Text

TMI-1 Reactor Containment Bldg Integrated Leak Rate Test
	ML20090H298
Person / Time
Site:	Three Mile Island
Issue date:	04/30/1984
From:	Roxanne Summers; GENERAL PUBLIC UTILITIES CORP.
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THREE MILE ISLAND NUCLEAR STATION UNIT 1 i

Euclear .

REACTOR CONTAINMENT BUILDING INTEGRATED LEAK RATE TEST APRIL 1984 PREPARED B JJ* Cf R. L. Summers GPUN V

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REVIEWED BY # ~7 1984 g ,,

APPROVED BY ejj h 7- /9-6 /

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TABLE OF CONTENTS

- 1984 ILRT REPORT 1.0 . INTRODUCTION 2.0 GENERAL AND TECmICAL DATA 3.0 ACCEPTANCE CRITERIA 4.0 TEST INSTRtMENTATION 5.3 TEST PROTDURE 6.0 METHODS OF ANALYSIS

^

7.0 DISCUSSION OF RESULTS 8.0 TYPE B AND C LEAKAGE RATE HISTORIES APPENDICES .

A. ILRT INSTRtNENTATION SCHEMATIC B. AVERAGE TEMPERATURE AND MAXIMUM TEMPERATURE DIFFERENTIAL TABLE C. 24 HOUR TEST REDUID LEAKAGE RATE DATA - TABLE D. 24 HOUR TEST - MASS / TEMPERATURE PLOT E. 8 HOLF. SUPERIMPOSED LEAK TEST - REDUCED LEAKAGE R ATE DATA - TABLE F. 8 HOUR SUPERIMPOSED LEAK TEST - MASS / TEMPERATURE PLOT G. 1982 LOCAL LEAK RATE TEST REPORT H. 1983 LOCAL LEAK RATE TEST REPORT I. 1984 LOCAL LEAK RATE TEST REPORT L-

o

1.0 INTRODUCTION

The Three Mile Island Nuclear Station Unit 1 reactor containment building was subjected to a periodic integrated leak rate test during the period f rom April 15, 1984 to April 19, 1984. The purpose of this test was to demonstrate the ac ptability of the building leakage rate a t the calculated design basis accident pressure of 50.6 psig

, (Pa . The allowable leakage is defined by the design basis

- a dent, appaled in the safety analysis, in accordance with site emosure guidelines specified by 10 CFR 100. For Three Mile Island Nuclear Station Unit 1, the maxinum allowable integrated leakage rate a t the design basis accident pressure is 0.10 per n t by weigh t per da y (La) -

Testing was performed in a ccordance with the requirements of 10 CRF 50, Appendix J, ANSI N45.41972, ANSI /ANS-56.8-1981 and the promdural requirements as stated in GPU Nuclear Corporation Three Mile Island Nuclear Station Unit 1 Surveillance Procedure 1303-6.1.

This procedure was recommended for approval by the Three Mile Island Nuclear Station Unit 1 Plant Review Group and approved by the Operations and Maintenanm Director TMI 1 prior to the commencement of the test. All testing was performed by GPU Nuclear Corporation with the technical assistanm of Volumetrics, Inc. Procedural and calculational methods were witnessed by Nuclear Regulatory Commission personnel and audited by the GPU Nuclear Corporation site Quality Control staf f.

The combined loml leakage rate f rom the reactor containment building isolation valves and penetrations, required to be tested by 10 CFR 50 Appendix J, was less than 40 per nt of the maximum allowable leakage rate (L a), well below the allowable value of 60 permnt, a t 50.6 psig prior to the caninencement of the integrated leak rate test.

The 1984 " As Found" leakage through containment isolation valve IC-V3 was about 24,000 scan. Subsequent repair prior to the ILRT failed to correct the leakage. The leakage actually increased to abou t 70,000 s can. During the ILRT IC-V3 was maintained shut with its inboard isolation valve IC-V2 also shu t in accordance with the normal ILRT valve lineup requirements. No special consideration or valve lineup was performed on the affected penetration during the ILRT. Repair was a ccomplished af ter the ILRT and the " As Lef t" leakage through IC-V3 was 7813 scan, which was acceptable.

The calculated leakage rate based on the mass point method of analysis, for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was found to be 0.0374 percent by weight per day a t 50.6 psig. The leakage rate at the upper bound of the 95 percent confidence interval was 0.0405 percent by weigh t per day which is well below the allowable leakoge rate of 0.075 percent by weight per day a t 50.6 psig. The low integrated leakage rate from the reactor containment building, approximately one-half of the allowable 1 0006U L.

leakage rate, provides assurance tha t the reactor containment building is capable to perform its intended safety function.

Since the Industrial Cooler System was in operation during the integrated leak rate test, addition of the local leakage rate of the system isolation valves, RB-V2A and RB-V7, (1321 sccm and 1201 sem respectively) to the measured integrated leakage rate must be

" mnsidered.' The combined local leakage rate of these isolation valves

. is 0.0015 permnt by weight per day. The addition of these values changes the calculated leak rate a t the upper bound of the 95%

confidenm level to 0.0420.

The stpplemental instrumentation verifistion a t P ac was 15.8% which is within the 25 permnt of L requirement a of 10 CFR 50, Appendi x J,Section III A.3(b) 2 0006U

2.0 GD4ERAL AND TECFNICAL DATA 2.1 GE9ERAL DATA Owner: General Public Utilities Nuclear Corporation Dod<e t No. : 50-289 Lomtion: Three Mile Island near the East Shore of the Susquehanna River in Dauphin County, Pennsylvania.

Containment

Description:

Reinforced concrete structure composed of cylindrim1 walls (prestressed with a post-tensioning tendon system in vertical and horizontal directions),

with a flat foundation ma t (conventional reinforcing) and a shallow dome roof (prestressed

' tilizing u a three-way past tensioning tendon system). The inside surface is lined with a 3/8" thick carbon steel liner.

Date Test Completed: April 19,1984

-2.2 TECFNICAL DATA Containment Net Free Volume: 2 x 106 cubic feet Design Pressure: 55 psig Design Temperature: 2810F Calculated Accident Peak Pressure: 50.6 psig Calculated Accident Peak Temperature: 2810F 3 0006U

d-3.0 ACCEPTANCE CRITERIA Acceptanm criteria established prior to the test and as specified by 10 CFR 50, Appendi x J, ANSI N45.4-1972 and ANSI N56.8-1981 are as follows:

a. The measured leakage rate (km) at the caloJlated design basis

, accident pressure of 50.6 psig (Pac) shall be less than 75

. percent of the maximum allowable leakage rate (L a ), specified as 0.10 permnt by weight of the building atmosphere per day at the upper bound of the 95 per nt confiden level. The ac ptance criteria is determined as follows:

La = 0.10%/ day 0.75La = 0.075%/ day

b. The test instrumentation shall be verified by means of a supplemental test. Agreement between the containment leakagc measured during the Type A test and the containment leakage determined during the supplemental test shall be within 25 per nt of L a-4 0006U

+

4.0 TEST INSTRlNENTATION 4.1 SLNMARY OF INSTRLNENTS Test instruments employed are described in the following subsections:

4.1.1 Temperature Indicating Syste_m

. Resistan Temperature Detectors Quantity 23 Manufacturer Yellow Spring Instr.

Type YSI Model, 4150-1/4-6-3-138-AW-G1/2-QR (platinum)

Range, OF- 60-120 Accuracy, OF 210 Sensitivity, OF 2001 4.1.2 Dewpoint Indicating System Dew 11 Elements Quantity 10 Manufacturer Fo > boro Type BD154WB, Lithium Chloride Range, OF 40-100 Accuracy, OF 10 2 Sensitivity, OF 11 0 4.1.3 Pressure Monitoring System Precision Pressure Gauges Quantity 2 Manufacturer Te>as Instruments (Modified by Volumetrics to interface with ILRT System)

Type Model 145.02 5 0006U

Range, psia 0-10 0 Accuracy, psia zu n'S% of indicated pressure Sensitivity, psi 10001 4.1.4 Supplemental Test Flow Monitoring System

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. Flowmeter Quantity 2 Manufacturer Sierra Type Model 14636 Range , scf m 0.0 - 20.0 Accuracy, 12% reading Sensitivity, scfm 105% of full scale 4.1.5 Inputs from the aforementioned sensors (with the e xception of the output from the mass flowmeters) is forwarded to the Data Acquisition System (DAS) for conversion, display and forwarding to the computer.

The installed DAS unit was a Model A-100, manufactured by Volumetrics.

The DAS unit has the capability of monitoring over 100 channels. For the ILRT, the channels were utilized as follows:

a.) Precision Pressure Gauge - 2 channels b.) RTDs - 24 channels (Channels 1 to 24,1 installed RTD was not usable during the ILRT)

c. ) Dewcells - 10 channels (Channels 30 to 39)

Output from the DAS went to both a hardcopy printer and to a computer.-

4.1.6 Sensor input conditioning cards, precision pressure gauges, DAS unit, etc., were purchased as a rad < mounted unit from Volumetrics by GPUN in March,1984. Details of the unit and equipment specifications are available onsite for review.

4.1.7 Af ter the DAS unit converted the signal input into the desired

- parameter of temperature, pressure, etc., these values were printed out on a hardcopy printer and forwarded to the onsite computer for processing. The computer used averaging and weighting factors as delineated below:

6 0006U

1'

, a.) Pressure - average of the two inputs b.) Temperature - an equal weighting f actor of about 4.35% was

- assigned to each of the 23 operable RTDs. TMI-l has 24 installed RTDs but one of these (TE-655V) could not be used due to a ground somewhere in one of the leads. This RTD was assigned a weighting f actor of zero (0.0%) which effectively removed the signal input

, from the average temperature determination.

c.) Dewcells - each of the 10 dewcells was assigned a weighting factor of 10%, which effectively resulted in a straight average of the 10 values.

4.1.8 The accuracy of the DAS unit with respect to the different monitored parameters is given below:

a.) Pressure - direct transfer of the number of counts from the precision pressure gauges to the computer.

b.) Dewpoint accuracy: 1 2.0% CF.

c.) Temperature: 1 0 10F. (607. to 12CPF. range) 4.1.9 All operable RTDs and dewcells were assigned equal weighting f actors.

This is because:

a.) There are' very few cubicles inside the Reactor Building, b.) There is free communication between all levels of the building and also between the cubicles and the Reactor Building.

c.) The air inside the Reactor Building is continually recirculated by the installed ventilation system.

d.) Almost all of the equipment in the Reactor Building, with the e>ception of the aforementioned recirculating f ans and required instrumentation, was deenergized during the test. This eliminated any heat producing equipment in the building which could cause local hot spots.

e.) No stratification has ever been observed during an ILRT.

'4.1.10 See Appendix A far an instrumentation layout followed by a listing of the computer weighting factors.

7 0006U

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l 4.2 - SCHEMATIC ARRANGEMENT The arrangment of the four measuring systems summarized in Section 4.1 is depicted in Appendix A.

The arrangment of temperature sensors can be grouped into five levels as follows:

. Level Elevation Sensors 1 287 feet TE-655R TE-655S TE-655T TE-655U 2 314 feet TE-655M TE-655N TE-6550 TE-655 P TE-655Q 3 346 feet TE-655A TE-655G TE-655I TE-655K 4 365 feet TE-6550 to TE-655J 405 feet TE-655L TE-655W TE-655X TE-655B 5 437 feet TE-655C TE-655E TE-655H' TE-655F The average Reactor Building temperature varied by only .200F during the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test. Of the 23 RTDs utilized, TE-655X (365' Elev. ) indicated consistently lov 'and TE 655N (314' Elev.) indicated consistently high. The average difference was 2.02 F. These results are su==arized in Appendix B.

This analysis demon'strates that there was no large regional temperature variation in the Reactor Building and also that no large temperature fluctuations occurred during this ILRT. Small fluctuations, as discussed in section 5.2.2, were noted. Operation of the three Reactor Building recirculation units provided satisfactory temperature equalization throughout the building.

8 0006U

4.3 CALIBRATION CHECKS Temperature, dewpoint, pressure and flow measuring systems were checked for calibration before the tes t in accordance with GPU Nuclear Corporation Promdure 1430-Y-23, as recommended by ANSI N56.8-1981.

The results of the calibration checks are on file at Three Mile Island Nuclear Station Unit 1. The supplemental test at 50.6 psig confirmed

,. the instrumentation acceptability.

4.4 INSTRUMENTATION PERFORMANCE The twenty-three temperature sensors, ten dewcells, two precision pressure gauges, flowmeters, and readout equipment performed satisfactorily throughout the integrated leakage rate test. No sensor or readout equipment malfunctions occurred during performance of the test.

4.5 INSTRUMENTATION SELECTION GUIDE VALUE Justifistion of instrumentation selection was accomplished, using manufacturer's sensitivity and repeatability tolerances stated in Section 5.1, by computing the instrumentation selection guide (ISG) value. Utilizing the methods, techniques and assumptions in Appendix G to ANSI N56.8-1981, the ISG was computed for the absolute method as follows:

a. Conditions La = 0.1%/da y P = 65 psia T = 72.30 OR = 531.990R dry bulb (typical)

Tdp = 67.20 F dewpoint (typical) t = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

b. Total Absolute Pressure: ep Sensor sensitivity error (E): 10001% of full scale Measurement system error (c): 10002% of full scale ep = 1 _ (Ep)2 + (cp)2 1/2/ no. of sensors 1/2

. ep = 1{(O' 001)2

. + (0.002)2- 1/2/ 2 h/2 ep=1 0.0016 psia 9 0006U

4

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c. Water Vapor Pressure: epy m.

Sensor sensitivity error (E): + 0.10F Measurement system error (c), excluding sensor: 1 0 10F

'At a dewpoint temperature of 67.2aF, the equivalent water vapor .

c

, pressure change (as determined from the steam tables) is 0.0114.9' ,

o psia /T.-

Ep_

y = + 0.10F (0.01149 psia /0F) . .

Epy.= 1 0.001149 psia '

cpy = 1 0 10F (0.01149 psia /0F) cpy = 1 0.001149 psia ,

_ _ _ _ ~~

epy = 1_(Ep y)2 + (cp y)2_ 1/ _ no. of sensors 1/2 epy = + (0.001149)2 + ( o,001149)2_ 1/2/

/ _ 10~ ll/2 .

epy = 1 0.0005138 psia ;_

d. Temperature: eT - /

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No. of Sensors: 23 ~

Sensor sensitivity error (ET ): .1 '0 010F = + 0.010R System Error (cT ): ' + 0.020F = 1 0.02Jh -

1/2 eT = + _Ef)( +

.(CT)2_ 1/2/ _no, of sens rs -

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1/2.

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eT = + (.01 )2 -+ ( 0.,02)F 1/2 / - j

/ __23_ ^'

n eT = + 0.0046620R- .~ , ,

,_~_ s, a

e. Instrumentation Selecticn Guide-(ISG) ' v- _T, ' A 2

1/2 s -

ISG = 12400 2 "P_ + 2 *Pv + "T 7 --.- I t _

P Pg, #

2(T ,s,

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-2 2 s 2 ,. , . 2 - 1/2 ,

0.0016 +2 0 000514 +2 0.00466- >

ISG = 1 2400 ..

24 65 65 531.99

."5 ISG = 1 0.00386 ,

The ISG does no t exceed 0.25.La (0.025%/ day) and it is ,

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1 therefore concluded that the instrumentation selected was acceptable for use in determining the reactor containment integrated leakage rate.

4.6 SUPPLEMENTAL VERIFICATION 4.6.1 8 Hour Superimposed Test

. In addition to the calibration ched<s described in Section 4.3, test instrumentation operation was verified by a supplemental 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow test subsequent to the completion of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months leakage rate test.

This test consisted of imposing a known calibrated leakage rate on the reactor containment building. Af ter the flow rate was established, it was not altered for the duration of the test.

During the supplemental test, the calculated leakage rate was Lc=Ls y +L a where, Lc= calculated composite leakage rate consisting of the reactor building leakage rate plus the imposed leakage -

ra te La= imposed leakage rate Lsy = leakage rate of the reactor building during the supplemental test phase Rearranging the above equation, t'=

v 'c - Lo The reactor containment building leakage during the supplemental test y can be alculated by subtracting the known superimposed leakage rate f rom the calaalated composite leakage rate.

The containment building leakage rate during the supplemental test (Lyi) was then compared to the calculated reactor containment building leakage rate during the preceding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test (Lam) to determine instrumentation acmptability. Instrumentation is considered acceptable if the agreement between the two building leakage rates .is within 25 percent of the maximum allowable leakage

, rate (L a)

11 0006U

i 5.0 TEST PROCEDURE 5.1 _ PREREQUISITES Prior to commencement of reactor containment building pressurization, the following basic prerequisites were satisfied.

,. . a. Proper operation of all test instrumentation was verified. As noted, one RTD was declared inoperable prior to the start of the test.

b. All reactor containment building isolation valves, with the ex ption of mose within the R. B. Cooling System and Decay Hea t System, were closed using the normal made of operation. All associated system valves were placed in post-accident positions.

The Reactor Building Cooling System was in servi for temperature control during the test and the Decay Hea t System was in servi to maintain the plant in a safe condition during the tes t.

c. Equipment within the reactor containment building, subject to damage, was protected from external differential pressures,

d. Portions of fluid systems which, under post-accident conditions become extensions of the containment boundary, were drained and vented.

e. The Penetration Pressurization and Fluid Blod< Systems were depressurized. Manometers were installed a t penetration pressurization manifolds to provide means for detection of leakage into the system.

f. Manameters or pressure gauges were installed on the purge valve interspaces, and access hatch interspaces to provide means for detection of leakage into such systems.

g. Local leakage rate testing of containment isolation valves and penetrations was concluded except that one containment isolation valve (IC-V3) had high local leakage prior to the ILRT and was repaired and retested af ter the ILRT.

h. Potential pressure sources were removed or isolated from the containment.

i. All accessible liner weld channels were vented to the containment a tmosphere.

J. A general inspection of the accessible interior and e xterior areas of the containment was completed.

I 12 0006U l

5.2 TEST PERFORMANCE 5.2.1 Pressurization Phase Pressurization of the reactor containment was started on April 16, 1984, a t 1700. The pressurization rate was approximately 1.9 psi per hour rather than the desired 2.5 psi per hour due to low compressor sizing. Building pressure and temperature were monitored hourly and

. the amperage required by the recirculation unit fans ( AH-E-1A,18,1C) was monitored about every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . When containment internal pressure readied 12 psig on April 17, 1984, pressurization was secured.

Subsequently at 0830, an inspection team entered containment to perform the 12 psig inspection. During the 12 psig inspection, one dewcell and two RTDs were found in their storage position. They were relocated to their proper test position. The 12 psig internal inspection was completely satisfactorily and pressurization was restarted a t about 1000 on April 17, 1984.

During pressurization to the 50.6 psig pressure level, the following observations were made:

a. The water in saveral of the penetration pressurization manometers was blown out. This was apparently due to slight leakage past the fuel transfer tube flange gasket.

b. Manometers on the purge supply and e>haust interspaces were replaced with the original pressure gauges. The purge e >haust and supply interspace showed an increase in pressure indicating slight leakage from the inner valve.

c. A very gradual decrease in pressurizer level was noted as pressure was increased. Concurrently, a gradual increase was noted in Borated Water Storage Tank level. The allowed pressurizer level band was broadened as required to a ccomodate this level decrease. No makeup water was added to the RCS until after completion of the ILRT. No increase in Reactor Building sump level was observed. This loss of water from the reactor building would cause the leak rate to appear greater than actual.

Final pressurization'was secured at 1119 on April 18, when Reactor Building pressure indicated between 50.7 and 50.8 psig. The stabilization period commenc.ed at 1130 on April 18. The stabilization criteria was satisfied and the twenty-four (24) hour test began a t 1600. Reactor Building pressure a t tha t time was 50.727 psig with an average temperature of about 72.47.

5.2.2 Integrated Leak Rate Testing Phase At 1600 on April 18, 1984, leakage rate testing was initiated. The following fif teen minute frequency test data was collected (See Appendix C) .

13 0006U

a. Pressure indicated by the two precision gages was recorded.

b. The twenty-three RTD temperatures were recorded.

c. The ten dewpoint values were recorded. The average of the ten values was converted to vapor pressure by a computer. This permitted correction of the total pressure to the partial pressure of air Ly subtracting the vapor pressure.

. The use of vapor pressure (Pwy), average temperature (T) and the total pressure (Pr) is described in more detail in Section 6.1. All original data is on file at Three Mile Island Nuclear Station Unit 1.

The plot of average tenperature and weight of air was maintained for each fifteen minute reading (See Appendix D).

At 0100, minor temperature fluctuations were noted due to cycling of the Industrial Cooler. Initially, the Shift Supervisor attempted to regulate air temperature by taking manual control of the system.

Small perturbations resulted, however. At 0700, the Shift Supervisor was told to maintain the Industrial Cooler Fans in the existing configuration in manual control. No further temperature perturbations resulted.

The aforementioned temperature and pressure indications were monitored at fif teen minute intervals by the Volumetrics DAS. The data was then forwarded to a local mini-computer and also to a hard copy printer.

The data was retained by the compater on a floppy disc. The computer also had a graphics display associated with it to allow the Shift Engineer to monitor and more easily trend ILRT parameters.

Periodically, the calculated and measured leak rate values were obtained. A graphic display of leak rate, temperature, pressure, etc., was also available. This graphics display was very useful.in monitoring and trending the inputs into the computer.

At 1600 on April 19, 1984, the integrated leak rate test was concluded. A final computer run for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of data collection resulted in a calculated containment integrated leakage rate of 0.0374 percent per day. At the 95 percent UCL, the containment integrated leakage rate (corrected for RB-V2A/7) was 0.0420 percent per day.

This confirmed the preliminary data and was significantl / below the 0.075 percent per day limit a t the 95 percent Upper Confidence Level.

5.2.3 Supplemental Leakage Rate Test Phase 8 Hour Superimposed Leak Rate Af ter the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months integrated leak rate test data was obtained and evaluated, the leakage rate found to be acmptable, and a release permit had been obtained,-a known leak rate was imposed at 2000 on 14 0006U

April 18, 1984, on the reactor containment building through a calibrated flowmeter for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . During this time, temperature, pressure, and vapor pressure were monitored as described above. The average superimposed leak for the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period was 3.27 SCFM. This value equates to an integrated leak rate of 0.0538.

w%/ day. If this value is added to the calculated leak rate of 0.0374 w% day, the leak rate is 0.0912 w%/ day. This is within 0.0158 w%/ day

,. of the 8 hr. leak rate of 0.0754 w%/ day, and, therefore, meets the requirements to be within 25% of La (0.025) w%/ day.

5.2.4 Depressurization Phase Af ter all required data was obtained and evaluated. The supplemental test results were found to be ac ptable. Permission from the Rad Con Department and O&M Director was obtained, and depressurization of the reactor containment building was started. A post test inspection of the building revealed no unusual findings.

U 15 C006U

6.0 METHODS OF ANALYSIS 6.1 GENERAL DISCUSSION The absolute method of leakage rate determination was employed during testing at the 50.6 osig pressure level. The Volumetrics computer program calmlates the permnt per day leakage rate using the mass

, point method of data analysis. The results presented are based on the

. mass point method.

The mass point method of computing leakage rates uses the following ideal gas law equation to caloJlate the weight of air inside containment for eadi quarter hour:

W= 144 PV = g RT T where, W = mass of air inside containment, lbm 2

K = 144 V/R = 5.3983 x 106 lbm - R - in .

lbf P = partial pressure of air, psia T = average internal containment temperature, OR V = 2.0 x 106 ft3 The partial pressure of air, P, is calculated as follows:

P = 'T1 + T2 - P where, PTl = true corrected total pressure from PI-390, psia PT2 = true corrected total pressure from PI-391, psia Pwy = partial pressure of water vapor determined by averaging the ten dewpoint temperatures and converting to vapor i pressure, psia.

The average internal containment temperature, (T), is calmlated as follows:

i T = sum of 23 R TDs + 459.690R 23 16 0006U

The weight of air is plotted versus time for the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test and for

- the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months supplemental test. The Volumetris computer prograc fits the locus of these points to a straight line using a linear least squares fit. The equation of the linear least squares fit line is of the form W = W o + Wi t where W 1 'is the slope in Ibm per hour and Wo is the weight at time zero. The least squares parameters are calaalated as f allows:

, . Et EW -Et It W s

xx

, y ,N Et W - Et g EW 1

S xx where, 2 2 Six N 'E t 1 - (E t t)

The weight percent leakage per day can then be determined from the following equation:.

wt. %/ day = 2400Wy w .

where the negative sign is used since W1 is a negative slope to

.emress the leakage rate as a positive quantity.

6.2' -STATISTICAL EVALUATION After performing the least squares fit, the following statistimi parameters were calaJlated:

a. ' Standard error of confidence for the curve fit (Se) .

b. Limits of the 95 per nt confidence interval for the curve fit.

The signifimnce of. the measured leakage rate can then be evaluated in

. view of the number of data points exceeding the limits of the 95 percent confidence interval and by the magnitude of the upper bound of the 95 permnt confidence interval for the leakage rate.

l' Standard error of confidenm is defined as follows:

E 7 gW1 - (W, +W 1 t)g 2-m g , t N-2 where, l

17 0006U

Wi = observed mass of air

(%+W1 t i) = least squares calmlated mass of air N = number of data points This parameter is an e >pression of the difference between an observed

" and a calrmlated (least squares) mass point. The 95 percent

o. confidenm interval of the fit is twi the standard error of confidence (2Se). The " degree-of-fit" is evaluated by determining the number of data points, iW , not falling in the interval (Wo+

Wi t) + 2Se .

The 95 permnt confidenm limit for the mass leakage rate was alculated per the fornulas in ANSI N56.6, N274. It is a measure of the un rtainty of. the measured leakage rate.

=

18 0006U

7.0 DISCUSSION OF RESULTS 7.1 RESULTS AT P 2 The method used in calculating the mass point leakage rate is defined in Section 6.0. The result of this calculation is a mass point leakage rate using absolute values of 0.0374 %/ day.

~

o The 95 permnt confidence limit associated with this leakage rate is 0.0031 percent per day. Thus, the leakage rate at the upper bound of the 95 percent confidence interval becomes Lam = 0.0374%/ day Lam (at 95% Confidence Limit) = L, + 0.0031 = 0.0405%/ day The measured leakage rate a t the upper bound of the 95 per nt confidence level is below the ac ptance criteria of 0.075 per nt per day (0.75 La ) . A comparison of eadi of the observed weights with the weights calculated using the least squares line reveals only four of the ninety-seven data points do not lie within the 95 percent confidenm interval. Therefore, reactor containment building leakage a t the calculated design basis accident pressure (Pa ) of 50.6 psig is considered to be acmptable.

7.2 SUPR_EMENTAL TEST RESULTS -

Af ter conclusion of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test at 50.6 psig, a mass flowmeter was placed in service and a flow rate, corrected for pressure and temperature conditions, of 3.27 SCFM was established. This flow rate is equivalent to a leakage rate of 0.0538 percent per day. Af ter the flow was established, it was not altered for the duration of the supplemental test.

The measured leakage rate (L c ) using absolute values during the supplemental test was calculated to be 0.0754 per nt per day using the mass point method of analysis. The 95 percent confidence interval associated with this leakage rate is 0.0150 permnt per day. Only two of the 33 data points were outside of the 95 percent confidence interval.

The building leakage rate during the supplemental test is then determined as follows:

Ls y =Lc-La Ls y = 0.0754 - 0.0538 %/ day Ls y = 0.0216 %/ day 19 0006U

( .

-Comparing this leakage rate with the building leakage rate measured during the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test yields the following:

b ~

am v' =

(0.0374-0.0216) = 0.158 L 0.1 The building leakage rates agree within 15.,8 percent of L which is below the acceptance criteria of 25 percent of La-Therefore, the acmptability of the tes t instrumentation is considered to have been verified.

O e e

1 9

20 0006U b _

-J8.0 ' TYPE 8 AND C LEAKAGE RATE HISTORIES Refer to Appendims G, H, and I for the report on Type B and C testing performed since the previous Type A test.

9 e

.c 6

21 00060

9.0 REFERENCES

1. SP 1303-6.1, " Reactor Building Integrated Leak Rate Test",

Metropolitan Edison Company Surveillance Procedure. -

2. ANSI N45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Reactors", American Nuclear Society, (March 16, 1972).

~

o. 3. Steam Tables, American Society of Mechanical Engineers, (1967).

4, 1430-Y-23, " Reactor Building Integrated Leak Rate Test Instrument Calibrations", Metropolitan Edison Company Procedure.

5. ANSI N56.8-1981, N274, " Containment System Leakage Testing Requirements", American Nuclear Society, (February 19, 1981).

6. Volumetrics Performance Test Reports. Document No . 83-02-7682.

Dated 4/2/84.

7. 10CFR50 Appendix J.

22 0006U

1 F.

6 APPENDIT S 0006U

{

O O

4 O

APPDDIX A INSTRlNENTATION SCHEMATIC AND VOLUME WEIGHTING DAT A

}

l 24 0006U

= - . _ _ _ _ _ _ _ . _ _ _. -

TMI-1 1984 INTEGRATED LEAK RATE TEST VOLUMETRICS PASO ROBLES CALIFORNIA ILRT SUS-VOLUME WEIGHTING PROGRAM TYPE OF CONTAINMENT DAS TYPE OF CCNTAINMENT DAS SENSCR WEIGHTING FACTOR (Il CHANNEL NO. SENSOR WE!EHTING FACTORtIl CHANNEL NO.

! RTD4 1 4.35 2 RIDI 2 4.:5 RTDs 3 4.35 4 RTDI 4 4.;5 3

RTDI 5 4.!5 6 RTDI 6 4.;5 5

7 RIDI 7 4.35 8 RTDI 8 4.35 9 RTDI 9 4.35 If RTDtti 4.:5 11 RTD411 4.35 12 RTDil2 4.!5 13 RTDe13 4.35 14 RTD414 4. 5 15 RTD015 4.35 16 RTD816 4.35 RTDt!7 4.35 18 RTDt!8 4.;5 17 19 RTD019 4.35 28 RTDt28 4.35 21 RTDt21 4.35 22 RTDf22 H.fi RID 423 4.35 24 RTD124 4.;5 23 25 DEW CELLI 1 If.N 26 DEW CELL 4 2 If.H 27 DEW CELLI 3 If.fi 28 DEW CELLt 4 18.18 29 DENCELLI5 If.N If DEW CELLt 6 ti.H 31~ DEW CELLI 7 ff.fi - 32 DEW CELLI 6 If.H 33 DEW CELLt 9 ti.N 34 DEW CELLitf (f.H NO FAULTY PRESSURE GA6ES THE CONTA!NMENT VOLURE IS f.2 NSE+f7 CUBIC FEET i' THE NUMOER OF RTDS IN USE ARE 23.

THE TOTAL PERCENT FOR RTDS IS 1N. Nd I THE NUMBER OF DEW CELLS IN USE IS 18.

THE TOTAL PERCENT FOR DEW CELLS IS IN. Nd I

Individual RTD total greater than 100% due to computer rounding.

26 0006U

, ; l.

1 .

APPI NDIX A - - - '- - ~ - - -

SCHEMATIC ARRANGEMENI' OF insTfti tit tW I C URFR TEST INSTRUMENTATION (SIMPLIFIED) ~'

TE-655 A RID /YSI x' ,

To-655 X 3 TE- 654 A DEWCELL /

ELE V437 *0 "

lit IV 11 I i PI-330

, PRESSURE IND./

PI-391 TEXAS INSTR.

-LEV 405 FI-ll0 MASS. FLOWMETER /

Fi-lll SIERRA ELEV 385 p LEGEND

@@ @ @ e, _n : 6 m Vsee

_, w .

O@O@ O TO ATMO5PHERE LLEV346 ILRT INSTRUMEHTATION ni iv i n C"5"UP^'^ ' MER l grt3 '

@@@@@@@ : C #(y) SEEtiYL g i LEV.314 i ni n iv i n ni

. __ _ ; oilSITE

'" ' '" " 'Y ' '

t L EV.281 _

io DEwcELLS

e 1

E a

APPDOIX 8 .

Average Temperature and Maximum

- y y.

Temperature Differential for the Reactor Building s

m O

v 27 0006U

""'**""rr**==-ww%,-

APPENDIX B Average Temperature and Maximum 'Iemperature Differential. in the Reactor Building TE655 TE655 N X DATE TIME AVG. TEMP. HI TEMP. LOW TEMP. ( Hi-Lo )

4/17/84 1600 72.370 73.530 71.640 1.8 90 1700 72.350 73.500 71.600 1 .90 1800 72.320 73.510 71.370 2.1 40 1900 72.310 73.490 71.590 1 .90 2000 72.330 73.510 71.520 1.9 90 2l00 72.310 73.470 71.400 2.0 70 2200 72.320 73.500 71.4 10 2.090 2300 72.300 73.480 71.340 2.1 40 4/18/84 2400/0000 72.360 73.500 71.680 1.8 20 0100 72.370 73.590 71.710(1) 1.8 80 0200 72.330 73.530 71.510 2.020 0300 72.250 73.430 71.390 2.0 40 0400 72.220 73.380 71.510 (2) 1.870 0500 72.230 73.420 71.310 2.11U 0600 72.420 73.570 71.580 1.990 0700 72.270 73.470 71.420 2.050 0800 72.260 73.440 71.290 2.1 50 0900 72.310 73.470 71.450 2.020 1000 72.280 73.470 71.3 50 2.120 1100 72.320 73.460 71.400 2.060 1200 72.280 73.430 71.390 2.040 28 0006U

1300 72.290 73. 4W 7 1.430 2.0@

1400 7 2.320 73.520 71.490 2.030 1500 72.330 73.500 71.4$ 2.040 1600 72.320 73.510 7 1.420 2 .090 Avg.: 72.310 73.493 71.4@ 2 . 0 10 (1) RTDs J and X both had the same reading.

(2) RTD J was low reading.

29 0006U

APPENDIX C Reduced Leakage Rate Data (24 Hour Test) 30 0006U

DIFFERENCE AVG. AVG. OBSERVED LOWER BOUND CALCULATED UPPER BOUND (OBSERVED-TIME TEMP. PRESS. MASS MASS MASS MASS CALCULATED) 0 532 06 64 729 656742 77 656666 33I 656737.83 656809 34 d.9398245 25 532.03 64.728 656769.66 656663.78 656735.28 656806.78 34.382045

.5 532.04 64.724 656716.73 656601.22 656732.72 656804.22 -15.991442 75 532.03 64.726 656749.37 656658.66 656730e17 656801.67 19.201928 1 532 04 64.724 656716.73 656656.11 656727.61 656799.11 -10.878342 1 25 532.03 64 725 656739 22 656653.55 656725.05 656796 55 14.168419 15 ~

532501 64.727 656784 2 656650.99 656722.5 656794 61.707929 1 75 532 01 64.725 656763 91 656648.44 656719.94 656791.44 43.970499 2 532 01 64.722 656733.47 656645.88 656717.38 656788.88 16.0Re079 2e25 532 04 64.724 656716.73 656643.32 656714.83 656786.33 1.9044065 25 532 01 64.721 656723 32 656640.77 656712.27 656783.77 11.052189 2 75 532 02 64.722 656721 12 656638.21 656709.71 656781.21 11.411579 3 532 64.719 656715 37 656635.66 656707.16 656778.66 L.2153513 3.25 532.05 64.717 656633.36 656633.1 656704.6 656776.1 -71.236122 3.5 532.01 64.719 656703 03 656630.~54 656702.04 65677,3.54 .98440899 3 75 532.03 64.725 656739 22 656627.99 656699.49 65677'0.99 39.733917 4 532 02 64.72 656700 83 656625.43 656696. 93 656768.43 3.9007297 4 25 532 64.72* 656766.11 656622.87 656694.37 656765.87 71.734002 4.5 532' 64.72 656725 52 656620.32 656691.82 656763.32 33.70183 4.75 532 64.723 656755 96 65e617.76 656689.26 656760.76 66.699922 5 532 64.72 656725 52 656615.2 656686.7 656758.21 38.81493 5 25 532 01 64.717 656682 73 656612.65 656684.15 656755.65 -1.413722 5.5 531 99 64 717 656707 42 656610.09 656681.59 656753.09 25.830613 5e75 532 64.716 656684.93 656607.53 656679.03 656750.54 5.8958575 6 532 01 64.718 656692 88 656604.98 656676.48 656747.98 16.402917 6 25 532 64.71 65662&.05 656602.42 65e673.92 656745.42 -49.874126 65 532 64 708 656603 75 656599.86 656671.36 656742.87 -67 611937 6.75 531.97 64.711 656671.22 656597.31 656668.81 656740 31 2.4164865 7 531.99 64.707 656605 95 656594.75 656666.25 656737.75 -60.303801 7 25 531 97 64 71 656661.08 656592.19 656663 69 656735.2

-2 6181666 7.5 532.26 64.747 656678.56 656589.64 656661.14 656732.64 17.421169 7.75 532.1 64.727 656673 11 656587.08 656658.58 656730.08 14.532445 8 532 05 64.714 656602.92 656584.52 656656.03 656727.53 -53.100357 l

l

Corrected for water vapor pressure

- Least squares fit i

I l

31 0006U

- - ~ -

DIFFERENCE ,

AVG. AVG. OBSERVED LOWER LOUND CALCULATED UPPER BOUND (OBSERVED- l TIME TEMP. PRESS. MASS MASS MASS MASS CALCULATED) 8 25 532.01 64 711 e56621 85 656581.97 656653.47 656724.97 -31.617063 8.5 532.09 64.719 656604.29 656579.41 656650.91 656722.41 -46.6P0227 8.75 532 22 64.736 656616 34 656576.85 656648.36 656719.86 -32.015009 9 532.06 64 718 656631 17 656574.3 656645.a 656717.3 -14.630781 9.25 532.05 64.715 656613 07 656571.74 656643.24 656714.74 -30.171381 9.5 532j03 64.712 656607 31 656569.18 656640.69 656712.19 -33.371346 9 75 532 64 707 656593 61 656566 63 656638 13 656709 63 -44.52397 10- 532.02 64.71 656599.36 656564.07 656635.57 656707.07 -36.210066 10 25 531.96 64.707 656642.98 656561.52 656633.02 656704.52 9.9607818 10.5 531.97 64.704 656600.19 656558.96 656630.46 656701.96 -30.269536 10.75 531.96 64.704 656612.53 656556.4 656627 9 656699.4 -15.369949 11 531.94 64.704 656637 22 656553.85 656625.35 656696.85 11.874067 11 25 531.91 64.702 656653.96 656551.29 656622.79 656694.29 31.167502 11 5 531 92 64.701 656631 46 656548.73 656620.23 656691.73 11.230369 11.75 531.93 64.702 656629 27 656546.18 656617.68 656689.18 11.591114 12 531.91 64.699 656623.51 656543.62 656615.12 656686.62 8.390659 12 25 531 89 64.691 656567.01 656541.06 656612.56 656684.07 -45.557024 12.5 531 9 64.695 656595 26 656538.51 656610.01 656681.51 -14 747927 12.75 531.92 64.696 656580.72 656535.95 656607.45 656678.95 -26.730415 13 531.92 64.699 656611 17 656533.39 656604.89 656676.4 6.2722546 13 25 531 86 64.695 656644.64 656530.84 656602.34 656673.84 42.302782 13 5 531.9 64.696 656605 41 656528.28 656599.78 656671.28 5.6273608 13.75 532 02 64.706 656558.78 656525.72 656597.22 656668.73 -38.449015 14 532.11 64.71 656488.31 656523.17 656594.e7 656666.17 -106.36115 14.25 532.09 e4.717 656584 656520.61' 656592.11 656663.61 -8.1105105 14.5 532.01 64.701 656520.38 656518.05 656589.55 656661.06 -69.173216 14 75 531.94 64 7 656596 63 656515.5 656587 656658.5 9.6290134 15 531.96 64.704 656612 53 656512.94 656584.44 656655.94 28.001397 15 25 531.96 64.699 656561.79 656510.38 656581.89 656653.39 -20.09177 15.5 531.95 64 701 656594.43 656507 83 656579.33 656650.83 15.103595 15.75 531.94 64.697 656566 18 656505.27 656576.77 656648.27 -10.589763 16 531.95 64.702 656604.58 656502.71 656574.22 656645.72 30.364829 32 0006U

i DIFFERENCE AVG. AVG. OBSERVED LOWER BOUND CALCULATED UPPER BOUND (OBSERVED-TIME TEMP. PRESS. MASS MASS MASS MASS CALCULATED) 16 25 '531.96 64.7 656571 94 656500.16 656571.66 656643.16 .28237221 16.5 531.94 64.695 656545.89 656497.6 656569.1 656640.6 -23.216763 16 75 531.97 64.693- 656488.56 656495.04 656566.55 656638.05 -77.981071 17 .532 64.699 656512 43 656492 49 656563.99 656635.49 -51.56147 17.25' 531.92 64.704 656661 91 656489.93 656561 43 656632.93 100.47713 17.5 531 64.698 656588.67 656487.38' 656558.88 656630.38 29.797695 17,75 531,.93

.95 64.699 656574 14 656484.82 656556.32 656627.82 17.816275 18 531.97 64.702 656579.89 656482.26 656553.76 656625.26 26.131452 18.25 531.94 64.699 656586.48 656479.71 656551.21 656622.71 35.272385 18.5 532' 64.696 656481 99 656477.15 656548.65 656620.15 -66.663713 18.75 531.98 54.698 656526 96 656474.59 656546.09 656617.59 -19.131338 19 532.01 64.701 656520.38 656472.04 656543.54 656615.04 -23.15532 19 25 531.99 64 702 656555.21 656469.48 656540.98 656612.48 14.230282 19.5 531.97 64.699 656549.45 656466.92 656538.42 656609.93 11.0?7493 19.75 532.02 64.704 656538.48 656464.3'7 656535.87 656607.37 2.6145816 20 531.97 -64.699 656549.45 656461.81 656533.31 656604.81 16.140592 20 25 532 32 64.703 656528 34 656459.25 656530.75 656602.26 -2.4191178 20.5 '531.97 64 701 656569 75 656456 7 656528.2 656599 7 41.549197 20.75 532.07 64.712 656557.95 656454.14 656525.64 656597.14 32.310918 21 531.98 64.699 656537.11 656451.58 656523.08 656594.59 14.025172 21 25 *532 03 64.705 656536.29 656449.03 656520.53 656592.03 15.760236 21 5 532 02 64.698 656477.6 656446.47 656517.97 656589.47 -40.370364 21,75 531.99 64 699 656524 77 656443.91 656515.42 656586 92 9 3536666 22 532.01 64.704 656550.82 656441.36 656512.86 656584.36 37.964247 22.25- 532.05 64.705 656511 61 656438.8 656510.3 656581.8 1.3069388 22 5 531 99 64.698 656514.62 '656436.24- 656507.75 656570.25 6.8759447 22 75 531.99 64.701 656545.06 656433.69 656505.19 656576.69 39.874608 23 532 02 64.097 .656467.45 656431.13 656502 63 656574.13 -35.177864

.23.25 532.03 64 706 656546 43 656428.57 656500.08 656571.58 46.359242 23.5 532.02 64.704 656538.48 656426.02 656497.52 656569.02 40.962828 23.75 532 02 64.7 656497.89 656423.46 656494.96 656566.46 2.9321821 24 532 01 64.698 656489.94 656420.9 656492 41 656563.91 -2.4652935 WO (1dE LEAST SQUARES EQ. INTERCEP7 POINT = W0 = 656737.83 W1 (THE LEAST SQUARES E0 SLOPE = W1 = -10 226199 l SEOC (STANDARD EPROR OF CONFIDENCE) = SEOC = 35.750628

- 95 PERCENT CONF;0ENCE LEAK RATE = LR95 = .0031471543 LEAKAGE RATE = LR = .037370891 5

33 0006U

\. - - -

< ~

r e

APPDOIX D -

MASS / TEMPERATURE PLOT (24 Hour Test)

I i

l j 34 0006U

1600 24 0 O 4/17/84 04C3 1600 4/17/ 84 4/18/84 4/13/84 0 1 2 3 4 5 6 7. 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 657005 6 e i 6 e i a a a 6 6 4 i i e # 4 e i i i I 657003 65690i: .. 1 7 * -

656900 d UPPER BOUND 95 PERCENT CONFIDENCE INTERVAL

[ 65680(j - %

' 656800 2 t

. * . .m~%

N 65670( * . '

E - ** -

656700

~-

7-

! 6566m y

1.- u .*. ,.... ...-

656600 c

. . . .* e I 65650( - ,,

-

w -- ,656500 8 IJMER BOUND 95 PERCENT -

U 65640g .-

CONFIDENCE INTERVAL '---.%

3 LEAST SQUARES FIT EQUATION W = 656 73 7.83 - 10.2 3 t 656400 w

" 1984 24 Il0VR lilTEGRATED LEAK RATE TEST TilREE HILE ISLAHD UillT 1 C 72,g _

L .

72.6 d

u y2,* ,

H

12. 5 m %

$s 72,t _

w .

s . *

- 72.4 e *.*.* .

g 77, , .

g * * * ,

, . .. - 72. 3 w . . * * * *

y 73,; ,

o 8 -

- 12.2 oW oO 72.I

&I _

- 72.1 c' W f g g t f t f I f f f f I f f f f f f f i I l O I 2 3 4 5 6 7 8 9 10 il 12 13 14 15 16 t/ 18 19 20 21 22 21 24 1600 2400 4/17/84 0400 4/11/E9 1600 EI.APSED TIME (ikMIRS) 4/18/84 4/18/84

r APPENDIX E RedJced Leakage Rate Data (8 Hour Superimposed Leak Test) s 1

\

N

-36 0006U

DIFFERENCE AVG. AVG. OBSERVED LOWER BOUND CALCULATED UPPER BOUND (OBSERVED-TIME TEMP. PRESS. MASS MASS MASS MASS CALCULATED)

'~ ^~

532 0[ 64 699 656500.09

~ ~ ~

5

~

656395 89 656462 03 656528.lf"38.058227 25 532.03 64.696 656444.97 656390.73 656456.87 656523.01 -11.90328

.5 532 03 64.697~ 656455 12 656335.58 656451.71 656517.85 3.4007087

.75 532 01 64.693 656439.21 656380 42 656446.56 656512.7 -7.3515699 1 532.01 64.69 656408 76 656375.26 656441.4 656507.54 -32.635159 1.25 532 64.694 656461.69 656370.1 656436.24 656502.38 25.449454 15 532 64.691 656431 25 656364.95 656431 09 656497.22 .16529339 1 75 532.05 64.693 656389 85 656359.79 656425.93 656492.07 -36.073741 2 532.02 64 695 656447 16 656354 63 656420 77 656486 91 26 390311

~

2.2% 532.04 64.695 656422.48 656349.47 656415.61 656481.75 6.8710826 2.5 532.02 64.695 656447.16 656344.32 656410.46 656476.59 36.705072 2.75 532 02 64.689 656386 28 656339.16 656405.3 656471.44 -19.018341 3 '532.04 64.694 656412.34 656334 656400.14 656466.28 12.196806 3.25 531.99 64.688 656413.15 656328.84 656.394.98 656461.12 18.164008 35 531.99 64.682 656352 26 656323.69 656389.83* 656455.97 -37.562838 3.75 532 64 689 656410 96 656318.53 656384.67 656450.81 26.287357 4 532.01 64.685 656358.03 656313.37 656379.51 656445.65 -21 481542 4.26 532.01 64.688 656388 47 656308.22 656374.35 656440.49 14.116808 4.5 532 64.685 656370 37 656303.06 656369.2 656435.34 1 1707761 4.75 532.01 64.686 656368 18 656297.9 656364.04 656430.18 4 1375893 5 532 03 64.685 656333 36 656292.74 656358.88 656425.02 -25.525742 5.25 532.02 64.68 656294 96 65e287 59 656353 72 656419 86 -56.765728 5.5 532 64.682 656339.93 656282.43 656348.57 656414.71 -8.6412434 5.75 532.01 64.684 656347.88 656277.27 656343.41 656409.55 4.4731317 6 532.02 64.685 656345 69 656272 11 656338.25 656404.39 7.4404087 6.2% 532.02 64.68 656294.96 656266.96 656333.1 656399.23 -38.136206 6.5 532.04 64.679 656260 14 656261.8 656327.94 656394.08 -67.796131 6.75 532 01 64.685 656358 03 656256.64 656322.78 656388.92 35 249643 7 531 98 64.682 656364.6 656251 48 656317.62 656383.76 46.9784 -

7 25 531 95 64.681 656391 47 656246.33 656312.47 656378.6 79.004167 75 532 64.675 .656268.9 656241 17 656307.31 656373.45 -38.412463 7 75 531.95 64 67 656279 84 656236 01 656302.15 656368.29 -22.310549 8 531.97 64.678 656336.35 656230.05 656296.99 656363.13 ,39.355288 do (THE LEAST SQUARES EQ. INTERCEPT POINTS W0 = 656462.03 bl (THE LEAST SOUARES EU. SLOPE' s W1 = -20.629522 SEOC (STANDARD ERROR OF CONFIDENCE) = SEOC = 33.069485 95 PEWCENT CONFIDENCE LEAK RATE = LR95 = .014987951 LEAKAGE RATE = LR = .07542074

Corrected for water vapor i cc L;ast squares fit 37 0006U

s s

/,

~ /

g

= N J*

. ~

~ APPENDIX F ,

s 3%

MASS / TEMPERATURE PLOT - - .

(8 Hour Supprimposed , Test) y

/

e j.#

s+ f on l

, - i 1 ,

i .

+ ,

+-

,e - t

/ /

s .1+

k

}

N w -

( j. - - - - .

~~

4 6

) #.

a

.,e

~

9 . r[ F 4

< ~

%}

/

M*

g i e WW

'y m

i ,$ f y

_g l *

./ " d s ' r

,- ,~ ,

A 4 . & '

4 .)*r 38 ,

0004J p* O t _ _

4 6 8 0 1 2 3 5 7 I I # s s a :

I UPPER IMlUNil 95 PERCENT

^ 656500d j

~~% ~~_ QAIFillEl3CE INTERVAL.

g 656500 0 _

9 ~%

656400 ~_

'~'~~ 65640Q y

g

~

~% d

, p= 9 7 L._ _ -_

~

656300 -

' ' - - - - , 656300 3 ~'M---. 9 a

" 656200

- IIMER BOUND 95 PERCENT WNFIDENCE INTERVAL I AST SQUARES FIT EQUmTION , _ 656200 3 li = 656462.03-20.63 t 4

o i

- +

die W #

=

198f1 8 HOUR SUPER lMPOSED LEAKAGE TEST u '

e TilREE MILE ISLAND UNIT 1 REACTOR BUILDING !

I e -

L S

6 a n.4 -

_ n.4 e

e . . . .

y n.a - * *

~

. _ n.3 E

, a 72.2 --

_ 72.2 o N 3

8 m 8 C

E .

i , , , , , ,

( g 0 2 3 4 s 1 6 7 g 2000 EIAPSED Title (HOURS) 0400 04/18/84 04/19/84

7;.

+. ,.

APPENDIX G TWEE MILE ISLAND UNIT 1

.9 1982 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT SP 1303-11.18 08/09/81 - 10/12/82 l

40 0006U

e-INDEX - 1982 R. B. LOCAL LEAK RATE TESTING REPORT

1. PURPOSE

2. StMMARY OF WORK ACCOMPLISFEC 2.1 Valve Testing / Repairs

. "2.2 Access Hatdies 2.3 Penetra tion Pressurization

3. METF0DS OF TESTING 3.1 Valves 3.2 Access Hatches 3.3 Penetration Pressurization

4. TEST EQUIPMENT USED 4.1 Valves 4.2 Access Hatches 4.3 Penetration Pressurization

5. SlMMARY AND INTERPRETATION OF DATA 5.1 Valves 5.2 Access Hatches 5.3 Penetration Pressurization

6. ERROR ANALYSIS 6.1 Valves 6.2 Access Hatches 6.3 Penetration Pressurization

7. REFERENCES

8. ATTACFNENTS 8.1 Results E* valuation Promdure/ Repair Criteria 41 0006U

REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT 1982 REFUELING FREQUENCY

1. PURPOSE 1.1 To provide analysis to the Nuclear Regulatory Commission

,. on the type B and type C leakage tests performed on the Three Mile Island Unit 1 Reactor Building in the interval between the sixth and seventh periodic refueling frequency leak tests.

This is in a ccordance with " Reactor Containment Leakage Testing for Water Cooled Power Reactors". Appendi x J, Part 50, Title 10 Code of Federal Regulations whidi required the contents of this summary report to become part of the Type A test report along with the details of any other type B and type C testing performed since the previous type A test (also required per technical specification 4.4.1.1.8) .

The unit has t;een in cold shutdown sin the March 28, 1979 Unit II accident.

2. S'J44AR Y OF WORK ACCOMR_ISHED 2.1 Valve Testing / Repairs During the time period covered by this report Appendix J Type B and C leak tests were performed on the components as listed below.

1. AH-VlA/18/lC/1D

2. Personnel Access Hatch.

l 3. Emergency Access Hatch.

i Monthly leak checks were done when convenient starting in l May 1982 in addition to annt-d tests of R. B. Purge valves AH-VlA/B/C/D due to recurring seat leakage. Results on these leak checks were used to e >pedite timely adjustments / repairs.

2.2 Access Hatch Testing / Repairs 2.2.1 Door Seals SP 1303-11.25 - Several door seal tests

were performed though not required in the cold

shutdown plant conditions by Technimi Specifications. No repairs were required.

l l

42 0006U i

r 2.2.2 Overall Hatch Test Semi-annual integrated type leak tests were performed on each access hatch in 1982 as required b y Technimi Specifistion 4.4.1.2.5. Due to higher than desirable leakage on the personnel access hatd1 for the begiming of year test some

, tubing fittings were tightened and a satisfactory retest was performed. Also it was suspected that the strongbacks for the doors had been tightened unevenly.

2.3 Penetration Pressurization SP 1303-11.24 Quarterly readings were recorded from the flow rotameters which supply air pressure or nitrogen pressure to reactor building medianical and electrical penetrations as

, required by Tedinical Specification 4.4.1.2.5. No penetration leakage problems were noted although flow mter malfunctions required meter repair, and occasional tubing leaks in the air supply system were found and eliminated.

e 43 0006U

, - . . , --__m.. - , _y.. ..m u- y .,

r i

3. METHODS OF TESTING 3.1 Valve Test Methods Testing was performed by use of TMI Unit 1 surveillance pro dure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This proa: dure gives detailed guidance on the tes t equipment and methods to be used for each

. penetration / valve. The following general philosophy is contained in the surveillance procedure.

3.1.1 Use air or nitrogen a t a pressure differential across the valve greater than Pa (caloJlated acciden t pressure) . 55 psig was normally used.

3.1.2 Assure that the pressure is e >erted in the accident test direction unless it can be demonstrated that pressurizing in the opposite direction is as mnservative. Butterfly valves AH-V1A/lB/lC/lD, and globe valves WDG-V14, SA-V3, and IA-V20 were tested in the reverse direction.

3.1.3 Assure that the test volume is drained of liquid so that air or nitrogen test pressure is against valve seats.

3.1.4 Assure that the test verifies valve pad <ing integrity in those cases where the pad <ing would be an R . B. leakage boundary.

3.1.5 Assure adequate time period for stabilization of test conditions.

3.1.6 Assure test equipment is calibrated and used in a manner consistent with the data accuracy desired (weekly meter standardization was performed to verif y meters accurate within -+ 4% full scale. --

MP 1430-Y-22) .

3.1.7 Assure that the fluid blo'd<ing system is drained and vented during tests on the associated containment isolation valves to prevent any effects it might have on the test results (most of the F. B. system piping is seismic 3).

3.1.8 Assure valves to be tested are closed by the normal method prior to testirg.

3.1.9 Document as-found conditions (prior to adjustments / repairs) and as-lef t conditions.

l 44 0006U t

3.1.10 Record tes t instrument scale readings prior to doing any data corrections.

3.1.11 Assure that system drains and vents which could serve as containment isolation valves, are closed and capped and tagged af ter completion of the test

, program. -

A training program prior to the refueling outage was performed to help assure that the above philosophy was understood by the personnel involved in the testing.

3.2 Access Hatch Test Methods 3.2.1 Door Seal Leak Tests-Method Door seal leak tests were performed by use of SP 1303-11.25. This procedure gives detailed guidance on the tes t equipment and methods to be used.

The door seal tests are performed by pressurizing the interspace between the double seals on each door with metered air a t the manufacturers remmmended test pressure of 10 psig. After stabilization the air rotameter indicates the rate of air input required to maintain the test pressure.

3.2.2 Overall Hatd1 Leak Test -- Semi-annual overall hatch leak testing was performed by use of THI Unit 1 Surveillance Pro dure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This pro dure gives detailed guidance on the test equipment and methods to be used. The overall integrated leak test verifies the integrity of all of the following barriers:

a. Hatch shell/ welds,

b. Rubber door seals,

c. Teflon operating shaf t pad <ing,

d. Bulkhead electrical penetrations,

e. Penetration pressurization check valves,

f. Emergency air flange and associated "0" rings on outer bulkhead, 45 0006U

r

g. Bulkhead equalizing ball valves and associated mounting flanges /"0" rings.

The overall leak tes t is performed by pressurizing the hatch to greater than calculated accident pressure and observing the rate of pressure drop on a high accuracy (Heise) pressure gage.

. Pressure corrections are made by reference to a barometer. Minimum test duration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months af ter a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months stabilization period.

3.3 Penetration Pressurization - Method Quarterly readings were taken on the flow rotameters which are permanently installed in the penetration pressurization system. These readings represent the air / nitrogen makeup rate required to maintain approximately 60 psig in mechanical penetrations and 30 psig in electrical penetrations. High meter readings have occasionally accurred but these have bean attributed to leaks in the compression fittings in the penetration pressurization system or to malfunctioning (stuck) rotameters. Testing was per plant surveillance procedure SP 1303-11.24.

l 46 0006U l

L _ _ . _ _ . _ . . _ . . _ ___

E

4. TEST EQUIFNENT USED 4.1 Valve Test Equipment (See Figure l_)

a. Rotameters - Sets of 3 Mfgr. - Brooks Inst. Co.

~

Model - 1114 Full View Ranges:

Float Mat'l. Tube No. Range Pyrex R-2-150 8-1,120 SCCM __

Saphire R-2-15C 100-12,000 SCCM Carboloy R-6-158 1,000-142,000 SCCM Accuracy 3 2% full scale industrial accuracy

b. Temperature Indimtors (as follows or similar)

Mfgr. - Ashcrof t Model - EH or AH / 3" or S" Dial Range - 303-1300F Accuracy - + 20F

c. Pressure Indimtors (as follows or similar)

Mfgr. - Ashcrof t Model - 1279 1/2" Dial

- Range 60 or 0-100 psig Accuracy - 1 2 psig

d. Pressure Regulator (as follows or similar)

Mfgr. - Union Carbide Corp.

Model - UPG 3-75-580 Range 100 psi output / 0-3000 psi input 47 0006U

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I" A' I 800v2 N V3 ( VALVE

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p SNISE IINAIN MAINTAINFDs 6BACK-FRF.S5UME V9

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ARRANGEMENT FOR TEST v41VES i y.m n- g l

' TI2 VI VI-Nitrugen Bottle Regulater e l lpgg . I r*,

V2/V3/v6/vF-3/4 or I/2" Sall valvo < b i corrAINMEstr .

M %y V4-Toggle Type Globe Valve ISOLATION yAt ys g *- l v5-454 S tesy valve i

} PkEShelltE ltNitCATous ril-0 E. 60 pass u=.se s.ose witti 2 est tacte==ats I -

l

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Fl2-Hano.wtur with 36 inch lotis scale f y 9 l l

0P It T r?irFR A l tlN b ltHilCA rum S y TII/Til-0-100*Tumase sange wi tte 2*F inc r=== sis.

b -

t jin rAtie.r t at s gg __,

l l

. - u..se -

k Tul.e Float g 0 pd3 ~ d 'S fpi# g ,_,_ [__,,_.___ ]

[o~ ~ R"- i ' i(SO Black Class 8-)S05CCM ' 4 5-I,lIOSCCH 100! ! [060'5dC7i'

~

Hei' a-i'-iS TC ijs'.~S! lo6~3 30dsCCn~

O jj[gg {jjil Ca'sTo}j{jEiie ,-. j!@ghj@jjgH, W ,0do$dCl1 _ INEAN11MFAM VFulFICAT18'N klG -

r,,,,,,,,~ ~ r>

O g -IswMIav hwr 7EsrBa -

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r

e. Calibration Rotameters (Set of 2)

Mfgr. - Brooks Inst. Co.

Models - ll10-05K2B1Z49,1110-08K2B1Z06 Ranges 16,000 SCCM, 3,600-234,000 SCCM

. Repeatability - 1 1/4% of instantaneous Accuracy - 1 1% instantaneous

f. Flow rate Calibrator Mfgr. - Brooks Inst. Co.

Model - 1056A Range O to 2,400 SCCM Accuracy - 1 0.2% of indicated volume 4.2 Access Hatch Test Equipment

a. Precision Pressure Gage (as follows or similar)

Mfgr . - i else Model - CM Range - 0.60 psig Resolution - 0.25 psig Accuracy - 0.1%

b. Barometer (as follows or similar)

Mfgr. - Pennwalt Model - FA185260A Range - 10.8 - 15.5 psia Resolution - 0.005 psia Accuracy - 0.1%

49 0006U ws

7 -.

, 4.3 Penetration Pressurization Test Equipment

a. Flow Rotameters -- (Permanent System Equipment)

Mfgr. - Brooks Inst. Co.

Model - 1114

"~

, Range -'0-10 SCFH a t 60 psia air Accuracy - + 2% Industrial accuracy O

d e

'. ^

50 0006U

5.- SlM4ARY AND INTERFRETATION OF DATA 5.1 Valve Test Results There was no complete leak test program during the time period aavered by this report. Followir.g are the results for those few valves / penetrations whid1 were tested:

. .l. AH-VlA/lB Not pressurizable to test pressure **

2. AH-VlC/1D Not pressurizable to test pressure _,**

t-5.2 Access Hatch Test Results SCCM Personnel Emeraency Beginning Year 6164/1205 2970/2970 1/30/82 11/27/81 Mid Year 1813/1813 1107/1107 5/29/82 5/21/82 5.2.1 Door Seal Leakage - SP 1303-11.25 None of the door seal leak tests e xceeded the 3 SCFH administrative leakage limit. Typically, the leakage was less than 1 SCFH.

5.3 Penetration Pressurization Leakage - SP 1303-11.24 Leakage Rates __

Electrical Mechanical Da te 12/13/81 Data Missing

Data Missing

03/15/82 0. 8' 3.6 0 6/12/82 0.6 >10.7 H 08/26/82 0.9 1.92 o 0 9/11/82 0.5 2.7

The data was missing but a surveillance record sheet was found indicating that results met the acceptance criteria.

- The purge valves were disassembled for repair the res t of r the year thereby preventing retest in 1982.

l 51 0006U

'N*-'f8 w .y-r-7 9 y,, -.m+,,%,.,7.y4%9,.m.- ,p_ ,e--- -> wee, ,,_em em-m.,e -,__.m g--__w.-_-_,- y .. ., - .,,,e

r Occasional meter problems were found and repaired and occasional leaks of tubing / pipe fittings in the PP system were located and eliminated. No penetration leakage problems were identified.

There is no technica1 specification limit on penetration pressurization system leakage.

L 52 0006U L

6. ERROR ANALYSIS 6.1 Valve Testing Errors (For purge valves see Section 6.2)

The flow meters used in the field have normal industrial a ccuracies of + 2% f ull scale in the 10-100% (15-150 mm) scale range. Prior to use mm versus sccm graphs were developed for each meter by 10 point calibrations using

. high accuracy (f 1% instantaneous) lab rotameters. During the leak test program weekly 3 point standardizaticns were performed on the field rotameters to verif y continued a ccuracy. The acceptance criteria for these standardizations was a variance of no more than 4% from the calibration graphs. If meters were repaired or the 3 point standardization ex eded the inaccuracy limit a new 10 point calibration was performed. Scale readings on the leak rate procedure (SP 1303-11.18) data sheets were evaluated and corrected using the methods in Attachment 1. Conservative bias was introduced into the results by assuming 15 mm (10% of scale) as the minimum scale. Half of the test results actually showed a lower scale reading. More involved error corrections were not considered meaningful based on the very high total leakage as-found and the low total leakage as-lef t.

6.2 Access Hatch and Purge Valve Testing Errors The measured pressure drops were corrected by adding the minimum scale increment of the gage used for both the helse gage and the barometer. This conservatively corrected for the resolution and repeatability errors.

Gages used were recently calibrated. A minimum one hour temperature / pressure stabilization period was used prior to ead1 pressure drop test. The access hatches and purge valves are not instrumented to allow temperature corrections.

6.3 Penetration Pressurization Testing Errors These test results are used for information only and do no t count toward the total leakage limit for Technical Specification cor::armance. The meters, installed permanently in the system, have 1 2% full scale industrial accuracy.

53 0006U

[1 .

7. REFERENCES J

7.1 1430-Y-22 Standardization of Flow Rotameters 7.2 SP 1303-11.18 Reactor Building Local Leak Rate Testing (Rev.19) ~

7.3 Three Mile Island Unit 1 Technical Specification 4.4.1 7.4 TMI Surveillance File (for Data sheets) 7.5 SP 1303-11.24 R. B. Local Leakage Penetration Pressurization (Rev. 0) 7.6 SP 1303-11.25 R. B. Local Leakage Access Hatd1 Door Seals (Rev. 4).

e 54 0006U L -' _ _ _ _ _

e o

ATTACWENTS i

e 55 0006U

F ATTACWENT 1 RESULTS EVALUATION PROCEDURE (SP 1303-11.18 Enclosure 9)

(

I 56 0006U t

I

r Attachment 1 R. B. LOCAL LEAK RATE TESTING RESULTS EVALUATION The vent rotameter reading will be used if it can be demonstrated by the test datothat all signifimnt CIV leakage is being accounted for. If CIV packing, fluid block deck valve, or gasket leakage was evident the supply rotameter results will be used unless this non-seat leakage was measured reliably and documented.

FOR USE OF SUPPLY ROTAMETER DATA: FOR USE OF VENT ROTAMETER OATA:

Pro dure: Procedure:

a) Record supply meter reading in (1) a) Record v_eg meter reading in below*._ Also identify the meter (1) below.*

used by tube # in (8) below and the metering pressure in (9). b) Record downstream verification meter reading in (2) below.

b) Convert meter units in SCCM units Also identif y the respective using latest lab meter calibration meters used in (8) below and curve. Enter in (3) below, the metering pressure in (9) .

c) Correct results for temperature. c) Convert meter units to SCCM Enter supply temperature in (4) units using latest lab meter b elow. calibration curve. Enter in (3) below.

Calculate and enter in (7) below.

d) Correct results for temperature. Ente r ven t temperature (OF) in (4) below.

the n -

l Calculate and enter in (5) below e) If measurements of any other significant leakage paths

If meter scale reading was less (fluid block check valve, than 15 mm (minimum scale) use packing) are being claimed 15 mm in calculations. enter corrected flow (SC04) in (6) below.

57 0006U 1

r e

Attachment 1 (Continued)

(

._MM) (SCCM) 530

, ( + ,) convert (- + )XA + 460 = SCCM (1) (2) (3) \ (4) (5)

+ SCCM (8) (Identify meters used) (6)

(9) (Meter Pressures) = CIV Leakage SCCM (7)

I r

L 58 0006U

Attactinent 1 (Continued)

" 0006U L-

7. _-

Y e

APPENDIX H THREE MILE ISLAND UNIT 1 1983 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT SP 1303-11.18 10/12/82 - 12/31/83 i

l:

60 0111U D

m.

IPOEX - 1983 R. B. LOCAL LEAK RATE TESTING REPORT

1. PLIIPOSE

.2. SlMMARY OF WORK ACCOMPLISED a-2.1 Valve Testing / Repairs 2.2 Access Hatmes 2.3 Penetration Pressurization

3. METIOOS OF TESTING 3.1 Valves 3.2 Access Hatches 3.3 ' Penetration Pressurization

4. TEST EQUIPMENT USED 4.1; Valhes 4.2 Access Hatmes 4.3 Penetration Pressurization

5. SIMMARY AND INTERPRETATION OF DATA 5.1 Valves.

5.2 Access Hatches 5.3

Penetration' Pressurization

6. ERROR ANALYSIS 6.1 . Valves

, 6.2 Access Hatches 6.3 Penetration Pressurization

7. REFERENCES

8. ATTACH 4ENTS 8.1 Results Evaluation Pro dure / Repair Criteria 8.2- Tabulation of Individual Test Repairs 61 0111U

_-,,_,_.,__,,,~m. . ~ . .

7 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT 1983 REFUELING FREQUENCY

1. PURPOSE 1.1 To provide analysis to the Nuclear Regulatory Commission on the seventh periodic type B and type C leakage tests performed on the Three Mile Island Unit 1 Reactor Building.

This is in accordance with " Primary Reactor Containment Leakage Testing f or Water Cooled Power Reactors".

Appendix J, Part 50, Title 10 Code of Federal Regulations whim required the contents of this summary report to become part of the Type A test report along with the details of any other type B and type C testing performed since the previous type A test (also required per tedinial specification 4.4.1.1.8).

The unit has been in cold shutdewn since the March 28, 1979 Unit II accident e xcept for periods of hot functional testing.

2. SUMAR Y 0F WORK ACCOMPLISHED 2.1 Valve Testing / Repairs Appendix J Type B and C leak tests were performed on the components as listed in TMI Unit 1 Temnical Specification

4. 4.1. In addition the following components were leak tested though not yet listed in the Tednica1 Specification.

p 1. HM-V1A/B, 2A/B, 3A/B, 4A/B - New System

2. IC-V16/18, NS-Vll - Check valves no t previousl y tes ted. NRC request to add.

Monthly leak decks were done when convenient'in addition to annual tests of R. B. Purge Valves AH-VlA/B/C/D due to recurring seat leakage. Results on these leak ched<s were used to e 4edite timely adjustments / repairs.

Repairs were initiated on the following mmponents due to higher than desirable leakage.

1. CF-V12A - Seats lapped

2. DH-V64 - Disc replaced j 62 Oll1U

3. DH-V69 - Disc replaced

4. m-V4A - Manual opening devi misadjusted

5. HP-V1 - Sof t seats installed.

6. HP-V6 - Sof t seats installed

7. IC-V3 - Packing readjusted - Valve cycling / retest program run to verify consistent seating.

8. IC-V4 - Nail under valve wedge prevented closing

9. LR-V49 - Sof t seats installed

10. NI-V27 - Seats lapped

11. WDG-V4 - New type valve installed - Upgrade Repairs were mandatory due to gross leakage During a November 1982 valve inspection the ethylene propylene rubber seats in AH-V1D were found to have cracked. The crad<ing did not affect leakage and was not judged to be a short term safety concern. It was evaluated by the valve vendor as e x ssive mold release agent. The vendor provided new seats for all four purge valves. The new seats were installed in AH-VlA/1B/1C/1D in March 1983.

In August 1983 the seats on AH-V18 were found to be cracked and a sample of the material was once again sent to the vendor for analysis. Once again the cracking did no t affect valve leakage. The seats for AH-V1B were replaced with material f rom the same batch as that which failed.. The vendor committed to replacing the seat material af ter development of a more suitable f abrication method. The new sea t material was no t received ye t as of 12/31/83. The vendor, Pratt Valve Company, submitted a

-10CFR21 type report to the NRC on the sea t crad<ing

- problem. The cracking was now a ttributed to delamination between plies in the rubber and the promised new seats are to be one piece (no plies).

2.2 Access Hatch Testing / Repairs 2.2.1 Door Seals SP 1303-11.25 Door sealleak tests were performed a t the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months f requency as required by Technical Specification 63 0111U

i t,

- t 4.4.1.2.5 while hot functional testing was being performed. All of the seal tests satisfied the surveillance procedure administrative ac ptance criteria.

o 2.2.2 Overall Hatch Test Semiannual integrated' type leak tests were .

performed on each access hatch in 1983 as required A by Temnical Specification 4.4.1.2.5. Due to higher than desirable leakage on the emergency access hatch for the mid-year test some tubing fittings were tightened and a satisfactory retest

, was performed.

2.3 Penetration Pressurization SP 1303-11.24 Quarterly readings were recorded from the flow rotameters

-which supply air pressure or nitrogen pressure to reactor building medanical and electrical penetrations as .

required by Tednical Specification 4.4.1.2.5. No penetration leakage problems were noted although flow meter malfunctions required meter repair, and occasional tubing leaks in the air supply system were found and eliminated.. ,

s

\

A L

64 0111U i

3. METHODS OF TESTING 3.1 Valve Test Methods Testing was performed by use of TMI Unit 1 surveillance

. procedure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This pro dure gives detailed guidan on the test equipment and methods to be used for each penetration / valve. The following general philosophy is contained in the surveillance procedure.

3.1.1 Use air or nitrogen a t a pressure differential across the valve greater than Pa (calculated accident pressure) . 55 psig was normally used.

3.1.2 Assure that the pressure is e x!rted in the accident test direction unless it can be demonstrated that pressurizing in the opposite direction is as conservative. Butterfly valves AH-V1A/lB/lC/lD, and globe valves WDG-V4, S A-V3, and IA-V20 were tested in the reverse direction.

3.1.3 Assure that the test volume is drained of liquid so tha t air or nitrogen test pressure is agains t valve seats.

3.1.4 Assure that the test verifies valve pa&ing integrity in those cases where the pa&ing would be an R. B. leakage boundary.

3.1.5 Assure adequate time period for stabilization of test conditions.

3.1.6 Assure test equipment is calibrated and used in a manner consistent with the data cucuracy desired (weekly meter standardization was performed to verif y meters accurate within + 4% full smle. --

MP 1430-Y-22) .

3.1.7 Assure that the fluid blo&ing system is drained and vented during tests on the associated containment isolation valves to prevent any effects it might have on the test results (mos t af the F. B. system piping is seismic 3).

3. 3 . 8 Assure valves to be tested are closed by the normal method prior to testing.

3.1.9 Doa; ment as-found conditions (prior to adjustments / repairs) and as-lef t conditions.

65 0111U

7 ,

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3.1.10 Re cord ~tes t.lastrument'fical'e readidg s'prio r30 "

doing any data corrcct'.ons. '"

w ~n s y 3.1.11 Assur[tha't system drainsTand;vbot's whichbuld

are closcu

~ and

'serveds-contairment capped andetagged afteYcorrdletion isola tion'yalves,'o f-the tes,t, ,

program. , - , ,

n,

,, A trai_ning' program prior to the re' fueling outage [

Meperformed to help' assure tha t the above philosoptjy was understood by the personnel involved

. , pin th,e testing. ~ , ,

. r .

/

3.2 Acce'sS Hatch Te st Metbods ' ,

i,

.e 3,2.1 Door Seal L'eak Thst,s-Method ,< , ,

./ ,

s . . y ,- _ ,

s.

~ Door sealleak tests were perforndiy use of.

SP 1303-11.25. This pro dure gives_ dettailed -

f ~

.)

guidance on the tes t equipment and nethods to be n

[

, used. s

/

- The door seal testa are p0rformed by epylssuyizing ;

the interspace betwesn,the double Aeals on/ach "

~ j , door with metered air a t the minufy.turers/

-I / ' re' commended test pressure of 10 pwy. After stabilization the air rotarbtsr indimtes the rate of air.inpu t required to thaintain the te's t pressure. '

i 3.2.2 Overall Hatch Leak Tes,t -y Semisennual overall '

hatetfleak testing wah pcrforrNd by use of TMI Unit

~

1 Surveillanm Promdure SP 1363-11.18 Reactor .

Building Local Leak Rate Testingr This pro dure ..

gives Aetailed guidan . on the test equipment and

~

't , '

4 methods to be used. The overall' integre.te d leak

,1 , te'at verifies the integrity of all of the following ?' [ '

U' barriers:

~

/

. . / , , .

a. Hatch shell/ welds, _

^

s p .

, b. Rubber door seals, -

c. Teflon operating shaf t pad <ing, , ; M g d. < Bulkhead electrical penetrations, . . "

,q , ,

e. Penetration pressurization check valves, s e, .,

l ,- f. Emergency air flange and associated "0" rinos on outer bulkhead, 5 *, -

9 ,

66. Oll1U e

l. : .-

~ -

_l., '

8.-

g. Bulkhead equalizing ball valves and associated mounting flanges /"0" rings.

The overall leak tes t is performed by pressurizing the hatch to greater than caloJlated accident

- . pressure and observing the rate of pressure drop on a high accuracy (Heise) pressure gage.

Pressure corrections are made by reference to a barometer. Minimum test duration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months stabilization period.

3.3 Penetration Pressurization - Method Quarterly readings were taken on the flow rotameters which are permanently installed in the penetration

~

pressurization system. These readings represent the air / nitrogen makeup rate required to maintain approximately 60 psig in mechanical penetrations and 30 psig in electrica1 penetrations. High meter reading s haye o ceasionally occurred but.these have been attributed to leaks in the compression fittings in the penetration pressurization system or to malfunctioning (stuck) rotameters. Testing was per plant surveillance promdure '

SP 1303-11.24.

L.

t f

l- .

h L

k -

l-L 67 0111U L

b - ,,+*,v-ww-r----www---*

4 TEST EQUIFNENT USED 4.1 Valve Test Equipment (See Figure 1)

a. Rotameters - Sets of 3 Mfgr. - Brooks Inst. Co.

Model - 1114 Full View Ranges:

Float Mat'1. Tube No. Range Pyrex R 150 8-1,120 SC04 Saphire R-2-15C 100-12,000 SCCM_

Carboloy _ R--6-158 l',000-142,000 SCCM Accuracy 1 2% full scale industrial accuracy

b. Temperature Indicators (as follows or similar)

Migr. - Ashcrof t Model - EH or AH / 3" or 5" Dial Range - 3CP-1300F Accuracy - 1 20F

c. Pressure Indicators (as follows or similar)

M, gr. - Ashcrof t Model - 1279 1/2" Dial Range 60 or 0-100 psig Accuracy - 3 2 psig

d. Pressure Regulator (as follows or similar)

Mfgr. - Union Carbide Corp.

Model - UPG 3-75-580 Range 100 psi output / 0-3000 psi input 68 0111U

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e. Calibration Rotameters (Se t of 2)

Mfgr. - Brooks Inst. Co.

Models - ll10-05K2BlZ49, ll10-08K2BlZ06 Ranges 16,000 SCCM, 3,600-234,000 SCCM Repeatability - 3 1/4% of instantaneous Accuracy - 1 1% instantaneous

f. Flow rate Calibrator Mfgr. - Brooks Inst. Co.

Model - 1056A Range O to 2,400 SCCM Accuracy - 1 0.2% of indimted volume 4.2 Access Hatch Test Equipment

a. Precision Pressure G' age (as follows or similar)

Mfgr. - Heise Model - CM Range - 0.60 psig Resolution - 0.25 psig Accuracy - 0.1%

b. Barometer (as follows or similar)

Mfgr. - Pennwalt l

Model - FA185260 A Range - 10.8 - 15.5 psia Resolution - 0.005 psia L Accuracy - 0.1%

l l 70 0111U

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- 4.3 Penetration Pressurization Test Equipment a.

Flow Rotameters - (Permanent System Equipment)

Mfgr. - Brooks Inst. Co.

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Model - 1114 Range 10 SCFH a t 60 psia air Acc.fracy - + 2% Industrial accuracy 4

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71 0111U F.

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5.

SUMMARY

AND INTERPRETATION OF DATA 5.1 Valve Test Results As-Found/As-Left Leakage - Also see tabulation of

- individual results in Attachment #2. The as-found leakage greater than acceptance criteria was not considered to require a License Event Report due to the long term plant shutdown condition during which the leakage was found.

Total leakage Tech. Spec. Limit % Tech. Spec. Limit As-Found >305,863 SCCM 104,846 SCCM >100%

As-Lef t 32,716 SCCM 104,846 SCCM 30%

NOTE: The total shown above is cumulative by penetration and not the total of all valve leakages. i.e.,

Only the highest valve leakage on each penetration is cnunted. This number is labeled as " PENT 0TAL" on the tabulation of results in Attachment 2.

EXAMPLE: Penetration XYZ has three containmen t isolation valves inside the reactor building in parallel and one outside. The leakage from the three inside totals 500 SCCM and the outside valve is 1000 SCW. The penetration leakage is counted as 1,000 SCCM no_t 1,500 SCW .

5.2 Access Hatch Test Results 5.2.1 Overall Hatch Leakage - SP 1303-11.18 - See the computer tabulation of 1983 leak rates in Attachment #2. The leakages were considered to be satisfactory exmpt for the mid year tes t on the emergency ac ss hatch. Fittings were tightened to reduce that leakage.

5.2.2 Door Seal Leakage - SP 1303-11.25 None of the door sealleak tests ex ede d the 3 S&H actninistrative leakage limit. Typially, the leakage was less than 1 SCFH.

1 72 0111U

5.3 Penetration Pressurization (PP) Leakage - SP 1303-11.24 Leakage Rates - SCFH Mechanica1 Electrica1

~

Da te 12/12/82 0 10.5 First Quarter 83 Not performed

Not performed

06/12/83 18.5 1.3 0 9/30/83 16 LR-Vl/49 2.3 Sea t Leakage 10/04/83 42.5 LR-V1/49 .8 Sea t Leakage 12/12/83 3.9 1.5

Not performed due to system condition.

LR-V49 seats were found to be damaged and were replaced in March 1984. Otherwise no penetration leakage problems were identified. Occasional meter problems were found and

. repaired and occasionalleaks of tubing / pipe fittings in the PP systera were lomted and eliminated.

There is no technical specification limit on penetration pressurization system leakage.

73 0111U

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

6. ERROR ANALYSIS 6.1 Valve Testing Errors (For purge valves see Section 6.2)

The flow meters used in the field have normal industrial

~ a ccuracies of + 2% f ull scale in the 10-100% (15-150 mm)

~

scale range. Prior to use mm versus scem graphs were developed for ead1 meter by 10 point calibrations using high accuracy (f 1% instantaneous) lab rotameters. During the leak test program weekly 3 point standardizations were performed on the field rotameters to verif y continued a ccuracy. The acceptan criteria for these standardizations was a variance of no more than 4% from the calibration graphs. If meters were repaired or the 3 point standardization exceeded the inaccuracy limit a new 10 point calibration was performed. Scale readings on the leak rate procedure (SP 1303-11.18) data sheets were evaluated and corrected using the methods in Attachment 1. Conservative bias was introduced into the results by assuming 15 mm (10% of scale) as the minimum scale. Half of the tes t results actually showed a lower scale reading. More involved error corrections were not considered meaningful based on the very high totalleakage as-found and the low total leakage as-lef t.

6.2 Access Hatch and Purge Valve Testing Errors The measured pressure drops were corrected by adding the minimum scale increment of the gage used for both the helse gage and the barometer. This conservatively corrected for the resolution and repeatability errors.

Gages used were recently calibrated. A minimum one hour temperature / pressure stabilization period was used prior to each pressure drop test. The a ccess hatches and purge valves c:e no t instrumented to allow temperature corrections.

6.3 Penetration Pressurization Testing Errors These test results are used for information only and do no t count toward the total leakage limit for Technical Specification conformance. The meters, installed permanent 1y in the system, have 1 2% full scale industrial accuracy.

1 i

I 74 0111U b

7. REFERENCES 7.1 1430-22 Standardization of Flow Rotameters 7.2 SP 1303-11.18 Reactor Building Local Leak Rate Testing

, (Rev. 22) 7.3 Three Mile Island Unit 1 Technimi Specifimtion 4.4.1 7.4 TMI Surveillance File (for Data sheets) 7.5 S P 1303-11.24 R. B. Lomi Leakage Penetration Pressurization (Rev. 3) 7.6 SP 1303-11.25 R. B. Lomi Leakage Access Hatd1 Door Seals (Rev. 6).

'l 75 0111U

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ATTACENENT 1 RESULTS EVALUATION PROCEDURE (SP 1303-11.18 Enclosure 9) 77 0111U

Attachment 1 R. B. LOCAL LEAK RATE TESTING RESULTS EVALUATION a-The vent rotameter reading will be used if it can be demonstrated by the test data that all signifimnt CIV leakage is being accounted for. If CIV pa d<ing, fluid block check valve, or gasket leakage was evident the supply totameter results will be used unless this non-seat leakage was measured reliably and documented.

FOR USE OF SUPPLY ROTAMETER DATA: FOR USE OF VENT ROTAMETER DATA:

Procedure: Procedure:

a) Record sucoly meter reading in (1) a) Record vent meter reading in below*. Also identify the meter (1) below.*

used by tube # in (8) below and the metering pressure in (9). b) Record downstream verification meter reading in (2) below.

b) Convert meter units in SCCM units Also identif y the respective using lates_t lab meter calibration meters used in (8) below and curve. Enter in (3) below, the metering pressure in (9) .

c) Correct results for temperature, c) Convert meter units to SCCM Enter supply temperature in (4) units using lates t lab meter b elow. calibration curve. Enter in (3) below.

Calculate and enter in (7) below.

d) Correct results for tempera ture. Ente r ven t temperature (OF) in (4) below.

then Calculate and enter in (5) below e) If measurements of any other significant leakage paths

If meter scale reading was less (fluid block check valve, than 15 mm (minimum scale) use packing) are being claimed 15 mm in calculations. enter corrected flow (SCCM) in (6) below.

78 0111U

~

DRAFT Attachment 1 (Continued)

(MM) '(SCCM) 530 e.

'( -+ ) convert ( + )X + 460 = SCCM (1) (2) (3) (4) (5)

+ SCCM (8) (Identify meters used) (6)

(9) (Meter Pressures) = CIV Leakage SCCM (7)

?

+

i 79 0111U L. .

L- . :

Attachment 1 (Continued) 80 0111U

ATTACENENT 2 DATA 1983 TYPE C REACTOR BUILDING LEAK RATE TESTING i

81 0111U

PAGE = 1 0F 3 LOCAL LEAK RATE TEST RESULTS THREE MILE ISLANO UNIT 1 REACTOR BUILDING 1983 1983 1983 1983 1983 1983 1983 1983 ,

RESULTS:GIVEN IN STO. CUBIC CENTIMETERS PER MINUTE (SCCM)

I,TEMS ' TAGS _DESCS OPERS SIZE ASFOUND COMMENTS ASLEFT DATES ooooo eeeeeeee seeeeese oeseo e se e e- eeeeeee seeeeeee eeeeee eseeoeoo 1 AH=VIA/S 8 FLY P/Mc 48 1413 NEWSEAT- 741 6/4/83 2 AH=VIC/D 8 FLY M0/P 48 2360 NEWSEAT 1790 5/27/83

3 CA.V1 GLOSE MO 1 76 OK 76 12/16/82 ,

t 4 CA=V2 GATE P 1 1930 HIGH 1930 12/9/82

5 CA=v3 GLOSE MO 1 76 OK 76 12/14/u2 6 CA V4A GLOBE MO 1 '76 OK 76 12/2/82 7 CA V48 GLOBE MO 1 124 OK 124 12/3/82 ,

8 CA-VSA GATE P 1 378 OK 378 12/2/82 9 CA-V58 GATE P 1 . 825 OK 825 12/3/82

'- 10 CA=V13 GLOSE MO 1 76 OK 76 12/14/82 11 CA-V189 GATE P 2 1591 HIGH 1591 1/15/83 12 CA=V192 LFT CHK- N/A 2 109 LOW 109 1/15/83 4

13 0 0 0 14 0~ 0 -

0 15 CF=V2A GLO8E MO 1 119 ok 119 1/17/83 16 CF.v28 GLO8E H0 1 328 OK 328 1/18/83 17 CF-V12A LFT CHK N/A 1 5166 HIGH 750 2/2/03 18 CF=V128 LFT CHK N/A 1 382 OK 382 1/10/83 19 CF-V19A GATE P 1 1976 HIGH 1976 1/12/83 20 CF=V198 GATE P ,1 253 OK 253 1/12/83 21 CF=V20A GATE P 1 1144 HIGH 1144 1/17/83 22 CF.V208 GATE P -

1 849 HIGH 849 1/18/83 23 CM-V1 SALL P 1 58 OK 58 12/29/82 20 CM-V2 SALL P 1 323 HIGH 323 12/29/82 i

25 CM-V3 BALL P 1 58 OK 58 12/29/82 26 CM-V4 BALL P 1 75 OK 75 12/29/82

'27 OH=V64 GLOBE HW 2 39850 FAILED 70 6/22/83 28 OHeV69 STOP CHK HW 2 97533 FAILED 209~ 6/24/83 29 0 0 0 -

30 FTTEAST FLANGE N/A 30 162 HIGH 162 12/13/u2 31 FTTWEST FLANGE N/A 30 152 HIGH 152 12/13/82 32 HMeV1A GLOSE S ~

.5 69 NEWVALV 69 3/3/83

-33 HMeV1B GLO8E S .5 69 NEWVALV 69 3/3/83 34 HM-V2A GLOBE S .5 69 NEWVALV 69 3/2/83 35 MM-V28 GLOBE S -.5 69 NEWVALV 69 3/3/83 36 HM-V3A GLO8E S .5 31 OK 31 6/30/83 37 HM-V38 GLOSE S .5 31 OK 31 6/30/83 38 HM-V4A GLO8E 5 .5 2692 HIGH 31 7/11/83 39 HM-V48 GLOBE S 5 31 OK 31 e/30/83 40 Hp=V1 GATE HW 6 7273 NEWSEAT 50 2/17/83

+1 MP.V6 GATE HW 6 1345 NEWSCAT 50 2/le/83 42 HR-V2A/8 GLOBE HW 2 58 OK 58 2/6/83 63 HR.V4A/B GLOBE HW 2 58 0x 58 2/4/83 44 HRv22A/8 GLOSE S 2 58 Ok 58 2/4/63 45 HR-v23A GLOSE S 2 58 Ok 58 2/4/83 82 0111U a- - , - w,,,e -, ,,,c,,-,,--,,;_-,-- _nne-r.-- - - , - - , - , m,w---,--,--,,, - , , , - _ - ,. ,e-

I PAGE = 2 0F 3 LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 PEACTOR BUILDING 1983 1983 1983 1983 1983 1983 1983 1983 RESULTS GIVEN IN STO. CUSIC CENTIMETERS PER MINUTE'(SCCM) ,

ITEMS TAGS DESCS OPERS SIZE ASFOUND COMMENTS ASLEFT OATES o,ooes eeseeee. ........ .

...... ....... ...... ....... eeeeee ........

46 HR-V239 GLOBE S 2 58 ok 58 2/4/83 47 IA-V6/20 GLOSE HW 2- 150 OK 58 2/10/83 i 44 IC-V2 GATE MO 6 894 OK 894 11/5/82 49 IC-V3 GATE P- 6 13500 FAILED 4764 12/7/82 50 IC-V4 GATE P 6 144000 FAILED 3893 11/24/82 51 IC-V6 GATE P 3 71 OK 71 11/2/82 52 IC=V16 CHECK N/A 4 87 LOW 87 11/3/82 53 IC-V18 CHECK N/A 6 71 LOW 71 10/29/82 54 LR-v1/10 GATE HW 6 199 OK 199 2/3/83 55 LR-v4 GLOSE Hw .75 70 OK 70 10/12/82 >

56 LR-V5 GLOSE HW 2 70 OK 70 10/12/82 57 LR-V6 GLO8E hw 2 70 OK 70 10/12/82 58 LR-V49 GATE HW 6 3970 OK 418 7/12/83 59 MU-V2A GLOBE MO 2.5 36 OK 36 1/28/83 60 MU-V28 GLOSE Ho 25 36 OK 36 1/28/83 61~ MU-V3 GATE P 2.5 36 OK 36 1/28/83 62 HU-V18 GATE P 2.5 130 LOW 130 2/1/83 63 MU-V20 GATE P 4 77 OK 77 12/7/82 64 MU-V25 GLOBE Ho 4 225 OK 225 1/6/83

~6 5 - MU-V26 GATE P 6 28 LOW 28 1/6/83 66 MU-V116 PIST CHK N/A 1.5 272 LOW 272 12/6/82 67 NI-V27 GLOSE HW 1 572 HIGH 36 3/11/83 68 NS-V4 GATE N/A 1.5 58 LOW 58 1/8/83 69 NS-V11 CHECK N/A 8 581 LOW 581 1/9/83 70 NS-V15 GATE MO 8 124 LOW 124 1/9/83 71 NS-V35 GATE MO 8 58 LOW 58 1/8/83 72 PENET104 SLK FLG N/A 2 70 Ok 70 8/4/83 73 PENET105 BLK FLG N/A 10 58 OK 58 2/9/83 74 PENET106 GLK FLG N/A 4 58 OK 58 2/8/83 75 .PENET210 SLK FLG N/A 2 70 OK 70 8/4/83 76 PENET211 SLK FLG N/A 2 70 OK 70 8/4/83 77 PENET241 SLK FLG N/A 18 50' OK 50 8/8/83

-78 RS.V2A GATE MO 8 926 OK 926 10/19/82 79 R8-V7 GATE. MO 8 71 OK 71 10/20/82 l 80 3A-V2/3 GLOSE HW . 2 50 OK 50 8/9/83 181 SF-VR3 GATE HW 8 36 OK 36 1/24/83 8? WOG-V3/4 GL/GA M0/50L 2 2835 HIGH 218* 3/24/83 83 WOL-V303 GLOSE MO 4 84 OK 84 12/23/82 84 WOL-V304 CATE D 4 497 OK 497 12/23/82 85 WOL-V534 GATE P 8 58 LOW 58 1/16/83 86 W0t.-V535 GATE P 8 58 LOW 58 1/16/83 '

87 EQPFLG FLANGE N/A 216 52 OK 52 88 PERACCES MISC. W/A , 96 991 OK 991 ~5/29/83 2/22/83 89 PERACCES MISC N/A 96 1813 OK 1813 11/26/83 l 90 EHEACCES. MISC N/A 96 10693 HIGH 2822 5/21/83 91 EHEACCES MISC N/A 96 1107 OK 110T 11/24/83 83 0111U

' FAGE

3 0F 3 LOCAL LEAK RATE TEST RESULTS THREE WILE ISLAND vNIT 1 REACTOR BUILDING 1983; 1983- 1983 1983 1983 1983 1983 1983 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM)

-TAGS ASFOUNO COMMENTS ASLEFT oooeeees eeeeeee eeeeeeee eeeeeee

. TOTAL , 352365 36879 PENTOTAL 305663 FAILED 32716 ACC CRIT .104846 104846 FOLLOWING 15 THE TERMINOLOGY USED IN THE PREVIOUS COMPUTER DATA 8 i 1) el .(ALONE) OR ANY OTHER NUMBER OTHER THAN ZERO IN THE FIRST DECIHAL PLACE MEANS TEST SCHEDULED.

.01 (ALONE) MEANS NO DATA AVAILABLE FOR THE YEAR OR THAT

.THE TEST WAS DELAYED. (E.G. VALVE NOT INSTALLED YET OR NOT IN PREVIOUS TESTING SCOPE.)

3)'e001= (OR ANY NUMBER OTHER THAN ZERO IN THE THIRD DECIMAL PLACE) AFTER A LEAK RATE ( 1-.E. 5950 0. 0 01 )

MEANS ACTUAL LEAK RATE WAS GREATER'THAN MEASURED / RECORDED VALUE. .

6) ASFOUND- LEAK RATE (SCCM) IN THE AS- FOUND CONDITION BEFORE ANY REPAIRS OR ADJUSTMENTS.

5) ASLEFT- THE LEAK RATE (SCCM) AFTER ANY A0JUSTMENTS/ REPAIRS.

6) DATES- DATE OF THE LAST ACCEPTABLE TEST RESULTS FOR THE ITEM '

.7) DESC- DESCRIPTION OF THE VALVE OR PENETRATION.

B) OPER- TYPE OF VALVE OPERATOR (ACTUATOR).

9) NOTEST- THE TECH SPEC SCOPE DID NOT REQUIRE THIS VALVE TO BE TESTED OURING THE RESPECTIVE YEAR.

10) NOVALVE- THIS VALVE WAS INSTALLED DURING A LATER REFUELING OUTAGE .

i 11) COMMENTS- COGNIZANT ENGINEER COMMENTS ABOUT THE RESULTSI

A) FAILED- EXCEEDED THE PLANT ESTABLISHED LEAKAGE ,

RATE LIMIT FROM SP 1303-11.18 ENCLOSURE 10 WHICH MADE REPAIR / ADJUSTMENT NECESSARY.

8) HIGH/ LOW- SUBJECTIVE JUDGEMENT OF COGNIZANT ENGINEER. REPRESENTS THE RESULTS WITH RESPECT TO THE LEAKAGE WHICH THE TYPE OF F LEAKAGE BARRIER (E.G GATE VALVEe GLO8E VALVE CHECK VALVE, FLANGE, ETC.) IS CONSIDERED TO BE CAPABLE OF WINT0UT EXTRA 0ROINARY REPAIR / ADJUSTMENT.

C)OK- NO PROBLEMS WITH LEAKAGE D) OTHER- E.G. NEWVALVEs NOVALVEe NOTESTs REPACKED .

SEATWORK, STEM 8 ENTE ETC. (SELF-EXPLANATORY)

12) SIZES- THE NOMINAL PIPE SIZE FOR THE LEAKAGE BARRIER.

84 OlllU s

J ie .._..--._,--,y--.-----y.~.._-y-v, - y , m _ m m m,w w , _ , - , _ ----ww,,-mm.__ ---,.___--.m-, -

-_e-- .- - - - - - - - -

6 q.

APPE!OIX I THREE MILE ISLAND UNIT 1

-1984 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT SP 1303-11.18 01/01/84 - 07/18/84 85 0111U

It0EX - 1984 R. B. LOCAL LEAK RATE TESTING REPORT

1. PURPOSE

2. - SLMMARY OF WORK ACCOWLISED e

2.1 Valve Testing / Repairs 2.2. Access Hatmes 2.3 Penetration Pressurization

3. METICOS OF TESTING 3.1 Valves J 3.2 Access Hatmes 3.3 Penetration Pressurization

4. TEST EQUIPMENT USED 4.1 Valves 4.2 Access Hatmes 4.3 Penetration Pressurization

5. SLNMMtY ant > INTERPRETATION OF DATA 5.1 Valves 5.2 Access Hatmes 5.3 Penetration Pressurizatic.s

6. ERROR ANALYSIS 6.1 Valves 6.2 Access Hatmes 6.3 Penetration Pressurization

7. REFERENTS

8. ATTAC$ENTS 8.1 Results Evaluation Pro dure / Repair Criteria 8.2 Tabulation of Individual Test Repairs 86 0111U

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

- w ,.

REACTOR' BUILDING LOCAL LEAK RATE TESTING REPORT

- 1984 REFUELING FREGRJENCY 1.: ' PtRPOSE

~~

-- l '.1 To provide analysis to the Nuclear Regulatory Commission -

on the eighth periodic type B and type C leakage tests performed on the Three Mile Island Unit 1 Reactor Building.

t' This is in accordance with " Reactor Containment Leakage Testing for Water Cooled Power Reactors". Appendix J,

. Part 50, Title 10 Code of Federal Regulations mich required the contents of this sumary report to become part af the Type A test report along with the details of any o ther type B and type C testing performed since the previous type A test (also required per technical specification 4.4.1.1.8).

3-The unit has been in cold shutdown since the March 28, 1979 Unit II accident e xcept for periods of hot functional tes ting.

'2. SLM4AR Y OF WORK ACCOMPLISHED 2.1 Valve Testing / Repairs Appendix J . Type B and C. leak tests were performed on the .

components as listed in TMI Unit 1 Tedinical Specification 4 . 4.1. In addition the followirg components were leak tested though not yet listed in the Technimi Specification. A Tech. Spec. change was pending to add these:

.l. W-V1A/B, 2A/B, 3A/B, 4A/B - New System y

. 2. IC-V16/18, NS-Vll - Check valves-no t previousiy tested. .NRC request to add.

2 Monthly non-tech spec leak checxs were done when convenient in addition to annual tests of R. B. Purge Valves AH-V1A/B/C/D due to recurring sea t leakage.

Results on these leak ched<s were used to e medite timely

( adjustments / repairs.

, Repairs were initiated on the following components due to higher than desirable leakage.

1. AH-V1B & AH-V1D - Adjusted seats to regain L acmptable leakage.

87 0111U iL

2. CA-V5A/5B - Hand wheel type override device was in a position that prevented full valve closure.

Permanently removed override provision to prevent recurrence of problem.

3. IC-V3 - Extensive seat / wedge work.

4. LR-V49 - Mechanical damage to teflon sea t inserts.

Replaced seat rings.

During a November 1982 valve inspection the ethylene propylene rubber seats in AH-VlD were found to have cra &ed. The cra&ing did not affect leakage and was no t judged to be a short term safety concern. It was evaluated by the valve vendor as exmssive mold release agent. The venoor provided new seats for all four purge valves. The new seats were installed in AH-V1A/lB/lC/1D in March 1983.

In August 1983 the seats on AH-VlB were found to be cra&ed and a sample of the material was once again sent to the vendor for analysis. Once again the crading did no t affect valve leakage. The seats for AH-V1B were replamd with material from the same batch as that which failed. The vendor cunnitted to replacing the seat material after development of a more suitable f abrication method. The vendor, Pratt Valve Company, submitted a 10CFR21 type report to the NRC on the seat cra&ing problem $1ch was now attributed to delamination between plies in the rubber and the promised new seats are to be one piece (no plies). Testing of materials by the vendor has delayed the shipment of new type seats. They have not ye t been received as of 07/18/84 2.2 Access Hatch Testing / Repair _s 2.2.1 Door Seals SP 1303-11.25 Door sealleak tests were performed a t the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months f requency as required by Technical Specification 4.4.1.2.5 while ho t functional testing wa s being performed. All of the seal tests satisfied the surveillance procedure administrative acceptance criteria.

2.2.2 Overall Hatch Tes t Semiannual integrated type leak tests were performed on each access hatch in 1984 as required by Technical Specification 4.4.1.2.5. There was 88 0111U L

E-acceptable but higher than desirable leakage on the emergency acmss hatch and parts are on order to repair the bulkhead equalizing valve.

2.3 Penetration Pressurization SP 1303-11.24 Quarterly readings were recorded from the flow rotameters which supply air pressure or nitrogen pressure to reactor building mechanical and electrical penetrations as required by Technical Specification 4.4.1.2.5. No penetration leakage problems were noted although flow ,

meter malfunctions required meter repair, and oceasiona1 tubing leaks in the air supply system were found and eliminated.

O e 9

89 0111U i

p

3. METHODS OF TESTING 3.1 Valve Test Methods Testing was performed by use of TMI Unit 1 surveillan procedure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This procedure gives detailed guidanm on the tes t equipment and methods to be used for each penetration / valve. The following general philosophy is mntained in the surveillance procedure.

3.1.1 Use air or nitrogen a t a pressure differential across the valve greater than Pa (calculated accident pressure) . 55 psig was normally used.

3.1. 2 Assure that the pressure is emrted in the a ccident test direction unless it can be demonstrated that pressurizing in the, opposite direction is as conservative. Butterfly valves AH-V1A/lB/lC/10, and globe valves WDG-V4, SA-V3, and IA-V20 were ,

tested in the reverse direction.

3.1.3 Assure that the test volume is drained of liquid so tha t air or nitrogen test pressure is agains t valve seats.

3.1.4 Assure that the test verifies valve packing integrity in those cases where the packing would be an R. B. leakage boundary.

3.1.5 Assure adequate time period for stabilization of test conditions.

3.1.6 Assure test equipment is calibrated and used in a manner consistent with the data accuracy desired (weekly meter standardization was performed to verif y meters accurate within + 4% full scale. --

MP 1430-Y-22) .

3.1.7 Assure that the fluid blod<ing system is drained and vented during tests on the associated -

containment isolation valves to prevent any effects it might have on the tes t results (mos t o f the F. B. system piping is seismic 3).

3.1.8 Assure valves to be tested are closed by the normal method prior to testing.

3.1.9 Document as-found conditions (prior to adjustments / repairs) and as-lef t conditions.

90 0111U

3.1.10 Record test instrument scale readings prior to doing any data corrections.

3.1.11 Assure that system drains and vents which could serve as containment isolation valves, are closed,

. capped, and tagged af ter completion of the test program.

A training program prior to the refueling outage was performed to help assure tha t the above philosophy was understood by the personnel involved in the testing.

3.2 Access Hatch Test Methods 3.2.1 Ooor Seal Leak Tests-Method Door sealleak tests were performed by use of SP 1303-11.25. This promdure gives detailed guidance on the tes t equipment and methods to be used.

The door seal tests are performed by pressurizing the interspace between the double seals on each door with metered air a t the manufacturers recommended test pressure of 10 psig. Af te r stabilization the air rotameter indimtes the rate of air input required to maintain the tes t pressure.

3.2.2 Overall Hatch Leak Test -- Semi-annual overall hatch leak testing was performed by use of TMI Unit 1 Surveillanm Promdure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This pro dure gives detailed guidan on the test equipment and methods to be used. The overall integrated leak test verifies the integrity of all of the following barriers:

a. Hatch shell/ welds,

b. Rubber door seals,

c. Teflon operating shaf t packing,

d. Bulkhead electrical penetrations,

e. Penetration pressurization check valves,

f. Emergency air flange and associated "0" rings on outer bulkhead, 91 01110

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g. Bulkhead 'equalizin'g bal1 valves and associated mounting flanges /"0" rings. . , s

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The overallleak tes t is performed'by pressurizing the hatch to greater _than calaJlated accident

, pressurezand observing the rate of pressure drop on a high'aco; racy (Heise) pressure gage.

Pressure corrections are made by reference to a barometer. Minimum test duration is'4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 'arter a l' hour stabilization period.

3.3 Penetration Pressurization lethod Quarterly readings were taken on'the flow rotameters whidi are permanently installed in the penetration pressurization system. These readings represent the air / nitrogen makeup rate required to maintain approximately 60 psig in medianimi penetrations and 30 psig in electrical penetrations. High meter readings have 't occasionally occJrred but these have ,been a ttributed to <

leaks in the corraession fittings in the penetration pressurization system or to malfunctioning (stud ()

rotame ters. Testing was per plant surveillance procedure SP 1303-11.24

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92 9111U

'4. TEST EQUIPMENT USED 4.1 Valve Test Equipment (See Figure 1)

a. Rotameters - Sets of 3 Mfgr. - Brook s Inst. Co.

Model - 1114 Full View 1

Ranges:

Float Mat'l. Tube No. Range Pyre x R 150 8-1,120 SCCM _

Saphire R-2-15C 100-12,000 SCCM _

Carboloy R-6-158 1,000-142,000 SCCM Accuracy 1 2% full scale industrial accuracy

b. Temperature Indicators (as follows or similar)

Mfgr. - Ashcrof t Model - EH or AH / 3" or 5" Dial Range - 3CP-1300F Accuracy - 1 20F

c. Pressure Indicators (as follows or similar)

Mfgr. - Ashcrof t Model - 1279 1/2" Dial Range 60 or 0-100 psig Accuracy - 1 2 psig

d. Pressure Regulator (as follows or similar)

Mfgr. - Union Carbide Corp.

Model - UPG 3-75-580 Range 100 psi output / 0-3000 psi input 93 Olllu

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e. Calibration Rotameters (Set of 2)

Mfgr. - Brooks Inst. Co.

Models - 1110-05K2B1249,1110-08K281Z06 Ranges 16,000 SCCM, 3,600-234,000 SCCM Repeatability - + 1/4% of instantaneous Accuracy - 1 1% instantaneous

f. Flow rate Calibrator Mfgr. - Brooks Inst. Co.

Model - 1056A Range O to 2,400 SCCM Accuracy - + 0.2% of indicated volume 4.2 Access Hatch Test Equipment

a. Precision Pressure Gage (as follows or similar)

Mfgr. - Heise Model - CM Range - 0.60 psig Resolution - 0.25 psig Accuracy - 0.1%

b. Barometer (as follows or similar)

Mfgr. - Pennwalt

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Model - FA185260A Range - 10.8 - 15.5 psia Resolution - 0.005 psia Accuracy - 0.1%

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4.3 Penetration Pressurization Tes t Equicmen t

a. Flow Rotameters - (Permanent System Equipment)

( Mfgr. - Brooks Inst. Co.

Model - 1114 Range 10 S7H at 60 psia air Accuracy - + 2% Industrial aCQJraCy

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96 0111U

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5.

SUMMARY

AND INTERPRETATION OF DATA 5.1 Valve Test Results As-Found/As-Left Leakage - Also see tabulation of

, individual results in Attachment #2. The as-fornd

. leakage greater than acceptance criteria was not considered to require a License Event Report due to the long term plant shutdown condition during which the leakage was found.

Total Leakage Ted. Spec. Limit ' % Tech. Spec. Lifnit As-Found >249,090 SCCM 104,846 SCCM >100%

As-Left 3 9,783 5CCM 104,846 5CCM 38%

NOTE: The total shown above is cumulative by penetration and not the total of all valve leakages, i.e. ,

Only the highest valve leakage on each penetration is counted. This number is labeled as "PENTOTAL" on the tabulation of results in Attachment 2.

EXAMPLE: Penetration XYZ has three containment isolation valves inside the reactor building in parallel and one outsioe. The leakage from the three valves inside totals 500 SCCM and the outside valve is 1000 SCCM. The penetration leakage is counted as 1,000 SC04 ng 1,500 SC04.

5.2 Access Hatch Test Results

5. 2.1 Overall Hatch Leakage - SP 1303-11.18 - See the computer tabulation of 1984 leak rates in Attachment #2. The leakages were considered to be satisfactory though future repair /retes t is planned on the emergency access hatcn after the bulkhead equalizjng valve seats are replaced.

5.2.2 Door Seal Leakage - SP 1303-11.25 None of the door sealleak tests ex eded the 3 S7H adninistrative leakage limit. Typimlly, the

. leakage was less tnan 1 SCFH.

y U 5.3 Penetration Pressurization (PP) Leakage - SP 1303-11.24 Leakage Rates - SCFH Mechanical Electrica1 Da te 03/14/84 0.5 6.1 No penetration leakage problems were loentified.

97 01110

. W_ M L

Occasional meter problems were found and repaired and oceasional leaks of tubing / pipe fittings in the PP system were lomted and e]iminated.

There is no tedinial specification limit on penetration pressurization aystem leakage.

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98 0111U

6. ERROR ANALYSIS

, 6.1 . Valve Testing Errors __(For purge valves see Section 6.2) l-

" The flow meters used in the field have normal industrial o accuracies of + 2% full scale in the 10-100% (15-150 mm) scale range. Frior to use mm versus sccm graphs were developeo for nost of the meters by 10 point calibrations using high accurscy (f 1% instantaneous) lab rotameters.

During the leak test program weekly 3 point standardizations were performed on the field rotameters to yerify continued accuracy. The acceptan criteria for these standardizations was a variance of no more than 4%

from the calibration graphs. If meters were repaired or the 3 paint standardization exmeded the inaccuracy limit a new.10 point calibration was performed. Scale readings on the leak rate procedure (SP 1303-11.18) data sheets were evaluated and correcteo using the methods in Attachment 1. Conservative bias was introduced into the results by assuming 15 mm (10% of scale) as the minimum

' scale. Half of the test results actually showed a lower scale reading. More involved error corrections were not considered meaningful based on the very high total leakage as-found and the low total leakage as-lef t.

Several of the meters, however, did not receive the above described calibration prior t.o use sin the calibration standard had not been returned from the vendor until late in the test program. The affected field meters were calibrated af ter use and leak rate data was currected accrrdingly.

. 6.2 Access Hatch and Purge Valve Testing Errors The measured pressure drops were corrected by adding the minimum scale increment of the gage used for both the helse gage and the barometer. This conservatively

, corrected for the resolution and repeatability errors.

Gages used were recently calibrated. /\ minimum one hour temperature / pressure stabilization period was used prior to Jad1 pressure drop test. The a c ss hatches and purge valves are not instrumented to allow temperature corrections.

6.3 Penetration Pressurization Testing Errors These test results are used for information only and do no t count toward the totalleakage limit for Technical Specification conformance. The meters, installed permanently in the system, have 1 2% full scale industrial accuracy.

99 OlllU

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7. REFERENCES 7.1 1430-22 Standardization of Flow Rotameters

-7. 2 SP 1303-11.18 Reactor Building Local Leak Rate Testing

.. (Rev. 26) 7.3 Three Mile Island Unit 1 Technimi Specification 4.4.1

.7.4 TMI Surveillance File (for Data sheets) 7.5 - S P 1303-11.24 R. B. Local Leakage Penetration Pressurization (Rev. 5) 7.6 SP 1303-11.25 R. B. Lomi Leakage Ac ss Hatd1 Door Seals (Rev. 9).

4 1 00 0111U

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101 0111U

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ATTACEND4T 1 RESULTS EVALUATION PROCEDURE (SP 1303-11.18 Enclosure 9) 102 0111U L_

Attachment 1 R. B. LOCAL LEAK RATE TESTING RESULTS EVALUATION The vent rotameter reading will be used if it can be demonstrated by the test data that all significant CIV leakage is being accounted for. If CIV packing, fluid block check valve, or gasket leakage was evident tne supply totameter

.. results will be used unless this non-seat leakage was measured reliably and documented.

FOR USE OF SUPPLY ROTAMETER DATA: FOR USE OF VENT ROTAMETER DATA:

Promdure: ,

Procedure:

a) Record supply meter reading in (1) a) Record vent meter reading in below*. Also identify the meter (1) below.* .

used by tube # in (8) below and the metering pressure in (9). b) Record downstream verification meter reading in (2) below.

b). . Convert meter units in SCCM units Also identif y the respective using lates_t lab meter calibration meters used in (8) below and curve. Enter in (3) below. the metering pressure in (9) .

c) Correct results for temperature, c) Convert meter units .o SCCM Enter supply temperature in (4) units using lates t lab meter b elow. calibration curve. Enter in (3) below.

Calculate and enter in (7) below.

. d) Correct results for tempera ture. Enter ven t g temperature (OF) in (4) o below.

then Calculate and enter in (5) below e) If measurements of any other significant leakage paths

If meter scale reading was less (fluid block check valve, than 15 mm (minimum scale) use packing) are being claimed 15.mm in calculations. enter corrected flow (SC04) ir. (6) below.

103 0111U

G ATTACif4ENT 2 DATA 1984 TYPE C REACTOR BUILDING LEAK RATE TESTING 104 01110 m

PAGE = 1 0F 3 LOCAL LEAK RATE TEs7 RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILOING 1984 1984 1984 1984 1984 1984 1984 1984 RESULTS GIVEN IN STO. CUBIC CENTIMETERS PER MINUTE (SCCM)

ITEMS LAGS DESCS OPERS SIZE ASFOUND COMMENTS ASLEFT DATES

.a... ........ ........ ..... ..... ........... ........ ...... .......

1 AH-VIA/B BFLY P/MO 48 3583.001 HIGH 983 4/5/8*

2 AH.VIC/D OFLY M0/P 48 51873.001 FAILED 1940 */6/84 3 CA-V1 GLOBE MO 1 60 OK 60 3/15/84 4 CA-V2 GATE P 1636 1 HIGH 1636 3/15/94 5 CA-V3 GLOBE M0 1 60 OK 60 3/15/84 6 CA-v4A GLOBE MO 1 50 OK 50 2/28/84 7 CA-V48 GLOBE MO 1 50 OK 50 2/27/94 8 CA-VSA GATE P 1 118040 FAILE0 80 2/29/84 9 CA-V58 GATE P 1 11252 FAILED 50 3/1/04 10 CA-V13 GLOBE MO 1 60 OK 60 3/15/8*

11 CA-V189 GATE P 2 2140 NIGH 2140 3/26/04 12 CA-V192 LFT CHK N/A 2 69 LOW 69 3/26/84 13 0 0 0 14 0 0 . 0 15 CF-V2A GLO8E MO 1 61 OK 61 3/16/84 16 CF-V28 GLOBE MO 1 61 OK 61 3/15/86 17 CF-V12A LFT CHK N/A 1 737 OK 737 3/16/84 18 CF-V128 LFT CHK N/A 1 60 LOW 60 3/15/8* '

19 CF-V19A GATE P 1 761 OK 761 3/17/84 20 CF-V198 GATE P 1 52 OK 52 3/17/84 21 CF=V201 GATE P

. 197 OK 197 3/16/84 22 CF-V200 GATE P 1 a68 OK 368 3/15/84 23 CM-V1 BALL P 1 50 OK 50 2/23/84 24 CM-V2 8ALL P 1 236 HIGH 238 2/23/84 l 25 CM-V3 BALL P 1 50 OK 50 2/23/84 26 CH=V4 BALL P 1 50 OK 50 2/23/84 27 OH-V64 GLOBE HW 2 587 HIGH 587 3/4/6*

28 GHaV69 STOP CHK HW 2 65 LOW 65 3/2/84

. 29 0 0 0 .

30 FTTEAST FLANGE N/A 30 166 HIGH 168 3/17/84 31 FTTwEST FLANGE N/A 30 81 OK 81 3/17/84 32 HMsVIA GLOBE S .5 72 OK 72 3/20/84 33 HMsV10 GLO8E S 5 46 OK 46 3/18/84 34 HM-V2A GLOBE S .5 68 OK 68 3/20/84

= 35 HM-V28 GLOBE S .5 46 OK 46 3/1d/84 36 HM-V3A GLOSE S .5 69 O K~ 59 3/20/84 37 HM-V3B GLOBE S .5 46 OK 46 3/18/84 38 HM-V4A GLOBE S .5 69 OK 69 3/20/84 39 HM-V48 GLOBE S .5 46 OK 46 3/18/84

= 40 HP-V1 GATE HW 6 69 LOW 69 3/23/06

_- 41 MP-V6 GATE HW 6 69 LOW o9 3/23/8*

42 HR-V2A/8 GLOBE HW 2 309 OK 309 3/19/84

= 43 HR-V4A/8 GLO8E HW 2 294 OK 294 3/19/84 44 HRV22A/8 GLOSE S 2 279 OK 279 3/19/84 j 45 NR-V23A GLOSE S 2 69 OK $9 3/18/84 los 0111c

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PAGE = 2 0F 3 LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR SUILDING 1984 1984 1984 1984 1984 1984 1984 1984 RESULTS GIVEN IN STO. CUBIC CENTIMETERS PER MINUTE (SCCM)

ITEMS TAGS OESCS- OPERS . SIZE ASFOUND COMMENTS ASLEFT OATES opoos eeCoeees eeeeeseo eeeeee oeseeee eeeeeeees seeeeeee seeeeeeee seeeeee C6 HR-V238 GLOBE S 2 69 OK 69 3/18/84 47 IA-V6/20 GLOSE HW 2 46 OK 46 3/22/84 48 IC-V2 GATE MO 6 486 OK 486 3/9/84

, .49 IC-V3 GATE P 6 24242 FAILED 7813 5/13/84 l' 50 IC-V4 GATE P 6 3438 HIGH 3438 3/10/84 i 51 IC-V6 GATE P 3 172 OK 172 3/11/84 52 IC=V16 CHECK N/A 4 134 LOW 134 3/11/64

'53 IC-V18 CHECK N/A 6 56 LOW 56 3/10/84 l 54 LR=V1/10 GATE HW 6 69 HIGH 69 3/23/84 55 LR-V4 GLOSE HW 75 69 OK 69 3/18/84 56 LR-V5 GLO8E HW 2 69 OK 69 3/18/84 57 LR-V6 GLOSE HW 2 69 OK 69 3/18/64 58 LR-V49 GATE HW 6 11252 FAILED 69 3/23/84 59 MU-V2A GLO8E MO 2.5 64 OK- 64 4/7/84 60 MU-V2B GLOBE MO 2.5 64 OK 64 4/7/84 61 MU-V3 GATE P 2.5 38 OK 38 4/7/86 62 MU-V18 GATE P 2.5 247 OK 247 3/29/e4 63 MU-V20 GATE P 4 56 OK 56 3/13/84 64 MU-V25 GLOSE Mo 4 149 ok 149 3/14/84 65 MU-V26 GATE P 6 60 OK 60 3/14/84 66' MU-V116 PIST CHN N/A 1.5 382 LOW 382 3/13/64 67- NI-V27 GLOSE- HW l . 69 OK 69 3/25/84 68 NS-V4 GATE N/A 15 107 LOW 107 3/13/84 69 NS-V11 CHECK N/A 8 1007 OK 1007 3/12/84 70- NS-V15 GATE MO 8 376 OK 376 3/12/8*

71 NS-V35 GATE MO 8 99 LOW 99 3/13/86 i 72 PENET104 BLK FLG N/A 2 52 OK 52 4/12/84 i 73 PENET105 SLK FLG N/A 10 46 OK 46 3/23/84 l 74 PENET106 SLK FLG N/A 4 46 OK 46 3/23/84 l

75- PENET210 SLK FLG N/A 2 52 OK 52 4/12/84 76 PENET211 SLK FLG N/A 2 52 OK 52 4/12/04 t 77 PENET241 SLK FLG N/A 18 46 OK 46 3/23/64 L 78 RS-V2A GATE MO 8 1321 OK 1321- '3/27/84 l

79 RS-V7 GATE MO 8 1201 HIGH 1201 3/22/84 80 SA-V2/3 GLOBE HW 2 46 OK 46 3/23/84 81 -SF-V23 GATE MW 8 135 OK 135 3/25/84 82 WOG-V3/4 GL/GA M0/50L 2 922 HIGH 922 3/11/84 i 83 ~WOL-V303 GLOBE MO 4 109 OK 109 3/17/84 84 WOL-V304 GATE O 4 52 OK 52 3/17/84

[

85 . WOL-V534 GATE P 8 194 LOW 194 3/24/84 l

86 WOL V535 GATE P 8 259 LOW 259 3/24/84 87 EQPFLG FLANGE N/A 216 47 O K, 47 3/19/84 88 PERACCES MISC. N/A 96 1123 OK 1123 6/2/94 l 89 PERACCES MISC N/A 96 .01 .01 l

90 EMEACCES MISC N/A 96 10248 HIGH 10248 5/16/84 91 EMEACCES MISC N/A 96 .01 .01 106 0111U y,wr <----ww-, -w-----,----- r- ---- --

- - - - _ _ _ _ _ _ - - _____.__._.-.---__.--_____________._-.--,,,.,-.y_,,-- -- _. =rn,--c-.-,-yrvr-r-e--------,v---

l PAGE = 3 0F 3 LOCAL LEAK RATE TEST RESucTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1984 1984 1984 1984 1984 1984 1984 1984 RESULTS GIVEN IN STD. CUSIC CEtlTIMETERS PER HINUTE (SCCM)

TAGS" ASFOUND COMMENTS ASLEFT oeeeeeee eeeeeee eeeeeeee eeeeeee TOTAL 252850 43543

-PENT 0TAL 249090 FAILED 39783 ACC CRIT 104846 104846 FOLLOWING IS THE TERMINOLOGY USED IN THE PREVIOUS COMPUTER DATAI

1) .1 - (ALONE) OR ANY OTHER NUMBER OTHER THAN ZERO IN THE FIRST DECIMAL PLACE MEANS TEST SCHEDULED.

.01- (ALONE) MEANS NO DATA AVAILABLE FOR THE YEAR OR THAT THE TEST WAS DELAYED. (E.G. VALVE NOT INSTALLED YET OR NOT IN PREVIOUS TESTING SCOPE.)

3)'.001- (OR ANY NUMBER OTHER THAN ZER0 IN THE THIRD DECIMAL PLACE) AFTER A LEAK RATE (I.E.59500.001).

MEANS ACTUAL LEAK RATE WAS GREATER THAN MEASURED / RECORDED

'vALUE.- .

6) ASFOUND- LEAK RATE (SCCM) IN THE AS- FOUND CONDITION BEFORE ANY REPAIRS OR ADJUSTMENTS.

5) ASLEFT- THE LEAK RATE (SCCM) AFTER ANY ADJUSTMLNTS/ REPAIRS.

6) DATES- DATE OF.THE LAS1 ACCEPTABLE TEST RESULTS FOR THE ITEM

7) DESC- DESCRIPTION OF THE VALVE OR PENETRATION.

8) OPER- TYPE OF VALVE OPERATOR (ACTUATOR).

l

~

9) NOTEST- THE TECH SPEC SCOPE DID NOT REQUIRE THIS VALVE To BE TESTED DURING THE RESPECTIVE YEAR.

~

10) NOVALVE- THIS VALVE WAS INSTALLED DURING A LATER REFUELING -

OUTAGE

! .11) COMMENTS- COGNIZANT ENGINEER COMMENTS ABOUT.THE RESULTSI Al FAILED- EXCEELL.) THE PLANT ESTABLISHED LEAKAGE

' RATE LIMIT FROM SP 1303-11.18 ENCLOSURE 10 WHICH MADE REPAIR / ADJUSTMENT NECESSARY.

8) HIGH/ LOW- SUBJECTIVE JUDGEMENT OF COGNIZANT l ENGINEER. REPRESENTS THE RESULTS WITH RESPECT TO THE LEAKAGE WHICH THE TYPE OF LEAKAGE BARRIER (E.G GATE VALVE, GLOBE VALVE CHECK VALVE, FLANGE, ETC.) 15 CONSIDERED TO 8E CAPABLE OF WINTOUT EXTRAORDINARY REPAIR / ADJUSTMENT.

C10K- NO PR08LEMS WITH LEAKAGE D) OTHER= E.G. NEWVALVE, NOVALVE, NOTEST, REPACKED i

' SEATWORK, STEMBENT, ETC. (SELF-EXPLANATORY)

12) SIZES- THE NOMINAL PIPE SIZE FOR THE LEAKAGE BARRIER.

107 0111U

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

ML20090H298

Contents

Text

1.0 INTRODUCTION

Description:

9.0 REFERENCES

m- o - ap g l } 5'*Si os A oCV -

SUMMARY

SUMMARY

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ML20090H298

Text

1.0 INTRODUCTION

Description:

9.0 REFERENCES

m- o - ap g l } 5'*Si os A oCV -

SUMMARY

SUMMARY

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