IR 05000309/1988016

Insp Rept 50-309/88-16 on 881006-1109.No Violations Noted. Major Areas Inspected:Plant Operations & Outage Activities, Including Followup on Previous Insp Findings,Review of Special Repts,Licensee Event Followup,Maint & Surveillance

ML20196E546
Person / Time
Site:	Maine Yankee
Issue date:	12/01/1988
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20196E543	List: IR 05000309/1988016 ML20196E538
References
50-309-88-16, GL-87-12, IEB-88-008, IEB-88-009, IEB-88-8, IEB-88-9, NUDOCS 8812120032
Download: ML20196E546 (14)

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U.S. NUCLEAR REGULATORY COMMISSION Region !

Report No.: 50-309/88-16 License No.: OPR-36 Licensee: Maine Yankee Atomic Power 83 Edison Drive Augusta, Maine 04336 Inspection At: Wiseasset, Maine Conducted: October 6, 1988 through November 9, 1988 Inspectors: Cornelius F. Holden, Senior Resident Inspector R1charpJ. reudenberger, Resident Inspector

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Approved: / .i Il [88 L' owe l l E. T pp, Chief Oste Reactor Projects Section No. 3A Suma ry: Inscection on October 6.1938 to November 9,1988 (Report he. 50-309/83-16 Areas Inspected: Routine resident inspections of plant operations and outage activities including: followup on previous inspection fiadings, review of special reports, lira m e event follewup, operational safety verification, maintenance, surveillance, physical security and radiation protectio The inspection included bacbhif t inspections on October 22 and November 16, 198 Results: Overall, 'ontrol of outage activities appears goo . Planning and outage coordinati .r activities appear to be working well and the reducticn in control room ac','vities has benefited the operating crew. Major surveillance tests such as e utainment integrated leak rate testing and eddy current inspec-tion of number 1 steam generator had good oversight. The plant has identified a number of witing discrepancies during this outage. The residents will review the individual resolution of these discrepancies and the root cause/causes of these problem G12120032 G31005 PDR ADOCK 0500030P o PDC

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OETAILS 1. Persons Contacted Within this report period, interviews and discussions were conducted with various licensee personnel, including plant operators, maintenance tech-nicians and the licensee's management staf . Summary of Facility Activities At the beginning of the report period the plant was operating at 76 per-cent powe A shutdown was initiated on October 15 for the cycle 11 refueling outage. The plant remained in a refueling outage for the remainder of the report perio . Followup on Previous Inspection Fin ings (Closed) Unresolved Item 50-309/32-18-04 involved the venting and draining of closed systems that penetrate the containment during Integrated Leak Rate Testing (ILRT). During the performance of an ILRT during this report period, the inspector reviewed the licensee's actions to resolve the ite A summary of this review is inclrded in section 5 of this report. Based on the findinas of the review, l'.iresolved Item 50-309/82-18-04 is close . Operational Safety Verification (IP 71707)

l On a daily basis during routine facility tours, the following were checked: manning, access control, adherence to procedures and LCO's, instrumentation, recorder traces, protective systems, centrol room annun-ciators, radiation monitors, emergency power source operability, control room logs, shift supervisor logs, and operating order On a weekly basis, selected Engineered Safety Features (ESF) trains were verified to be operable. The condition of the plant equipment, radiological controls, security and safety were assessed. On a biweekly frequency the inspector reviewed a safety-related tagout, chemistry sample results, shift turn-overs, portions of the containment isolation valve lineup and the posting of notices te worker Plant housekeeping and cleanliness were also evaluate The inspectors witnessed a variety of activities a s soc ii. d with the refueling outage. These included: control of heavy loads in the vicinity of the reactor vessel, eddy current testing of the steam generator, reactor vessel head removal in preparation for refueling, refueling and spent fuel pool activities, motor operated valve testing, portions of the l reactor coolant pump rotating assemb'y change-out, turbine replacement and refurbishment, Emergency Diesel Generator Db1A overhaul, and circulating water pump house repairs.

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The inspector observed selec .*>4 as of the plant's operations to deter-mine compliance with the NRC' regulations. The inspector determined that the areas inspected and the licensee's actions did not constitute a health and safety hazard to the public or plant personne The following are noteworthy areas the inspector reviewed: ,

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Work Control Activities

' During a refueling shutdown the licensee employs contractors for rou-tine preventive and corrective maintenance as well as installation of Engineering Design Change Requests (EDCR's). During peak periods approximately S00 additional personnel are on site. The increased workload would severely stress the licensee's routine work control t system. The licensee has implemented a variety of control measures which are utilized during the outage to accommodate the increased "

workloa During routine operations, Deficiency Reports / Repair Orders (DR/RO's) ,

are processed through the control room. After receiving Engineering and Quality Assurance review, a priority is assigned by the plant o shift supervisor (pSS) and the DR/R0 is forwarded to Maintenance for work package preparation and scheduling. The DR/R0 is returned to the control room for tagging the system out-of-service and authoriza-tion prior to work. Upon completion of the activity, the work pack-age is returned to the control room for tag removal, functional test-ing and return to servic For the outage, the licensee eliminated the excess Mffic in the control room by establishing an Operations Control Center (OCC) above the maintenance offices. The OCC has responsibility for tagging of I systems. Operations crews change from six crew rotation to four crew

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rotatio Two crews are utilized for outage coordinators and OC All tagging functions previously conducted by the control room are now conducted from OCC. The OCC and the control room operators coor-dinate the removal of systems from servic Auxiliary operators assigned to OCC hang the approved tagouts. Additionally, some oper-ators are detailed to the Plant Engineering Depertmen l The inspectors conducted a review of the outage tagout system. DR/R0

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packages were prepared in advance and the OCC reviewed the work scope and tagged systems to accommodate multiple RO's under a single tag-out. Feedback to the Planning Section was frequent and provided for better coordination of system unavailabilit Tagoutt were prepared in advance and the control room maintained a copy of all active tag-outs in addition to the master tagout book maintair ed by OCC. An active worklist was maintained in the control room sia a dedicated telephone and computer terminal linked with OCC. The espectors con-cluded this initiative provided good control of plant egipment and  ;

alleviated unnecessary congestion in the control room, r

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b. Lead Shielding On October 14, the inspector reviewed the status of the lead shield-ing installed on safety related piping systems. The installation is controlled by the Discrepancy Report / Repair Order (OR/RO) procedur The repair order includes an evaluation prepared by the Plant Engi-neering Department of the effect of the additional weight of the lead on the piping system or structure on which it is installed. At the time of the review, the plant was operating at 75 percent powe There was a minimal amount of shielding installed. A plant tour was conducted to identify any shielding that may have been installed in preparation for the refueling shutdown. No installation of lead shielding was ide ified which could impact the operability of safety related systems or structures thst did not have the associated evalu-ation in a DR/R0 complete c. Service Water System The service water system is the safety related cooling system which provides cooling water from the ultimate heat sink to the primary and secondary component cooling heat exchangers. The system consists of four pumps, four heat exchangers and two supply headers. Any one pump discharging through either header to a single heat exchanger provides sufficient heat removal capability during design basis acci-dent situations. Pumps "A" and "B" are normally aligned to discharge through the south supply header and "C" and "D" through the north supply header. Following an automatic start upon receipt of an emergency diesel generator load program start signal, all four ser-vice water pump breakers are shu Pumps "A" and "C" are powered from the same emergency bus, therefore controller logic causes the

"C" pump to trip automatically if the "A" pump is running. Similarly the "B" pump will trip if the "D" pump is running. If the "A" or "0" breaker fails to shut its backup pump ("C" or "B") will not tri The controller logic provides for the tripping of extra pumps to minimize the load on the emergency diesel generators while maximizing the reliability of the service water syste Flow from the supply headers passes through check valves just prior to entering the heat exchanger inlet heade All valves in the heat exchanger inlet header are normally open to allow flow from either supply header to provide flow to any heat exchanger. The outlet valves of the heat exchangers are operated to control which heat exchangers are in ser-vice. Of the two heat exchangers associated with each component cooling subsystem (PCC or SCC), usually one is in service and one is in standb Therefore, the heat exchangers in service providing cooling to both safety trains of the component cooling system are supplied from a co-mon inlet heade _ - _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

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Operation of the Service Water System in this configuration has been reviewed by the licensee and determined to provide more reliable cooling flow than operation with the inlet header cross connect valve close Service Water pumps powered from opposite safety busses dis-charge through each supply header (north and south.) Also, there are check valves installed at the junction of the supply headers to the inlet header, therefore eperation in the configuration described above allows pumps powered f rom diverse sources to supply cooling flow through either supply header to any heat exchange During the past operating cycle, one of the Service Water pumps was removed from service for an extended period for maintenance. During such periods, if a design basis scenario were to occur which included a Loss of Coolant Accident (LOCA) coincident with a Loss of Offsite Power (LOOP) and a single failure of the emergency diesel generator which supplies power to the Service Water train which has both peeps in service, it 's possible for the Service Water System to be in a condition that it would at provide sufficient cooling flow to the component cooling subsystem in service. A single Service Water pump discharging to two heat exchangers does not provide sufficient heat removal capability during the Recirculation phase of incicent condi-tion Emergency Operating Procedures require a verification that two Service Water pumps are operating as part of the initial post incident respons Licensee evaluation of this situation was still in progress at the close of this inspection perio Inspection of this topic is not complete. The inspector will review this area further in a future inspectio d. Wiring Discrepancies During the report period, the licensee identified several wiring dis-crepancies which have an impact on the ope ation of safety-related system On October 21, the Secondary Cceponent Cooling Non-Safe-guards Isolation Valve t-1p circuitry was identified as having been miswired such that the valve would close on low pressure sensed by one of the two suction pressure switche The trip system was designed to close the valve on a two-out-of-two logi On November 1, the Plant Engineering Department identified a wiring dis-crepancy associated with the pressurizer pressure inputs to the reactor coolant subcooling margin monitor which resulted in the sub-cooling margin monitoring system indication being unreliable. (The licensee will submit 'n LER on this topic). On November 3, the Plant Engineering Department identified a wiring discrepancy associated with the containment hydrogen analyzers. The output from the "A" train powered analyzer was connected to the input of the "B" train powered control board recorder. Similarly, the output from the "B" train powered hydregen monitor was the input to the "A" train powered control board recorder. Also, an ongoing Engineering Design Change to replace incnre instrument cables, has identified rumerous incon-sistencies between field installation and print _ _ - - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _

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i The licensee is evaluating each of these instances independentl !

Also, the possibility of common root causes will be addressed. This is considered an Unresolved Item pending the completion of the above noted evaluation and further inspection. (UNR88-16-01),

a , Integrated Leak Rate Testing 1 During October 18 through 21, 1988, the licensee performed a periodic j Integrated Leak Rate Test (ILRT) as required by 10 CFR 50, Appendix i The test was performed in accordance with Procedure 3.17.1, "Integrated !

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Containment Leak Rate Test," Revision 9. The inspector reviewed the pro- '

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cedure, witnessed preparations, various portions of the test and observed penetration realignments following the tes . The inspector reviewed the test procedure for technical adequacy and to .

] verify compliance with the commitments the licensee has made to the NRC

regarding the venting and draining of closed systems which penetrate the j

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containment buildin '

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. The procedure review identified a discrepancy between the test pressure i used during a full pressure test and the peak predicted accident pre;sure ,

j as explained in the Final Safety Analysis Raport (FSAR). 10 CFR 50, !

Appendix J. Paragryh II. I, defines "Pa" as the calculated peak contain- !

ment internal pressure related to the cesign basis accident and specified

! in the technical specification or associated bases. Technical specifica- l j tion 4.4 assigns a value of 50 psig for Pa which was the test pressure '

specified by procedure 3.1 The Inputs and Assumptions Source Document i

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(IASD), which is a compilation of operator-controlled plant parameters, '

instrumentation setpoints administrative operating limits, and safety l analysis assumptions, identifies the maximum predicted pressure following -

) an incident to be 49.5 psig. Figure 14.15-1 in FSAR section 14.15 1

"Reactor Containment Pressure Analysis" appears to agree with this value.

The inspector oncluded that the Technical Specification value of 50 psig .

i was selected allowing conservative margin above the peak predicted con-l tainment pressure of 49.5 psi The containment is conservatively designed to withstand 55 psig, therefore, the designation of Pa assures i testing consistent with expected accident pressures.

l l Paragraph 14.15.3.2 of the FSAR states that analyses performed for a power i

uprate from 2,440 kdt to 2,630 Mdt indicate that the peak predicted con-

! tainment pressure would increase by 1.77 est for operation at 2,630 Mdt.

J The inspector questioned the conservatism of the value Pa as currently

defined in the technical specifications and used for testin Further

information was provided by the licensee as to the origin of the data

' included in the FSAR. The tables in section 14.15 are based on analyses performed during original plant itcensing. The Safety Analysis was per-formed assuming a core thermal power of 2,611 Mdt which conservatively oounded the original licensed maximum core thermal power of 2,440 Mdt. In l 1978, a sensitivity study on the containment pressure response was per-J ferred by the licensee to assess the a1fect of a core power increase from l

2,440 Mdt to 2,630 Mit. The sensitivity study used a different computer l

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code which modeled a more conservative transient than the original '

analysi Increases in peak containment pressure predicted by the power ,

upgrade analysis cannot be directly related to the original analysis; therefore, the licensee did not change the valve of Pa as defined in the

' : :hnical Specification The inspector will review the basis for the l u lculated peak accident containeent pressure in a future inspectio !

The inspector also reviewed venting and draining of systems during leak rate testin CFR 50, Appendix J. Paragraph !!!.a.1.(d), states, in part, that those portions of the fluid systems that are part of the reactor coolant pressure boundary (RCPB) and are open directly to the con-  !

tainment atmosphere under post-accident conditions and becon:e an extension of the boundary of the containment shall be opened or vented to the con-tair, ment atmosphere prior to and during the tes CFR 50, Appendix J Paragraph !!I a.1.(d) further states that, (with certain exceptions given therein) portions of closed systems inside con-tainment that penetrate con +einment and rupture as a result of loss of coolant accident shall be vented to the containment atmosphere; and that all vented systems shall be drained of water or other fluids to the extent necessam . ' assure exposure of the system containment isolation valves to l containmens air test pressure and to at,sure they will be subjected to post [

accident differential pressur ;

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Region I Inspection Report 50-309/82-18 Detail 3.5 identifies an unre-solved item concerning the venting and draining of systems which are sub-  !

ject to these requirement As described in the Inspection Report the licensee committed to complete the required analysis and necessary plant modifications to comply with the requirements of Appendix J or submit an Appendix J exemption reques To address this unresolved item the inspector reviewed the following cor-respondence between the NRC and the licensee:

- Maine Yankee Atomic power Company letter dated January 11, 1984,

"Containment Leak Testing".

- Nuclear Regulatory Commission letter dated August 24, 1984,

"Safety Evaluation Containment Isolation Valve Type C Leak Testing Program and Containment Liner Weld Test Channel Integrity and Testing."

- Maine Yankee Atomic Power Company letter dated January 8, 1985,

"Containment Leak Rate Testing Program."

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U.S. Nuclear Re wlatory Commission letter dated May 23, 1986, "Safety i Evaluation - C antainment Isolation Valve for Type C Leak Testing

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Program."

Tht above noted safety evaluations established those peratrations which i are exempt from Type C leak testing. A valve can be exempted from Type C testing if it can be shown that the valve does not constitute a potential i containment leakage path during or following a loss of coolant acciden Valves which are classified as requiring Type C testing are also required to be opened or vented to the containment atmosphere prior to and during the test. Of the 77 piping systems which penetrate the containment build-ing, there are 37 which are exempted from Type C testing and 40 required to be tested it. accordance with 10 CFR 50, Appendix To ensure that valves aligned to vent and drain penetrations were not i manipulated during the test ILRT, informational tags were hung on the valyes when aligned to the test configuration and removed when the valves were realigned. Af ter the test the inspector observed the realignnent of several penetrations in cc-tainmen One valve (PV-23) associated with penetration 42 was found to be in the closed position anoarently due to personnel error. Therefore penetrction 42 was not vented to the contain-

.nu.t atmosphere during the test. Penetration 42 will be Type C Local Leak l Rate Tested, the results from the Type C test wi'1 be added to the ILRT result Of the 40 piping penetrations which require Type C testing, there are 4 i

penetrations that were net able to be vented and drained during the Type A

, tes One penetration (PEN 63A) is aligned to provi.ie indication on the ,

instrumentation used to gather test dat The remaining three penetra-tions, Pen 24, Pen 39 and Pen MI-A, require modification to enable them to be properly vented and drained. All three of the penetrations require the addition of a manual valve to establish a test boundary b9 tween the con-tainment integrity valve and the system operatirig pressure. tiodifications f te many of the other penetrations have been completed previously. For  :

this Type A Containment Integrattd Leak Test the results from the Type C Local Leak Rate Test of the four penetrations listed above will be used to adjust the ILR1 results. For future Type A tests tte licensee has commit-ted to complete the modifications to penetrations 9EN 24, PEN 39 and PEN 1 MI-A prior to the next performance of a Type A tes Unresolved Iten

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6. Plant Maintenance (IP 62703)

The inspe: tor observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative and main-

tenance proc dures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualifications, radio-logical controls for worker protection, retest requirements, and reporta-i bility per Technical Specifications.

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'. l 8 l Discrepancy Date Report Number Description 10/26 3656-88 Relay testing 10/26 0401-88 Battery #2 Replacement 10/27 3801-88 Cavity Seal Ring Insta'ilation .

The insper. tor reviewed Repair Order 3656-88 which involved relay testing i for safety bus Procedure review and review of relay testing was con- l ducted. The technician performing the work had identified a discre. nancy :

between the Maine Yankee procedure and the relay setting book raintained by Central Maine Power (CMP) Co. The inspector reviewed the discrepancy with the cognizant engineer. The discrepancy was the manner in which the time delay thermal overToad setting was tested. The actual setting for the thermal overloads was incorrectly reflected in the procedure number '

5-23-2 due to a transcription error but correctly set on the re' The

relay trip time band specified in the procedure was 36-44 seconds with an

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amperage of 16.7 to 20.5 amps. The actual time / current band should have been 42.3 to 51.7 seconds at 13.5 to 16.5 amps. The licensee corrected the settings in the procedur The inspector had no further question l

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The inspector reviewed Repair Order 0461-88, Battery Number 2 Replacement, t Quality Control inspectors were observed to witness tore;;ing of seismic '

support brackets. Seismic design questions were reviewed with Yankee Atomic Engineers. The inspector had no further questiens.

7. Surveillance Testing (Ip 61726)

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The inspector observed parts of tests to assess performance in accordance

, with approved procedures and LCO's, test results, removal and restoration

of equipment, and deficiency review and resolution.

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The following surveillances were reviewed:

i Date Procedure Number Title 10/13 3.6.2.2.20 Reactor Trip Breaker Time Testing 10/21 3.1.20 - Sec Safeguards Valve Testing - P25A and P25C 5.4.8b Full Flow and Str.ne Test - Cold Shutdown Conditio /11 3.17. Purge Supply Ducs Leak Test 10/17-21 3.1 Integrated Containment Leak Rate Test

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The inspectors verified that the surveillance tests were performed in accordance with the procedure, that second verifications were conducted when required and that the technicians were knowledgeable of the equipment they were operating. No violations were identifie . Observations of Physica'i Security (IP 71707)

Chocks were made to dctermine whether security conditions met regulatory requirements, the physical security plan, and approved procedures. Those checks included security staffing, protected and vital area barriers, vehicle searches and personnel identification, access control, hadging, and compensatory measures when required. Minor discrepancies involving escorting of personnel were observad; these were corrected. The int.pector had no further commen . Radiological Controls (IP 71707)

Radiological controls were observed on a routine basis during the report-ing period. Areas reviewed included Organization and Management, external radiation exposure control and contamination control. Standard industry radiological work practices, conformance to radiological control proced-ures and 10 CFR Part 20 requirements were observe Independent surveys

, of radiulogical boundaries and random surveys of nonradiological points throughout the facility were taken by the inspector.

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Additionally, Radiation Work Permit (RWP) 88-10-329 was reviewed in detail. New fuel was transferred from the new fuel storage area into the spent fuel pool in accordance with this RWP. The inspector revisweo the RWP and observed a portion of the work performed. Radiological Control boundaries were adjusted by the Rad-Con technicians prior to the start of work to f acilitate the work to prevent movement into and out of contam-inated areas. The changes were properly reflected in surveys included with the PWP. The inspector noted no discrepancies asseciated with the posting, surveys or protective clothing requirements established by the RWP, or the workers adherence to the RWP requirement . Generic Lecter 87-12 A review was conducted of the licensee's response to NRC Generic l.etter

87-12, Loss of Residual Heat Removal (RHR) while the reactor coolant sys-tem (RCS) is in a partially filled configuration. The licensee's response was dated October 26, 1987. The response covered actions taken during routine refueling operations and under abnormal circumstances. The inspec-tor reviewed Operations Procedure OP-1-17-6, Lowering of the Reactor i Vessel level (-12 inches) for head removal, in light of the response to this Generic Letter. This procedure celineates eperator actions for rou-tine refueling. During this routine refueling evolution, reactor vessel

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lovel is lowered 12 inches below the head flange in preparation for head detensioning and removal. Two sources of level indication were available to the operators, one wide range (WR) cavity level instrument, LT-105, and one narrow range (NR) cavity level instrument, LT-104. The vessel flange is ct elevation 20 feet and the top of the hot leg is at elevation 15 f The procedure contains precautions for periodically stopping the draining process to verify level indication is in agreement. Operations Abnormal Operating Procedure (AOP) 2-34 was recently changed to address the loss of RHR cooling. Some of the precautions listed included a restriction against starting a second RHR pump if one pump has tripped due to cavita-tion, and establishing the ability to close the containment equipment hatch within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of loss of RH The licensee's response indicated that specific procedures addressing operations with the vessel partially drained will be developed prior to conducting those operations. The irspector had no further question During the review, the inspector identified one Procedure Change Request which was reflected in a controlled procedure as require During the refueling the Primary Inventory Trend System (PITS) was removed from ser-vice in order to perform major system upgrade All references to the PITS in active procedures were removed by Procedure Change Request (PCR)88-144 As a result of the identification of this discrepancy the licen-see conducted a review of all active PCR's. A weekly audit of PCR's was establisheo for the cutage since PCR's are more frequent during plant out-ages. This was the only PCR identified as as a problem during the review conducted by the license The inspector had no further question . NRC Bulletin Review NRC Bulletin 88-08 "Thermal Stresses in Piping Connected to Reactor Coalant Systems" The NRC issued Bulletin 88-08 "Thermal Stresses in Piping Connected to Reactor Coolant Systems" on June 22, 1988, requesting that licen-sees review their reactor coolant systems to identify any connected, unisolable piping that could be subjected to temperature distribu-tions which would result in unacceptable thermal stresses and take action, where such piping is identified, to ensure that the piping will not be subjected to unacceptable thermal stresse Two supple-ments to Bulletin 88-08 were issued, Supplement 1, on June 24, 1988, and Supplement 2 on August 4, 198 __ _ _ _ _ _ _ _ _ __-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

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The licensee performed an evaluation to determine whether unisolable sections of piping connected to the Reactor Coolant System are sub-jected to stresses from temperature stratification or oscillations that could be induced by leaking valves. The Reactor Coolant, Chem-ical and Volume Control, Safety Injection, Loop Drain, Sample, Reactor Water Level Indication and Reactor Vessel Gas Vent Systems were included in the review. The inspector verified that the appli-cable piping sections were evaluated. The licensee's review con-cluded that there are no unisolable sections of piping connected to the reactor coolant system that may have been subjected to excessive thermal stresses induced by leaking valves which have not been pre-viously evaluated in the piping design analysis.

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Bulletin 88-08 is closed.

, NRC Bulletin 88-09 "Thimbh Tube Thinning" The NRC issued Bulletin 38-09, July 26, 1988, concerning thimble tube thinning in Westinghou e reactors. Thimble tube thinning was the result of flow-inducec vibration. Since the thimble tubes are exposed to reactor sy stem pressure, they form a portion of the

reactor coolant system boundar There were some instances of thimble tubes experiencing leaks. Although Maine Yankee was not required to respond to this Bulletin, the inspector reviewed Maine Yankee's design similarities with the problems noted in the Bulleti The inspector reviewed Maine Yankee's design for thimble tubes and their resolution of NRC Inforuation Notice 87-44 dated September 16, 1987, which dealt with the same subject. Maine Yankee has a fixed incore monitor system that utilizes a system similar to

Westinghouse reactors with some distinct differences. The incore

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instruments (ICI's) extend from the seal table, through the bottom of l the vessel and into guide tubes within the fuel elements similar to

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the Westinghouse design. The Westinghouse design utili:es a single tube with the detector mounted internally. The Maine Yankee design utilizes a double tube. Four detectors and a core exit thermocouple are mounted between tae inner and outer tubes. The inner tube is dry and capable of accepting a calibration probe from a movable incore detector system similar to the Westinghouse design. In order for the ICI to leak, fretting of the outer sheath and rhodium incore detector leads would have to take plac The failure of the incore detectors would be information available to the operator In effect, Maine Yankee has an additional boundary to RCS pressure in the ICI system =

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Additionally, Maine Yankee replaces ICI on a rotating basis with the ;

entire set of ICI's replaced every three refueling cycles. Westing-house ICI's are designed for extended service. Maine Yankee has not experienced any fretting of ICI's. The licensee does not have a pro-gram for examining used ICI's to determine if thimble tube fretting has occurred since the detectors are disposed of after us The inspector had no further question Bulletin 88-09 is close . Steam Generator Eddy Current Testing Steam generator eddy current testing (ECT) is being performed en the number 1 steam generator during this outage. The inspector reviewed the i selection process for choosing steam generator number 1 for testing with

! the engineer responsible for the ECT. Industry experience demonstrates

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that under some circumstances, the operating conditions in one or more steam generators may be more severe than those in other steam generator Previous testing of the Maine Yankee generators had indicated that the

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number 2 steam generator may be subject to slightly more severe operating conditions than the number 1 or 3 steam generator However, the most recent testing of number 2 steam generator indicated there was minimal degradation since the prior tes The licensee decided to perform eddy current testing on the number 1 steam generator in accordance with the requirements of Technical Specification 4.10. The inspector reviewed the tube inspection sample selection. Areas where plant history and industry experience indicated a higher potential for tube damage had a higher con-centration of tubes examined with a random pattern for the examination of j

the remaining areas. An indication was identified in a tube in the first

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sample of 514 tubes which exceeded the plugging criteria established in Technical Specification 4.10, resulting in a C-2 classification. An

expanded Inspection sample of 1028 additional tubes was chosen and reviewed by the inspector. A second indication exceeding the plugging

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limit was identified in a tube in the second sample, requiring an addi-l tional sampic of 2056 tubes. At the end of the report period eddy current testing was still underway. The inspector will review the final results of the eddy current examination in a future inspectio . Containment Temperature Profiles (TI 2515/98)

The containment air temperature is monitored by 12 temperature detectors l

located throughout the containment building. The inspector verified the

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physical location of the detectors, reviewed the basis for the weighting

- factors, and verified the accuracy of the weighting factors in the plant I

compute The temperature detectors are located in various compartments of the containment building, three in the loop areas, one in the head lay down area, one in the pressurizer cubicle, three on the charging floor,

_ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ .

~ .

.

three in the outer annulus (near the suction of containment recirculation fans), and one on the polar cran The output from the detectors is idjusted by a weighting factor then averaged to arrive at an overall weighted average containment temperature. The weighting factors are based on an approximate air volume for which the detector is considered to be representative. The weighting factors bases were reviewed. The inspector considered the locations of temperature detectors in combination with the weighting factors for the average containment air terrperature to be a representative method for monitoring the containment temperature. Infor-mation on the monthly weighted average containment temperature is included

! in Region I Inspection Report 50-309/88-19, detail TI 2515/98 is close .

'

14. Exit Interview (IP 30703)

Meetings were periodically held with senior facility management to discuss the inspection scope and findings. A summary of findings for the report period was also discussed at the conclusion of the inspection. The licen-  !

see did not identify 2.790 materia ,

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