ML19210A111

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Reactor Containment Bldg:Integrated Leak Rate Test,Apr 1978
ML19210A111
Person / Time
Site: Crane Constellation icon.png
Issue date: 07/14/1978
From: Klingaman R, Shirk R
GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT, METROPOLITAN EDISON CO.
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{{#Wiki_filter:- T THREE MILE ISLAND NUCLEAR STATION UNIT 1 REACTOR CONTAINMENT BUILDING INTEGRATED LEAK RATE TEST APRIL 19'78 l METROPOLITAN EDISON COMPANY SUBSIDIARY OF GENERAL PUBLIC UTILITIES CORPORATION I ^ r. s '741b PREPARED BY C R. E. Shirk ILRT Engineer Gilbert Associates, Inc. E APPROVED BY / Ase 7jh Y y k. M. *J.1'dganarf / Terap on Enginedring i f Janager-Ce Metropolitan Edison Co.

79. tog 40 m._,._

1489 276 m

1 TABLE OF CONTENTS Section Item Title Paga 1.0 SYNOPSIS 1

2.0 INTRODUCTION

2 3.0 ACCEPTANCE CRITERIA 3 4.0 TEST INSTRUMENTATION 4 4 4.1

SUMMARY

OF INSTRU}E.4TS s 4.2 CALIBRATION CHECKS 8 4.3 INSTRUMENTATION SELECTION 8 4.4 SUPPLEMENTAL VERIFICATION 10 4.5 INSTRUMENT ERROR CORRECTIONS 12 5.0 TEST PROCEDURE 13 5.1 PREREQUISITES 13 5.2 GENERAL DISCUSSION 14 5.3 TEST PERFORMANCE 15 6.0 METHODS OF ANALYSIS 19 6.1 GENERAL DISCUSSION 19 6.2 STATISTICAL EVALUATION 21 7.0 DISCUSSION OF RESULTS 23 7.1 RESULTS AT P 23 a 7.2 SUPPLEMENT TEST RESULTS 23 8.0 TYPE B AND C LEAKAG:1_ RATE HISTORIES 25

9.0 REFERENCES

26 APPE:OICIES REDUCED LEAKAGE RATE DATA B. WEIGHT OF CONTAINMENT AIR AND AVERAGE CONTAINMENT TEMPERATURE VERSUS TIME C. REPORT OF 1978 REFUELING R.E. LOCAL LEAK RATE TESTI::G ~ D. REPORT OF MISCELLANEOUS LOCAL LFlK TESTI::G APRIL 1977 TO APRIL 1978 C.!!:';/Ccm 33ari'th 1489 277

t 1.0 SYNOPSIS The Three Mile Island Nuclear Station Unit i reactor containmen building was subjected to a periodic integrated leak rate test during the period from April 12, 1978 to April 15, 1978. The purpose of this test was to demone,trate the acceptability of the building leakage rate at an internal pressure 50.6 psig (P ). Testing was performed in accordance with the requirements of 10 CFR 50, Appendix J and AN51 N45.4-1972. The measured leakage rate based on the mass point met *.od of analysis and using absolute values corrected for instrument error was found to be 0.061 percent by weight per day at 50.6 psig. The leakage rate at the upper bound of the 95 percent confidence interval is 0.064 percen: by weight per day which is below the allowable leakage rate of 0.375 percent by weight per day at 50.6 psig. Since the industrial cooler syster was in operation during the integrated leak rate test, addition of the local leakage rate of the system isolation valves (RB-V2* and RB-V7) to the measured integrated leskage rate must be considered. In addition containment isolation valve IC-V4 could not be opened for draining. The combined local leakage rate of these isolation valves was 0.007 percent by weight per day. The addition of this value increases the total integ ated leakage rate to 0,.071 percent by weight per day. The supplemental instrumentation verification at P was 3.0 percent, a well within the 25 percent requirement of 10 CFR 50, Appendix J, Section III A.3.b. All testing was perfocmed by Metropolitan Edisen Company with the_ technical assistance of Gilbert Associates, Inc. Procedural and calculational methods were witnessed by Nuclear Regulatory Commission personnel and audited by the Metropolitan Edison Company site Quality Control staff. c.wto m~ne +

h kk dued'jd J\\ h 'd W l 20 INTRODUCTION The objective of the periodic integrated leak rate test was the, verification of the overall leak tightness of the reactor containment building at the calculated design basis accident pressure of 50.6 psig. The allowable leakage is defined by the design basis accident applied in the safety analysis in accordance with site exposure guidelines specified by 10 CFR 100. For Three Mile Island Nuclear Station Unit 1, the maximum allowable }ntegrated leakage rate at the design basis accident pressure of 50.6 psig (P,) is 0.10 percent by weight per day (L,). Testing was performed in accordance with the procedural requirements as stated in Metropolitan Edison Company Three Mile Island Nuclear St.ation Uni: ; Surveillance Procedure 1303-6.1. This procedure was recommended for approval by the Three Mile Island Nuclear Station Unit 1 Plan: Operations Review Committee and approved by the Unit Superintendant prior to the commencement of the test. The combined local leakage rates from tne reactor containment building isolation valves and penetrations required to be tested by 10 CFR 50, Appendix J, was less then 60 percent of the maximum allowable leakage rate (L,) at 50.6 psig prior to the commencement of the integrated leak rate test (Refer to Appendix C). Leakage rate testing was accomplished at the pressure levt.1 of 50.6 psig for a period of 44.5 hours. The 44.5 hour period was followed by an 8 hour supplemental test for a verification of test instrumentation. e b.tFt / NP af 3@ 1489 ?79

s 3.0 ACCEPTANCE CRITERIA Acceptance criteria established prior to the test and as specified by 10 CFR 50, Appendix J and ANSI N45.4-1972 are as follows: The measured leakage rate (L,,) at the calculated design basis a. accident pressure of 50.6 psig (P,) shall be less than 75 (L,), specified percent of the maximum allowable leakaga rate as 0.10 percent by weight of the building atmosphere'per day. The acceptance criteria is determined as follows: L,= 0.10%/ day 0.75L,= 0.075%/ day b. The test instrumentation shall be verified by means of a supple =antal test. Agreement between the contain=ent leakage measured during the Type A test and the containment leakage deter =ined during the supplemental test shall be within 25 percent of L,. m., m_,,o 14R9 780

t 4.0 TEST INSTRUMENTATION 4.1

SUMMARY

OF INSTRUMENTi' The sensor locations were the same as those used for the preoperational ILRT in 1974. Test instruments employed are described, by system, in the following subsections. 4.1.1. Temperature Indicating System Overall system accuracy: 0.19 F Overall system repeatability: 0.19 F l Components: a. Resistance Temperature Detectors Quantity 20 Manufacturer Rosemount Type Model 104 AAN, 100 ohm, l, plantinum Range, F 60-100 Accuracy, F 0.1 e Repeatability, F ! 0.1 O 1 I i l ew.:u,- n 1489 281

b. Bridge Cards Quantity 20 Manufacturer Rosemount Type Model 440-L3 Range, F 60-110 Accuracy, F ! 0.25% of span Repeatability, F t 0.25% of span c. Digita_ Indicator Quantity 1 Manufacturer Weston Type Model 1230

  • Range, F 60-110 Accuracy, F

! 0.1 Repeatability, F i 0.1

  • Modified for direct digital temperature readout 4.1.2 Dewpoint Indicating System Overall system accuracy:

! 1.12 F Overall system repeatability: ! 0.52 F

pw. tem :.,. m 140V 2d2

r Components: f a. Deweell Elements i Quontity 10 Manufacturer Foxboro Type Model 2711AG, 18 c'arat gold Range, F 0-100 Accuracy, F 1 1.0 Repea: ability, F t 0.5 b. Dewpci.: Recorder Quantity 1 Manufacturer Foxboro Type Model Y/ER312

Range, F

0-100 Accuracy,,F t 0.5% of span Repeatability, F t 0.15% of span 4.1.3 Pressure Monitoring System Overall system accuracy: ! 0.015% of indicated pressure Overall system repeatability: 0.001 psia ta'DM /Uf"f"!1.1 A eglL4 6 }dSQ 907

Precision Pressure Gauges Quantity 2 Manufacturer Texas Instruments Type Model 145-01 Range, psia 0-100 Accuracy, psia ! 0.015% of indicaced pressure Repeatability, psia 0.001% of full scale 4.1.4 Supplemental Test Flow Monitoring System Overall systa: accuracy: 1 1% of full scale l Flow meter Quantity 1 Manufacturer Brooks Type Model 1114-08 Range, scfh at 0 psig and 100 F 30.9 - 309 Accuracy, scfh 1% of full scale I Uf w 7 n 1489 284

I ?. 2 CALIBRi?.sa CHECKS Temperature, dewpoint, pressure and flow measuring systems were checked for calibration before the test in accordance with Metropolitan Edison Company Procedure 1430-Y-23, as recccmended by proposed ANS-56.8, N274, Draft No. 2, February 1, 1978. 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. l. 3 INSTRUMENTATION SELECTION Justification of instrumentation selection was accomplished, using manufacturar's repeatability tolerances stated in Section 2.1, by computing tha i.strumentation selection guide (ISG) formula. j Utilizing the machods, techniques and assumptions in Appendix G to proposed ANS-56.8, N274, Draft No. 2, dated February 1, 1978, the ISG was computed for the absolute method as follows: a. Conditions L,= 0.1%/ day P = 65.3 psia T = 74.5 F = 534.2 R dry bulb Tdp = 61.8 F dewpoint t = 24 hours (minimum expected test duration) b. Total Absolute Pressure: 2~ h - h e =2 (0.001) - -2 p e = ! 0.00071 psia CcertICommonwnu 1489 285 8

1 c. Water Vapor Pressure: e PV Sensor repeatability error (E): 1 0.5 F Measurement system error (c), excluding sensor: ! 0.13 F 1 0.5 F (0.0096 psia / F) E = y 1 0.0048 psia E = 1 0.15 F (0.0096 psia / F) c = py ! 3.0014 psia c = py ~ I 2 2 h- _ b t (E ) + (c ) no. of sensors e = i 2 2 h 1 (0.0048) + (0.0014) 10 e = 0.00158 psia e = py d. Temperature: eT Sensor repeatability error (E): 1 0.1 F = 1 0.1 R Measurement system error (c), excluding sensor: 1 0.160 F = 1 0.160 R 2 2 4 y n. f sensors (ET ) + (CT) e = T 2 2 4 h (0.1) + (0.160) 20 e = T 1489 286 Gltet/Cammonwe14 s e,

a e = 1 0.0422 R T e. Instrumentation Selection Guide (ISG) 2400 'o "pv "T +2 +2 ISG = ! 2 p p ~ 2 2 2400 0.00071 0.00158 .0422 +2 + 24 65.3 ) 65.3 534.2 ISG = 1 0.012%/ day The ISG does not exceed 0.25 L, (0.025%/ day) and it is therefere concluded that the instrumentation selected was accepti:la for use in determining the reactor containment integra:ai leakage rate. !. 4 SUPPLEMENT.C. VER1?ICATION In addition to the calibration checks described in Section 4.2, test instrumentation operation was verified by a supplemental test subsequent to the completion of the 44.5 hour leakage rate test. This test consisted of imposing a known calibrated leakage rate on the reactor containment building. Aft.er the flow rate was established, it was not altered for the. duration of the test. During the supplemental test, the mensured leakage rate was L =L,+L c v o D))j jj'Mii))'[h ~ m l i c.micum. r 1489 287 1

where, L = measured composite leakage rate consisting of the reactor e

building leakage rate plus the imposed leakage rate L, = imposed leakage rate ~L, = leakage rate of the reactor building during the y supplemental test phase Rearranging the above equation, L,=L -L tJ v c o The reacter :ntainment building leakage during the supplemental test can be calculated by subtracting the known superimposed leakage rate from the measured conoosite leakage rate. l The containment building leakage rate during the supplemental test (L,) was then compared to the measured reactor containment building y leakage rate during the preceding 44.5 hour test (L ) to determine instrumentation acceptability. Instrumentation is considered acceptable if the difference between the two building leakage rates is within 25 percent of the maximum allowable leakage rate (L,). m,~...._, 1489 288

4.5 INSTRUMENT E".ROR CORRECTIONS Subsequent to the 44.5 hour integrated leak rate test and the 8 hour supplemental leak rate test, calibration curves for the precision pressure gages, the RID's (including readout) and the deweells (including readout) were developed. These curves were developed using manufacturer's calibration data (for the precision pressure gages) and Metropolitan Edison Company Procedure, 1430-Y-23. Using these curves, each precision pressure gage reading, each RTD reading, and each deweell reading was corrected for instru=ent error. In addition, subsequent to testing, the Brooks flow meter was returned tc the manufacturer for recalibration. Using this informati:2, :ha flow meter readings taken during the supplemental leak rate tes: were corrected for instrument error. I I I J J l ~ Ce rt/Co r r.u s 12 1489 289

5.0 TEST PROCEDURE D D (\\ Q% i o 5.1 PREREQUISITES j d[j (- as y Prior to commencement of reactor containment building pressurization, the following basic prerequisites were satisfied: a. Proper operation of all test instrumentation was verified. b. All reactor containment building isolation valves were closed using the normal mode of operation. All associated system valves were placed in post-accident positions. c. Equipment within the reactor containment building, subject to damaga, was protected from external differential pressures. d. Portions of fluid systems which, under post-accident conditions become axtensions of the containment boundary, were drained and ventes. e. The penetration pressurization and fluid block systems were depressurized. Gauges were installed at penetration pressurization manifolds to provide means for detection of lekage into the system. These gauges were removed and the manifolds were vented prior to the start of the test. f. Pressure gagues were installed on closed systems within containment to provide means for detection of leakage into such systems. g. Local leakage rate testing of containment isolation valves.,and penetrations was concluded. cimicommanunu jk8h 9 13

E h. Pctential pressure sources were removed or isolated from the containment. i. All accessible liner weld channels (approximately 35 percent of the totall were vented to the containment atmosphere. j. A general inspection of the accessible interior and exterior areas of the containment was completed. "[@L %)S[ 5.2 GENERAL DISCUSSION D S " "D he J o Following the satisfaction of the prerequisites stated in Section 5.1, the reactor containment building pressurization was initiated at a rate :f approximately 2.5 psi per hour. Building internal temperature as =aintained at apprioximately 74 F. Building pressure and temperature were monitored hourly and the amperage required by the recirculation unit fans (AH-E-1A, IB and IC) was monitored etery 5 psig. Leak rate testing was initiated at the 50.6 psig pressure level. Atmospheric pressure at tine of leak rate test initiation was 14.37 psia. Forty-four and one half hours of data were collected at the 50.6 psig pressure level. During the test the following occurred at half-hour intervals (See Appendix A1: a. Pressures indicated by each. of the two precision gages vera recorded and the average calculated. b. The twenty RTD temperatures were recorded and the average calculated. c. The ten dewpoint values were recorded. The average of the ten values was converted to vapor pressure using steam tables. This permitted correction of the total pressure to the partial pressure of air by subtracting the vapor pressure, unicmrmn - 1489 291 u

The use of vapor pressure (P ), average temperature (T) and the total pressure (P ) is described in more detail in Section 6.1. All T original data is on file at Three Mile Is'and Nuclear Station Unit 1. The plot of average temperature and weight of air was pc.fo'rmed half hourly (See Appendix B). When convenient, the available half-hourly values of P , T and PT were~ transmitted via on-site portable computer terminal to the Gilbert Associates, Inc. home office for analysis using the CLERCAL computer program. Computer program results, including a least squares fit of the data, were returned to the site via the termir.al. A final computer run was made after data for a full 44.5 ho e period was available. Subseque.: :: the 44.5 hour leak test, a superimposed leakage rate was estab'is'..ed for an additional 8 hour period. During this time, temperatura, pressure and vapor pressure were monitored as descri',ed above. 5.3 TEST PERF0EMCE & 6 t } 5.3.1 Pressurization Phase Pressurization of the reactor building containment was started on April 12, 1978 at 1200. The pressurization rate was approximately 2.5 psi per hour. When containment internal pressure reached 12 psig, at 1807 on April 12, 1978, pressurization was secured. An inspection team entered containment to perform the 12 psig inspection. The 12 psig internal inspection was completely satisfactorily ani. pressurization was restarted at 1920 on April 12, 1978.

  • *^-

GSert /Cc-mme3:th I4dy 292 15

i During pressurization to the 50.6 psig pressure level, the following observations were made: F } D D A slight buildup of pressure on several of the pressure gad'ges a. installed on penetration pressurization manifold indicated a small amount of leakage from the fuel transfer tube flanges, the personnel hatch, and emergency airlock door seals. b. A decrease in the pressurizer level and an increase in,the RB su=p level was observed. This was investigated during the 12 psig inspection and was attributed to a ruptured tygon level tube which had been attached to the reactor coolant system cold leg. The tubing was then isolated which did not affect the integrated leak rate test. l c. The in:arspace between LR-V2 and the associated blind flange 1 (on one of the two leak rate depressurization lines) was l prescurized to containment pressure. No leakage was evident l from the blind flange. d. Pressure was slowly increasing in the interspace between LR-V3 and the asso: fated blind flange (on the second leak rate depressurization line). To prevent leakage into this interspace, which would show up initially as containment leakage, LR-V3 was opened with the containment pressure at approximately 50 psig to equalize the interspace pressure. LR-V3 was then closed and no leakage was evident from the blind flange. Pressure was also slowly increasing in the purge chaust interspace. This interspace was purposely equalized, using air from the purge exhaust air tank. When containment internal prescure reached 50.7 to 50.8 psig at 1233 on April 13, 1978, pressurization was secured. All penetration pressurization system tenporary manifold pressure i gauges were removed. I i i G h t/C.~ na W.h 1489 295 16

Dn1 D D {' h n 5.3.2 Integrated Leak Rate Testing Phase i I b UJ u a i Af ter waiting 4 hours, leak rate testing was started at 1700 on April 13, 1978. From 1700 on April 13, 1978 until 1700 on April 14, 1978, a leakage rate at the upper bound of the 95 percent confidence Laterval of 0.069 percent per day was indicated by the data collected. With the addition of the local leakge rates for RB-V2*, RB-V7 and IC-V4, the total integrated leakage rate was 0.076 percent per day. Since this value exceeded the acceptance criteria of 0.075 percent per day and since the data had ndt been corrected for instrument error, the test was continued. The pressurizar level was slowly decreasing and at 2025 on April 13, 1978, the casing drain on make-up pump IC was found to be open. Water was f'_:w from the drain at a rate of several gallons per minute to tha auxiliary building sump. This water was leaking from the reactor ccolant system through the make-up system check valves. These valves are not leak tight at low pressures. The loss of water from containment had a conservative effect on the indicated containment building integrated leakage rate. Leak detection was initiated and potential leakage paths such as the outside purge exhaust valve were investiagted. No major source of leakage was discovered. At 0045 on April 15, 1978, during a valve lineup verification, it was discovered that control room indication for IC-V2 (laside containment isolation valve) was indicating open. IC-V2 was closed, however this had no effect on the containment leakage rate as the outside containment isolation valve, IC-V3, was holding. At 1330 on April 15, 1978, the integrated leak rate test was concluded with an indicated containment integrated leakage rate of 0.064 percent per day based on 44.5 hours of data. The associated 95 percent confidence interval was 0.003 percent per day. 1489 294 CritErt /7,,cmT3c4g3d 17

4 5.3.3 Supplemental Leakage Rate Test Phase Af ter the 44.5 hour integrated leak rate test data was obtained and evaluated, and the leakage rate found to be acceptable, and a release permit had been obtained, a known leak rate was imposed on the reactor containment building through a calibrated flowmeter for a period of 8 houts. 5.3.4 Depressurization Phase j After all required data was obtained and evaluated, and the supplemental test results were found to be seceptable, and permission from the health physics department and unit superintends : was obtained, depressurization of the reactor containment hilding was started. A post test inspection of the building ravsded no unusual findings. 5.3.5 Post-Test Laakage Repair After the integrated leak rate test had been completed and during plant heatup, valve MS-V60A on the "A" steam generator was found to be blowing steam around its body / bonnet seal. The steam was blowing in a 360 degree are to a distance of approximately eight feet. This secondary system leak path undoubtedly had a large effect on tbe observed containment building integrated leakage rate. The valve had been replaced immediately prior to the integrated leak rate test and apparently its bonnet gasket was not installed. The leakage was corrected by injecting sealant into the bonnet joint.

wucmmes 1489 295 3

6.0 METHODS OF ANALYSIS 6.1 GENERAL DISCUSSION The absolute method of leakage rate determination was employed during testing at the 50.6 psig pressure level. The Gilbert Associates, Inc. CLERCAL computer code calculates the percent 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 mathe' of computing leakage rates uses the following ideal gas law equation to calculate the weight of air inside ontainment for each half hour: I W l #- I'I. K D**D [D) ~ )[a

where, W = mass of air inside containment, Ibn 6 lbm -

- in. K = 144 V/R = 5.3983 x 10 P = partial pressure of air, psia T = average internal containment temperature, R 6 3 V = 2.0 x 10 ft The partial pressure of air, P, is calculated as follows: PT1 + PT2 p, _p 2 vv !4h9 2}j c.wuc:mn.em 19 1

l

where, P

= true corrected total pressure from PI-390, psia Tl r P = true e rrected total pressure from PI-391, psia T2 P = partial pressure of water vapor determined by averaging the ten dewpoint temperatures and converting to vanor pressure with the use of steam tables, psia i The average internal containment temperature, T, is calculated as follows: D**D yfdj!$j[y] l'lfG [ T = su= of 2 + 459.69 R ow] .L 2 0 The weight Of air is plotted versus time for the 44.5 hour test and for the 8 hour supplemental test. The Gilbert Associates, Inc. CLERCAL computer code 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 + W t where W is the y 1 slope in lbs per hour and W is the weight at time zero. The least squares parameters are calculated as follows: 2 It EW -Et Ec W y f SXx y 1 1 - Et EW net W 1 f 1= 3xx Gac.iQmm:nwera 20 1489 297

f

where, 2

2 S nit - (I t ) f The weight percent leakage per day can then be determined from the following equation: vt. %/ day = 0 l is a negative slope to where the negative sign is used since Wg express the laakr.ge rate as a positive quantity. 6.2 STATISTIC.C. TJALUATION After perfor-ing the least squares fit, the CLERCAL computer code calcule.es the following statistical parameters: Standard error of confidence for the curve fit (S,). a. b. Limits of the 95 percent confidence interval for the curve fit. c. Limits of the 95 percent confidence interval for the leakage rate (C ). The significance 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 percent confidence interval for the leakage rate. e I 't U / L70 y j e e ah

1 Standard error of confidence is defined as follows: h E .W - (W +b t). 2 S, = y g' N-2 D g

where, g

W = observed mass of air

1) = least squares calculated mass of air (W, + Wg t

N = number of data points This parame:ar is an expression of the difference between an l l observed and a calculated (least squares) mass point. The 95 I i percent cenfidence incerval of the fix is twice the standard error of confidence (2S,). The " degree-of-fit" is evaluated by determining the number of data points, U, not falling in the g interval (R + W t) !2S. y The 95 percent confidence limit for the mass leakage rate is calculated as follows: S + (E t ) + L"U95 e NS

where, tudent's t Estr nution d th N-2 degrees of freeO _.

t = 95 This parameter is an expression of the uncertainty in the measured leakage rate. G.F.rtlComm u

  • w a 1489 299

~

7.0 DISCUSSION OF RESULTS 7.1 RESULTS AT Pa 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 corrected for instrument error of 0.061 %/ day. The 95 percent confidence limit associated with this leakage 7 ate is 0.003 percent per day. Thus, the leakage rate at the upper bound of the 95 percent confidence interval becomes L, = 0.051 + 0.003 %/ day L,,= 0.064 %/ day The measured leakage rate at the upper bound of the 95 percent confidence level is below the acceptance criteria of 0.075 percent per day 00.75 L,). A comparison of each of the observed weights with the weights calculated using the least squares line reveals only two of the ninety data points do not lie within the 95 percent confidence interval. Therefore, reactor containment building leakage at the calculated design basis accident pressure CP,) of 50.6 psig is considered to be acceptable. 7.2 SUPPLEMENTAL TEST RESULTS After conclusion of the 44.5 hour test at 50.6 psig, flowmeter FI-111 was placed in service and a flow rate, corrected for instrument error, of 193.9.SCFR was established. This flow rate is equivalent to a leakage rate of 0.053 percent per day. After the flow was established, it was not altered for the duration of the supplemental test. Gt rtlCcm: wee 3 1489 300

Themeasuredleakagerate(L)usinga.bsolutevaluescorrected[or instrument error during the supplemental test was calculated to be 0.111 percent per day using the mass point method of analysis. The I 95 percent confidence interval associated with this leakr.ge rate is 0.037 percent per day. None of the 17 data points is out of confidence. The building leakage rate during the supplemental test is then determined as follows: L,=L -L v c o D "r D 60 o f] r[,- 5 L, = 0.111 %/ day - 0.053 %/ day y d I _b Uu dj, 22 L, = C.:35 %/ day y Comparing this leakage rate with the building leakage rate measured i during the 24 hour test yields the following: bam - b '! l(0.061) - (0.058)l v 0.03 = L 0.10 a .he building leakage rates agree within 3.0 percent of L,which is well below the acceptance criteria of 25 percent of L,. Therefore, the acceptability of the test instrumentation is considered to have been verified. s. I e t i c.umumun ,c 1489 301

8.0 TYPE 3 AND C LEAKAGE RATE HISTORIES Refer to Appendicies C and D for the report on Type B and C testing performed since the previous Type A test. 1 ADO 709 C.wt / Commen*sn' i 't U / JUL 25

9.0 REFERENCES

1. SP 1303-6.1, " Reactor Building Integrated Leak Rate Test", Metropolitan Edison Company Surveillance Procedure. 2. Code of Federal Regulations, Title 10, Part 50, Appendix J, (1-1-75). 3. ANSI N!.5.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Reactors", A=erican Nuclear Society, (March 16, 1972). 4. Stea: Tables, American Society of Mechanical Engineers, (1967). 5. CLERc.c,

=puter Code, Gilbert Associates, Inc.

l I 6. 1430-Y-23, " Reactor Building Integrated Leak Rate Test Instrura:- Calibrations", Metropolitan Edison Company Procedure. 7. ANS 56.8, N274, "Contain=ent System Leakage Testing Requirements", American Nuclear Society, (Draf t No. 2 - February 1,1978). f3t0If $Nbb t i, i I I G ':ert lCa mn.esta 1489 303 26

APPENDICES m,um_.." 1489 304

APPENDI7 A REDUCED LEAKAGE DATA

~

l-- cc:ruceme.n::.s 1489 305

APPENDIX A REDUCED TEST DATA Containment Partial Pressure Containment Weight of Total Pressure Water Vapor Temperature Containment Air Time (psia) (psia) ( R) (lbm) O 4/13/78 1700 65.050 0.292 534.2 654405.94 1730 55.046 0.294 534.2 654344.29 1800 C5.035 0.291 534.1 654386.96 1830 6S.032 0.291 534.1 654356.64 1900 65.025 0.289 534.1 654306.10 1930 65.027 0.291 534.1 654306.10 2000 65.025 0.290 534.1 654295.99 9 2030 65.018 0.290 534.1 654225.24 3 2100 65.012 0.289 534.0 654297.21 9 2130 65.007 0.288 534.0 654257.79 y 2200 65.005 0.290 534.0 654216.34 i 2230 65.004 0.288 534.0 654227.46 S 2300 64.999 0.286 534.0 654197.13 2330 64.994 0.285 533.9 654279.22 2400 64.980 0.285 533.9 654218.55 llg 4/14/78 0030 64.984 0.286 533.8 654290.5% 0100 64.974 0.285 533.8 654199.53 0230 64.967 0.284 533.7 654261.42 0200 64.962 0.283 533.7 654220.96 0230 64.964 0.281 533.7 654261.42 0300 64.961 0.282 533.7 654220.96 0330 64.956 0.282 533.6 654292.98 0400 64.956 0.281 533.6 6543La.10 654231.27 };' 0430 64.947 0.279 533.6 CX) 'd) ~. U CD Ch

APPENDIX A (Cont'd) REDUCED TEST 9ATA Containment Partial Pressure Containment Weight of Total Pressure Water Vapor Temperature Containment Air Time (psia) (psia) ( R) (Ibm) 0 0500 64.947 0.281 533.6 654212.05 0530 64.941 0.275' 533.5 654332.65 0600 64.233 0.276 533.5 654242.59 0630 64.927 0.283 533.5 654112.06 0700 64.924 0.281 533.4 654224.57 0730 64.919 0.281 533.4 654173.97 08CS 64.912 0.279 533.3 654245.01 U OJ30 64.908 0.279 533.3 654204.52 5 0900 64.902 0.283 533.3 654104.31 S 0930 64.898 0.274 533.3 654152.90 [l 1000 64.892 0.277 533.2 654185.47 1030 64.889 0.275 533.2 654174.34 ~ 11.00 64.884 0.278 533.2 654094.35 1130 64.886 0.278 533.2 654114.60 1200 64.887 0.275 533.3 654031.43 lll 1230 64.884 0.277 533.2 654104.48 1300 64.882 0.278 533.2 654074.11 1330 64.886 0.278 533.3 653991.95 1400 64.890 0.275 533.3 654061.79 1430 64.892 0.274 533.4 653969.53 1500 64.891 0.274 533.4 653959.41 1530 64.893 0.274 533.4 653979.65 1600 64.892 0.275 533.4 653959.41 1630 64.894 0.276 533.4 653970.55 A C33 (>4 CD N

APPENDIX A (Cont'd) REDUCED TEST DATA Containment Partial Pressure Containment Weight of Total Pressure Water Vapor Temperature Containment Air Tina (psia) (psia) ( R) (lbm) 1700 64.897 0.273 533.5 653907.66 1730 64.898 0.276 533.5 653888.44 1800 64.896 0.277 533.5 653858.08 1830 64.893 0.273 533.5 653867.19 1900 64.889 0.272 533.4 653958.40 1930 64.884 0.270 533.4 653927.03 2000 64.883 0.270 533.4 653916.91 S 2030 64.882 0.272 531.4 653887.56 d 2100 64.883 0.273 533.4 653888.57 ? 2130 64.881 0.272 533.4 653877.44 s 2200 64.881 0.271 533.4 653887.56 I: 2230 64.881 0.273 533.4 653868.33 2300 64.881 0.273 533.4 653868.33 2330 6*e.880 0.273 533.4 653858.21 2400 64.879 0.273 533.4 653848.09 4/15/78 0030 64.875 0.273 533.4 653807.61 0100 64.871 0.275 533.4 653746.88 0130 64.867 0.271 533.3 653868.45 0200 64.863 0.273 533.3 653808.73 0230 64.861 0.273 533.3 653788.49 0300 64.859 0.275 533.3 653748.00 0330 64.855 0.274 533.2 653840.23 ~ 0400 64.852 0.274 533.2 653809.86 0430 64.846 0.271 533.2 653778.47 CD CD

APPENDIX A (Cont'd) REDUCED TEST DATA Containment Partial Pressure Containment Weight of Total Pressure Water Vapor Temperature Containaent Air Time (psia) (psin) ( It) (lbm) 0 0500 64.844 0.271 s33.2 653758.23 0530 64.843 0.272 533.2 653737.98 0600 64.840 0.271 533.1 653840.35 0630 64.834 0.271 533.1 653779.60 0700 64.830 0.272 533.1 653728.97 0730 6^.826 0.269 533.1 653717.83 0800 64.825 0.269 533.1 653707.70 7; 0830 64.820 0.266 533.0 653808.07 3 0900 64.813 0.269 533.0 653708.81 g 0930 64.803 0.269 532.9 653730.18 y 1000 64.794 0.266 532.8 653790.06 { 1030 64.79'O 0.266 532.8 653749.53 u ?'00 64.776 0.266 532.7 653730.38 1130 64.//z 0.268 532.7 653670.59 1200 64.767

0. 67 532.7 653630.06 1230 64.765 0.266 532.7 653618.91

( ) 1300 64.757

0. '.6 5 532.6 653670.68 1330 64.758 0,269 532.6 653642.30 SitP!1RIMPOSED TEST 4/15/78 2100 64.708 0.265 532.4 653419.40 2130 64.708 0.267 532.4 653400.13 2200 64.705 0.265 532.4 653378.84

(.., 2230 64.701 0.265 532.3 653480.29 2300 64.694 0.266 532.3 653390.03 2330 - 64.690 0.266 532.3 653349.46 C._) ' i

APPENDIX A (Cont'd) REDUCED TEST DATA Containment Partial Pressure Containment Weight of Total Pressure Water Vapor Temperature Containment Air Time (psia) (psia) ( R) (Ibm) 2400 64.685 0.266 532.2 653421.51 g 4/16/78 0030 64.679 0.265 532.2 653370.79 W 0100 64.672 0.268 532.2 653270.38 0130 64.671 0.266 532.1 653402.28 0200 64.666 0.265 532.1 653361.70 0230 64.665 0.264 532.2 653248.06 0300 64.662 0.267 532.1 653301.84 0330 64.659 0.268 532.1 653261.26 S 0400 64.654 0.266 532.1 653229.81 E 0430 64.650 0.265 532.1 653208.50 ? 0500 64.645 0.265 532.1 653157.78 t S 8 CD a

APPENDIX B k*EIGHT OF CONTAINMENT AIR AND AVERAGE CONTAINMENT TEMPERATURE G-'ty lComm~anB __ _ 1489 311-

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APPEllDIX C THREE MILE ISLAtlD U LIT 1 1978 REFUELIflG REACTOR SUIL0itlG LOCAL LEAK RATE TESTIriG REPORT SP 1303-11.18 1489 313 .=.-.

APPE:10IX C THREkMILEISLATIDUilIT1 1978 REFUELIt!G REACTOR SUILDIflG LOCAL LEAK RATE TESTIrlG P.EPC' SP 1303-11.18 e 95 1489 514 1

IllDEX 1. PURPOSE 2.

SUMMARY

2.1. Testing 2.1. Valve Repairs 3. METH005 ?

4. TEST EQUIPMEtiT 5.

AtlALYSIS OF RESULTS - AS FOUtID/AS LEFT 5.1. Interpretation of Data 52. Error Analysis 6. REFEREi!CES ATTACHiiEllTS I. Results Evaluation Procedure II., Data 2 1489 315

.-l-REACTOR B'JILDIflG LOCAL LEAK RATE TESTIllG ilRC REPORT 1978 REFUELIllG {D 1. PURPOSE 1.1. To provide analysis to the fluclear Regulatory Commission on the third periodic type B and type C leakage tests performed along with the second periodic integrated leak rate test of Three tlile Island Unit I reactor building. This is in accordance with " Primary Reactor Containment Leakage Testing for Water Cooled 9:. er Reactors," Appendix J, Part 50, Title 10 Code of Federal Regulatic. s,hich 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) 1489 316 3

SUMMARY

OF WORK ACCOMPLISHED f.q;qgJg @]$ lD'I3-)('6 2* jj)<1"*lD jn-uJh 2.1. Testing 2= Reactor building refueling frequency local leak rate testing was performed on the containment isolation valves and penetrations listed in the technical specifications and those additionally ccmmitted to be tested per Reference 2. Twelve (12) valves (IA-V6/20, SA-V2/3, LR-V1/2/3/4/5/6/4g) and four other devices (equipment access flange, Penetration 241, Fuel Transfer Tube Flanges) which were previously tested by quarterly penetration pressurization system flow meter readings were tested this year by Type C (Appendix J) test methods. A total of approximately eighty-one (81) seat and/or packing leak tests were perforced, nine (9) as retests after repairs. Three (3) of the containment isolation valves had 'igher seat and/or packing leakage than the cognizant engineer could accept and repairs were performed. 1489 317 4

, 2 ; 2.. VALVE REPAIRS Two (2) gate valves (.IC-V4, LR-V2) required refinishing of seating surfaces. The seats in one ball valve (CM-V2) and one but'.:-fly valve (AH-V1B) uere replaced. AH-V1B had satisfactory leakage as-ica.id but the rubber seats had slight cracking. The packing was replaced in seven (7) valves' (LR-V1, 2,3,4,5,6,49) s 5 1489 718

3. METh005 0 .aTING Testing was performed by use of TNI Unit 1 surveillance procedure SP 1303-11.18 Reactor Building Local Leak Rate Testir.g. This procedure gives detailed guidance on the test equipment and methods to be used for each penetration / valve. ' The following general philosophy is contained in the surveillance procecure. 1. Use air or nitrogen at a pressure differen'tial across the valve greater than Pa (Calculated accident pressure) 2. Assure that the pressure is exerted in the accident test direction unless' it can be demonstrated that pressurizing in the opposite direction is as conserva tive. 3. Assure that the test v:lume is drained of liquid so th.t air or nitrogen ' test pressure is agr.ns valve seats. D D 0 l i /j I 4. Assure that the test verifies valve packing integrity. e 5. Assure adequate tics period for stabilization of test conditions. 6. Assure test equipment is celibrated and used in a manner consistent with the data accuracy desired. (Weekly meter standardization was performed to verify meters accurate within + 5% full scale. MP 1430-Y-23 ) ~ 7. Assure that the fluid. blacking system is drained and vented durin5 tests on the associated containment isolation valves to prevent any effects it might have on the test results. (The majority of the F. B. system is seismic 3 ) 8. Assure valves to be tested are closed by the narcal method prior to testing. g. Document as-found conditions (prior to adjustments / repairs) and as-left conditions. 10. Record test instrument scale _ readings prior to doing any data corrections. 1489. 319

11. Perform test rig bypass valve tests weekly.

12. Assure that system drains and vents which could serve as containment is61ation valves, are closed and capped and tagged af ter completion of the test program.

A training program prior to the refueling outage was.also perforced to help assure that the above philosophy was understood by the personnel involved in the testing. 1489 320

4, TEST Fn" ~'NENT (SeeFigure1) ~ Brooks Model 1114 01F A1A rotemeters were used to measure the supply and/or vent flow rate for each valve and penetration (except for the purge valves which were tested by p'ressure drop methods). These flow meters are fitted with 0 - 150 mm scales and have quick-disconnect couplings to allow switching meters for proper scale. The range of the meters for both.zero and fif ty five psig metering conditions is given on Figure #1, which also ~ shows the valving, tubing and other controls for the testing apparatus. The flow rotemeters were standardized once a week against identical lab meters which had been factory calibrated ~ prior to the outage. (See Reference 1) The testing apparatus also included calibrated pressure gages for regulation of proper test pressure and thermometdrs to allow correction of ~ readings for significant variations frca calibration conditions. v D** . i~f f'19 3 - ' on uw u[s s'[ O 9 8 IbObl321:

T~0.5PECiM M l'~ _] ] VEN7/NG 4qCh l I 1 I G I 7' i i. 5 w.9~W:sw. I fj l. ( l .\\ smd 1 a .,.o ,,,rcisia l ' VENTING LOCATION I I +{[0- .y. g l e i 6-l

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5. AtlALYSIS OF RESULTS AS-FOU?ID/AS-LEFT (See Attachment II for data)

"As-found" leakage data were recorded on an individual data sheet for each valve / penetration tested. The data sheet was signed by the Test Forema'n'and a Cognizant Engineer. Retesting was performed for those v&lves which were repaired. 5.1. Interpretation of Data 5.1.1 As-found leakage Results (Also see Attachment II) The "as-found" total Reactor Building local leakage is shown in the below table along with a comparison to Technical Specifications criteria. AS-F0utlD TOTAL REACTOR BUILDIllG LOCAL LEAKAGE Type Total Tech. Spec / Percent Tech. Test Leakace FSAR Limit Spec./FSAR Limit Remarh f12/ Air 69,179 seem 104,846 sccm 66% tt0TE: The totals shown are cumulative by penetration and not the total of all valves, i.e., highest valve (s) on penetrations added. Example: Penetration XYZ has one containment isolation valve inside tha reactor building and one outside the reactor building. One valve leaks 500 scem and the other leaks 1,000 seca. The leakage for the penetration is 1,000 sccm not 1,500 secm. The maximum leakage which can be forced through the worst vaives at a pressure of Pa is still 1,000 secm. m 1489 323

. 5. l. 2 As-lef t Leakage Results (Also see Attachment II) ~ (Cubsequent to repair / maintenance) The existing ccmbined reactor building. local leakage is shown below. Comparison to FSAR limit is also given. AS-LEFT REACTOR BUILDING LOCAL LEAYAGE Type Total Tech. Spec. Perccot Tech. Test Leakage Limit Spec. Limit Remarks N2/ Air 50,I63 sccm 104,846 scca 47.8% The total shown is cumulative by penetration and not the total of all valves tested. (See discussi:n note of section 1.1) 1489 324

5.2. Error Analysis The flowmeters used in the field have normal industrial accuracies of + 2% full scale in the 10-100% scale range. However. weekly comparisons of these meters with lab meters were done to verify better than 15% full sc[le ac.- curacy. The lab meters were certified as 11% full scale accuracy frem 10-100% F.S. by the manufacturer. See Ref. 6.1. for the meter Standardization Procedure. The usable scale range for the field meters and the lab meters bas 15-150 m millimeters. The relationship used to detemine ueter accuracy from standardization data was as follows: 2 2 % Field Meter Accuracy (Labmeteraccuracy) + (Largest deviation) 2 or (Industrial Accuracy) whichever is largest In cases where this calculated value exceeded 5%, (it was normally approximatel; 3%) or where the meter float did not m' ave freely when the mater was turned alternately upside dcwn and then right side up, the meter was dissasembled, cleaned, repaired, and then reassembled and retested. The scale readings on the data sheets were evaluated using SP1303-11.18 (See Attachment I). e 1 1489 3g5

6. REFEREtlCES 6.1. 1430-Y-23 Standardiration of Flow Roto.netars 6.2. Met-Ed to flRC Licensing Letthe 9/17/75 - Comparison of TMI 1 Tech. - Spec. with Appendix J - 10 CFR 50 6.3. SP 1303-11.18 Reactor Building Local Leak Rate Testing ' ~ 6.4 Three Mile Island Unit 1 Technical Specification 4.4.1 6.5. TMI Surveillance File (for Data Sheets) a V '3 1489 32.6

W O O 't e ATTACHMEllTS e t 0 e 1489 327 14

? ATTACHMEllT I RESULTS EVALUATI0tl PROCEDURE (SP1303-11.18 Enclosure 9) O e e e 4 S 1489 328 15

R.B. LOCAL LEAK RATE TESTIilG -(ErlCLO5URE g-SP1303-11.13) RESULTS EVALUATIO:t d ) 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 ' ], the supply rotameter results wil'1 be,used unless this non-seat leakaga '. ~' was measured reliably and dccumented.. O D Dh ~ Fbr" N aAm FOR USE OF SUPPLY FOR USE OF VErlT ROTAMETER DATA: ROTAMETER DATA: Procedure: Procedure: a) Record supply meter reading in (1) a) Record vent meter reading' in (1) belou*. Also identify the meter below*. ~ used by tube # in (S) below and the b) Record downstream verification metering pressure in (9). meter reading in (2) below. ~ b) Convert mater unit: Also identify the respective to SCCM units usin; htest lab meter cieters used in (8) below and calibration curve. En er in (3) the metering pressures in (9). below. c) Convert meter units to SCCM c) . Correct results for amperature. Units using latest lab meter I Enter supply temperature in (4) calibration curve. Enter in b elow. (3) below. d) Correct results for temperature. Calculateandenterin(7)below. Enter vent temperature (9 ) in F (4) below. then - ~ Calculate and ente'r in (5) below. c) If measurements of any other - significant leakage paths (fluid block check valve, packing) are being claimed e enter corrected flow (SCCM) . in (6) below.

  • If meter scale reading was less than 15mm (minimum scale) use 15mm in cal cul a tions.

(mm) (SCCM) A[530 )Xkl ) C "V*#. ( + ( + + 460 = SCCM (l) (2) (3) (4) (S), (8)(Identify meters used) + SCCM (6) 0 ~3 (9)(t eter Pressures = 'lV Leakage SCCM d (7) ~ 1489 32.9...

ATTACHMENT II DATA 197 8 REFUELIllG REACTOR BUIleDING LEAK RATE TESTING-e e ' S e S e e e e e O e e e e e b 1489 330 17

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~~KtT2VI A7B "~T5'00 I 693-)(ELTSE6fs 2613 27-T8 AH-ViC/D 1'500 3393 .3393 I~3/78 -^.c ~ CA-Vi 150. . ' 170 3/25/78'l ~ 170 ~CTFQ2 600 101 101-3/31/78 CA-V3 150 170 170 3/25'/78' CA-V4A 150 129 129 3/27/78 -"C'A -0 W 150 109 109 3727/78 ' CA-VSA 600 q'238 ,.238 '3/27/78. CA-V5B 600 ~ 238 '238'.3/28/78' TA VT3 1 50 102 R U W ITIC D-~ 128 128 3/31/78 T CA-ViS9 600 1497 HIGli '1497 3/19/78' CF-V2A 75 102 102 3/22/78 ~Cl' (/2B 7$ fD2 "TOD 3722/78 ~-- CF-V12A 225 fi O2 '102 3/22/78',' CF-V123 225 102 '102 3/22/78 ~~CFUTi% 2D 2 mfd FYIDH 2288 3/22778 ~ CF-Vi?3 300. 3? 350.3/22/78 CF-V20A 300 i:2 - C r~~V.TO T-- ~~760~~~ '. 2 ~~ 102 3/22/78 102 37f2/Y8 ~ ~~- Cii-Ut 100 i33 't 'i S8 3/24/78 CH-V2 100 2964 ' REPAIRED 129 129 3/29/78 Cif:0a T O'0 - - i F3 i FJ-3724/78 Cli-V4 100 165 168 3/24/78 DH-Vd4 75 50 50 4/8/78 - Im -TT69 275'- 7d3 . 260 W/8/78 HP-Vi 400 229 229.3/19/70 lip-V6 400 102 i02 3/19/78 ifi-Va/20 100 50 50 4/10/78 IC-V2 800 1544 HIGH 1544 3/2I/78 ,IC-V3 800 ' 131 131 3/21/78 ~~1 CCT,-'~~ 800

i'Tf56 RiEPITIRED 30170

. 30000 -11 91 a4/11/78 ~ IC-V6 400 , q-Q; 50 ' n, .,,f48,j j3/26/78.,. 48-LR-Vi 500 ' 50 3/2G/78 ~[R-V2 500 ?686 REPAIRED 55400 856S 8565 4/7/78 LR-V3 500 1005 1005 3/28/78 LR-V1 ~~ [R l> 5 ~~ 50 452 HIGH 452 3/28/78 C 5'O 492 ilIGH 492 3/28/78,. LR-V6 50 95 95 3/23/78'6 LR-V4? 500 50 50 3/20/78 tiU-V2ii 225 159 159 3/19/78 D h ]nk hMM liU-V2 D 225 102 liU-V3 900 102 L UN . dl /d } i i02 3/20/78' I \\ 102 3/19/78 liU-Vi8 900 408 408 3/16/73 ~ liU--V 20 600 102 102 3/19/73 liU-V25 150 102 i02 3/2i/73 t'iij - V 2 a 600 102 102 3/21/78 iiU-V i i /, 225 301 301 .3/20g73 . t:S - V 4 000 294 294 3/31/7B HS-V i.'. 800 6V 69 3/26/70 !!S-V5 5 800 294 294 3/31/70 f.;0--v 2 A - G00 296 296 3/3i/78 ,im.-V / - Ocv ,101T0 FII GU ~~ iOiTO -3/25/fd h, '.' 2 < a 100 50 4 / i W73 33 1489 331

.. I a daEO ' TAG Tr.RGET ASFD78 COMMENTS RETEST) RETEST 2 thSLT78 ASLTDATE u n nu m. n. nununs nunnuun unnunnun nununnn Hununun ununnun ununnn++ __SF-V23 6OO 10;' 1 0.' J?/13/JE__ UDG-V3/4 400 5886 .HIGH 5886 3/25/78 WDL-V303 600 130 130.3/28/78 Jij)L-V304 600 130 1 3.0 3/22/78 WDL-V534 1600 130 .130 4/2/78 UDL-V535 1600 130 130 4/2/78 'PEFJE.T_iO4 0 0 O 4/J i /_79 PENETiO5 .O O O.. 4/7/78 PENET106 0 O ...;,.O' 4/7/78* '. ' o"4/7/79 PENET2io O O PENET21i O O O 4/7/78 PENET241 O O O 4/4/78 FUEAST O 95 25 4/7/33 FTTWEST O 95 95 4/7/78 EQPFLG O 51 Si 1849'f4/4/78 PERACCES 1400 1949' 6/_'25/7 7 EMEACCES 1400 4163 4163 5/17/77 TOTAL 30625 71425 52350 PENTOTAL 69179 50163 ACC CRIT 104S46 104846 ^ ~? J,RTERMS - TERMINOL0GY USEJL.JLC_Q?1P.tJ.T.FR PRQ.CEAij.f_0E_LQC(A._kEs.K RATE TESTIt!G RESULTS. 1) .01 (ALONE) MEANS NO DATA AVAILABLE.

2). Oi - LOR ANY DF&ltif,L UALUE) (CER.J_kE.GK_EG.T_E_LI.<E,.595_OO_~O 1 )

MEANS ACTUAL LEAK RATE GREATER THAN MEASURED / RECORDED. VAL'E. U

3) TAROET-ADMINISTRATIVE LEAKAGE LIMIT. BASED ON' TESTING EXPERIENCE.

COMPLETE EXPLANATION GIVEll..IL spi 3Qj _i_i_._iB.u.___

4) ASFD___ - LEAK RATE (SCCM) IN THE AS-FOUND VALVE CONDITION, BEFORE ANY REPAIRS OR ADJUSTilENTS.

FOR THE DESIG* LATED YEAR.

5) ASLT,_

- LEAK RATE (SCCM) ATTAINED AFTER ANY ADJUSTMENTS / REPAIRS. 'h i", '

6) DESC - DESCRIPTION OF VALVE OR PENETRATION.

. w. ~ d '..... T -4. '.

7) SIZEI -- PIPE DI AMETER :(INCHES) FOR VALVE / PENETRATION.
8) RUMTOTAL-RUNNING TOTAL'. THIS II THE LIST OF' LEAKAGES UHICH' IS USED FOR DETERMINATION'0F REPORTADILITY.

A NEW ASFD LEAKAGE REPLACES THE PREVIOUS YEARS ASLT LEAKAGE. RETEST RESULTS ARE NOT INCLUCED.U@L Agg_ogr@m m WG_P;_EcygeAijy, ^ .r'

i,.,:-

.-.-Q ' '}, h.^^. . - ~ -?- 6 o uuu uuJLUU.::2 J WWuu e ..-M.*i ~ 1489 132 )g

' '~ APPEtIDIX D T}{REE MILE ISLAtl0 UtlIT 1 MISCELLAt1EOUS LEAX TESTIflG SItlCE PREVIOUS TYPE A TEST (APRIL 1977 TO APPIL 1978) REACTG?. BUILDIllG LOCAL LEAX RATE.TESTIflG REPORT 1489 .533 20

APPEllDIX D IflDEX 1. PURPOSE 2. SUP.F.ARY 3. METHODS 4 TEST EQUIPtiEliT 5 Af!ALYSIS OF RESULTS - AS F00 TID /AS LEFT 6 DATA e e e e e e 1489 334 21

APPEtiDIX D

1. PURPOSE To provide analysis to the fluclear Regulatory Ccmission on the various non refueling frequency Type B and Type C leakage tests perfomed on Three Mile Island Unit I reactor building since the previous Type A' test.

0 9 e e e e o e 9 e 4 S e

  1. 4 1489 335 22

A PFEi10IX D

2. SU:iFARY OF TESTIllG PERFORiE0 Atl0 REPAIRS B.

1. Penetration Pressurization System quarterly flow meter readings SP 1303-11.24 TESTI:lG Quarterly readings of the installed system flow rotometers were taken and ~ compared to the acceptance criteria specified in the proccoure. The acceptance criteria deals with: a. Limits on individual manifold flows I b. On the total system flow and c. On the correspondan:e bat,ieen the sum of individual manifold flow indicators and the indicatica en the supply to the system. Test dates and results are shown in section 6.1. REPAIRS The-repairs perfomed on the system were in every case repairs to the penetration pressurization system and not to the penetration seals. The repairs normally consistad of tightening or replacing tubing fittings. 1489 336 23

APPEtIDIX D 2.2. Access Hatches (Air Locks) Door Seal Tests TESTIt'G Door seal tests were performed routinely af ter door usage as required by Technical Specification 4.4.1.2.5b. These tests were performed per.' SP 1303-11.25 which requires pressurization to Pa (calculated accident test pressure) and the reading of a supply rotometer. The supply flow raeter readings were compared against an arbitrary 3 scn1 target criteria to determine the need for repairs. Reportability was based on indication of leakage through one of the doors of greater than 60 SCFH (The range of the installed flow indication. REPAIRS Failures of the door seals were generally followed by cleaning, lubri-cation and inspectio.-::f seals. If that did not eliminate the leakage problem, door adjust: ants were made, or seals were replaced. Due to a test pressure requirement which exceeds the manufacturer's reccmmended pressure, the door seals quite likely have failed leak tests at times when they would have been very capable of sealing in the accident pressure direction. The periodic door seal tests are not a realistic mock--up of accident conditions but are much more demanding on the equipment. 0 $lDl ~ 1489 337....

n

23. Access Hatches-Integrated Leakage (all seals)

TESTIflG_ Integrated hatch tests, by pressure drop, were performed semi-annually as required by Technical Specification 4.4.1.2.5b.,These tests were perfon=ed per SP 1303-11.18C which requires pressuriza-tion to Pa.(calculated accident test pressure) and a four hour pressure drop test. 1 The dates of testing and results are shown in report section 6.3 The calculated leak rates were used to update the total for Reactor Building local Leakage and the updated total was the basis for reportability. A target criteria of 1400 SCCM was established to give some basis for requiring repairs. REPAIRS Failures of the hatch integrated tests (as determined by the cognizant engineer) were always followed by tubing fitting adjustnents and door seal cleaning an'd lubrication. Tightening of shaf t seals was performed on the basis of local lea!: che r'<s. ' pgg @hb I489 338 25

. ~ e '3. METHODS OF TESTII;G 3.1. Penetration Pressurization System ~ Installed system flow rotemeters are read once per quarter and the' readings are ccmpared to the procedure acceptance criteria., S.2. Access Hatches - Door Seals Door seal tests are performed by pressurizing between the concentric rubber seals with 55 psig metered air. The supply flow rate is read, recorded, and evaluated. High leakage could be indicative of door seal problems, or problems with numerous other hatch penstrations which are pressurized from the same supply. 3.3. Access hatches - Integrated-Integrated hatch tests are performed by pressurizing the hatch interior to Pa (calculated accident pressure) or greater and observing the pressure drop ever a four hour period. O 9 -~ e e e e 9 9 ,e e e e e e a 1489 339 26

APPEflDIX D ~

4. ' TEST EOUIPMEffr_

41. Penetration Pressurization System Flow Rotemeters Manufacturer - Brooks Inst. Co. Model #1114 Range 0 - 9 SCFH (individual manifolds) 0 - 60 SCFH (Supply to system)

4. 2.

Access Hatches - Door Seals Flow Rotemeters_, Manufacturer - Brcoks Inst. Co. Model #1114 ~ Range 0 -9 SCFH Accuracy j; 2% Full Scale Pressure Regulater Manufacturer - Fisher Control Co. /L 3. Access Hatches - Integrated Leakage Pressure Gauge (Teaporary for Integrated Test) ~ .~ Accuracy f; 0.25% or better Range 0 - 60 psig with 0.01 psi scale divisions Barometer (Temporary for Integrated Test) Accuracy j;0.05 in Hg I487 440 27

APPENDIX D

5. AtlALYSIS OF RESULTS e

5.1. Penetration Pressurization System--Analysis of Resul ts There were no instances of excessive leakage on any containment boundary constantly pressurized by the penetration pressurization system. Where high flow meter readings were noted, it was always due to leakage in system piping, usually requiring tightening of fittings. 5.2. Access Hatches - Door Seals - Analysis of Results The number of tests and failures for each access hatch are shohn in the following table: TESTS / FAILURES Personnel 170/2' Emergency 29/C There were no inscances of c.oncurrent excessive leakage on both doors of either hatch Osreby providing assurance of one of the two series leakage barriers. When excessive leakage was found, the doors were ~ tested independently and the or.e with a good seal was locked closed pending repairs / retesting of the other door. 5.3. Access Hatches '- Integrated - Analysis of Results 'llone of the hatch integrated tests yielded results which would cause the updated total nf local leakage to exceed the 0.6 L criteria of 3 Tech. Spec. 4.4.1. Where the test results were.significantly above the target criteria, repairs were promptly performed and the hatch was retesced. eg b e 1489 341 23

DATA - MISCELLAtl0US LEAK TESTS (4/20/77 - 4/12/78) . r,. 1. Penetration Pressurization Quarterly Flow Meter Readings (SCFH) Test Date As-Feund As-Left 6/8/77 less than Satisfactory

  • 60*

9/10/77 0 0 12/14/77 18.5 20 2/26/78 19 19 ~ Data sheet lost 6.2. Access Hatch Door Seal-Meter Readings-Periodic fl0TE: Only tests which exceed the target criteria (3 SC511) are listed here. Hatch Test Date As-Found As-Left Personnel 6'27/77 4 less than 3 Personnel ci /73 4.1 less than 3 63. Access Hatch Integrated-Semiannual Pressure Drop Test Date As-Found As-Left Personnel 6/21/77 38000 761 Emergency 5/17/77 3700 3700 Personnel 12/12//7 1849-1849 Emeroency 12/7/77 8183 4163 e g o e t 9 29 14@O '47 .}}

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