IR 05000289/1985012

Insp Rept 50-289/85-12 on 850408 & 0506.Major Areas Inspected:Shutdown Plant Activities,Including Steam Generator Tube Repair,Hot Functional Testing & Related Event Followup

ML20127P683
Person / Time
Site:	Three Mile Island
Issue date:	06/11/1985
From:	Conte R, Urban R, Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20127P661	List: IR 05000289/1985012 ML20127P656
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM 50-289-85-12, NUDOCS 8507020488
Download: ML20127P683 (21)

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U.S. NUCLEAR REGULATORY COMMISSION REGION I I Report N /85-12 Docket N License N DPR-50 Priority -- Category C Licensee: GPU Nuclear J Cor aration Post office Box 40 Niddletown, PennsyTvania 17057

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Facitity At: Three Mile Island Nuclear Station, Unit 1 Inspection At: Middletown, pennsylvania Inspection Conducted: April 8, 1985 - May 6, 1985 Inspectors: [ k R. Conte,5enioNNd$ntInspp/2kTN~19 tor

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R. Urban, Reactor Engineer date l

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F. Young,ResidentInspector/(TMI-1)

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date Approved By: a E. Conner, ChT f, e r _veactorProjects

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Section No. 1A, Projects Branch No. 1 Division of Reactor Project inspection Summary l This routine safety inspection (203 hours0.00235 days <br />0.0564 hours <br />3.356481e-4 weeks <br />7.72415e-5 months <br />) reviewed routino shutdown plant activities, including those related to steam generator tube repair hot functio.1al testing and related event followup; emergency feedwater operability; plant modifications including those related to decay heat post-accident sampling and plant shielding; control of examinations; radiological exposure in excess of administrative limits; quality assurance assessment, restart readiness including valvo lineups; and licensee action on previous inspection findings,

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Results Licensee Management and t'e Quality Assurance Department continued their detailed involvement in plant activitie Overall, procedures were properly implemented during the Hot Functional Test period. A misunderstanding of the design basis operability requirements for backup air supply (banks of air bottles) for operation of emergency feedwater system valves was reviewed and resolved. Modifications were properly installed; but, in certain instances from a human factor viewpoint, more reliable equipment could have been provided in order to minimize radiological exposure. The licensee continued to implement hearing related commitments on the control of examinations. A radiological exposure event was properly reviewed by licensee personnel. The annual quality assurance assessment provided licensee management with information pertinent to performance strengths and weaknesses. The licensee continues to work on making the plant ready for restart. The licensee either initiated appropriate action or completed commitments related to previously identified inspection findings,

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DETAILS

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1.0 Introduction This inspection report documents the activities conducted by the resident j

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inspectors assisted by region-based personnel. The overall purpose of the inspection was to assess the licensee's activities as they relate to reactor safety and worker radiation protection for the shutdown mode i and to assess plant readiness for the restart of TMI- '

The inspectors made this assessment by reviewing information on a sampling basis through licensee interviews, actual observation of activities (where  ;

i possible), measurement of radiation levels, and review of listed documents i or records. Within each area, the inspector listed the specific purpose

! of review (or verification), scope of the review (or specific inspector '

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activity) and finding .0 Plant Operations During Long Term Shutdown '

2.1 Routine Review

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The resident inspectors periodically inspected the facility to deter-j mine the Itcensee's compliance with 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 i including plant housekeeping and fire protection measures; l

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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 log kooping 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 practico [

t The inspectors focused on the following areas

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the control room during regular and backshif t hours which .

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included the selected sections of the shift foreman's log and

control room operator's log for the period April 8, 1985, l through May 6, 1985, and selected sections of other control room daily logs;

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areas outside the control room on April 8 to 14 (daily) 16, 17, 24, 26, 27, 29 and May 3, 1985; and, '

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selected licensee planning meeting Based on the review of the various licensee activities noted above and, in particular, those activities noted in paragraphs 2.2, 2.3, and 2.4, the inspector identified no conditions adverse to nuclear safety or regulatory requirement Personnel stationed in the control room presented a posture of I overall control of daily activities, including problem areas that needed resolutio The planning meetings indicated attentiveness to proceed safely with daily activities, including surveillance e d maintenance, and to resolve any inter-department interface problems, i Licensoo upper management continued their detailed involvement in  !

site activitie .2 OnceThroughSteamGenerator(OTSG) Repairs Between April 2, 1985, and April 12, 1985, the licensee plugged the i remaining group of tubes with indications greater than 404 through- l wall wastage. The indications were determired from Eddy Current Testing that was completed in the beginning of this yea The inspector reviewed portions of Job Ticket packages CF849, CF850, C0348, and CG34 The review was to ensure that applicable admints-trativo and maintenance procedures were established and implemented to address tubo plugging. In addition, the work packages were reviewed to ensure that required post testing was performe Review of the work packages indicated that the licensee adequately performed the work and properly documented the task, llowever, the inspector noted that soveral calculations for percent wall thinning

, taken on one shift inside containment had boon written in pencil and then written over in ink at a later timo . The use of pencil for for-mal records was inconsistent with licensoo past practico. Discussion with the engineer taking the data indicated that he did not have an ink pen insido containment on that shif The inspector reviewed records from other shifts and found all the records to be written in ink. The calculations written in pencil were accurate and consistent with the values from other shifts. The inspector considered this an ,

isolated caso. Licenseo representativos acknowledged the inspector's [

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Ingeneral,theinspectornotedQualityAssurance(QA) involvement  ;

in field observation and resolutions of technical issues that aroso  ;

during that phase of OTSO repai ,

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l Within this period, the licensee performed OTSG Hot Functional Testing (HFT). The HFT was to determine a new OTSG base line primary to secondary leakage. At the completion of the reporting period, the licensee was in the process of evaluating this data but preliminary calculations show that OTSG 1eakage to be less than one gallon per hour. Test results were preliminarily evaluated by inspectors as documented in Inspection Report 50-289/85-16. The final review of this information will be done in a subsequent report after the licensee submits a report on this testing in accordance with Licensee Condition No. 2.C. .3 High Ta11 pipe Temperature on the power Operated Relief Valve Between April 12 and 14, 1985, a high differential temperature (approximately 50' - 60*F) occurred for the tailpipe connected to the Power Operated Relief Valve (PORV). The differential temperature indicator measures the temperature difference between the Reactor Building ambient and the wall of the PORV tailpipe; the associated instrument string provides an alarm for differential temperature in ;

excess of 30'F. The acoustic monitor and differential pressure instru- s mentation on the PORV tailpipo provided no indication of PORV leakag !

This coupled with a slight increase in Drain Tank Temperature and RCS leakrate calculations, led Itcensee representatives to conclude that leakage occurred past the main disc of the POR Several attempts to better seat the main disc by PORV cycling failed to reduce the minute leakage (tailpipe differential temperature to less than 30'F).

Subsequently, on April 14, 1985, licensee representatives shut the block valve in accordance with Emergency Procedure 1202-29. In accor-dance with the RCS cooldown pracedure (conducted April 16-17,1985),

licenste representatives repor:ed that they cycled the PORV and successfully seated the main disc such that PORV tailpipe differential temperature was less than 30* During and subsequent to those events, the inspector reviewed EP 1202-29, Revision 26, March 7, 1985, " Pressurizer System Failure" to assure that Ilconsee representatives:

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properly implemented the applicable section (A) of the emergency proceduro; and,

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provided sufficient technical guidance to the operators in handling symptoms associated with PORV operability problem The inspector noted an inconsistency in Section A of the procedure for a leaking PORV. The immediate action requires the block valve to

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be shut without evaluating related symptoms to assure that thoro is l definito leakage from the PORV. However, follow-up actions provide

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steps to evaluato and conclude whether a leaking PORV oxists. The ,

follow-up actions were more consistent with the intent of TS 3.1.12; -

it indicates that the block valve may be shut to reduce RCS leakago

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to within the requirements of TS 3.1.6. At the time of the PORV leakage, the licensee calculated RCS leakrates to be well within TS requirement Concurrently, the Manager of Plant Operations identified the incon-sistency and initiated a procedure change request (PCR No. 1-05-85- l 0284) to clarify the procedure. The Plant Review Group approved the PCR on April 30, 1985; the revised procedure will be issued shortl The inspector concluded operators properly implemented EP 1202-29 consistent with applicable T3 requirements. Management involvement on the suspected PORV leakage was evident, and they took appropriate action to improve the applicable procedural guidance for a leaking POR '

2.4 InadvertentDrainDownofBoricAcidMixTank(BAMT)

On or about March 29, 1985, licensee representatives started Job Ticket (JT) CG-342 to troubleshoot a valve operator malfunction for a Makeup and Purification Valve (MU-V51). Electricians determined

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that the malfunction was due to a faulty diaphram which necessitated a transfer of the JT to Instrument and Control (!&C). The I&C Technicians stopped work on the valve for the weekend of March 30-31, 1985. The !&C Foreman discussed the stop work with the Operations

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Shift Foreman. Apparently, transfer of information on the status of the valve, which was open, were not clear. Further, tho accompanying

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switching and tag order was not adequate to support isolation of i

MU-VS1, due to personnel erro During the same weekend, licensee representatives depressurized and drained down the RCS in preparation for OTSG tube plugging. When RCS water level dropped below DAMT level, gravity flow occurred from the BAMT through pumps CA-P1A/0, valves MU-VS1 and MU-V78 and into the RCS through valves MU-V17 and 18. Eventually the BAMT level instru-mentation provided an alarm on low level. Operators responded to the '

alarm and subsequently identified and isolated the improper flow path into the RC l Shortly thereaf ter, Operations Management initiated a " Plant Incident" review on the event in accordance with Administrative Procedure 1029, " Conduct of Operations." Licensee representatives completed that review and documented the results in Plant Incident .

Report (PIR)No. 1-85-003, dated April 2, 198 The PIR identified that maintenanco personnel failed to properly use the tag application, that tag isolation was inadequate to isolate and the job (reviewed communications bynot were maintenance and operations proper to adequately personnolj, reflect the o pen" status of the valvo. Corrective actions included a review of the event with Operations and Maintenanco Department personne l

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The inspector first became aware of the event on April 2, 1985, by I reviewing the March 30 event log entry in the Shift Foreman's Lo St.bsequent to the HFT, and during a review of HFT activities, the inspector discussed this event and other matters with the Plant Operation Manager who provided the subject PIR to the inspecto Operations Management expressed concern over the event and noted that the report identified no programmatic deficiencies and provided sufficient corrective action The inspector concluded that there was adequate management attention and involvement in the event along with proper documentation of the event in a PIR for self review and corrective action. Being familiar with the licensee's program for protection of personnel and equipment, the inspector acknowledged the non-identification of a programmatic deficiency and attributed the event to a lack of attention to detail on the part of certain individual Licensee management oriented their corrective action toward the personnel who worked for the l maintenance and operation departmen The inspector had no additional comment .0 E_quipment Operability The inspectors reviewed other selected areas which involved safety related equipment operability during the HF In particular, while inspecting areas in the Emergency Diesel Generator ,

Building, the inspector noted that one of the banks of air bottles ("A" Bank) for the Emergency Feed Water Two-Hour Backup Air Supply System was depressurized due to a pressure regulator malfunction. The Backup Air Supply System supplies air to selected EFW valves during an emergenc The valves are normally supplied air by the Instrument Air (IA) or Service AirSystems(SA). Both IA and SA systems were in continuous operation during HFT. The inspector then questioned the operability of the EFW loops supplied by the "A" air bank. Licensee representatives stated that . I they relied on the operability of the two instrument air compressors, two

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service air compressors (not considered to be safety grado) and a small capacity AC powered compressor (also not to be relied upon).

The TS definition states, "a system, sub-system, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of per-forming its specified function (s) and when all necessary attendent instru-mentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that aru required for the system, sub-system, train, component, or device to perform its function (s) are also capable of performing their related support function (s)." In light of this definition, the licensee's position appeared to be correct. However, the section of the FSAR addressing the Two-Hour Backup Air Supply System stated that the system was designed to be operable during a design basis earthquake with a loss of site A The only reliable source of air available to operate the l

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EFW valves would be the Two-Hour Backup Air Supply System assuming signi-ficant reactor decay heat generation. In order for two trains of EFW to be operable, both banks (A and B) of backup air would be required. Because the RCS heat input was due primarily to Reactor Coolant Pumps (and not reactor decay heat), the inspector determined that the EFW system would perform its safety function with the "A" bank depressurized. However, the inspector questioned the licensee representative's understanding of the operability requirements with respect to critical operation After review of this event by the licensee's Plant Review Group, the licensee agreed to require both banks of backup air to be operable before both trains of EFW are considered operabl However, the licensee stated however that they may still pursue a TS change to clarify this matte .0 Modification Review 4.1 Post-Accident Sampling (PAS) Capability NUREG-0737, Item II.B.3, specifies that licensees shall have the capability to promptly collect, handle, and analyze post-accident samples which are representative of conditions existing in the reactor coolant and containment atmosphere. Implementation of Item II.B.3 was inspected in Inspection Report 50-289/84-03. At that time the NRC staff questioned the licensee's ability to collect a repre-sentative Reactor Coolant System (RCS) sample under all accident conditions and modes of operation. The inspectors noted the flow through the PAS system relied entirely on RCS pressure. The licensee agreed to review this item (289/84-03-01) for atmospheric RCS pres-sure condition The licensee completed their review and modified the system. The modification provided the capability to obtain a post-accident sample from the Decay Heat Removal System via the shielded reactor coolant sample line in the nuclear sampling roo The inspector reviewed the new modification and numerous licensee letters against the criteria identified in NUREG-0737. The inspector determined that the PAS meets the basic requirements and adequately addressed the intent of the NURE In addition to review of the modification documentation, the inspector witnessed a RCS sample drawn via the Decay Heat / Reactor Coolant Sample Cross Tie line. The licensee was able to obtain the required sample; however, during the valve lineup, several valve handles became loose and fell off in the operator's hand. Discussions with licensee representatives indicated that the valve handles were maintained in place by " allen" screws. It was noted that these screws quickly became loose. Because the operator would be in a high radiation fleid when drawing a sample, a problem of loose handles could add to his radiation exposure or require additional significant exposure on a post-accident situation to fix the handwheel problem. The licensee

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is acknowledged the inspector's concern and stated that a possible solution would be to stake the' allen screws. This is unresolved pending licensee corrective action to assure the handwheel remains in

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place during sampling evolutions. This action will be reviewed in a subsequent NRC" inspection (289/85-12-01).

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Plant Shie'iding Modifications

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NRC Inspection Report No. 50-289/82-13 documented the in pl' ant-review in su'pport of the NRC safety evaluation for Task Action Plan TAP) Item II.E.2, Plant Shielding. In addition to the review of the

' (licensee's shielding study for adequacy, the inspector conducted a walkthrough of procedures used by operators in handling post- acci-dent response activitie The inspector identified that for the g evolution'of boron precipitation control (long term recirculation),

'the operator needs to manually operate valves in the decay heat g sault or at the decay heat shielded areas that'would have prohibitive

' radiation field Exposures could result in excess of the 5 Rem guideline with potentially highly radioactive water in OH piping due to the Reactor Building Sump Isolation Valves (DH-V6A/B) being ope The licensee's initially proposed resolution to the problem was to revise applicable procedt.res to preclude the prchibitive exposures

.(as a short term fix) and to modify manually operated valves (DH-64, 12A/B, 19A/B, and DC-V 2A/B, 65H/B) with remote operators by Cycle 6

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'! Pefueling (as a long term fix). Inspectors verified the procedural changes in NRC Inspection 50-289/83-01 and the Commission accepted the Cycle 6 modification commitment as noted in SECY 384A, dated

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December 6, 198 After various proposals and counter proposals, the NRC staff. accepted a simplified resolution to the boron precipitation cc-ntrol . problem as noted in its final TMI-1 safety evaluation, dated December"26, 198 ~

The' staff accepted the<fnstallation of a reach rod for DH-V64, Auxi-s liary Pressurizdr Spray ^ Isolation Valve, to utilize existing shielding in that area,_of the plant for the "A" DH loco recirculation. Further the staff acdepted the commitment to lock open DH-128, Tie Isolation Valve from the RCS Drop Line, for RCS letdown (by gravity) to the RB Sump and "B" loop recircufation of water to the reactor core. The a staff's acceptance was contingent on a post implementation inspection (by NRCser gion I) of the conformance of the shielding review (for

, DH,-V64) to NUREG-0737 requi ment 'NRC Inspection Reports 50-289/84-03 and 84-16 documented the review of proper shielding for the post-accident sampling system (TAP II.B.3). In March 1985, the licensee essentially completed those commitments noted above for DH-V64 and DH-V128 (TAP II.B.2).

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In addition to discussions with cognizant licensee personnel and observations, in the plant, the inspector verified the proper

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reviewing the related modification package for DH-V64 for proper documentation in accordance with Administrative Procedure 1043;

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again performing a sampling review and walkdown of applicable emergency and operating procedures (substantially revised since NRC Inspections 50-289/82-13 and 83-01) to assure that the results of the previous shield study and inspector verifications were not invalidated by procedure revision; and,

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verifying procedure changes to assure that DH-V128 is locked ope The inspector reviewed the following specific documents / records:

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Modification Packages related to Budget Account (BA) 412394, DH-V64 Reach Rod;

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Operating Procedure 1104-4, Revision 50, March 5, 1985, " Decay Heat Removal System;"

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Abnormal Transient Procedure (ATP) 1210-6, Revision 5, March 8, 1985, "Small Break LOCA Cooldown;"

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ATP 1210-7, Revision 6, March 8,1985, "Large Break LOCA Cooldown;"

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Annunicator Procedure E-1-8, Revision 2, January 4, 1984,

" Borated Water Storage Tank Low Level;"

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Emergency Plan Implementing Procedure (EPIP) 1003.9, Revision

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4, December 3, 1984, " Radiological Control During an Emergency;"

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OP 1104-13, Revision 16, April 2, 1985, " Decay Heat Closed Cycle l Cooling . System;" .

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OP 1102-1, Revision 75, March 20, 1985, " Plant Heatup to 525 F;"

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OP 1103-32, Revision 14, July 9, 1984, " Decay Heat River Water

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System."

The DH-V64 Modification Package was complet It reflected proper l installation, drawings, and specification Pre-operational testing ( confirmed that consistent torque was applied to the valve using the

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newly installed reach rods. The test also demonstrated that l containment isolation valve local leakrate was consistent with values assumed for normal operations. A walkdown of the reach rod assembly identified no deficiencies, and it confirmed that the

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licensee utilized the same shielding that inspectors previously I found to meet NUREG-0737 requirements as noted in NRC Inspection l Report 50-289/82-1 l As noted in previous inspections, the inspector found that applicable procedures effectively cautioned or warned operators on the hazard of operating certain manually operated isolation valves during the post-accident period and, in particular, during the long term boron prect- l pitation control evolutio The procedures required that operators position other potentially inaccessible valves before opening DH-V6A/B (letting RB sump water into DH piping). The licensee revised the DH system emergency standby line up procedure to require DH-V12B to be locked ope The inspector noted that the precautions and limitations section of DH system operating procedure (1104-4) were confusing with respect to operator guidance on throttling DH flow. It appeared that this section was inflexible on the use of valves for throttling when the DH Pump took suction from the RB sump; namely, it stated that DH- l V19A/B were to be used to prevent pump runout. However, the boron l precipitation control section of this Operating Procedure (0P) rightly l cautions against the use of DH-V19A/B since they are manually operated '

and are in a potentially inaccessible area during the post-accident boron precipitation control evolutio Based on discussions with licensee personnel, the inspector learned that engineering personnel cautioned against the use of DH-V4A/B (the alternate means of throttling DH flow), since these valves are gate valves not normally designed for '

throttling. The legitimate engineering concern was not clearly stated l in the procedur The inspector concluded that licensee management did not completely provide limitation / precaution guidance to operators in the operating procedure on use of DH-V4A/B versus DH-V19A/B considering radiologi-cal hazards and engineering concerns for the various preplanned

, evolutions in this procedur Licensee representatives acknowledged

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the inspector's comments, and they initiated a revision to the operating procedure to clarify the guidance to the operators (PCR ( Nos. 1-02-85-295 and 296).

I i- The inspector had no additional comment Based on the above and

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previous reviews of licensee actions related to TAP II.B.2, Plant l Shielding and TAP II.B.3 (in part) related to Post-Accident Sample System Shielding, the inspector considered TAP II.B.2 to be closed.

i 5.0 Control of Examinations i On or about April 2, 1985, the licensee reported that a microfiche copy of l TMI-1 auxiliary operator excminations had been found in the motorcycle l parking lot near the TMI-2 Administration Buildin In conjunction with

! Training Department Management, the Director of TMI-1 immediately con-firmed that the security of the examinations were not compromised since l l

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the microfiche were records of graded final examinations in accordance with Procedure 6200-ADM-2600.01. However, the Director of TMI-1 expressed concern to the TMI Information Management Department (IMD), that a review should be performed to identify the circumstances that led to the parti-cular microfiche being in an uncontrolled stat The IMD documented their review in internal memorandum No. 7132-85-057, dated April 11, 1985. The microfiche contained April and May 1984 examina-tions for 19 TMI-1 auxiliary operator requalification examinations along with answer sheets, seating charts, review sheets, and attendance form The archival copy and working copy of the same microfiche were in the records storage vault at the TMI-2 Administration Building. The training department copy of that microfiche was also in the vault awaiting distri-bution to the Training Department. The copy found in the parking lot was an extra. The IMD never determined why personnel made the extra copy (perhaps for better quality to be discarded later at the local waste receptacle). The licensee representatives classified the information on the microfiche as " sensitive," apparently because of personal data on the forms, not for examination security purpose The IMD corrective actions included the establishment of an internal procedure to destroy by shredding or other means all extra copies of such documents. They placed the subject microfiche on file in the vault along with the above noted IMD repor The inspector discussed the event with cognizant licensee management. He reviewed 6200-ADM-2600.01, Revision 2-00, dated November 30, 1984,

" Control of Examination," and the above referenced internal memorand 'The inspector co.1cluded that licensee management properly reviewed the event and took appropriate corrective actio Management showed initia-tive in the timely reporting of the matter to the NRC resident offic Based on this review, the inspector concluded the licensee met the require-ments of the control of examination procedure and, thereby, continued to implement their commitments made to the applicable Licensing Board in this area. The IMD developed adequate corrective actions to preclude loose copies of the GPU classified " sensitive" document In a related event, the licensee reported that a TMI-2 contractor Security Guard was caught seeking help from another individual during a General Employee Training Examination on April 15, 1985. The training instructor /

proctor immediately confiscated the examination and sent the guard back to the work supervisor. The licensee later reported that the individual was sent back to the contractor as unacceptable for employment at TM The inspector discussed the event with the Director of TMI-1. The inspec-tor concluded that licensee representatives properly implemented the Control of Examination Procedur The inspector has no further comments on these matter _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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6.0 Radiographer Exposure in Excess of Administrative Limits On February 8, 1985, the Radiological Controls Manager reported to the NRC Resident Office that a contractor radiographer's whole body exposure was in excess of a licensee administrative quarterly limit (1000 mrem) based on the reading of the January 1985 TLD (Thermoluminescent Dosimeter). Also, he initially reported that the " suspect" TLD reading was inconsistent with self-reading dosimetry. Licensee management restricted the radiographer from performing additional work in an RWP area until a radiological engineer completed an investigation into the matter (documented in Radiological Investigation Report (RIR) No.85-001, dated February 7,1985).

During this inspection the inspector reviewed RIR 85-001 to assure that:

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the circumstances leading to the event were clearly identified and documented along with root causes; and,

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adequate corrective action was proposed or taken along with any necessary measures to preclude recurrenc This review included discussions with cognizant licensee personne The RIR provided a sequence of events leading to the identifications of the subject exposure. During January 1985, the NES (Nuclear Energy Ser-vices) employee worked in two primary functions as an NDT (Non-destructive Test) technician performing inspections inside the OTSG and in the TMI-1 Intermediate Building at the EFW (Emergency Feedwater Piping) piping. For each of the functions the licensee provided him with separate dosimetry consis' ant with past practice to segregate OTSG exposure from other plant work exposure. In addition, he had NES supplied dosimetry. On February 6, 1985, licensee representatives read the "other plant work" TLD for the NES worker and entered the data into the computerized system. Da an attempted entry into a TMI-2 RWP area, the licensee identified the NES employee's exposure for the quarter was 1150 mrem (in excess of the administrative limit of 1000 mrem but well less than the NRC limit of 3000 mrem). The individual was denied access to the TMI-2 area since licensee management had not approved exceeding the 1000 mrem exposure limi They referred the matter to management for further revie The RIR reflects extensive investigation by the licensee in an attempt to correlate SRD and TLD readings for both contractor and licensee supplied dosimetry. Licensee representatives verified proper operation and cali-bration of the TLD reading equipment. The SRD and TLD readings were cor-related by 10% for the NES employee's OTSG work with an assigned dose of 457 mrem. The licensee's TLD for other plant work substantially disagreed with contractor dosimetry and licensee SRD readings along with co-worker exposure results. Further review revealed that the suspect TLD was not at the processing center for TMI-1 on two nights in January 1985; no one (including the NES employee) can adequately account for its whereabout Analysis of data from the TLD processing indicated that the exposure on

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the suspect TLD was due to a lower photon energy source from that used in radiography work (Ir-192). The licensee surmised that the suspect TLD was inadvertently exposed to an Ir-192 source. To be conservative, licensee management assigned the detected 693 mrem from the suspect TLD to the individual's exposure record The inspector found licensee review of the event to be adequate; licensee management continued to exhibit their detailed attention and involvement in matters affecting personnel radiological protection. Investigative actions were extensive and reasonably complete leading to plausible con-clusions. The licensee was conservative in its final dose assessment for the contractor employee. The inspector acknowledged the licensee's con-clusion that the event was not reportable with respect to 10 CFR 20, 21, 50, and the Unit 1 TS. Since this appears to be an isolated case of poor control by an individual of his assigned dosimetry, the inspector concluded that no programmatic problem existe .0 Quality Assurance Effectiveness Review On April 25, 1985, the inspector attended the Quality Assurance Department (QAD) presentation to licensee management on the 1984 Quality Assurance Annual Assessment. The QAD section leaders made presentations covering the following areas; QA engineering, site welding, site inservice inspec-tion, quality control, operations QA, site audits, and QA system engineerin In addition to QAD representatives, a majority of the GPUNC Vice Presidents attended the meeting, including the Senior Vice President for GPUNC and Director /Vice President of TMI- Although statistics and bar charts were available as performance indicators, the main topics of discussion focused on licensee performance strengths and weaknesses as viewed by each of the QA sections. There was an exchange of information for licensee management to understand the points being made especially for improvement in weak areas. Assignments were made for cor-rective actions in weak areas. The overall conclusion was that the organi-zation possesses more strengths than weaknesses but the weaknesses need to be worked on in their resolve for excellenc The inspector noted that the presentation encompassed the findings noted in the most recent TMI-1 SALP. However, QAD's presentation and discussion was much more detailed because of QAD's detailed involvement in site /

corporate activities. The inspector concluded that the annual assessment continued to improve and provided licensee management with pertinent information with respect to organization performance strengths and weak-nesse .0 Restart Readiness During the inspection, the resident inspectors assisted by region-based inspectors initiated a specific hardware review of selected areas to assess the readiness of the plant for startup. The selected areas were:

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licensee's prerequisite list for hot functional testing / criticality (Flag

"2B"); open material non-conformance reports or quality deficiency reports; surveillance program open exceptions and deficiencies and related out-standing regulatory retest equipment (so tagged); and important to safety system valve lineups. The objective was to identify equipment operability problems that would adversely affect safe operation of the facilit A similar review was conducted in the preoperational test area and was documented in NRC Inspection Report 50-289/85-16. Other areas such as open job tickets and outstanding modification incomplete work list items will be reviewed during future inspections closer to criticality, if approve Results of this review are documented below:

8.1 Prerequisite List The inspector reviewed the GPUN restart package titled "TMI-1 Restart Readiness Prerequisite Listing Flag 2B Hot Functional /

Critical Testing." It is an extensive and updated package identi-fying those items that need to be addressed prior to restart. The licensee has tentatively scheduled the Flag 28 meeting for May 9, 1985. At that time, all items in the restart package are scheduled to be complete and therefore signed off. To date, many items have already been completed. However, any decisions made by the Commission could impora further restart requirements and this could be routinely reviewed by Region '

8.2 Quality Assurance Hardware Items Within the scope of this review, the inspector specifically reviewed the status of Material Non-Conformance Reports (MNCRs), Quality Defi-ciency Reports (QDRs), Exceptions and Deficiencies (E&Ds), and Regu-latory Retest Tags (RRTs). MNCRs deal with hardware related problems and QDRs deal with software related problems. All MNCRs and QDRs ,

affecting restart have been reviewed and closed out by the license E&Ds and related RRTs are created during surveillance testing. Deft-ciencies are equipment problems or malfunctions or failing to complete a surveillance test by the late performance date. An exception is a procedure change that does not alter the scope or intent of the procedur RRTs are tags displayed in the control room on equipment that requires testing at a time later than scheduled. The RRT system is used as a backup to the E&Ds so that operations personnel will question the operability of that particular piece of equipment. The inspector determined that approximately 50 E&Ds were open and that certain E&Ds affected operability of important to safety equipmen Licensee representatives acknowledged the inspector's comments and stated that the outstanding E&Ds list will be reviewed before criticality to assure no adverse condition exists with respect to i

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important to safety equipment operability. This area will continue to be routinely reviewed by the NRC Resident Offic .3 Valve Lineup Verifications As part of the validation of the TMI-1 readiness for restart, the NRC staff independently verified the position of safety related valves. The inspector, with the aid of an Auxiliary Operator, verified the position of valves listed in the following operating procedures:

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Operating Procedure 1104-38, " Reactor Building Emergency Cooling Water System (RBECW);"

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Operating Procedure 1104-30, " Nuclear River Water System (NRW);"

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Operating Procedure 1104-32, " Decay Heat River Water System (DHRW);"

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Reactor Building Integrity per Operating Procedure 1101-3,

" Containment Integrity and Access Limits;" and,

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Startup Breaker Checklist per Operating Procedure 1107-2,

" Emergency Electrical System."

In general, the inspector found the valve list to be accurate and it reflected the proper position for valves checked. The inspector did note several inconsistencies as discussed belo Breaker verification of Procedure 1101-3 found all breakers to be in their proper positions except the breaker for three welding receptacle This lineup lists the position of breakers for systems when contain-ment integrity is required (plant temperature greater than 200 F).

However, since the plant was in a shutdown condition, the inspector expected the welding receptacle breakers to be in the closed positio The inspector also noted several inconsistencies in how components were listed on the checklist. The licensee acknowledged the incon-sistencies and is revising Procedure 1101-3 to address these discrepancie In the review of the NRW valve lineup, several deviations from normal system valve arrangements were noted related to the low heat loads on the system. In addition, the inspector noted that a mechanical jumper (temporary cross connect piping from NRW outlet Header Vent to the DHRW Loop A Vent) was present. The cross connect is used to supply cooling water to the DHRW Loop A heat exchanger since the DHRW system cannot be adequately throttled to handle the unusually low decay heat load. All deviations were clearly explained to the inspecto . .

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Several deviations in the lineup of the DHRW system were note Most were due to the unique plant conditions. However, valves DR-V-38A/B on DR-P-2A and 2B minimum flow lines were found close These pumps supply bearing lube water to the DHRW pumps. Since lube water was not being supplied by these pumps but rather by the filtered water system (non-safety grade), these valves did not need to be open for the shutdown condition. The subject valves were opened by the auxiliary operator so that the lube water for the pumps was supplied by their normal water supply, rather than the non-safety grade filtered water syste In general, the inspector found valve lineups / breaker position to be in accordance with licensee's plant procedure. The inspector had no further questions concerning the valve lineup .0 Follow-Up on Previous Inspection Findings The following items were reviewed to assure that the licensee took adequate corrective action in a timely manner and/or met their commitments as stated in applicable inspection report .1 (Open) Unresolved (287/84-07-03) Review Licensee's Management Submittal to NRC Staff - 0TSG Preoperational Testin See paragraph .2 (Closed) Unresolved (289/82-13-03) and Task Action Plan Item II.B.2:

Manually Operated Valves with Remote Operators for Shielding Consideration during Post-Accident Long Term Recirculation See paragraph .3 (Closed) Unresolved (289/84-03-01) and Task Action Plan Item II. (in part): Provide Capability to Obtain a Post-Accident RCS Sample

at Low Pressure Conditions See paragraph 4.1.

l 9.4 Closed) Inspector Follow Item (289/84-19-02): Provide Additional ( Training on Entire Electro-Hydraulic Control and Nuclear Instrumentation An inspection conducted on August 27-30, 1984, concluded that ade-quate augmented training had not been provided by the licensee in two of the thirteen topical areas identified during the February 1984 Operational Readiness Evaluation (50-289/84-05). An inspection of

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these two areas, control functions of the Electro-Hydraulic Control System (Item 10) and predicting indications on Nuclear Instrumenta-tion (NI) during a reactor startup (Item 13), included a review of l

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lesson plans, training received in the Basic Principles Trainers (BPT)

and weekly quizzes. The following is our assessment of licensee con-ducted training for the two topic area Control functions of the EHC System: Lesson Plan, No. 11.2.0 , Electro-Hydraulic Control System, provided adequate aug-mented training in this area. The lesson plan details the func-tion of the EHC system and components, turbine trips, set points, flow paths, and control functions. The retention of this training by the operators was demonstrated by the results of the weekly quizzes which were found to be comprehensive and adequate to evaluate the level of knowledge of the operators. One operator who failed the quiz was provided with retraining and passed a second qui .

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Predicting indications on Nuclear Instrumentation (NI) during a reactor startup: Lesson Plan No. 11.6.01.002, Approach to Cri-ticality - Peak Xenon and manipulations on the BPT provided adequate augmented training in this area. The lesson plan includes NI response to reactivity changes due to boron or rods during an approach.to criticality. During the BPT manipulations, the operators predicted and observed the results of reactivity additions on the NI indication for a sub-critical reacto Operators have received adequate (performance oriented) augmented training in both topic areas (Items 10 and 13). The weekly quizzes were well written and adequately assessed the operators' knowledge of the subject .5 (0 pen) Unresolved (289/85-08-01): Evaluate Limitorque Operator Deficiencies NRC Inspection Report 50-289/85-08 documented a review of various licensee actions in response to IE Information Notice No. 84-10,

" Motor-Operated Valve Torque Switches Set Below the Manufacturer's Recommended Value." Although the licensee internally reviewed test results for December 1984 testing of selected valves, the licensee performed no formal safety evaluation of the apparently significant deficiencies (Code 1 Category) identified by its vendor. They com-mitted to perform such an evaluation within 90 days of March 8, 198 As of March 22, 1985, the licensee completed that review; during this inspection, the inspector reviewed that evaluation in conjunc-tion with additional discussions with cognizant licensee personne The vendor classified the deficiencies identified during the torque and torque / limit switch testing of December 1984 into four code cate-gories with Code 1 being the most significant- "strongly recommend that the condition noted be corrected immediately in order to assure continued reliable functioning of the valves (s)." The other code categories were problems that were more minor in nature in that they could wait for the next shutdown (assuming an operating plant),

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warranted further review / evaluation during the next scheduled main-tenance, or the information was provided for information purposes only indicating a'long range degradation potential. Accordingly, the licensee used the system provided by Administrative Procedure 1044,

" Event Review and Reportability Requirements," to evaluate all of the vendor identified Code 1 deficiencies. Licensee representatives discussed the evaluation at Plant Review Group Meeting No. 85-16, dated March 11, 1985. Licensee conclusion of the evaluation was that no unreviewed safety question existed and that the deficiencies were not reportable in accordance with 10 CFR 50.72 or 50.7 At the time of the PRG meeting, the licensee resolved all Code 1 deft-ciencies for the specific safety and non-safety related valves teste Licensee representatives reported that even with the Code 1 defi-ciencies the valves still operated properly. A specific valve problem was backseating which was corrected by limit switch adjustments at the time of testing. (In general, the motor control circuit stops the motor in the open direction by limit switch actuation.) Another individual valve problem was a suspected loose lock nut on DH-V2 which was inspected and found to be tight. An additional individual valve problem was RC-V1, Pressurizer Spray Valve, not seating. This is by design because of its frequent open/ shut use during operatio The three remaining common problems were:

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bypass limit switch setting adjustments in a majority of the valves tested;

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torque switch setting adjustment in all valves tested; and,

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grease in the spring pac The licensee's qualitative evaluation addressed each of these common problems in terms of valve operability. The bypass limit switch (LS) bypassed the open torque switch to prevent an inadvertent shutdown of the motor in case a high torque was needed to get the valve off-of its seat. The LS adjustment was to increase the length of time the bypass was in effec Licensee representatives reported that none of the higher than normal opening torques for any of the valves exceeded the torque switch setting; therefore, the valve would have continued to open even if the LS bypass was not in effect. The TMI-1 past practice of having open torque switch settings slightly greater than the closing torque switch settings was condusive toward valve opening on demand independent of the bypass limit switch settin The licensee also evaluated the torque switch adjustment problem as not adversely effecting operability. Of the 22 valves tested, 18 valves had torque switch settings that were above minimum valve for closing thrust (independently determined by another vendor for design

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basis conditions in which the valve must be operated) and below maxi-mum closing thrust allowed. (In general, the motor control circuit stops the motor in the close direction on torque switch to assure proper seating of disc.) Two other valves closing torque switch settings were 5-10% below minimum, which was evaluated as not signi-ficant since the valve properly operated in this condition. One other valve was approximately 30% low, which was high enough to question its operability; that particular valve had no engineered safeguard function and it did operate properly during testing. All torque switch adjustments were made by the vendor to assure a mid-position (" fine tuning") of the torque switch operating lever which serves to actuate either open and close direction torque switche The licensee's evaluation acknowledged the problem of potential for a hydraulic lock of the torque operating mechanism due to grease in the spring pack. This has not been a problem at TMI-1, apparently, due to preventative maintenance checks on these operators since one valve had the problem many years ag The licensee representatives concluded the design / engineering problems of certain operators do not effect operability. They stated that the generic problems with valve operators will be resolved and that additional industrial experience will be factored into this resolutio The inspector questioned the licensee on the aggregate of the defi-ciencies (although evaluated as not adversely affecting operability)

for the set of valves tested. Licensee representatives stated the sampling of valves testod was not random but prejudiced toward those particular valves that exhibited problems most frequently in the pas Pending additional " state of the art" valve testing developments, the licensee expressed confidence that deficiencies adversely affecting valve operability do not exist for the remaining valves (those not tested in December 1984) and they reiterated their past position that the preventive maintenance and surveillance programs were adequate to .

detect valve operability problem Based on the above, the inspector verified that the licensee fulfilled its commitment to properly evaluate the Code 1 valve operator deficiencie This item remains unresolved pending additional review and/or testing by the licensee to assure that valves with "Limitorque" operators are operable during design basis condition .0 Exit Interview The inspectors discussed the inspection scope and findings with licensee management at the exit interview conducted on May 6, 1985. The following licensee personnel attended the meeting:

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E. Eisen, Project Engineer

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C. Hartmen, Manager, Plant Engineer

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H. Hukill, GPUN, Director, TMI-I

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C. Incorvati, GPUN, TMI Audits Supervisor

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R. Neidig, Jr. , TMI-1 Communications

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M. Nelson, Plant Review Group

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S. Otto, TMI-1 Licensing Engineer

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L. Ritter, Administrator, Plant Operations

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M. Ross, Plant Operations Manager

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C. Smyth, TMI-1 Licensing Manager

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R. Toole, Operations & Maintenance Director, TMI-1 As discussed at the meeting, the inspection results are summarized in the cover page of the inspection report. The licensee representatives indicated that none of the subject matter discussed contained proprietary information. Also, discussed were licensee plans for making the plant physically ready to support criticalit Unresolved Items are matters about which information is required in order to ascertain whether they are acceptable items, violations or deviations. Unresolved item (s) discussed during the exit meeting are documented in paragraphs 4.1 and I l

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