IR 05000289/1987024

Insp Rept 50-289/87-24 on 871206-880109.No Violations or Deviations Noted.Major Areas Inspected:Plant Operational Items,Reactor Bldg Emergency Cooling Safety Sys Alignment & Health Physics Technician Allegation

ML20149M501
Person / Time
Site:	Three Mile Island
Issue date:	02/17/1988
From:	Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20149M482	List: IR 05000289/1987024 ML20149M481
References
50-289-87-24, IEB-86-002, IEB-86-003, IEB-86-2, IEB-86-3, NUDOCS 8802260081
Download: ML20149M501 (15)

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O.S. NUCLEAR REGULATORY COMMISSION

REGION I

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Docket / Report Nr. 50-289/87-24

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Licensee: DRp-50

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Licensee: GPU Nuclear Corporation #

P. O. Box 480 i

Middletown, Pennsylvania 17057

Facility: Three Mile Island Nuclear Station, Unit 1  !

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Middletown, Pennsylvania 1 '

i Dates: December -6,1987 - Janutry 9,1988

Inspectors: " Conte, Senior Resident Inspector (TMI-1) l'

D. Johnson, Resident Inspector (TMI-1)

Accompanied by: S Peleschak, Reactor Engineer, Region I (RI)

A. Sidparal Resident Inspector (TMI-1)

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• Reporti nspector Approved by: /NN JE _

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, C. Cowgil4, Chi g Reactor Projects Section lA Inspection Summary:

) Areas Inspected: The NRC staff conducted safety and safeguards inspections (66  !

1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) during power operations. Plant operational items reviewed were: diesel l j generator exhaust manifold fire; housekeeping; equipment foundation bolting; ex- l l pansion joint installation; steam generator level instrumentation repair; and, '

, reactor building emergency cooling safety system alignment. Items in other func-

} tional areas included health physics technician qualification ellegation and reac- i tor coolant pump trip criteria. Also reviewed were licensee act1 ns 9 in response !

to NRC Bulletin Nos, 86-02 and 86-0 *  !

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Results: The licensee's substantial efforts to monitor and maintain the plant is  !'

i reflected in full power operation throughout the inspection period. However, cer-i tain problems were noted that did not have impact on full power operations. Most '

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significant was an inspector-identified corrective n'ntenance error in the in-j stallation of the wrong expansion joint in the "B" loop of the reactor building

! emergency cooling water syster. Although the licensee justified continued power i j operations through review and analysis, the root ch of this error will be ex- l j amined later. Inspector observations in housekeepi odicated that some improve- I

! ments are not betog aggressively pursued by the lice .ee . In general, maintenance  ;

j and surveillance implementation were supportive of operation J i

i In the plant operaM ons area, the safety-related reactor building emergency cooling l

] system was properly aligned to standby status. In general, procedures were pro- '

J perly implemented. The licensee effectively translated commitments and require-1 ments for the reactor coolant pump trip criteria issue into procedural requirements l j and training needs which were properly implemented (SIMS Item 67.4.1).

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l Licensee response and action related to the following items were adequate: NRC Bul-

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letin Nos. 86-02 and 86-03 and an allegation on a health physics technician quali-j ficatio i PDR ADOCK 05000209  :

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OETAILS 1.0 Introduction and Overview 1.1 NRC Staff Activities The overall purpose of this inspection was to assess licensee activities during the power operations mode as they related to reactor safety, safe-guards, and radiation protection. Within each area, the inspectors docu-mented the specific purpose of the area under review, acceptance criteria and scope of inspections, along with appropriate findings / conclusion The inspectors made this assessment by reviewing information on a sarrp-ling basis through actual observation of licensee activities, interviews l with licensee personnel, measurement of radiation levels, or independent calculation and selective review of listed applicable document .2 Licensee Activities During this period, the licensee operated the plant at full powe .0 Plant Operations 2.1 Criteria / Scope of Review L

The resident inspectors periodically inspected the facility to determine 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 paraneters for abnormal trends;

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plant status from a maintenance / modification viewpaint, including 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 Most of the inspectors' emphasis during this inspection period was in i the maintenance and surveillance areas. Findings are delineated in the next section of this repor l

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U f , o, f ' 4 No' co'nditfons adverse to safety were noted in the plant operations-are '

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The good operating record' appeared to be due to licensee's substantial

efforts to maintain and monitor the-plant. For the components' reviewed, a /, ~

y sa fety-related_equiprent remained operable and in proper sa.fety alignmen J .For the areas reviewed, facility procedures.were properly implemented;

, , , ., Site Wpper management and the quality assurance den M ment continued

, .their detailed. attention and in'61vement v in routine' activities.

.T 4 3.0 M_ajhtenance/ Surveillance - Operability Review

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'3/1" piteria/ Scope of Review s ., ,

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The' ins'pectors reviewed selected activities to verify proper implementa-

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tiyn of the applicable portions of the maintenance and surveillance pro-grams. The inspector used the general criteria listed under.the plant operations sectiep of this report. ~ Specific areas of review are. listed i' 2 in Attachment 1-. A more detailed-review of equipment operability.is also

, addressed bs M .2 Findings / Conclusions 3. Housekeeping During this period, several inspections were conducted in order to assess licensee's housekeeping. While the overall quality of housekeeping continued to-be'high, several areas were.-iden-tified where it can be further improved. For example: (1).the circulating water bui1 ding has water leaks at several places; (2) the intake screen' and pump house also has water leaks, causing corrosion degradation of equipment foundation (the lic- i ensee is currently upgrading these foundations);. and, (3) the sump pump area in the. heat exchanger vault of the auxiliary building had similar problems due to frequent' intermediate closed cycle cooling system' draining last year because of the letdown cooler leakage proble Also, during these observations, some radiological deficiencies

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were noted in the auxiliary. building. These 1ncluded:.(1) con-taminated face-masks were stored in improper containers and/or did not have a radiation survey tag; and, (2) one bag contain-ing contaminated clothing lacked an identification tag and l radiation survey dat These deficiencies did not cause any significant safety-hazar .

They were reported immediately to the shift personnel. The licensee was responsive in correcting, as well as planning, appropriate actions. However, these observations indicate-that longstanding deficieni areas are not being aggressively pursue :

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3. Foundation Bolts During the routine inspections of the facility, the inspector found: (1) scaffolding in the spent fuel pump area that had some loose bolts even though it had been inspected a few days earlier; (2) the baseplate bolts on the "B" d.c. diesel fuel transfer pump were either missing, loose, or apparently of improper size; (3) three of the six baseplate bolts on one of the sump pumps in the heat exchanger vault of the auxiliary building were missing; and, (4) one main steam line tie rod support in the intermediate building was loos The licensee corrected items (1) and (4). The equipment noted in items (2)

and (3) did not involve safety-related applications. Licensee management acknowledged these finding The inspectors had no additional comment .2.3 Expansion Joints on the Reactor Building Emergency Cooling System Pumps The expansion joints on the reactor building en,ergency cooling system pumps were installed during September of 1987. During the NRC inspections of the intake screen and pump house early in this inspection period, a discrepancy in the shapes of both joints was noticed by the inspector. The licensee was advised of the problem immediately. Based on licensee's preliminary assessment of the problem, an error was made in the installa-tion of the joint for the "B" pum The licensee provided an engineering evaluation on December 18, 1987, which supported continued plant operation. This evaluation noted that the maximum system operating pressure is approximately 138 psig, the design and hydrostatic test pressure of the joint is 100 psig, and the burst pressure of the joint is 400 psig. One of the deficiencies identified by the inspector is that the original design criteria for the system requires hydrostatic testing at 225 psig. The joint was tested to 100 psig. The manufacturer, however, has assured the licensee that since the burst pressure is about three times the maximum system operating pressure, the joint integrity is not compromised. Also, the licensee expects w receive con-firmatory data from the manufacturers on the hydrostatic test nressure The licensee reviewed the applicable Final Safety Analysis Re-port (FSAR) section. The FSAR does not address the design criteria for the expansion joint This was independently verified by the inspector. The licensee, however, has agreed to perform a 10 CFR 50.59 review to address the design criterie questio ;

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The licensee has-initiated a special' review to determine.wh the installation error occurred. The~results' are expected by mid- January 1988. -The resident inspectors will review this area again when the-additional licensee information is avail-abl .2.4 Once-Through Steam Generator Level ~ Transmitter Repair The inspector reviewed the circumstances surrounding the fail- ,

ure of the power supply for two Once-Through-Steam Generator (OTSG) level transmitters (LT-1044/1046). .On December 22g.1987, the failure of a transistor and resistor in a common current-to-voltage converter in the power supply for both instruments occurred during transfer of the power supply for the instru-ment Plant operations / maintenance personnel were transfer-

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ring the power for the vital bus, which supplies the instru-ments, from the "E". vital inverter-to the "A"' vital inverte The Heat Sink Protection System (HSPS) logic bypass switches -

for several instruments in the HSPS cabinets were in bypast per Operating Procedure (0P) 1107-2, Revision 52, dated Sep-tember 25, 1987. The "A" inverter had previously experienced '

synchronization problems and had been repaired. The "E" in- ,

verter had been used in the interim. Upon re-engerizing the '

instruments by placing the HSPS bypass switches to normal, it was noticed that for LT-1044 and 1046, ind;vidual channel actuations had occurred, indicating low OTSG 1eve Plant maintenance personnel noted that a low output from LT-1044/1046 existed. Subsequent investigation revealed-the -

failed transistcr/ resistor in a d.c., chopper circuit and an open fuse, F-3, in the same circui The power supply module was replaced by Job Ticket (JT) CP-467 on December 22, 1987, and LT-1944 and 1046 were returned to service. The licensee has been in contact with the vendor (Foxboro) to attempt to determine the cause of the failur As yet, no positive determination has been mad The inspector reviewed JT CP-464 for the inverter repair and JT CP 467 for the HSPS level transmitter power supply repai No discrepancies were noted. The-inspector also reviewed HSPS 1 weekly surveillance data that is taken per Surveillance Proce- I dure (SP) 1301-4.1, "Weekly Surveillance Checks." The data for the "blind" HSPS level transmitter channels was reviewed for the past several months, sirce the failure, of LT-1047 on ,

September 19, 1987. No discrept Mies were noted and the in- l spector had no safety concerns on this ite i l

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3. Surveillance of-Fire System Diesel Batteries The inspector ivitnessed performance.of weekly surveillance on--

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the fire system diesel batteries;1.specifically, thefbatteries'

for'the fire pumps .in the circulating water building,,as well asLthe intake-screen and pump house butiding. .The purpose:for'

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witnessing this activity was to assess procedural compliance, adequacy 'of the procedure,' calibration of. test equipment ..

, training of the personnel involved, And documentation'.- The-

. procedure. involved was 3301-W2, Revision 8,. dated December 3, 1987. The surveillance was also witnessed by? licensee's QC

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staf The surveillance activity was properly implemented in' accord-ance with the approved procedure. 'Even though it.is.no technical specification requirement, more specific information to establish the water level in the batteries would be an enhart.ement.

i The QC participation was quite effective. For example,' the

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QC inspector pointed out the' proper method for correcting the~ ,

recorded data and was very attentive to assure the surveillance ,

was performed in compliance with the procedur .3 Operability Summary The inspectors' observations ~ in.the area' of housekeeping and preventive maintenance (foundation bolt discrepancies) indicated that improvements were not being aggressively pursued. Further licensee and NRC1 review

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is warranted on the reactor building emergency cooling system expansion joint replacement. The licensee was responsive to identified deficien-cies either noted by themselves or by'NRC inspectors. Also, maintenance

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and surveillance activities were supportive of plant operation .0 Reactor Building Emergency Cooling Water System / ment Review  ;

4.1 Review ,

The inspector reviewed Operating Procedure (0P)-1104-38,' Revision 16,- '

, dated June 26, 1987, "Reactor Building Emergency Cooling Water System,"

to ascertain whether it is in accordance with regulatory requirements (ANSI 18.7-1976) and whether its technical adequacy is consistent with desired actions and modes of operation. As part of this review,.the~

inspector also examined the following document GAI Drawing C-302-202, Revision 30, "River Water System"

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GAI Drawing C-302-203, Revision 27, "Screenwash and Sluice System" r

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r GAI'Orawing C-307-610, Revision 40, "Nuclear Services Closed Cycle

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-Conling Water"

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CP 1104, Rey:sion 26, '! Nuclear Services Closed Cycle . Cooling Water"

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OP 1104-248, Revision'4, "Intake Screen and. Pump House Ventilation"

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OP 1104-30, Revision 26, "Nuclear River Water"

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OP 1104-48, Revision'6, "Screenhouse Ventilation-Equipment River Water"

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TMI Operations Plant Manual, Section 04,- Revision 1, Rx Bld Emergency Cooling River Watsr" In addition to the above, the inspector reviewed _readily accessible areas-of the system. The inspector performed a sampling of valve lineups dur-ing the week of December 28, 1987, with the assistance of an auxiliary operato .2 Findings / Conclusions The inspector determined that the procedure's stepwise instructions were compatible with checklist information and provisions for signoffs were evident. Appropriate limitations and precautions were incorporated into the procedure. The startup valve checklist and the Piping'and Instrument

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Diagram (P&ID) were compatible and in agreemen During the inspection of the reactor building emergency cooling system, the inspector f cund accessible areas to be clean. Valve lineups were as require The inspector determined that OP 1104-38, Revision 17, dated November 24, 1987, "Reactor Building Emergency Cooling System" is adequate.to control safety-related operations within applicable regulatory require-ment If called upon, this procedure can be used to limit post-accident containment pressure and temperature. No adverse conditions were found that would affect safet .0 Safety Issue Management System Item Verification 5.1 Introduction The inspcctor verified proper implementation, on a sampling basis, of licensee actions related to the below-listed NRC Safety Issue Management System (SIMS) item. The generic inspection approach for the SIMS item was:

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research various licensee and NRC merespondence, including safety evaluation reports (SER's) to V ify key assumptions, commitments, or other licensee actions to be .4 Ken to resolve the safety issue;

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identify any additional items which need to be verified as deline-ated in the related NRC Temporary Instruction or other inspection procedures;

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verify proper implementation of the items planned above; and,

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assiss licensee performance related to that implementation and re-lated to dissemination of the issue and its resolution to licensee personnel who need to know, such as by procedural upgrading and trainin .2 Reactor Coolant Pump Trip Criteria (SIMS No. 67.4.1)

5. Background / Review The TMI Action Plan (TAP) Item II.K.3.5 of NUREG 0737 required all licensees to consider solutions pertinent to tripping reactor coolant pumps (RCP's) under transient and loss of coolant accident (LOCA) conditions. The SECY 82-475 addressed the summary of industry programs regarding RCP trips. Generic Letter 83-10 outlined RCP trip requirements. The B&W Owners Group developed generic RCP trip criteria based upon a loss of sub-cooling margin and provided general information to be applied for plant specific implementation. Generic Letter 86-05 directed eac applicant to provide RCP trip criteria and substantiating in mation. GPUN letter of July 31, 1986, satisfies the response to 86-05. The NRC SER, dated August 11, 1987, concurred with all the licensees action The inspector n yiewed the above correspondence as delineated in paragraph 5.2.1 abov . Finding / Conclusions GPUN has complied with the requirements of Generic Letter 86-0 Accordingly, all the reactor coolant pumps (RCP's) will be manually tripped if the indicated subcooling margin falls below 25 degrees. They have identified the required iristrumentation and their redundancy and reliability under normal and adverse environmental conditions. The containment isolation will have no impact on RCP operation, since the required RCP water ser-vices will be maintained. The manual trip of the RCP's is accomplished from the main control room. That means the cri-tical components required to initiate the trip are not located

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in a harsh environment. The necessery procedures as mentioned ,

below, have been developed and the operators have been ade-quately traine ^

Also, the GPUN letter _ referenced above included _a' commitment to provide an implementation schedule for-September -1986. The formal response was, however, transmitted by March 11, 1987, '

which was based on several earlier discussions with the NRC staff. .This response did not specifically address an imple- ,

mentation plan for developing new operating. procedures as well as the training of operator :

Abnormal test procedures (ATP's) have been developed and the necessary training has been provided to all the operator ,

The following procedures are noted:

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ATP 1210-4 covers the procedure to re dore primcry to '

secondary heat transfer to achieve 25 degree sub-cool ,

margin (SCM),

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ATP 1210-10 addresses 25 degree SCM.as required with natural circulation; and,

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ATP 1210-9 addresses the recovery from solid operations to achieve 25 degree SC The necessary training was provided during the 1987 training '

cycles 1, 2, and i In summary, all the assumptions made in the SER of August 11, .

1987, are satisfied and this item is close '

6.0 Health Physics Technician Qualification Allegation An allegation was received on October 16, 1987, from an individual at another nuclear facility alleging that a junior health physics (HP) technician became a senior HP technician at TM The inspector reviewed the qualifications for the HP technician in questio In order to be qualified as a senior HP technician at TMI, the individual must have two years of working experience. A year of working experience is defined i as 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of work in not less than forty weeks. The HP technician has ;

met this requirement with over 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in eighty weeks. The individual t has also passed the required screening and qualification exam !

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In a letter dated December 14, 1987, the licensee provided a response to the ;

allegation, stating that the individual meets American National Standard.In- !

stitute (ANSI) Standard 18.1-1972, "Selection and Training of Nuclear Power Plant Personnel," to which TMI is committed in their Technical Specifications (TS). After review of the individual's training record and qualifications, !

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. 1 10 I the inspector determined that the: individual does meet this requirement, ~ The

. qualifications. for RF technicians at the site with whom the alleger is' as-sociated are apparently not the same as at TM During review of the calculations.of the hours of experience: acquired by the HP technician, a discrepancy between the. inspector's calculation and the-lic-ensee's calculation was noted. The discrepancy involved an overestimate of approximately five weeks (300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> with overtime) of experience .in the lic-ensee's calculation. The individual still meets the standards of ANSI 1 with the five weeks deleted from his total experience. The inspectoriinformed the licensee of this discrepancy. The inspector was informed that the HP technician will perform work only as a junior HP. technician until he success-

' fully passes an oral qualification board. This process normally takes ap-proximately six months to complet The inspector concluded that the indivijual in question has met NRC qualifi-cation requirements as a HP technician. Also, upon completion of La oral qualification board, he will meet licensee's internal requirements to be classified as a senior HP technician at TMI. This allegation was found to "

be unsubstantiated for TMI- .0 NRC Bullctin Verification 7.1 Introduction l

The inspector verified proper response and implementation, on a sampling i basis, of licensee actions related to the below-listed NRC Bulletin The generic inspection approach was to: '

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verify that the licensee's written response was within the time period stated in the bulletin, included the information *equired to be reported, and included adequate corrective action commitments based on information presented in the bulletin and licensee's re-sponse; and, -i

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independently verify, on a sampling basis, that licensee actions were implemented on site as noted in-the following; licensee man-agement has forwarded copies of the written response to appropriate on-site management representatives; information discussed in the licensee's written response was accurate; and, corrective action tak-an by the licensee was as described in the written respons .2 (Closed) NRC Bulletin No. 86-02 (289/86-BU-02): Static 0-Ring Differential pressure Switches NRC Bulletin No. 86-02 ' addresses Static "0" Ring (SOR) differential pres-sure switches erratic behavior problems. Two NRC Information Notices (IN) related to this bulletin, IN 86-47 "Erratic Behavior of Static "0" Ring Differential Pressure Switches," and IN 87-16, "Degradation of ;

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Static "0" Ring Pressure Switches," were also reviewed.

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In a letter dated July 29, 1986,. the licensee responded to Belletin 86-0 The licensee determined that SOR Series 102 and 103 Static "0" Ring dif-ferential pressure switches are not used at TMI-1.in' safety-related applications in distinction to pressure switches which'are used in s 2fety-related applications. This also precludes IN 86-47 from applying to TMI- On April 27, 1987, the licensee. received a preliminary'10 CFR Part '21 notice; and, on August 18, 1987, an interim 10 CFR Part 21. notice from-Static "0" Ring, Inc. was received. These Part 21 notices concerned a potential problem with gauge pressure switches and referenced NRC.In-formation Notice 87-16.. IN 87-16 addresses a concern with degradation of Kapton diaphragms due to the presence of gases permeating the dia-phragm. The licensee proceeded to generate a list of all SOR pressure-and differential pressure switches using-an equipment database fil Of the 108 SOR switches used in the plant, twenty-four switches were found to be nuclear safety'related (NSR). Three of the twenty-four NSR switches had Kapton diaphragms and were used to monitor the make-up; pump-lube oil pressure. SOR Inc. had recommended replacing the degraded switches and an increased test frequency, but the test frequency (six months) was arbitrarily conservative. In a licensee internal memorandum dated July 31, 1987, it was noted that there is no record of Kapton dia-phragms being adversely affected by any lubricating oil; and, as a result, the licensee did not replace the switche The licensee currently per-forms the calibration / test on a refueling cycle basis, since the manu-facturer's recommendation was arbitrary.

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The inspector expressed a concern that pressure switches used to isolate main feedwater on a steam line rupture are Static "0" Ring pressure ,

switches (Model No. 9 TAB 45NXCIAJJTTX6) in a safety-related application but do not appear in the equipment database as such. By discussions with

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cognizant personnel, it m determined that the SOR pressure switches had been disabled and its function was replaced with qualified equipmen In conclusion, the licensee's internal organization was found to be in receipt of the bulletin and information notices and took prompt, appro-priate actions. A reasonable search of computerized equipment lists was performed for applicability of the above-noted equipment problems. This bulletin is close .3 (Closed) NRC Bulletin No. 86-03 (289/86-BU-03): Failure of Multiple Emergency Core Cooling System Pumps The licensee submitted a response to IE Bulletin 86-03, "Potential Fail-ure of Multiple Emergency Core Cooling System (ECCS) Pumps Due to Single Failure of Air-Operated Valves in Minimum Flow Recirculation Line," on November 10, 198 It was determined that there is one situation where this bulletin applies. The recirculation line for the high pressure

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-12 injection (HPI) pumps contains two-motor-operated,_ normally open valves in series which are isolated automatically on an engineered safeguards-actuatio The. inspector reviewed-the following correspondence'as it relates to thi bulleti GPU Nuclear memorandum, dated May 30,:1986, from J. Bashista to J. Colitz

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GPU Nuclear memorandum, dated-November 7, 1986, from J. Randazzo to B. Thompson

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GPU Nuclear letter, dated November 10, 1986, to J. Taylor, Director, Office of Enforcement and Inspection, USNRC ,

The licensee reviewed all applicable ECCS pumps to ensure that, under a single active failure, minimum flow requirements are achieved. It was determined that all ECCS pumps, except .for the HPI pumps, will not vio-late minimum flow requirements due to a single active failure. Further-more, the HPI pumps are of sufficient size to provide enough flow at maximum Reactor Coolant System (RCS) pressure to protect the pump. It was also noted that a loss of power does not isolate'the recirculation line. The Independent On-Site Review Group (IOSRG) determined that the analysis which takes credit for operator action to open the recirculation valves for the HPI pumps is valid. Training has been emphasized on this raquiremen Information Notice 85-94, "Potential for loss of Minimum Flow Paths 1 Leading to ECCS Pump Damage During a LOCA," was also reviewed by the licensee. A detailed list of ECCS pumps that were considered for loss of minimum flow was provided by the licensee. Each pump was reviewed

. independently and determined to meet minimum flow requirements for ECCS pumpt, during a small break loss of coolant acciden '

The inspector independently verified that the emergency feedwater system (EFW) was in conformance with this bulletin. The EFW system is not part of ECCS at TMI. The inspector determined the recirculation line, which is also known as the condensate storage tank de-ice line, would maintain the minimum flow necessary for recirculation. A-single active failure would not hinder minimum flow recirculatio In conclusion, the inspector determined the licensee received the bulle-

tin and took prompt action to review the bulletin and provide a respons The licensee completed a review of both motor-operated and air-operated valves, although the bulletin addresses only air-operated valve failure The licensee took appropriate corrective actions in all' cases where the failures in the bulletin apply, i

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8.0 Exit Interview The inspectors discussed.the inspection scope and -findings with licensee man-agement.at a' final exit meeting conducted Jar.uary 11, 1988. Senior licensee personnel attending the final _ exit meeting included the following: ,

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.P. Ahern, NSCC Senior Staff Assistant, THI-1 M. Barnisin, NSCC Consultant,.TMI-1 T. Broughttn, Operations and Maintenance Director, TMI-1 J..Colitz, Plant. Engineering Director, TMI-1 H. Hukill, Director, TMI-1 G. Kuehn, Radiological Controls Director, TMI-1 L. Robinson, Media Representative D. Shoviin, Plant Material Director, TMI-1 D. Tuttic, Radiological Field Operations Manager, TMI-2 A representative of the Commonwealth of Pennsylvania, Ajit K. Bhattachayyra, also attended the meetin The inspection results as discussed at the meeting are summarized in the cover page of the inspection report. Licensee representatives did not indicate that any of the subjects discussed contained proprietary or safeguards informatio !

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ATTACHMENT NRC INSPECTION REPORT NO. 50-289/87-24 ACTIVITIES REVIEWED Plant Operations

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Control room operations during regular and back shift hours, including f're-quent observation of activities in progress and periodic reviews of selected sections of .the shift foreman's log and control room operator's -log and

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selected sections of other control room daily 'og Areas outside the control roo Selected licensee planning meeting Reactor building emergency cooling system standby alignmen During this inspection period, the inspectors conducted direct inspections' during :

the following back shift hours: Saturday, December 19, 1987, 7:30 a.m. to 11:30

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a.m.; Saturday, December 26, 1987, 8:15 p.m. to 9:15 p.m.; Wednesday, December 30, '

1987, 6:30 a.m. to 7:00 a.m. ; Thursday, January 7, 1988, 5:00 a.m. to 7:00 a.m.;

and, Saturday, January 9, 10:00 a.m. to 12:00 Maintenance ,

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Equipment mounting and foundation bolts, including preventive maintenance on a main steam line tie rod supports.

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Reactor building emergency cooling expansion joint installation '(JT No. CK j 212). -

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Steam generator level repair (JT No. CP 467). l

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Inverter repair (JT No. CP 464).  !

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Surveillance

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Steam generator level weekly checks (SP 1302-4.1).

Reactor Coolant System (RCS) Leak Rate The inspector selectively reviewed RCS leak rate data for the past inspection period. The inspector independently calculated certain RCS leak rate data re-

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i viewed using licensee input data and a generic NRC "BASIC" computer program "RCSLK9" l as specified in NUREG 1107. Licensee (L) and NRC (N) data are tabulated below.

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-. .. -, . . ... . -. . - _ . - - .

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TABLE

'

RCS LEAK RATE DATA ALL VALUES GPM

DATE/ TIME (NUREG,1107) CORRECTED

- 0URATION Lg Ng Ng N I U 11

'

12/10/87- 0.6541 0.66 -0.09 0.01 0.0107 ,

2 Hours' ,

12/15/87 -0.0134 -0.02 -0.23 -0.15- -0.1421 4 2 Hours

- 12/18/87 0.3443 0.34 0.46 -0.02 -0.0144  :

2 Hours '

12/21/87 0.3731 0.37 0.44 0.03 0.0333  :

2 Hours ,

'

i

'

12/25/87 0.1149 0.11 0.19 0.02 0.0300 2 Hours ,

12/28/87 0.1151 0.12 0.11 0.11 0.1101 2 Hours

]

"

01/01/88 0.5060 0.51 0.47 0.14' O.1425  !

2 Hours '

01/06/88 0.2234 0.22 0.33 -0.01 -0.0038' )

2 Hours

i

G = Identified gross leakage U = Unidentified leakag L = Licensae calculated N - NRC calculated Columns 1 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 1107u N due to total purge flow!

~

)

through the N'.. 3 seal from RCP' 'i

^

,

j j

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