IR 05000289/1988003

Safety Insp Rept 50-289/88-03 on 880207-0305.No Violations Noted.Major Areas Inspected:Plant Shutdown & Startup Activities,Dhr River Pump Strainer Foundation Bolt Problem & Shift & Daily Surveillances

ML20151V658
Person / Time
Site:	Three Mile Island
Issue date:	04/20/1988
From:	Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20151V649	List: IR 05000289/1988003 ML20151V644
References
50-289-88-03, 50-289-88-3, GL-83-28, NUDOCS 8805030137
Download: ML20151V658 (20)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket / Report No. 50-289/88-03 License: OPR-50 Licensee: GPU Nuclear Corporation P. O. Box 480 Middletown, Pennsylvania 17057 Facility: Three Mile Island Nuclear Station, Unit 1 Location: Middletown, Pennsylvania Dates: February 7, 1988 - March 5, 1988 Inspectors: R. Conte, Senior Resident Inspector

• D. Johnson, Resident Inspector Accompanied by: S. Peleschak, Reactor Engineer, Region I (RI)

A. Sidpara, Resident Inspector M. Banerjee, Reactor Engineer, RI

• Report n Inspector Approved by:

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C. Codgillgief, Reactor Projects Section 1A Date Inspection Summary:

Areas Inspected: The NRC staff conducted routine safety inspections during routine operations. Plant operational items reviewed were: plant shutdown and startup activities, decay heat removal river pump strainer foundation bolt problem, shift and daily surveillances, and emergency feedwater pump inservice testing. Items reviewed in other functional areas included: engineering planning for Regulatory Guide (RG) 1.97 modifications; Safety Issues Management System (SIMS) items related to reactor trip system modifications, natural circulation cooldown, and emergency feedwater (EFW) upgrades; and, licensee fitness for duty polic Inspection Results: The licensee continued to operate the plant safel The short outage for a main generator stator coolant problem and the related plant shutdown and startup activities were conducted in an acceptable manner. Outage planning was good which resulted in the completion of repairs to NI-2 and several signifi-cant unisolable Once-Through Steam Generator (OTSG) skin valve leak The seismic concern associated with the defective bolts on the 1A Decay Heat River Pump strainer foundation was tentatively resolved by engineering personnel. Region I will review the licensee calculations in a future repor One unresolved item resulted from this revie l 0805030137 880422 PDR ADOCK 05000289 Q MD

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Licensee corporate engineering personnel appeared adequately prepared and knowl-edgeable on upcoming modifications to comply with the provisions of RG 1.9 SIMS items reviewed resulted in the closecut of two issues -- SIMS Item No. 75 (B-80), Reactor Trip System - Vendor-Related Modifications, was adequately ad-dressed, and SIMS Item No. MPA-8-66 Natural Circulation Cooldown Issue, was close SIMS Item No. II.E.1.2, EFW Upgrades, was reviewed and the status of remaining open items was documente The licensee responded to NRC staff inquiries concerning fitness for duty policy and provided answers to the staff questions. Licensee action on previous inspec-tion findings was acceptable.

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DETAILS 1.0 Introduction and Overview 1.1 NRC' Staff Activities-The_ purpose-of this' inspection was to assess licensee act'ivities during the power operations mode and during transition periods as they related to reactor safety, safeguards, and radiation. protection. Within each

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area, the inspectors documented the specific purpose of the area,under review, acceptance criteria and scope of. inspection, along with appro-priate findings / conclusions. The inspectors made this assessment by reviewing information on a sampling basis through actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, or independent calculation and selective review of listed applicable documents.

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1.2 Licensee Activities During the report period, the plant operated at full power with the ex-ception of a three-day outage. The plant was shut down on February 16, 1988, and restarted on February 19, 198 Full power was reached on L February 20, 198 During the outage, repairs were made to the main generator stator cooling system, which had previously showed signs of fouling. The licensee also

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completed several repairs to Once-Through Steam Generator (OTSG) un-isolable valve bonnet leaks and completed repairs to a channel of source range nuclear instrumentation, NI-2.

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As of_8.;3 a.m. on March 5, 1988, TMI-1 was operating at full power with the reactor coolant system (RCS) at normal operating temperature (579 F average) and pressure 2155 psig).

L 2.0 plant Operations (71707, 71715)

l 2.1 Criteria / Scope of Review

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j The resident inspectors periodically inspected the facility to determine I the licensee's compliance with the general operating requirements of Section 6 of the Technical Specifications (TS) in the following areas:

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review of selected plant parameters for abnormal trends;

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plant status from a maintenance / modification viewpoint, including i plant housekeeping and fire protection measures;

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control of ongoing and special evolutions, including control room personnel awareness of these evolutions;

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control of documents, including logkeeping practices;

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implementation of radiological controls; and,

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implementation of the security plan including access control, boundary integrity, and badging practice The inspectors focused on the areas listed in Attachment .2 Findings / Conclusions 2.2.1 Plant Shutdown / Outage /Startup In the early part of February 1988, the licensee noted indication of poor heat removal transfer across the main generator in the nator cooling system. There had been an apparent buildup of copper oxides at the heat transfer surfaces in the main generator stato Subsequent to licensee evaluation of the problem, the licensee started to decrease power to hot shutdown condition at 7:00 p.m.,

February 16, 1938. In conjunction with its vendor, General Electric, the licensee conducted a chemical cleaning and backflush process of the syste Safety-related work was also completed because of the outage oppor-tunity, such as "Furtraniting" steam leaks in the reactor building, cable replacement for one-of-two channels of source range nuclear instrumentation, and leak repair of a reactor water level indication sensing lin On Friday, February 19, 1988, the licensee returned TMI-1 to power operations after the unplanned outage of three days. The generator stator cooling system was cleaned and returned to normal. Th9 cause of the fouling remains to be determined. The plant reached 100 percent power on Saturday, February 20, 198 During the power transition periods, the resident inspectors pro-vided 24-hour inspection coverage. The purpose was to more ade-quately stay abreast of plant status and to assess licensee per-formance during the transition period There were overall good direction and control of the plant shutdown and startup. The plant startup procedures were used and strictly followed. Operators were alert and attentive and they were re-sponsive to alarms that were both expected or unexpected. Overall, the licensee's performance was professional with respect to entry into and out of the "unplanned" outag The inspectors identified no unacceptable condition ,-________- ._ -- - - . _ _ _ .

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2.2.2 Decay Heat River Pump Strainer Foundation Bolts During the evening of February 25, 1988, the Manager of Safety Re-view informed the-senior resident inspector of a problem noted on the foundation support bolts for the "A" loop decay heat removal river. water (DR) system strainer (DR-S-1A). During foundation re-pair work, the licensee found that an approximate 4-inch section of a 36-inch foundation bolt ~ (embedded in the strainer pedestal)

was sheared off, apparently due to corrosion. Subsequent examina-tion of the studs revealed that during a previous modification, the anchor studs were lengthened by approximately four and one-half inches by welding to the existing stud. The resulting weld ex-hibited partial penetration and this caused the stud failur By ultrasonic testing, a similar problem was noted for the remaining three bolts on DR-S-1A and four bolts on DR-S-18. Testing on other river water system strainers did not reveal a similar proble The' licensee considered themselves to be on a 72-hour time clock with respect to T5 3.16, "Shock Suppressors (Snubbers)" as the-bolts provided support for the strainer. The TS also required an engineering evaluation as to the affect on system operability and/or to make repairs as necessar Corporate-based personnel performed that evaluation and on February 26, 1958, concluded that the strainer foundation bolts were not needed with respect to the seismic installation of that H etion of the DR system (strainer and adjacent piping). The licensee provided the resident office a copy of their evaluation, which also included a 10 CfR 60.59 evaluation and this was tentatively found to be ac-ceptable. This is unresolved pending Region I specialist review of this information (289/88-03-01).

2.3 Plant Operations Summary The plant continued to be operated safely. The plant shutdown, outage, and startup activities related to the repair of the main generator stator coolant problem exhibited good planning and overall control, Response to minor plant control system transients was good and the operating staff was able to control the events without major problem .0 Maintenance / Surveillance - Operability Review (61726, 62703)

3.1 Criteria / Scope of Review The inspectors reviewed selected activities to verify proper implementa-tion of the applicable portions of the maintenance and surveillance pro-grams. The inspector used the general criteria listed under the plant operations section of the repor Specific areas of review are listed in Attachment 1. A more detailed review of equipment operability is also addresse _

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3.2 Findings / Conclusions

~3. Surveillance Shift and Daily Checks

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The inspector. verified by direct observation of licensee activities .

that surveillances are being conducted in accordance with Technical-Specification (TS) requirements. The aspects which were reviewed'

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conformance to TS's;

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calibration of instrumentation;

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conduct of the surveillance test;

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adequacy of documentation review;

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accuracy and completeness of test data;

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qualification of personnel performing the test; and,

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performance of the test within the required frequenc The inspector observed the performance of Surveillance Procedure (SP) 1301-1, Revision 69, dated September 4, 1987,."Shift and Daily Checks." The surveillance procedure was found to be in conformance with TS requirements. The conduct of the test was satisfactor Adequate reviews were evident. The results of the surveillance were accurate, complete, and contained the required surveillance accept-

, ance criteria. The personnel who performed the surveillance were L found to be qualified. .By review of surveillance records, it was

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determined that the test was completed within the required frequenc The inspector observed that calibration records of instrumentation were difficult to track. The inspector did not find any instrumen-

, tation which had not been calibrated at the required frequency, but j it is noted that control of calibration records is not uniform for

important-to-safety (ITS) and not important-to-safety (NITS) systems.

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Currently, numerous surveillances exist for calibration of instru-l mentation used in ITS system These records are kept either in

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the control roem or are kept on microfiche in the records department, f

Calibration records for instrumentation in NITS systems (and some ITS) are kept in a library outside the control room. The system,

master test index (MTX), by which these records are kept is not l controlled. Calibration records may, therefore, be retained in j several locations, i

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S In summary, the performance of the. surveillance was within require-ments. The personnel performing the test were knowledgeable and-qualified. Although difficulty was experienced in verifying cali-bration of instrumentation, all instrumentation chosen to be veri-fied was found to be properly calibrate . Inservice Testing Activities The inspector witnessed surveillance being performed on motor-driven emergency feedwater pumps EF-P-2A/ The surveillance was performed on February 12, 1988, in accordance with Surveillance Procedure (SP) 1300-3F to ensure compliance with Technical Specifi-cations, Section 4. This requires testing of pumps and valves in accordance with Section XI of ASME (American Society of Mechani-cal Engineers) boiler and pressure vessel code, as well as 10 CFR 50.55 a(g).

Recording of the~ test data, procedure compliance, and communication  ;

between the personnel at the pump site and the control room was  :

thorough and effectiv .3 Operability Summary

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Maintenance and surveillance activities continue to be accomplished in a satisfactory manner and housekeeping is adequat 'i 4.0 ' Corporate Inspection / Briefing (Regulatory Guide 1.97) (37700) I F

. 4.1 Background / Scope of Review During the upcoming outage (7R), presently scheduled for June 1988, the licensee plans to complete various modifications, calibrations, and licensing actions to meet commitments concerning the provisions of RG i 1.97, "Post-Accident Monitoring Instrumentation." The inspectors were briefed by corporate licensing and engineering personnel on February 23, 1988, on the status of completed modifications, planned modifications, i and anticipated licensing actions concerning RG 1.97. The inspectors  ;

reviewed various licensee /NRC correspondence concerning this issue in ,

order to assure that the licensee was adequately prepared to complete t l the required actions to ensure compliance with commitments for RG 1.97 ,

at this time, f 4.2 Findings The license- had previously completed several modifications during the *

6R outage taat related to RG 1.97. Those included: the Reactor Coolant Inventory Tracking (RCITS) and void function monitoring system, effluent  ;

monitoring from the main condenser offgas system, safety grade condensate  ;

water storage tank (CST) level indicating system, and redundant OTSG i pressure indication in the control room.

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These modifications have been evaluated by the NRC staff in previous inspections. The only major concern remaining is the acceptability of the accuracy of the CST level indicating system. The NRC staff had a concern that the level indicators give a lower than actual reading when the EFW pumps are running due to flow induced pressure changes at the point where the level is monitore The licensee has addressed this issue and has evaluated the approximately 3/4-foot lower than actual reading as acceptable and has documented this review in a letter to NRR, dated June 29, 1987. Remaining action of this item consists of NRR staff approval via a supplement to the Safety Evaluation Report (SER) for RG 1.9 Several open questions remain concerning five other design issues raised during the initial NRR SER. Those are:

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accumulator tank level or pressure;

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pressurizer level;

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pressurizer heater status;

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containment sump water temperature; and,

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containment heat removal system instrumentation classificatio The issues surrounding the acceptability of the type of indication for the safety injection accumulators and the containment sump water tempera-ture are being evaluated by NRR on a generic basis. The acceptability of the licensee response will be based on this review. The type of in-dication available for pressurizer level and whether or not it is tem-perature compensated with environmentally qualified instrumentation is still in question. The type of instrumentation to be used for pressuri-zer heater status is still unresolve NRR is presently requiring in-struments to measure current draw vice just using heater switch positio The last issue involves the type of instrumentation used for monitoring the performance of the reactor building cooling river water system, which is presently Category 3. NRR is presently requiring Category 2 instru-ment The licensee is relying on containment temperature / pressure to monitor containment heat removal performanc During 7R, the licensee plans to complete the following modifications:

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Reactor Coolant Drain Tank (RCDT) temperature range extension;

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Reactor Coolant System (RCS) pressure range extension;

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redundant wide range neutron flux monitoring (NI-11/12); and,

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_ control room indication (qualified) for low pressure indication (LPI) flow, high pressure indicator (HPI) flow, and borated water storage tank (BWST) leve "

During 8R.the licensee will complete modifications for RG 1.97. :This will consist of providing "on demand" recordings of several important plant variables.via the new plant compute The inspectors have reviewed-the system design descriptions for the new instrumentation and instrumentation modifications. Those areas will be examined as they aro completed in the upcoming outag The inspectors are presently tracking three unresolved items related to RG 1.97. These are: (1):289/85-21-03, which is the NRR review of design adequacy of the new non-nuclear instrumentation (NNI); (2) 289/87-09-03, condensate storage tank (CST) level indication oscillations; and, (3) 289/87-09-06, adequacy of differential / pressure (DP) instrumentation used in high pressure. service. The resolution of these issues is await-ing NRC staff review.as the licensee has completed their action on these item .3 Conclusion The licensee is aware of the remaining issues to be resolved by NRR prior to achieving full compliance with RG 1.97. Modification status for those items to be accomplished during 7R appears to be on schedule.

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5.0 Safety Issue Management System Item Verification 5.1 Introduction

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l The inspector verified proper implementation of licensee actions related

to the below-listed NRC Safety Issue Management System (SIMS) ite The generic inspection approach for the SIMS item was:

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research various licensee and NRC correspondence, including safety i evaluation reports (SER's) to identify key assumptions, commitments, l or other licensee actions to be taken to resolve the safety issue;

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identify an additional items which need to be verified as delineated

in the related NRC Temporary Instruction or other inspection proce-l dures; and, i

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verify proper implementation of the items planned above.

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5.2 Generic letter 83-28, Reactor Trip System - Vendor-Related Modifications

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(SIMS No. 75 (B-80)) (25591)

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5.2.1 Background / Review

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The Generic Letter 83-28, Item 4.1 required that all vendor-recom-

-mended modifications on the reactor trip breakers be reviewed to

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verify that each modification has, in fact, been implemented or a written evaluation, including justification for not performing the modifications,. exist .2.2 Findings / Conclusions The reactor trip breakers utilized at TMI-1 are of type AK-2-25 and AK-2-15 manufactured by General Electric (GE). A GPU Nuclear Cor-J poration (GPUN) letter of November 8,1983, references a GE letter

to Babcock and Wilcox (B&W), dated September 7,1983, concerning vendor-recommended modifications on reactor trip breakers. GE did not recommend any modification to their subject breakers. The let-p- ter recommended that the shunt trip should be actuated simultane-ously with the undervoltage trip mechanism for the reactor trip function. This function was recommended by Generic Letter 83-28, Item Subsequent to this, the B&W Owners Group (BWOG) undertook a progrcm to determine long-tern actions to improve reliability of the reactor trip breakers; in particular, the undervoltage trip device. The ensuing recommended action, documented in their April 8, 1985, let-ter to the NRC, included two modifications to the reactor trip breakers. The BWOG recommended that the utilities replace the trip shaft and latch roller bearings with bearings lubricated with Mobil 28 and that Mobil 28 lubricated bearings be used exclusively in GE AK-type reactor trip breakers. The second recommendation was the addition of a reactor protection system (RPS) type signal to the d.c. shunt trip device. A GPUN letter, dated August 23, 1985, endorses the recommended equipment upgrade and stated that they were i

implemented in THI- The inspector reviewed implementation of the first recommendation (about Mobil 28 lubricated bearings) at TMI-1. The licensee stated that the existing (installed and in stock) reactor trip breakers were shipped to the vendor for bearing replacement. GE performed the replacement work and B&W coordinated the replacement work, in-ciuding necessary qualification and certification of the modified breakers. The bearings were not easily accessible for inspectio However, based on discussions with licensee personnel and review

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of the purchase order associated with this modification, the in-spector concluded that the work was performed as require .

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The second BWOG recommended modification was reviewed and found to be acceptably completed by the NRC under Generic Letter 83-28, Item The inspector also reviewed *.fo vendor aavisories issued by GE for the AK-Series reactor trip eakers during 1985. Advisory No. 300, dated September 26, 1985, deals with: (1). improper application of paint on the mating surface of the-armatures and pole pieces for certain undervoltage trip devices made between mid-1978 and May 1985; and, (2) insufficient clearance between the armature and the under-voltage device mounting studs for devices with assembly number 568B309G. Licensee's follow-up actions for advisory were reviewed under 10 CFR Part 21 report (289/85-PT-07) and was found satisfac-tory in Inspection Report No. 50-289/87-23, dated January 27, 198 The second GE advisory (number 9.21, dated February 7, 1985) deals with the.use of improperly heat treated shunt trip paddles in the shunt trip assembly on AK-Series reactor trip breakers. The sus-pected trip paddles were used in production lots from February 1983 through April 1984. The affected devices include AK-Series reactor trip breakers which have code dates from F307 ta F419, replacement front frame assembly series which included the shunt trip assembly and replacement trip paddles.

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The inspector discussed the advisory with licensee personnel. The licensee stated that the reactor trip breakers were all purchased during the early 1970's. The TMI-1 breakers have a bill of mate-

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rial date of March 27, 197 The TMI-2 reactor trip breakers, now utilized in TMI-1 as spares, were also purchased before 1979. The reactor trip breakers were purchased before the date of concern.

l The trip paddle problem did not exist in the originally purchased j reactor trip breakers. The licensee did not believe that shunt trip paddles were purchased as spare parts during the time of concern.

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The licensee was asked to confirm that the shunt trip paddles of

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concern were not introduced in TMI-1 as spare parts for the reactor trip breaker The licensee reviewed machinery history files to verify that no shunt trip paddles were replaced before, during, or after the dates I

mentioned in the service advice letter. All shunt trip paddles presently installed were verified to be original equipmen The shunt trip paddle is not kept in stock at TMI-1. Based on the above, this item is close .3 (Closed) Safety Issue Management System Item (289/MPA-B-66): Natural Circulation (NC) Cooldown (71707)

l The NRC Inspection Report No. 50-289/88-01, paragraph 5.3, documented the verification of licensee actions related to the subject multi plant action item (MpA-B-66). There was a residual concern with respect to

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the adequacy of instructions to the operators on void formation at the reactor vessel upper head (RVUH) during emergency conditions (ATOG -

Anticipated Transient Operator Guidelines).

During this inspection period, the inspector discusseJ the concern with licensee engineering personnel at the corporate headquarters on February 23, 1988. Licensee reoresentatives stated that the AT0G procedures are adequate with respect to NC cooldown because they are symptom-oriented procedures. The head bubble itself is not a safety concern; its inter-ference with core decay heat removal is a concer The ATOG procedures provide for indication if a loss of decay heat removal (namely, the loss of thermal coupling between steam generator and the reactor) and these procedures treat those symptoms -- loss of decay heat removal -- rather than the event void function at RVU The licensee representatives also indicated that they had recent research ,

results that indicate a cooldown rate of 100 F/ hour for a steam generator tube rupture does not result in a head bubble or a bubble in the unaf-fected loop assuming proper Reactor Coolant System (RCS) pressure contro The inspector views the licensee's position on the ATOG procedures as reasonable. The NRC staff shares the licensee's concern about putting too much information in the ATOG procedures such that they are no longer symptom oriented (in distinction to event oriented). Also, based on the last inspection, it appears that the training program addresses the RVUH bubble formation concern and it appears that operators are aware of the problem. The inspector confirmed that Abnormal Transient Procedure (ATP)

l 1210-1, Revision 15, dated January 4, 1988, "Reactor / Turbine Trip,"

coupled with ATP 1210-2, Revision 8, dated June 13, 1986, "Loss of 25 F Subcooled Margin," and 1210-4, Revision 10, dated December 17, 1987,

"Lack of Primary to Secondary Heat Transfer," addresses indications of i and the necessary actions to correct a loss of decay heat removal.

l I In resolving this item, the inspector places no reliance on the above-l noted research results as stated by the license The inspector had no safety concerns on this issu .4 TAP II.E.1.2 (EFW Upgrade) (71710)

l l 5. Background / Scope of Review

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l The inspector reviewed the status of ooen items concerning the re-l cently completed Heat Sink Protection System (HSPS) installation i and EPd system upgrade to safety status. During past inspections,

many unresolved issues have been generated that affect the deter-l mination of the acceptability of licensee compliance with the re-l quirements of Task Action Plan (TAP) Item II.E.1.2. At the begin-

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ning of the review, there were twenty-five open issues which are l

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tabulated in Attachment 2. Some of these items were closed in this repor The status of each item and which NRC organization is re-sponsible for review.is also note . Findings / Conclusions

.. During this' inspection period seven of these. items were. closed and are noted in Section IV on Attachment 2. Discussions with the lic-ensee have noted that several of the remaining open items listed in Sections II and III are ready for review and that licensee action has been complete The insoector will review licensee action on these items in future inspection For the items to be reviewed by NRR listed in Section I, NRR has committed by memorandum from B. Boger to W. Kane, dated January 4, 1988, to complete a review of these items by May 31, 198 In addition to these previously identified open items, the inspec-tors expressed a concern about some other problems with the recently installed HSPS syste One problem is the inability of the control room operators to read all four channels of both operating (0P) and startup (SV) range OTSG 1evels. At present, for each 0TSG, only two of four channels of each range (SU and OP) are available in the control room. Since the channel being read in the control room is automatically selected by a logic system, confusion can exist as to what the OTSG level actually is and which channel is controlling automatic functions. This has caused at least one automatic initi-ation of emergency feedwater (EFW) in the past. The licensee has proposed some modifications to resolve this problem. The inspectors were briefed on these proposed modifications during the corporate inspection on February 22 and 23, 1988. The licensee intends to install a modification to improve the HSPS signal selection to the control room indicators. This modification is part of an overall upgrade for the Integrated Control System (ICS), which involves installation of Smart Automatic Signal Selection (SASS) modules in the IC The licensee also intends to install well labelled plug-type in-strument jacks in the HSPS cabinets to promote more reliable readout of the various OTSG 1evel sigrials from all four channels. The cross-check is presently accomplished by Instrument and Control (I&C) personnel, using test probes on terminal block, and has the potential to cause inadvertent shorting across adjacent circuit Prior to final resolution of this SIMS item, the remaining unre-solved items listed in Attachment 2 must be reviewed. The licensee has completed their action on these remaining open items. The NRC staff is tentatively planning on completing all reviews by July 198 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

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At~present, no conditions adverse to safety that would affect the presert safety grade status of the EFW/HSPS systems have been iden-tifie .5 SIMS Summary For the SIMS issues reviewed above, the licensee appropriately translated generic licensing actions into plant specific action Specifically, they, properly incorporated requirements and commitments into procedural requirements and/or training plan / exercis .0 Fitness for Duty Program / Survey (50-289 and 320/88-TI-01)

The NRC Region I issued Temporary Instruction (TI) No. 88-1, dated January 21, 1988, "Fitness for Duty - Drug Testing Information and Reporting." It reouired resident inspectors to obtain licensee response to certain cuestions about the licensee's Fitness for Duty Program. The Director of TMI-1 re-sponded to these questions and the answers were provided to Region I staff by separate memorandu This action closed the subject Regional Instruction (289 and 320/88-TI-01).

7.0 Licensee Actions on Previous Inspection Findirgs Introduction For these items listed below that were previously identified violations, the inspector reviewed the licensee's response and corrective action to:

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verify the licensee responc'ed in a timely manner;

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verify measures taken to correct the item and avoid recurrence were completed and within the specified time frame; and,

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verify licensee commitments were complete .2 (Closed) Unresolved Item (289/87-06-01): Labeling of HSPS Cabinets ,

This item concerned the possible Operator confusion that might arise due to the similarity of the multiple test switches in the various Heat Sink Protection System (HSPS) logic cabinet The licensee subsequently added additional labels to the outside cabinet doors and also added train /

channel designators to the inside test switch cover panels. The indi-vidual switches have all been changed to red (Train A) and green (Train B) colored labels to further eliminata confusion. As the HSPS surveil-lance procedures have been accomplished routinely and without incidents for the past year, it appears that the present labeling system is ac-ceptabl The inspector had no other concerns on this issue.

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7.3 (Closed) Unresolved Item (289/87-06-03): Verify Preventive Maintenace is Scheduled for Emergency Feedwater System Flow Control Block Valves EF-V-52A/8/C/0 This item involved the new manually operated block valves EF-V-52A/B/C/D that replaced the previously motor-operated block valves EF-V-52, 53, 54, and 55_in the emergency feedwater (EFW) systems. These valves are used for maintenance isolation and to manually isolate the EF-V-30 valves (flow control valves). It was noted that no preventive maintenance was

, in place for these valves. The licensee subsequently scheduled these valves for a three year open and inspect cycle commencing in 1990. Pro-cedure E-13 is being developed for this purpose, but it is not complete l at this time.

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The inspector verified that the valves are presently scheouled for main-tenance on the computerized system used by the licensee to schedule all plant maintenance. Additionally, the valves are being cycled during the monthly inservice testing (IST) for the motor-driven EFW pumps per SP 1300-3F. No problems have been noted with these valves. The inspector had no other concerns with this issu .4 (Closed) Violation (289/87-06-06): Failure to Properly Review and Verify a Design Calculation Associated with Once-Through Steam Generator Low Level Actuation Setpoints This violation concerned a licensee change made to a contractor design calculation No. 0370-129-001, Revision 0, which was then not properly reviewed by the contracto The licensee contractor subsequently reviewed tFe calculation which was design verified prior to use of the calculatio The licensee corrective action consisted of this correction in addition to counseling the per-sonnel involved. The intractor reviewed design calculations for other licensee modifications to ensure that they were properly reviewed and that changes were also properly reviewed. Two calculations, C8708-21 and C8706-021, both for Regulatory Guide (RG) 1.97 work, were reviewed and design verified by the contractor. No problems were note The inspector discussed with various licensee engineering personnel the method by which calculations were* changed. All personnel were aware of the requirements to review and design verify the calculations when changes are made. The inspector had no other concerns on this issu .5 (Closed)_ Unresolved Item (289/87-06-07): Heat Sink Protection System Testing This item concerned the initial testing of the HSPS logic system prior to use af ter the startup from the 6R outage. The licensac opted to per-form preliminary surveillance pro:edures via the Special Temporary Pro-cedure (STP) process in order to verify the operability of the HSPS system logic. The inspec, ors witnessed portions of that testing in March

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1987. Subsequently, the inspectors reviewed the completed STP No , 0017, 0018, and 0019 and also reviewed subsequent completed version of the formal SP's 1303-11.36, 11.37A/8/C/D 11.37, 11.38, and 11.39. These procedures have been performed on a quarterly basis since startup in March 1987. Prior to startup, the licensee reviewed the data from the completed STP's and declared the HSPS operable. Based on tr,a inspector review of the above procedures and witnessing of selected por-tions of the formal surveillance procedures without problems at various times during the past ten months, this item is close .6 (Closed) Unresolved Item (289/87-06-10): Licensee Completed Audit on Design and Modificatien Control At the time of that NRC inspection, the licensee was in the process of completing a major audit (No. 0-TMI-86-11). The audit was extensive in that it spanned several months (September 8, 1986 - March 12, 1987) and it's purpose was to determine whether the NRC Performance Appraisal Team I (PAT I) inspection findings were generic to other Cycle 6R modifica-tions. One of the modifications included in the review was the Heat Sink Protection System (HSPS), which is the safety grade initiation and con-trol system for emergency feedwater. The licensee selections of HSPS :

and other modifications were representative of major work completed in 6 The audit report was issued April 9, 198 The summation finding of the audit was that there was no additional generic programmatic or technical issues. However, the audit report had some regulatory and technical issues of interest to the inspector so licensee corrective or follow-up actions were reviewed for adequac The audit identified two audit findings, three quality deficiency repcrts, two preliminary safety concerns (PSC) and twelve recommendations. The findings, deficiencies, and recommendations were appropriately acted upon and verified to be completed by Quality Control (QC) inspectors. The PSC's were also appropriately addressed, but certain actions remaine The PSC No.86-006, dated September 11, 1986, identified, in part, a concern about the nuclear services closed-cycle cooling water (NSCCW)

system in terms of withstanding a seismic event. Specifically, the surge tank fill valve has a manual isolation valve and there was an upstream Seismic I boundary. The licensee had tentatively decided to install a Seismic I check valve to retard leakage on upstream pipe breaks due to a seismic event until operator actions could shut the manual isolation valve. The check valve is scheduled for installation in this refueling outage, July 1988. Also, an internal engineering memorandum indicated that a plant-wide review would occur by February 29, 1988, to identify similar problems. This PSC remains open and the inspector established confidence that the licensee was heading toward proper resolution of this

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item. The inspector had no additioral comments.

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7.7 (Closed) Unresolved Item (289/87-13-02): NI-2 Replacement-in-Kind Issue This item involved a modification where NI-1 cable was replaced with one which was not the same as the original. The replacement-in-kind review was performed as require However, additional technical data was re-quired to clarify the justification for the NI-1 cable replacemen The licensee committed to do so during the similar replacement of cable on NI-2 instrumentatio During the last unplanr.ed shutdown from February 16-19, 1988, the NI-2 cable was replaced under Job Ticket (JT) CP-025, dated October 22, 198 The work involved replacement of the cable and the associated connectors from the pre-amp box to the NI-2 detector inside the reactor buildin Plant Engineering Evaluation 88-007-E, dated February 1, 1988, satisfac-torily provided the requested technical justification. Tre inspector reviewed the engineering evaluation as well as the JT. The inspector reviewed the adecuacy and completeness of the technical data, job plan-ning, testing, and acceptance by the operational staff and concluded that the licensee has satisfactorily resolved this ccncer .0 Exit Interview The inspectors discussed the inspection scope and findings with licensee man-agement an an interim exit meeting at the corporate office on February 23, 1988 on certain SIMS items and at a final exit meeting conducted March 4, 198 Senior licensee personnel attending the final exit meeting included the fol-lowing:

J. Colitz, Plant Engineering Director, TMI I H. Hukill, Director, TMI-1 C. Incorvati, TMI-1 Audit Manager R. Knight, Engineer, TMI-1 M. Ross, Plant Operations Director, TMI-1 The inspection results as discussed at the meeting are summarized in the cover page of the inspection repor Licensee representatives did not indicate that any of the subjects discussed contained proprietary or safeguards informatio Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable, violations, or deviation Unre-solved items discussed during the exit meetirl a e addressed in Section 7.

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AT,TACHMENT 1 NRC INSPECTION REPORT NO. 50-289/88-03

ACTIVITIES REVIEWED Plant Oper, .ans

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Control room operations during regular and back shift hours, including fre-quent observation of activities in progress and periodic reviews of relected sections of the shift foreman's log and controt room operator's log and selected sections of other control room daily log Areas outside the control roo Selected licensee planning meeting Reactor shutdown -- reactor startup activitie During this inspection period, the inspectors conducted direct inspections during the following back shift hour Day /Date Time 2/7/88 8:45 a.m. - 11:15 /15/88 8:00 a.m. - 12:00 Noon 2/16/88 4:00 p.m. - 12:00 Midnight 2/19/88 11:30 p.m. - 7:00 /20/88 9:30 a.m. - 10:30 Maintenance / Surveillance

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.? 1300-3F, Revision 30, effective March 20,1987, "Motor-Driven EFW Pump IST."

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JT CP-869, Repair of Condensate System Valves C0-V-16A/B Wor SP 1300-3H A/B, Revision 24, effective January 15, 1988, "Makeup Pump and Valve Functional Tests."

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SP 1302-3.1, Revision 49, effective January 5, 1988, "R.M.S. Calibration."

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JO 80842, Decay Heat River Pump Expansion Joint Wor Reactor Coolant System (RCS) Leak Rate

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The inspector selectively reviewed RCS leak rate data for the cast inspection period. The inspector independently calculated certain RCS leak rate data reviewed using licensee input data and a generic NRC "BASIC" computer program "RCSLK9" as specified in NUREG 1107. Licensee (L) and NRC (N) data are tabulated below.

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.' Attachment 1 2 TABLE RCS LEAK RATE 0ATA All Values GPM DATE/ TIME (NUREG 1107) CORRECTED DURATION L L G "O "U "U U 02/10/88 0.3366 "0.33 0.05 0.15 0.1601 12:40 Hours 02/15/88 0.2881 0.29 0.05 0.15 0.1470 1:14 Hours 02/20/88 0.2198 0.22 0.32 0.22 0.2242 9:58 Hours 02/27/88 0.4366 0.44 -0.06 0.04 0.0405 5:07 Hours 03/06/88 0.4383 0.44 -0.05 0.054 0.0531 1:10 Hours G = Identified gross leakage U = Unidentified leakage L - Licensee calculated N = NRC calculated Columns 2 and 3, 5 and 6 correlate + 0.2 gpm in accordance with NUREG 1107. N u

is corrected by adding 0.1044 gpm to the NUREG 11079 N due to total purge flow through the No. 3 seal from RCP' . ,

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ATTACHMENT 2 EFW/HSPS UNRESOLVED ITEMS (II.E.1.2) Unresolved Items to be Reviewed by NRR Reference: October 21, 1987, Memorandum W. Kane to F. Miraglia 86-03-08 - Breaker Coordination Study 86-12-12 - Pump Overcurrent Protection 86-12-13 - 115 Volt a.c. Circuits 86-12-14 - Minimum Motor Start Voltages 86-12-15 - Grounding Practices 87-06-08 - Mechanical / Structural Systems Review 87-06-09 - Electrical / Instrument Control (IC) Systems Review I Unresolved Items Requiring Region I Review 86-05-05 - EFW Nozzle / Cracking 86-12-17 -

RSP - OTSG 86-19-02 - Seismic LevelStudy Interaction Indication (Electrical Isolation)

Review 87-02-03 - Diesel Generator Loading (0APPER)

87-10-01 - Cable Grounding Practices III. Unresolved Items for Resident (On Site) Inspector Review 85-20-01 - Residual EFW Concerns from Restart Review 86-03-22 - Back-Up instrument Air Quality 36-12-10 - Modification Review of EFW Pump Recirculation Valves Block 87-06-02 - ATOG Procedure Enhancements - OTSG High Level Main Feedwater Isolation 87-06-05 - Feedwater Pressure Switch Calibration Orift I Unresolved items Closed in 289/88-03 or 289/88-06 86-12-09 - IST for Three-Way Back-Up Instrument Air Valves 86-21-04 - MS-V-9A/B Inspection Frequency 87-06-01 - Labelling of HSPS Cabinets "

87-06-03 - Preventive Maintenance for EF-V-52 Valves 87-06-07 - HSPS Testing Prior to Cycle 7 Startup 87-06-10 - HSPS Modification Auuit 87-06-06 - Violation Concerning HSPS Calculation Review / Approval

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