4-3, New Fuel Detail Inspection, Revision 7. during the period May - June 1984. Several discrepancies were identified during these inspection Vendor repairs were accomplished in accordance with procedure SPL '

10.7-218, Rework and Inspection of Batch 15 Fuel, Revision 0. The inspector noted three discrepancies in the completed inspection documents, including incorrect crane inspection dates and missing documentation of inspection finding dispositions. The licensee

' J provided supplemental infomation showing proper disposition of these items. The inspector noted'the need.forJnore critical supervisory re-

view of inspection packages to-insure that all dispositions.are: pro-perly documented. The complet.eness of receipt inspections will,.re-l ceive routine NRC reinspectio .2 Refueling Procedure Review A review of tho refueling and startup testing procedures was perfome This review included verification of procedure adequacy and implemen-tation of plant technical specification requirements. The following i procedures were reviewed:

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FP-CYW-R12 Connecticut Yankee - Refueling Procedure Cycle

XII - XIII

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N0P 2.1-7, New Core Initial Critical Approach, Revision 8,

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SUR 5.3-6, Control Rod Reactivity Worth Measurements, Revision 14

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SUR 5.3-3, All Rods Out Just Critical Baron Concentration, Revision 8 i --

SUR 5.3-2, Hot Rod Drop Time Measurements, Revision 7

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SUR 5.3-4, Zero Power Flux Map, Revision 4

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SUR 5.3-17, Operation of the Flux Hap System, Revision 8 l

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SUR 5.3-5, Isothermal Temperature Coefficient Measurement, Revision 8 4 --

SUR 5.3-19, Boration Requirements for Reactor Shutdown, Revision 10 l

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SUR 5.3-23, Excore-Incore Axial Offset Correlation, Revision 10 '

-- SUR 5.3-20., Reactivity Balance Procedure, Revision 5

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SUR 5.3-24, At Power Flux Maps, Revision 5

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SUR S.3-26, Excess Reactivity Balance, Revision 4

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-- SUR 5.3-39, Hot Rod Drop Time Measurements Using Digital Method, Revision 2

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k 4 -- SUR 5.3-40,' Core Quadrant Power Tilt Detemination Revision 0 l -- -SUR 5.1-25, Core Power Tilt Determination, Revision 1 L - . SUR 5.1-26. Incore Power Distribution Monitoring Axial offse l

. Revision 5 1- ,

Upon completion of individual ~ procedure review, the inspector dis-

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cussed the refueling and'startup test program procedures with a licensee representative. Several minor deficiencies were noted with ;

j the procedures, and these were indicated to the licensee for correctio ,

No unacceptable conditions were identified.

3.3 Core Reload Package Submittal i

i The inspector reviewed the licensee's submittals for modifications of

> TechnicalSpecifications,(licenseelettersdatedMay2and25, July 1 24 and August 31,1984), based on the Cycle XIII core reload. Amend-ments 59 and 60 to the plant operating license were approved on l October 15, 1984. The inspector verified the implementation of these j changes in appropriate plant documents. No inadequacies were identified.-

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l 3.4 Outage Control

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! The inspector reviewed the licensee's preparation for and coordination !

! of refueling-outage activities. Management level outage coordinators '

were assigned to provide 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage of outage activities to main-

, tain current critical path status. The licensee employs a full time j outage scheduling staff. . Perfomance of plant Nfueling operations

was contracted to a refueling services vendor. Coordination and con-l trol of these operations were maintained by the plant maintenance

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department. The inspector had no further questions in this are . Refueling Activities

] 4.1 The inspector observed the reactor disassembly process in preparation ,

4 for refueling operations. On August 21, 1984, prior to commencement

! of retueling activities, the refueling cavity seal failed, draining i j the pool to the centainment floor. This event was detailed in NRC i Inspection' Report 50-213/84-23. Plant recovery actions from this ,

I event are detailed in paragraph 6 of this report. Refueling operations '

l were resumed after NRC approval of an upgraded seal installation on i October 2, 1984. The inspector observed fuel handling operations and !

! accountability in the containment,' spent fuel pool, and control j room. The following items were reviewed: l i i

! -- Core monitoring perfomed

! -- Fuel handling, fuel accountability conducted in accordance with 4 approved procedures

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Core internals stored to protect against damage 1 -- Housekeeping and cleanliness conditions acceptable

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-- Make-up and qualification of refueling crews i --

Refueling Cavity and Spent Fuel Pool water level as required

} by the cavity seal failure analysis Boron concentration and make-up (shutdown margin) as required

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l by Technical Specification 3.13

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-- Constant direct communication maintained between the Control l Room and fuel handling personnel when core alterations were 1

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in. progress

!- -- RHR flow maintained to Refueling Cavity as required by Tech-

.nical Specification 3.13

-- Audible count rate monitoring functioning in the Refueling

Cavity Area and Control Room '

-- Radiation monitoring requirements l' -- Core Reloading Check-off sheets

-- Core Component Verification L Refueling concerns will be pursued under Inspection 50-213/84-23 followu .

5. . Maintenance Program

} 5.1 The licensee's maintenance program was reviewed to detemine the i

!- adequacy of measures implemented to ensure that equipment failures '

i are evaluated for frequency and root cause, and to identify maintenance -

errors, their cause and corrective action. The inspector also re- .

,7 viewed licensee record systems to determine if they are organized I

i to support the above function r I The licensee's failure evaluation program is embodied in management j control and corrective action programs. In addition to the initiation i of repair activities, component failures-result in management review i and assignment of currective action re3ponsibilities. A Plant  ;

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Infomation Report (PIR) is written for all significant failures.

! This report is reviewed promptly by management and establishes reprt-  !

ing requirements and assignment of event evaluation and development i of corrective action. These reports receive an initial and final review by the onsite review committee (PORC). It is through this mechanism that the causal analysis and effectiveness of corrective action are evaluated. The inspector reviewed component failure PIR's issued in 1983. The system was found to be effective for management 4 infomation, assignment of corrective action and tracking completion

! status. Internal discussion triggered by PORC review and intra-

! department assignment of responsibilities appeared to be effective in prompt identification of generic failure applicability, although  ;

these initial actions Wre not always well documente The activities of the plant maintenance department have been limited 4 to perfomance of preventive and corrective maintenance and to  !

j documentation of these activities. As such, the maintenance depart-

ment had no femal means or objective to perfom detailed evaluations >

. of corrected component failures. Maintenance supervisors have been ,

i effective in updating preventive maintenance schedules based on failure ,

l experience. Inspector review of plant maintenance activities during

} 1983 did not identify evidence of repeated failure events or of similar failures to redundant components. The licensee records re-viewed did not detail any component retest failure history. In addition, l previous inspection results have identified no significant maintenance ,

i problems. The recent 417 day continuous operating period also points r

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! to the quality of plant maintenance activities.

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-Plant maintenance records have recently undergone a significant i change. A computerized data system for scheduling, control and documentation of preventive and corrective maintenance was implemented ;

in November 1983. The system provides a. capability for trending

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component failure infomation which was previously not available,

, except for personnel recollection. The licensee has not yet imple - l

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mented any fomal trending program for this data base. The inspector !

also reviewed the charter and procedures for the corporate Nuclear !

Safety Engineering Group. This organization reviews PIRs to access -t

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the need for more generic action, primarily with regard to other :

licensee and utility facilities. The NSE gmup has trended PIR -

)_ failure dat !

i The inspector concluded that the licensee has implemented an e"ffective i maintenance program. Weaknesses in documentation as they apply to ,

maintenance history and trending have been recognized and new systems l

are being implemented. The inspector had no further questions in :

this area. Maintenance will receive routine NRC reinspectio '

i Reactor Cavity Seal Failure - Recovery Programs  !

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f In response to the August 21,1984, reactor cavity seal failure, the ,

i licensee comitted to implement a recovery program prior to resuming re-

! fueling operations. This program would review the design of the pneumatic 1 seal, detemine its failure mechanism, and implement corrective actions i necessary to prevent the recurrence of a catastrophic seal .

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failure and eliminate the potential for draining the spent fuel pool '

, below the top of the active fuel. In addition, a complete!) evaluation i of equipment wetted during the event was perfomed and corrective actions  !

were completed prior to plant startu i a ,

! The inspectors followed these activities and the implementation of the associated design changes and modifications after the licensee had presented ,

{) them to the NRC Office of Nuclear Reactor Regulation. At all times, the  !

j inspectors verified that data, conclusions, and the comitments made to i ( the NRC were accurate and were implemente '

e i The licensee's programs were defined in three principal documents. Project i

Assignment (PA)84-110."ConnecticutYankeePoolSealFailureAnalysis, l Resolution and the Return to Operations" and Plant Design Changes PDCR-672, I" " Fuel Transfer Canal Waterstop at E1. 22'-0" Reactor Cavity" and PDCR-673, t,

" Reactor Cavity Seal Modification". Additionally, Engineering Specifications .

SP-ME-474 and SP-CE-183 contain the construction specifications for the '

seal modifications and the fuel transfer canal waterstop, respectively.

l l Design criteria were established for the new cavity seal and its modifica-  !

i tions. These criteria included the requirements that: the failure of a j single component does not result in catastrophic consequences; the pneu-l matic seal be modified to provide substantial margin of safety against l failures; a backup seal be provided which is capable of withstanding seismic, ;

hydraulic and air pressure forces, assuming that the pneumatic seal failed; and that, the maximum x>ssible backup seal leak rate allow sufficient time to move all fuel assem:11es into safe positions without significtnt  !

dose rates to the personnel perfoming the fuel transfer, i

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, 7 6.1 Safety Analysis Corporate Nuclear Engineering and Operations Procedure (NE0-) 3.12

" Safety Evaluations" implements regulatory requirements for conducting safety evaluations. Revision 0 of this procedure dated March 31, 1983, was reviewed during this inspection and the implementation of its requirements for the above referenced design changes was verifie The engineering discipline safety evaluations were available to the inspectors who conducted independent verifications of aspects of their content. In addition, the inspectors attended several meetings of the Plant Operations Review Committee (PORC)

and a combined meeting of *.he Nuclear Review Board and PORC on September 13. (NRB 84-15). During those meetings they confimed that the review functions were being conducted in accordance with the

, requirements stated in Technical Specification Section 6. There were no unacceptable conditions identifie The inspectors observed that an in-depth technical audit was being conducted which actively challenged aspects of the analysis and the desig In accordance with procedure NE0-3.12, paragraph 6.3, the licensee published an Integrated Safety Evaltation. This evaluation was re-viewed by the inspectors. Various aspects were discussed with NRC:NRR personnel; however, there are no outstanding issue ! 6.2 Design Change Controls The licensee implements the regulatory requirements applicable to the control of facility modifications through Station and Corporate

Procedures. These procedures establish the methods for identifying the need for a modification; the definition of design criteria; the

. control of construction specifications; the control of design and analysis calculations; and, the preparation of design drawings and drawing change During the inspection of the associated modifications, the inspectors reviewed all or part of the following procedures: NE0-3.03, "Prepara-tion, Review and Disposition of Plant Design Change Requests"; NE0-3.04, " Preparation, Issuance and Control of Project Assignrients",

GenerationEngineeringandConstructionDivisionProcedures(GE&C)

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2.01, " Preparation, Review, Approval ... , of NUSCO GE&C Division Specifications"; GEAC-2.04,"NUSCO Field Change Authorization"; GE&C-4.03, " Safety Evaluation and Technical Review of Technical Specifica-tion Change Requests"; GE&C-4.04, " Preparation, Review..., of Design Analysis, Technical Evaluations and Manual and Computer Calculations";

and, GE&C-5.07, " Processing Nonconfomance Reports".

The inspectors reviewed the implementation of these modifications as

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presented in Speci fications SP-ME-474 and SP-CE-183 and design drawings:

16103-51112, Reactor Cavity Backup Seal; 16103-51095, Reactor Cavity Seal Ring and Seismic Support; 16103-51113, Fuel Transfer Canal

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.Waterstop; 16103-51112 _ Reactor Cavity Seal Cover Assembly; 16103-20017,

. Air Supply to Reactor, Cavity Seal; 16103-29618 Reactor Cavity Seal, NES/

. -Selasco Dwg 82053-1;'and, 16103-22034 Reactor Cavity Liner. In addi-i tion, the' following implementing procedures were available and were re-l viewed: CE No. 18767-RCE-802 " Guidelines for Reactor Pool Seal. Rein-j forcing Pin Installation;" NUSCO Welding Procedure Specifications WPS-

103, 105, and 61 The inspectors noted that the licensee's QC organization had identified

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several nonconfomances associated with the installation of the steel >

n reinforcing pins in the pneumatic seal material. Specifically, station

QC nonccnfomance reports (NCR)84-191, 192, and 193 documented the __

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findings that a modification to the seals had been made without a PCRC

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approved procedure or PDCR; the materials used (stainless pins) were not

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processed through the required receipt inspections to maintain identifi-cation and traceability, and work was perfomed under a non-safety-related .

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. work order. These.NCR's represent licensee identified violations of NRC

, and quality assurance program requirements. The inspectors expressed

, concern to the licensee about this apparent disregard for quality as-surance requirements. The licensee stated that these activities had 1 l been. undertaken on a risk basis pending acceptable resolution of the

NCR's. NCR resolution then found the installation acceptable by trac- ,

1 ing the pins to supplier receipt records, review of the installation  :

i procedure, and inspection of the modified seals. The inspectors had j no further questions on this concer .

s I f 6.3 Implementation Process t .

t The inspectors observed the modifications to the Reactor Cavity Seal  ;

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. throughout its implementation. They used, as references, the NRC Licen-sing submittals, engineering specifications, PDCR's, and design i j drawing ]

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k The installation of the two back-up seal cover place assemblies and seal i ring seismic modifications required that component fit-up be made to' ,

relatively close tolerances. In the case of the back-up seal, the total

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, fit-up clearance establishes the maximum leakage rate with a failed pri-

! mary seal. This leakage rate was a basic assumption of the safety analy-

{ sis.

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! A number of changes were made to the design in response to problems en-l- countered in obtaining the s acified tolerances. This aspect of the work,

) including seventeen Design C1ange Notices (DCNs), was reviewed by the in-j specto I 10 CFR 50, Appendix B, Criterion III requires design control measures l to insure that; field changes are reviewed by the organization that per-i formed the original design unless another organization is' designated to 4

! do that. The Northeast Utilities Quality Assurance Program connits to

design control measures which meet Regulatory Guide 1.64 and ANSI 1 N45.2.11, Quality. Assurance Requirements for the Design of Nuclear Power

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Plants. These standards require that changes to design documents be re-

{ ' viewed and approved by the same groups or organizations that reviewed j i / the original documents.' Procedure -

NE0 3.03 requires design verification,

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comittee review and management approval for all plant d; sign change Procedure GE&C 2.01 specifies a quality assurance and supervisory review of all design specifications. The revision of approved engineering de .

sign specifications, drawings and documents is permitted by GE&C-2.01, paragraph 6.6.4 and is controlled ia accordance with GE&C-3.05, para-graph 6.6. The inspector reviewed the seventeen DCN's within the series 257-84 through 284-84 to determine that their content and administration met these requirements. In each case the inspector found that the DCN was controlled in accordance with the above procedurt.s; however, the in-spector noted that these procedures did not require that field changes

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receive the same review and approval process applied to the original de-sign. Specifically, the design verification, committee review, and man-agement approval of DCNs were not required, nor were the quality assur-ance and supervisory reviews required. Although the licensee subse-quently incorporated all these DCNs to the modification into a revision of the original modification package and completed the required review cycle, the failure of the design control procedures to implement the re-quirements of the quality assurance program constitutes a violation (213/84-14-04).

The inspectors reviewed the work in progress on the seal ring to assure that the activities were in accordance with the approved design document NUSCO Bettement Construction S mcial Work Proceiure BC-SC-8, "Installa-tion of Reactor Cavity Back-up Sool Seal" along with NUSCO engineering calculation 84-110-315GP " Redundant Seal Leak Rate and Structural Integ-rity Calculation" were both reviewed by the inspector. These documents i provided the detailed instructions for assembly of the back-up seal along with the requirements for mea:;uring the leakage gaps in the back-up sea Because of problems encountered with excessive clearances in the outer seal ring back-up seal, this procedure was supplemented with DCN 269-84 which attempted to reduce the back-up seal channel to plate gap with

3/16" x 1 3/4" rolled bar. Also, DCNs 279-84 and 28184 removed certain back-up seal leakage clearances where overlapping closure plate surfaces

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were sealed with RTV. Upon completion of the back-up seal installation, the licensee contirmed the area for leakage by taking feeler gauge read-ings of the available gaps between the seal sections and the seating plates. The inspector observed the gap measurement in process and veri-fled the correct transmittal of leak area data. Based on the available gap measurements and the obstruction of certain areas using RTV verified capable of withstanding full cavity hydrostatic pressure, the licensee judged that the minimum leak rate criteria had been established. No

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unacceptable conditions were identifie .4 Quality Control Inspection The inspector reviewed the data available for the visual weld inspections perfomed in accordance with procedure QCI-CY-10.02. Weld data cards and NUSCO Construction QC Inspection Reports 84-110-001 through 029 were r available for the seismic alignment fixtures, and back-up seal attachments, the seal ring bearing plate, and the fuel transfer canal waterstop field welding. There were no unacceptable conditions identifie The inspector also reviewed the QC Honconformance Reports (NCR) issued by the station QC group and those issued by NUSCO Construction QC for these modifications. The following NRC's were reviewed 84-191, 192, 193, 225,

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226, 227, 228, 229, and CY-84-075, 077, 078, 079, 080, 083, 084, 086, 087, 088, 090 and 092. The inspector verified that the administrative controls for resoIution and disposition of these NCRs were being followed and that the findings for cause and prevention of recurrence appeared accurat There were no unacceptable conditions identified, f

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6.5 Wetted Equipment' Evaluation

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The licensee perfomed a complete inspection of the affected contain-ment areas to identify all components and structures which had been wetted during the seal failure event. An evaluation of each exposed item was completed and corrective action was completed. This ' pmgram and the results were detailed'in the licensee's submittal to NRC Licensing on October 16, 1984. The inspector reviewed the licensee inspection pmcedure, NU SAE-1, Revision 0. Independent inspection-of the wetted areas and repair work confirmed the results documented in'the licensee's submittal. The inspector verified that two open '

issues, inspection cf the containment liner and threaded conduit connections have been incorporated in outage schedules for the next refueling. The inspector had no further questions in this are .0 Licensee Event Reports (LERs)

The folicwing LERs were reviewed to verify that the details of the event were clearly reported, including the accuracy of the description of the cause and the adequacy of the correc'tive action. The inspector detemined whether further infomation was required, and whether there were generic implications. The inspector also verified that the reporting requirements of Technical Specifications and Station Adninistrative and Operating Procedures had been met, that appropriate corrective action had been taken and that the continued operation of the. facility was conducted within Technical Specification Limit Fire Door Latch Inoperable s

-- 84-07 Degraded Cable Penetration Fire Barriers

-- 84-08 Inoperable Fire Door

-- 84-09 Total Loss of Offsite Power / Reactor Trip - event detailed '

in NRC Inspection Report 50-213/84-19

-- 84-11 Containment Integrated Leak Rate Test Fia1ure - event detailed in NRC Inspection Report 50-213/84-25

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84-12 Containment Loi.a1 Leak Rate Test Failure - event detailed in NRC Inspection Report 50-213/84-25

-- 84-13 Reactor Cavity Seal Failure - event detailed in NRC Inspection Report 50-213/84-23 and in paragraph 6_of this report

-- 84-14 Loss of Offsite Power / Emergency Diesel Generator 2A Failed to Pickup Load - event detailed in NRC Inspection Report 50-213/84-22, 8.0 Post Accident Sample S.vstem Isolation Valves During a review of post accident sample system (PASS) operation, the -

inspector identified a potential pmblem with the operation of the

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PASS when conteinment integrity is established. The containment air i sample and retum line penetrations are isolated by redundant remotely l operated ranual valves. In order to take a PASS air sample, these

! valves must be open. TechnicalSpecification(TS)3.11iowever, requires manual containment isolation valves to be closed when reactor coolant temperature is greater than 2000 s i

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Although NRC standard technical specifications have allowed manipula- l tion of manual containment isolation valves under administrative control, the licensee had not requested or been granted such relie The inspector determined that the PASS isolation valves had in fact-been opened multiple times for test samples since October 1983, with nc' identification of the potential violation of TS. Upon notification, licensee ~ management required the valves to be tagged closed until the technical aspects of this problem are resolved. At the conclusion of the inspection the licensee had determined that several submittals to NRC had documented the existing condition of the PASS isolation valves, but no NRC concurrence or revision of TS 3.11 had been identified. Since the isolation valves were opened

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for short periods under approved procedusal control and this practice i

o has been routinely approved for operation of manual containment i-isolation valves, the inspector detennined that there was minimal safety significance associated with this potential violation. The inspector stated that this item would remain unresolved pending the i

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licensee and NRC determination of the correct containment isolation t requirement for these penetrations. (213/84-14-05) .

9.0 Unresolved Items

! Unresolved items are matters about which more information is required in order to determine whether they are acceptable items or violation ,

Unresolved items identified during this inspection are discussed in '

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paragraph ) 10.0 Exit Interviews l

During this inspection, meetings were held with plant management to j discuss the findings. No proprietary information related to this

inspection was identified.

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