IR 05000289/1988021

Insp Rept 50-289/88-21 on 880801-05.No Violations or Deviations Noted.Major Areas Inspected:Test Witnessing & Administration of Control of Local Leak Rate Testing & Followup of Previously Identified Open Items

ML20154E363
Person / Time
Site:	Three Mile Island
Issue date:	08/30/1988
From:	Eapen P, Joe Golla NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20154E361	List: IR 05000289/1988021 ML20154E358
References
50-289-88-21, NUDOCS 8809160296
Download: ML20154E363 (8)

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l U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /83-21 Docket N License No. OPR-50 Licensee: GPU Nuclear FT6. Cox 430 Riddletown, Pennsylvania 17057 Facility Name: Three Mile Island Nuclear Station, Unit 1 Inspection At: Middletown, pennsylvania Inspection Conducted: August 1-5, 1983 Inspector: k mff M eph K. Golla', Re~aMor Engineer

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Approved by: K. t Dr. P.K. Eapen, Chfef, Special Test

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Programs Section, EB, DRS Inspection Sur. mary: Inspection on August 1-5_,_193_3 (Inspection Report No. 50-289/89-21)

Areas _ Inspected: Routine unannounced inspection of test witnessing and administrative control of Local Leak Rate Testing (LLRT) and followup of previously identified open item ;

Results: No violations or deviations were identifie LLRT was found to be implerented adequatel Three open items were close i t

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DETAILS 1.0 Persons Contacted

• R. Barley, TMI-1, Manager of Plant Engineering J. Bashista, Plant Engineering T. Graham, QC Supervisor

• G. Gurigan, GPUN Licensing 0. Hosking, Operations QA Manager

• C. Incorvati, TMI Audit Manager M. R. Knight, Licensing Engineer R. Stoehr, Plant Engineering Department LLRT Cognizent Engineer R. Summers, Supervisor Plant Engineering O. Washko, Plant Engineering P. Webber, I&C Foreman Nuclear Regulatory Commissien

• 0. Johnson, Resident Inspector

• T. Moslak, Acting Senior Resident Inspector

• A. Sidpara, Resident Inspector

• Indicates those present at the exit meeting held on August 5 1933, 2.0 Lecal_ Leak Rate Testing (LLRT) Inspection Purpose and Scope The purpose of this inspection was to ascertain that local leak rate testing is being administered adequately and conducted in compliance with the requirements and commitments referenced in the following section. The LLRT procedures were reviewed for technical adequacy to perforn the intended activity. Other record keeping and LLRT related documentation were reviewed to determine adequacy of overall administrative control of the local leak rate test program, At the time of the inspection approximately 81 of 89 local leak rate tests scheduled for the current outage had been perforned. The total naximum pathway leakage at that point was not yet calculated because testing was still ongoing. The licensee must show that the combined maximum pathway leakage for all penetrations testable under the requirements of 10 CFR 50, Appendix J is less than .6La before start-ing up from this (7R) outag .2 References

• 10 CFR 50. Apperdix J, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactor .

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• TMI Unit 1 Technical Specifications, Section 4.4. * Final Safety Analysis Report (FSAR)

• ANSI /ANS 56.8-1981, Containment Systems Leakage Testing Requirement * USNRC IE Information Notice Nutber 85-71, Containment Integrated Leak Rate Test .3 Documents Reviewed a

Current and previous outage LLRT result * QC Plant Inspection Reports and related 0.' auditing documentatio * LLRT instrument calibration documentatio * Standardization Procedure of Local Leak Rate Testing Rotateters No. 1430-Y-22 Revision * Reactor Building Local Leak Rate Test.ing Procedure; No. 1303-11.18 Revision 42, 2.4 LLRT Procedure Review and Administrative Control The inspector reviewed the LLRT procedures to determine their technical adeauacy. The local leak rate test procedure is structured such that each penetration is treated by a subsection which covers plant status required for the test, test limitations and precautions, diagram of the penetration showing the valve lineup to be used fc" the test, and test equipment require Local leak rate test methods include mass flow-in for penetrations involving containeent isolation valves and pressure decay for airlock These methods are acceptable per 10 CFR 50 Appendix J and current industry practic The inspector verified that the licensee is calculating penetration leakage utilizing maximum pathway methodology per the industry standard ANSI /ANS 56.8-1981, Containment Systems Leakage Testing Reautrements. It is the NRC's position that penetration leakage be calculated in this manner. The inspector verified by an audit of records that containment isolation valves and other penetrations are being local leak rate tested at their required frequenc The inspector also verified that the licensee is performing as-found LLRT's en all penetrations and subsequent as-left LLRT's for those penetrations which received maintenance on some component or seal which affects the pressure boundar No unacceptable conditions were identifie _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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3 2.5 Test Witnessing The inspector witnessed the performance of test activities to verify that qualified test equipment and tools were used and that the test technicians were competent to perform their duties. The following mass flow-in tests were witnessed:

2. Instrument Air Valve No. 6 and 20 at containment penetration Nc. 109 on August 2, 1988, 2.5.2 Once Through Steam Generator (OTSG) Chemical Cleaning penetration Nos. 105 and 106 on August 3, 198 The inspectcr reviewed the ;ystem lineup for the tests witnessed and determined that they were in an effective test configuration. The tests passed their specific target criteria and the penetrations did not require adjustments or maintenance to pressurv retaining components. The test technicians followed approved procedures, utilized qualified test equipment, and were knowledgeable and competent in local leak rate testing. No unacceptable canditions were identifie .0 personnel Training and Qualifications The qualification and training of selected test personnel were discussed with a Itcensee representative. Ir, addition, the inspector evaluated the performance of test technicians during the test witnessing. It was determined that LLRT technicians attended a training class given by exper-tenced LLRT personnel. Attendance records were reviewed by the inspecto The inspector determined that the test technician's qualifications met the requirements specified in ANSI N 18 1-1971 "Selection and training of nuclear power plant personnel". They were knowledgeable of their respon-Oibilities and technical aspects of leak testin No unacceptable conditions were identifie .0 Test Instrument Calibratten The inspector reviewed calibration records for the LLRT instruments (pressure gages and rotameters) being used this outage. All instruments used were found to be in current calibration. The standards used to calibrate the instruments were in current calibration and certified to be traceable to the National Bureau of Standards (NBS). No unacceptable condition was identifie .

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5.0 Quality Assurance and Quality Control  !

l The inspector discussed coverage of local leak rate testing with  !

representatives from the QA and QC organizations. Stveral quality l control plant inspection reports and a quality assurance monitoring report covering local leak rate test activity monitoring this outage were .

reviewed by the inspector. It was determined from the discussions that l QC covers surveillances frem a itst provided by QA which is generated I from the weekly maintenance plan. QA and QC monitoring / inspection report  ;

findings were well documented. The QA and QC representatives interviewed t were knowledgeable of local leak rate testing and their duties as QA/QC inspector / auditors. It was noted that a QA monitor was present at the

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LLRT witnessed on August 3, 19S8, Section 2.5 of this report. The I attention that the QA and QC departments are giving to local leak rate  ;

testing appears to be adequate. No deficiencies were identified within the scope of this revie .0 Plant Tour '

The inspector made several tours of the plant facilities including the  !

intermediate building, turbine butiding, control room, and plant exterior  !

to monitor outage activities and housekeeping. All areas inspected were '

generally clean and free from debris. No unacceptable conditions were  !

identifie t 7.0 Follow-Up of Previously Identified Open Items I f

(Closed) Unresolved item 50-289/87-09-19: plant Modifications /Precedure L Changes to Allow Inservice Testing o(~ Components PTier to Startup From 7R l The licensee was required by NRR to make system modifications to allow I pump flow measurements to be taken per Sect, ton X1 of the ASME Boiler and l Pressure Vessel Code. The pumps involved are the fuel oil transfer pumps l DF-PIA, P18, PIC, P10, the control building chilled water pumps  ;

AH-P-3A/B, and the screenhouse ventilation equipment pceps SW-P-2A/ t Additionally, the testing of six check valves in the fuel of) transfer '

system, DF-V-23A/B, DF-V-7A/B, B/A, A/B, and B/B needed to be proceduralized. The inspector observed the newly installed flow meters in the system piping of the associated pumps identified above. He also reviewed newly revised !$T procedures which require and provide for flow measurements. The revised !$T procedures which implement the changes l are:

i "Emergency Power System" No. 1303-4.16, Revision 4 This procedure provides for the flow meatureeent of the fuel oil transfer pumps and IST of the six fuel oil transfer system check i

valves identified above.

l "IST of AH-P-3A/B and Valves" No. 1300-3N, Revision 1 This 1 procedure provides for the flow measurement of these control l building chilled water m ps.

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. "!ST of SW-P-2A/B and Valves" No. 1300-3M A/B, Revision 22. This procedure provides for the flow measurement of these screenhouse ventilation equipment pump The licensee indicated that the pump flow reference values required by Section XI of the ASME Code would be established after the first perform-ance of each test. The inspector reviewed the modification packages which administered the installation of the flow devices in the above systems. All provisions made for the flow measurement of pumps and the IST of fuel transfer system check valves to address this issue were found adequate. This item is close (Closed) Unresolved item $0-289/S6-21-01: Penetration Pressurization

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'ystem Check Valves at Containment Pl:rge Valves Interspace to Receive S

Corrective Maintenance Af ter Failing the November 1M Unit 1 Containment Integrated Leak Rate Test (CTLD)

During the initial stages of the November 1986, "as-found" CILRT performance the licensee identified and quantified several containment boundary leaks. Of these, the containment purge line manifold "J" was shown to be a major contributor to the boundary leakage which failed the as-found portion of the CILRT. Check valves PP-V-101 and 102 were icentified as the source of leakage at this manifold. These valves were subsequently isolated from the containment boundary and a CILRT was conducted for the purpose of establishing an as-left overall containment leak rate. Upon co'npletion of the CILRT the licensee was evaluating corrective and preventive measures for the aene+, ration pressurization check valves. This unresolved item was written to track completion of the above corrective / preventive measure The leaking check valves are associated with the automatic actuation of the penetration pressurization system into the interspace between reactor building purge valves AH-V-1A/1B and IC/10. A modification has been performed by the licensee whereby check valves PP-V-101/102/133/134 were converted to normally closed manual glove valves and previous automatic initiation valves were failed ope The check valve internals were removed from the valve bodies and renewable threaded globe valve seat rings and bonnet assemblies were installed. Check valves PP-V-101/102/133/134 were then renumbered PP-V-210/212/213/211 respectively. The licensee is maintaining the new globe valves in the normally closed positio The inspector while on site observed the new globe valves in plac Local leak rate testing of these valves was ongoing. The inspector also revie.ed the modification package associated with the above changes to the penetration pressurization system. He verified that a safety evaluation was written by the licensee per the requirements of 10 CFR 50.59, "Changes, tests and experiments" addressing the modification. The inspector reviewed the safety evalcation and found it acceptable. No credit is taken in the plant Technical Specifications or FSAR for the penetration pressurization __

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. r system performirg an active, automatic function, therefore, there will be '

no increase in the probability or consequences of an accident previously evaluated. The licensee indicated that leak testing history has demonstrated that the reliabilit,y of globe valves is much better than check valves and  ;

that containment integrity will be made more reliable by this modificatie The inspector found the licensee's assessment concerning improvement to l the containment boundary acceptable. The new globe valves will be added 7 to the plant Technical Specification 4.4.1.2.1 which identifies isolation '

valves testable as defined in 10 CFR 50, Appendix This item is close [

I (Closed) Unresolved Item 50-289/65-26-06: Nuclear

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Service Valve Penetration Design Adequacy  ;

i A question arost during a previous inspev. ton concerning the design ,

adequacy o' three containment penetrations. The penetrations in question  :

are Nos. 404, 405, and 406 of the Nuclear Services Closed Cooling Water (NSCCW) System. These penetrations are Nuclear Services (NS) Closed Cooling water rctura lines (stismic class !) from the Raactor Building Fan .

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Motor Coolers. Each of these return lines has a relief valve (NS-B-36A B & C) Scated between the outside containment wall and a control  !

velve. The supply lines have a single isolation valve at the outside I containment wall. The coolers are located inside containmen l The purpose of the relief valves is to provide overpressure protection I for the emergency fan motor coolers We to thermal expansion of trapped fluid. The thermal expansion would result from heating of the fluid in f '

the cooler motor with both the inlet and outlet isolation valves close This system is closed loop. The question which arose concerning the design adequacy of these penetrations addressed the implications of ,

having a stuck-open relief valve outside of contairment concurrent with a  !

l break of the NSCCV piping inside containment. The relief valves are i nominally set at 175 psig and tne lines are needed for motor coolin !

l Containment accident pressure is 53 psig. This piping system is *

seismically designed and needed for containment post,-accident mitigation.

This issue was addrecsed previously by a region based specialist ,

j inspector, who concluded that there was no safety concern. For  ;

i an accident scenario in which the cooler pressurs integrity is breeche '

i the maximum pressure applied to the understat of the relief valve would f l be the peak contairment accident pressure (53 psig) which is insufficient to lift the relief valve (set to lift at 175 psig). If the valve did j lift, with inlet and outlet isolation valves closed, to provide thermal relief it would be passing uncontaminated closed cooling water. This ((

! would imply that the cooler's pressure boun g was still intact, thus [

! precluding any escaperent of building atm1 sphere to the oatside. A i

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postulated NS pipe break with containment pressure at the relief sotting

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(175 psig) is beyond the design basis even ,

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Another question addressed the penetration design with regard to current general design criteria (GDC) No. 57 of 10 CFR 50, Appendix A. This criterion states that containment isolation valves for closed-cycle systems that penetrate containment either be automatic or locked closed or capable of remote isolation. A relief valve is none of the above. However, criterion 50 of the GDC states that penetrations, and heat removal systens shall be designed so that the containment structure can accommodate, without exceeding the desigi. leakage rate, the calculated pressure and temperature conditions, resulting frem any loss-of-coolant accident. The reactor building fans and the NSCCW to th se fan motors are needed for post accident mitigation. The normal 6 id post accident position of the isolation valves for each loop (there are three loops) is open. To require tha+

these penetrations conform to a requirement for valves that isolate during an accident would be incongruous. Furthermore, the design is in accordance with FSAR commitments on which the operating license was issued in 1974.

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Accordingly, this aspect ./ the unresolved item is close Finally, another question was raised regarding the inaccurate depiction of these NSCCW penetrations in the FSA The updated FSAR Figure 5.3-1, Valve Arrangement No. 23 did not show the subject relief valves on the

. return lines between the outside isolation valve and the containment j wall. A licensee representative indicated at the time that the system

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drawing in other sections of the FSAR was accurate and that Figure 5.3-1 drawings reflected general arrangements only. Since that time the licensee has revised Valve Arrangement No. 23 of Figure 5.3-1 to show the i relief valve in the return lines. The inspector reviewed the revised

Figure 5.3-1 and found it acceptable. This item is now closed, l

8.0 Exit MeeMng n Licensee management was informed of the purpose and scope of the inspection at the entrance intervie The findings of the inspection were pericdically ciscussed and were summarized at the exit meeting on August 5, 1988. Attendees at tho exit meeting are listed in Section of this report. At no time varing the inspection was written material provided to the licensee by the inspectors. The licensee did not indicate that the inspection involved any proprietary informatio l t

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