Text

Integrated Leak Rate Test of Reactor Containment Bldg.
	ML19312C807
Person / Time
Site:	Oconee
Issue date:	10/06/1977
From:	Cribbe T; DUKE POWER CO.
To:	;
References
	NUDOCS 8001090609
	Download: ML19312C807 (50)
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t DUKE poi #ER COMPANY OCONEE NUCLEAR STATION DITEGRATED LEAK RATE TEST OF THE ,

REACTOR CONTAINMENT BUILDDIG

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DUKE P O TJ E R C0MPANY OCONEE NUCLEAR STATION .

UNIT 2 -

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INTECRATED LEAK RATE TEST OF THE REACTOR CONTADDfENT BUILDING i .

Prepared By: 1.A.h . 0 8. 6A -

Terrel E.'Cribbe "

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Test Coordinator . i

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Approved B -- '

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dah0C i B.G. Davenport ,

- 1 i Test Engineer l .

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1 TABLE OF CONT 1hIS SECTION PAGE

1.0 INTRODUCTION

'l.0-1 2.0

SUMMARY

AND CONCLUSIONS 2.1-1 e

2.1 Synopsis 2.1-1 2.2 Test organization 2.2-1 2.3 Initial Test Criteria 2.3-1 2.4 Test Results 2.4-1 2.5 Error Analysis -

2.5-1 3.0 DESIGN INFORMATION 3.1-1 3.1 Reactor Building 3.1-1 3.2 Measuremert system 3.2-1 3.3 Pressurization System 3.3-1 3.4 Recirculation System 3.4-1 3.5 Computer Programs 3.5-1 4.0 CONDUCT OF TEST PROGRAM 4.1-1 4.1 Local Leak Rate Test 4.1-1 4.2 Integrated Leak Rate Test 4.2-1 5.0 LOCAL LEAK RATE TEST FAILURE DATA 5.1-1 )

5.1 Intreduction 5.1-1

' 5. 2 Yailure Data 5.1-1 1

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1.0 INTRODUCTION

The periodic Integrated Leak Rate Test (ILRT) of the Oconee Nuclear Station Unit 2 Reactor Building was satisfactorily completed on August 1,1977. The testing was conducted in accordance with the requirements of FSAR Section 5.6.2.1, Techaical Specification 4.4, BN-TOP-1 (Bechtel Testing Criteria for ILR Testing), ANSI N45.4-1972 and 10 CFR Part 50, Appendix J. The absolute method of testing was employed with containment temperatures measured at 24 locations and ccntainment dewpoint temperatures at two locations. Leakage was measured at half the design basis accident pressure .of s 29.5 psig. A measured induced leak was used to verify the results.

Interpretation and final analysis of the test data show results well within the specified limits for this containment, which has a maxi =um allowable leak rate of .176% per day. The leakage rate for Oconee Unit 2 Containment Building was found to be .0969% per day.

The mean of the leakage rate over the 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months test period was

.07499% per day.

l 1.0-1

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2.0 SU&fARY AND CONCLUSIONS 2.1 Synopsis The successful ILRT was performed in accordance with periodic test procedure PT/0/A/150/03 as approved for use on July 8, 1977. This procedure is similar to that used for the pre-operational ILRT.

Pressurization for the ILRT began at 0512 hours0.00593 days 0.142 hours 8.465608e-4 weeks 1.94816e-4 months on 7/29/77 with both air compressors. Pressurization was stopped at 10 psig and the building was entered and inspected for indi-cations of leakage. The only indication of leakage was around the gasket of the inner door of the personnel hatch. The outer door was shut per the procedure and the compressors were restarted. At approximately 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months with the building pressure at 12 psig a leak was noted around the gasket of the outer door of the personnel hatch. Since the hatch pressure was only .1 psig it was decided to pressurize the hatch ex-terna11y to see if the outer door would seat. At 2330 hours0.027 days 0.647 hours 0.00385 weeks 8.86565e-4 months the hatch was pressurized and the outer door seated at 1.8 psig. The air supply to the hatch was secured and the RB pressurization continued. The compressors were secured at 0715 on 7/30/77 with the contain=ent at approximately 30 psig.

It was noted at that time that the personnel hatch was at the containment pressure (The hatch had equalized with the con ,

tainment).

The 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months test period started at 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months on 7/30/77. At 2315 hours0.0268 days 0.643 hours 0.00383 weeks 8.808575e-4 months on 7/30/77 inspection of data and calculations indicated that the test was satisfactory and data aquisition was terminated.

At 0220 hours0.00255 days 0.0611 hours 3.637566e-4 weeks 8.371e-5 months on 7/31/77 a known leak note of approximately

.177% was induced for purposes of verifying the capability to measure a leakage rate. This induced leak rate test gave favorable results. The personnel hatch was depressurized and it was noted that the inner door was not leaking at the full test pressure. Depressurization of the containment building was begun at 1135 on 7/31/77 and completed at 0500 on 8/1/77.

Instrumentation consisted of one (1) precision pressure sensor, twenty four (24) temperature sensors and two (2) dewpoint sensors which were used to measure containment atmospheric conditions during the ILRT.

The gasket on the inner door of the personnel hatch was re-paired on 8/4/77. The hatch was locally tested on 8/8/77 with satisfactory results.

i 2.1-1

2.2 Test Organization

'?he Perfornance secticn at Oconee Nuclear Station had overall responsibility for this periodic ILRT. The testing activities were supervised by the test coordinator. See Figure 2.2-1 for organization chart. The test personnel were as follows:

A. Test coordinator (and one alternate) - T. E. Cribbe Responsible for all ILRT Activities - H. V. Pham B. Data Engineers (one per shift) - R. P. Rogers Responsible for Testing Activities on - W. G. Newman their assigned shifts - R. R. Carpenter C. Data Takers (one per shif t) - J. H. Rowe Responsible for reading and recording - W. M. Matthews all test data - D. J. Vito D. Computer Operators (one per shift) -

T. E. Evans D. W. DeNard G. F. Long E. Operators (normal shift) 1 I

2.2-1 il I'

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l 2.3 Test Criteria )

i Pressure: 29.5 psig Leak Rate: .176% Limiting Condition for Operation

j. .132% Test Acceptance Temperature: 60 -100 Volume: 1,910,000 cu. fc.

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2.3-1

2.4 Test Results As shown in the computer printouts and graphs in Section 4.2 of this report, the calculated leak rate, the mean of the calculated leak rates and the mean of the measured leak rates are all well below the allowable leakage rate limit. The leakage rate test results are tabulated below:

Maximum Allowable Calculated C e Test Leak Rate (%/ day) Leak Rate g 29.5 psig .176 .097% .132 For reporting purposes, a leak rate of .097% shall be used for this test. This leak rate was obtained by using a least squares linear fit of the 15 min. leak rate data taken during the ILRT. These cal-culations are based on the total time method as deucribed in ANSI N45.4-1972.

The verification test consisted of imposing a known leak rate on the containment at the end of the ILRT. Results from this supplemental test are acceptable provided the difference between the supplemental test data and the type A test data is within .25 L t. A summary of the verification test data is given below.

29.5 psig ILRT %/ day Leak Rate calc. during ILRT, L .097 Imposedverificationleakratef'L t *177 Total .274 Upper verification limit = .274 + .25 L g

.318 Leak rate cale. during verification test .278 Lower verification limit = .274 .25L g .230 i

l 2.4-1 l

2.5 ERROR ANALYSIS The Icakage rate in weight percent per day is co=puted using the absolute method by the fornula:

1 - To P LR = 2400-g T Po _ (1) htere: .

Po = Initial Reactor Building absolute pressure corrected for water vapor ,

pressure.

P = Fins 1 Reactor Building absolute pressure corrected for vater vapor pressure.

To = Initial Reactor Building cean absolute tenperature.

T - Final Reactor Building mean absolute te=perature.'

H = Nunber of hours held at test pressure. .

The change or uncertainty interval'in IR due to uncertainties in the measured variables is given by: -

(dLR . gT 32

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,g , dLR . a h2 dLR ,o p)2 [dLR.ah )2 i + dT l W dB p}j dPO j dTo j

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Where a 1s the standard error for each variable.

The error in LR after differentiating is: '

b g ,2400 To . ep 2

To P .e p -P .e\

g [ top. (3)

H T Po TPo j (TPo (TZPo EI

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'P " P

'Po Po ,

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To = #To *T T The analysis technique, based os equation (3) above, was verified by K. -1 Horoschek and E. Weipport in "Tigntness Investigations on Reactor Safety l Pressure Vessels," Vol, 13, No. 3, March 1961. ,

l For sm:111 values of LR:

T = To, P = Po and l l

P ~ *Po' 'T ~ "To l

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l Therefore, the equation then si=plifies to:

-P "P -T T eLR = 2400 + + -

+i

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1 L ') ')l ..

or ,_ _., ,

4 eLR = g

+ (4)

Where ep = err r in pressure which accounts for the error in the absolt$te pressure measurement instru=ent (ePT) and the water vapor measurement.

system (epy) and where e7 = error in te=perature.

r ..

, 4 ep = (ep7)2 c,pv)2 (5) e = Absolute Pressure Instru=ent Accuracy Error PT dPy g = Dew Point Sensor Accuracy Error (Number of Sensors) dT D/

g = RTD Accuracy Error (Number of RID's)h .

To develop a nu:ierical value for g it will be assumed that:

H = 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months RB mean absolute pressure = 29.5 psig RB mean absolute temperature = 529.7*R (70*F)

RB mean dewpoint temperature = 68 F Substituting the instrument accuracy values given in. Section 3.2.1 into the above equations yields:

eg = 0.0005 percent full scale * -

Value of full scale repeatability used. This number is more significant for the error analysis than the absolute accuracy.

eg = 1 0.5 F = 1 0.353 F

/ 2 From the ASME Steam Tables at 68 F:

epy = 0.00414 pei i

' 2.5-2

ep = / (0.0005 psi)4 + (0.00414 psi)4 ep = 1 0.00417 psi c,.,= 1 0.07'F ,

(24)

Valua of full scale repeatability used. Tnis nu=ber is more significant.for the error a,nalysis than the absolute accuracy. ,

- 0 cT = 2 0.0143

Substitution into equation (4) yields:

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+ 2 0.0143 2400_

eLR = + 2 0.00417 10

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44.645 530.00

= + 0'.0330% per day e

e 2.5-3 .

REACTOR BUILDING INTEGRATED LEAK RATE TEST ORGANIZATION TEST COORDINATOR w

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Y DATA DATA SilIFT INSTRUMENT TAKER ENGINEER SUPERVISOR SUPERVISOR C W WER INST. INST.

OPERATOR OPERATORS TECH. TECll.

______________________.._______j_

3.0 DESIGN INFORMATION 3,1 _

REACTOR BUILDING The reactor building is a reinforced and post-tensioned concrete the structure designed to contain any accidental release of radioactivity front reactor coolart system as defined in the Final Safety Analysis Report (Reference 1). ,

The structure consists of a post-tensioned reinforced concrete cylinder and dome connected to and supported by a cassive reinforced concrete

- foundation slab as shown in Figure 3.1-1. The entire interior surface of the structure is lined with a 1/4 inch thick welded ASTM A36 steel plate to assure a high degree of leak tightness. Numerous mechanical and electrical systems penetrate the reactor building wall through welded steel penetrations.

Principal dimensions are as follows:

116 ft.

Inside Diameter Inside Height (Including Dome) 208-1/2 ft.

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Vertical Wall Thickness 3-3/4 ft.

- Doma Thicknass 3el/4 ft.

6 Foundat1on Slab Thickness 8-1/2 ft.

Liner. Plate Thickness 1/4 inch Internal Free Volume 1,910,000 Cu. f t. ;

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3.2 MEASUREMENT SYSTEMS Instrumentation used for the Oconee Unit 2 ILRT is similar to that used on previous tests conducted by Bechtel. The leak rate test measurement system is shown schematically in Figure 3.2-1.

! Reactor Building pressure was measured by a Texas Instrument precision l pressure gage. The unit was factory calibrated before the test.

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Reactor Building tenperature was measured by twenty-four (24) calibrated RTD's and read on a Leeds and Northrup Nc=atron digital readout device.

Each RID was assumed to be representative of a fraction of the total containment volume. In addition to the twenty-four (24) LTD's used.

for the calculation of Reactor Building temperature, four (4) additional RTD's were used to measure the liner plate te=parature.

Reactor Building dewpoint temperature was measured by two (2) Cambridge De., point Hygrometers .

Air samples for the two (2) dewpoint sensors caen from two (2) of the a'wf1hry fans which continually circulated air in the Reactor Building during the leak rate test. The relar:ive location of the humidity sensors i

is shown in Figure 3.4-1. A 0-10.45. sefm Brooks rotometer was used in establishing a known leak rate.

3.2.1 INSTRUMENT LIST . . .

Specifications for the instrumentation used for the Oconee Unit 2 ILEr are listed in Table 3.2-1. -

3.2.2 TEMPEL \TURE SENSOR LOCATIONS .

The locations of temperature sensors within the Reactor Br'Nh* are shown in Figures 3.2-2 through 3.2-6. .

3.2.3 RTD AND DEWPOINT VOLUME FRACTIONS ,

Volume fractions were used for calculating the average temperature and the average dewpoint temperature in the containment. These fractions ,,,

were determined using an equivalent volume for each sensor. The free ~.

volume of the containment was divided into " cells" with a sensor. eenter in each. Volume fractions are given in Table 3.2-2 _ , , , ,

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3.3 PRESSURIZATION SYSTEM 2eactor Building pressurization was acco=plished by two (2) electric motor driven air compressors operating in parallel. These campressors, purchased for pressurization of the Oconee Reactor Building, also include aftercoolers as integral equipment. The discharge from the compressors passes through a single air dryer which reduces the moisture content in the air prior to its entry into the Reactor Building. The specifications for these components are as follows:

A. Two (2) electric driven Joy Turbo-Air (20V2) centrifigual type air

' compressors with a capacity of 2300 scfm 0 80 psig. .

3. Tm) (2) Basco size 22048 af tercoolers (Integral to Compressors),

type "ES" Fixed Tubesheet, with a capacity of 2100 scfm @ 14.4 psia and with a design pressure of 150 psig.

C. One (1) Hankison (Model H-15) refrigerator type air dryer with inertial impingement separator, and a capacity of 3750 scfm (100 F Sat, inlet) @ 100 psig.

Three valves, 3LRT-15, 3LRT-16, and 3LRT-17 are used to control pressurization and depressurization of the Reactor Building. The controls for these valves are located in the test panel. The pressurization system is shown schematically in Figure 3.3-1.

e 3.3-1 l

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1 3.4 RECIRCULATION SYSTEM The Reactor Building Air Recirculation Systen consists of four (4) -

auxiliary fans and three (3) Reactor Building cooling fans. The auxiliary fans take suction through ducts in the upper region of the Reactor Building and circulate it downward. The Reactor Building cooling fans take air from midheight in the Reactor Building and exhaust it through duct work down to the lower levels of the Reactor Zuilding.

This is shown schematically in Figure 3.4-1.

t 3.4-1 O

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3.5 COMPUTER PROGRAMS All calculations, summaries, and reports were perfor=ed using the Duke Power Company ILRT computer program.

3.5.1 DUKE COMPUTER PROGRAM The Integrated Leak Rate Program calculates the leak rate for a nuclear reactor containment vessel. The program computes the leak rate at a given time from input values of pressure, te=perature and vapor pressure.

The leak rate, as a function of ti=e, is determined by elementary linear least-square nethods.

The Integrated Leak Rate Program is designed to allow the user to evaluatei containment leak rate test results at the jobsite during containment leak rate testing. Interim leak rate test reports may be obtained at any time during the testing period. Each interim report consists of two printouts.

The first printout. called the total-ti=e computation, uses the initial and latest input data to compute leak rate. Each computed leak rate is statistically averaged using a linear least-squares fit. Early in the test this method of computation gives indication as to whether or not .

the leak rate is acceptable.

A second printout, called the point-to-point computation, is also provided.

The point-to-point method uses the data at a given hour and the data from the previous reading to compute leak rate. Each individually computed leak rate is then statistically averaged using a linear least-squares fit.

The results of the two methods used indicate that either method is a satisfactory method for computing containment integrated leak rates.

Both methods of computing the containment leak rate are presented in ANSI N45.4-1972. However, the total-time method is recommended by Appendix J to 10 CFR 50.

3.5.1.1 Explanation of Program 3.5.1.1.1 Purpose .

(a) Process raw data for use in leak rate calculation.

(b) Calculate leak rate on a point-to-point or total time basis.

(c) Calculate a statistical point-to-point or total time leak rate by a linear least squares fit of the data.

(d) Output data.

3.5.1.1.2 Process Raw Data ,

(a) Temperature (a.1) Unit - F.

(a.2) Number - 28 (Includes 4 on liner plate).

(a.3) Input to computer will be in F. The computer will apply necessary calibration corrections to the temperature.

(a.4) Each temperature (28) will be weighted by a volume fraction.

The calculation for the average building temperature is:

T = IT, . V, lt 3.5-1 ff . .

F Where: T = Average Building Temperature,

= Temperature at each point, a + 1-28, ,FThe volume repres

= Volume fraction for each point. Volume fractions must add divided by the total volume.

'- 8 to 1.

R.by adding 459.7.

The Average' Temperature must be' converted to The temperature on the liner plate will be available for output.

(a.5)

(b) Pressure

- - (b.1) Unit - psia.

o (b.2) Number - 1. The computer vill apply necessary calibration

.(b.3) Input is in psia.

correction to the pressure. The calculation (b.4) for theThe pressure will be weighted by a volume fraction.

average building pressure is: .

P=WyPt Where: P = Average Building Pressure, psia P" = = Pressure by Texas Volume fraction (mustinstrument, equal 1) psia .

W

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(c) Baremeter_

(col) Unit - Inches of Hg 6 32*F. .

(c.2) Number - 1.

(c.3) Input is not used by calculations, but vill be svailable for output.

(d) Devooint Teciperature (d.1) Unit - F.

- (d.2) Number - 2. F. A calibration curve will be as- '

(d.3) Input to computer will be insociated with each sensor The and the inp Each Dewpoint Temperatt>re is weighted by a volume fraction. *

(d.4) calculation for the average is:

+ WT4 v2 .

T, = W T3 yy

'Where: T Dewpoint Temperature. *F ,F -

T , == First Sensor Dewpoint Temperature, F Tg v2

= Sec nd Sensor Dewpoint Temperature,

= 1)

W = Volume Frr -n (W3+W4 From the Dewpoint Temperature (Saturation Temperature) the Vapor (d .' 5) ~

Pressure (Saturation Pressure) is determined from the steam -

The range required is 40 to 120 F from the stea:a tables.

P, = f (T,)

y 3.5-2 .

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3.5.1.1.3 Leak Rate Calculations -

(a) Leak Rate on Hourly Basis

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Leak rate vill be calculated on a point-to-point basis from data obtained at the previous point readings (data subscript 1) and data obtained at the present point rcadings (data subscript 2).

(a.1) Absolute Method

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1 , T1 (P2-Pv2[

La = 2_40,0 OE T2 (Py -P yy - .

O Where in (a.1) above ,

L = Percent leak rate per hour (La or Lx)

T1 = Average absolute temperature of Reactor Building air at start cf each test period, F T2 = Average absolute ta=perature,of Reactor'Euilding air at end of each test period, P P1 = Absolute pressure of Reactor Building at start of each test period, psia .

F2 = Absolute pressure of Reactor Building at end of. each test period, psia Pv1 = Vapor pressure of Reactor Building at start of each '

test period, psia Pv2 = Vapor Pressure of Reactor Building at and of each test period, psia at = Time interval between start and end of test period, hours (b) Statistical Point-to-Point Lenk Rate _

Linear least square fitting is used to calculate a statistical point-to- -

point leak rate for the Absolute Method. .

Li = a + bei Where: a = Eti E(ti 2) ,ggg ggggg .

N E(ti2) - (Eti)4 '.

b = NI11ti - Eli Eti NE (ti') - (E ti)'-

Li = Statistical leak race '

11 = Calculated leak rate (

ti = Time between data sets N = Number of points to be fit (ausber of data sets)

(c) L ak Rate From Initial Data See Leak rate from the initial data set to each successive data set is calculated by the following equation:

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3.5-3 ,

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O (c.1) Absolute Method Laa = 2400 1 - Ti (Pt - Pvt)

TI , Tt (P1 - Pvi{

Where: .

1 = indicates data from initial set.

t = Indicates data from each successive set TI = Time from initial data set .

(d) Statistical Leak Rate Fro:2 Initial Data Sec ,

The leak rate is calculated by a linear least squares fit over a period -

from the initial data set to each successive data set. -

(d.1) The equations are the same as in 3.5.2.1.3(b) . -

(d.2) The data used is obtained from 3.5.2.1.3(c). ,

l (e) 95 Percent Confidence Limits CLi = Li + T oi ,

Where: .

CLi = Confidence limits at time ti Li = Statistical leak rate at time ti c1 = Sta~sdard deviation T = Tabulated constant T = 1.95996 + 2.37226 + 2.8225 .

(N-2) (N-2)Z .

Where: N =_ Number of data points ,

c1 = Variance 1 + 1 + (to - t)7 N Z(ti-ii)4_ _

Where: to = 0 '

t =1Iti N ,

Variance = [I(11 - a - bei) ]+ [N-] , ,

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INSTRUMENT SPECIFICATIONS -

Pressure Digital Readout -

Serial No. 10132 2646 .

Mfg. Texas Instrument 145 Modal Precision pressure" gage .

Type '

Range .

0-100 psia or 100,000 counts '

full' scala ' -

Scabilf47 i .001 psi

Repeats illity 1 .0005 psi ,,

Resolut.on i .001 psi i .

Accuracr i .015% of reading

.Pr.ssure Cauge -

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Mfg. Heise

Range 0-100 psig -

Accuracy 0.1 psi Repeatability 0.1 psi Temperature Elements

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Leeds & Northrup .

Mfg. *-

8197 .'

Model RTD, Copper,,100 ohms ' -

Type ,

0-1500F Range Repeatability and hysterisis i .02 F

Accuracy 1 0.12 r . -

Tamperature Indications for Te=pera'ture Ele =ents Leeds & Northrup -

Mfg.

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Model 245 Nu: natron ,

Range 0-15007 Reprod ;tivity i .07 F from 60 F to 120 F Accur .:y i .12 from 0 600F to 120 F and.1 48 Below 60 F - -

' Dewpoint Temperature .

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Cambridge .

Mfg. - :

Model. $92-C1 - -

Rangs' -1000F to +200 F .

Accuracy i 0.5 F -

Serial No. 332 and 333 -

Table 3.2-1 -

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Flow Indicator .

Brooks Mfg. - Rotometer Type 1110-24 Model O to 10.45 scfm Range Accuracy i 1% of instantaneous reading Repeatability Better than 1/4% of instrument reading 7004-39848 .

Serial No. .

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l Table 3.2-1 (Cont'd)

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VOLIIME FRACTIONS . .

l Volume Tractions for RTD's .

.' Volume Trnetion RTD #

. 03- ,

. 1 .

.02 ,-

2 - .02

' 3 .05 4 .02 - .

5 . .03 6 .01 7

.08 -

c 8 -

.05 9 .05 .

10 .02 11

. 02 12 -

.01 - -

13 .02 14 ,.

.02 .,

15 01 .

16 .

.05 -

17 .09 --

, 18 11

. i 19 -

.01 20 .01.- .- i 21

' 22 . .

. 09 11 ). -

23 -

.07 '

j 24

Total 1.00 - -

. . . l Dev Point Sensors Volume Fraction .

Volume Traction Dew Point Sensor # .

0.4 1 (Azimuth 100 Elevation 850') 0. 6' '

2 (Azimuth 260 Elevation 850') '

Total 1.0 ;

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Table 3.2-2 .

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O REACTOR BUILDU:G BASCdJ.NT FLCOR

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4.0 CONDUCT OF TEST PROGRAM.

4.1 LOCAL LEAK RATE TEST _

The purpose of the Local Leak tiesting program was to systematically d ,

check the integrity of valves (seats and packing), flanges, h pipe r an electrical penetration welds, seals and conpression, fittinga t at a eThese tests part of the boundaries of the containment system.

by section 4.4.1.2 of the Technical Specifications, have a combined -

Acceptance Criteria of less than or equal to 0.125% of the reactorFinal building at=osphere per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

, leakage rates shows that the total penetration leakage. rate was approximately 14 percent of the allowable.

4.1.1 TEST E THOD All electrical and mechanical penetration, including locks and hatche 59 psig.

were tested by pressurizing to %baronetric pressure werei recorded d by be of test decemined by penetration volu=e) and the leak. rata determ ne the mass difference nethod.

4.1.2 PENETRATION TEST RESULTS . .

i Per Technical Specification 4.4.1.2.3, the total hrleaI The total neasured leak rate from all penetrations prior to 7 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Results of all local penetration tests this test was .018% per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

done since the last type A test are given in tables 4.1-1 through 4.1-2 G

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4.2 COMPUTER CALCUTATIONS AND RESULTS 4.2.1 29.5 psig ILRT The curves for the 29.5 psig ILRT are shown in Figure 4.2-1 through 4.2-4. Supporting sumary data is given in Table 4.2-1.

4.2.2 29.5 psig Verification Test The curves for the 29.5 psig verification test are shown in Figure 4.2-5 through 4.2-8. Supporting sumary data is given in Table 4.2-2.

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3 D**}D a o Ju o Ju. .a
TYPE B LOCAL LEAK RATE TEST RESULTS I Date Leak Rate (%/ Day) Penetration
~~-0 03/06/75 2.85 I 10 Electrical Penetration ~~~
06/07/76 3.72 I 10_5 07/15/77 2.12 I 10 10/27/73 0 Fue'. Transfer Tebe Covers 0 04/17/74 09/11/75 0 06/30/76 0 07/27/77 0 10/26/73 0 Equipment Hatch 0 04/23/74 09/03/74 0 10/17/74 0 , 09/08/75 2.14 I 10 "' ~ 07/05/76 2.14 I 10 07/26/77 0
-3 Personnel Hatch 10/29/73 2.3 X 10 05/05/74 3.29 I'10-3 -3 b.'/16/74 2.63 I 10 ~3 01/J4/75 6.26 I 10 05/15/75 0 09/29/75 0 ,g 01/23/76 1.05 I 10 - -3 07/05/76 4.61 I 10 -3 11/09/76 4.61 I 10 -3 03/16/77 2.63 I 10 ~3 07/28/77 2.63 I 10 ~3 08/08/77 1.28 I 10 -3 Emergency Personnel Hatch 10/27/73 7.57 I 10 01/21/74 0 ~3 01/22/75 1.04 I 10-2 06/18/75 5.76 I 10 -2 06/20/75 9.88 I 10 06/22/75 3.29 I 10 -3 11/20/75 6.41 I 10 03/23/76 0 07/27/76 0 -3 11/16/76 6.42 I 10 ~3 03/16/77 2.63 I 10 -3 07/08/77 5.26 I 10 i
Table 4.1-1
TYPE C LOCAL LEAK RATE TEST RESULTS Date Leak Rate (*C/ Day) i Penacration
~~-2 Mechanical Penetrations _ 09/11/75 4.77 X 10 -2 06/30/76 1.54 X 10 -2 10/13/76 1.76 X 10 -2 07/27/77 - 1.13 X 10 l~~
Table 4.1-2
5.0 LOCAL LEAK RATE TEST FAILURE DATA 5.1 Introduction It is required by 10CFR50, Appendix J " Primary Reactor Containment Leakage Testing for Water Cooled Power Reactor" to report failure data on type A, B and C leakage tests. This report contains all failure data on Oconee Unit II since July, 1973. 5.2 Failure Data ITDI DATE REASON FOR FAILURE CORRECTIVE ACTION HP-37 9/2/75 Drain valve leaking past seat Capped drain line HP-124 9/2/75 Drain valve leaking past seat Capped drain line PR-1 9/9/75 Leakage past seat Adjusted and lubricated valve seat PR-2 9/9/75 Leakage past seat Adjusted and lubricated valve seat PR-5 9/9/75 Leakage past seat Adjusted valve closure HP-285 9/9/75 Leakage past seat Lapped seat HP-286 9/9/75 Leakage past seat Lapped Seat HP-146 9/9/75 Leakage past seat Lapped seat PR-5 5/18/76 Leakage past seat Lubricated and adjusted seat HP-120 6/10/76 Leakage past seat Lapped seat HP-286 5/12/76 Leakage past seat (seat cracked) Replaced valve HP-147 5/21/76 Leakage past seat Replaced valve PR-1 5/21/76 Leakage past seat Adjusted and lubricated seat PR-2 5/21/76 Leakage past seat Adjusted and lubricated seat PR-5 5/31/77 Leakage past seat Adjusted and lubricated seat PR-6 5/31/77 Leakage past seat Adjusted and lubricated seat PR-1 5/31/77 Leakage past seat Adjusted and lubricated seas. PR-2 5/31/77 Leakage past seat Adjusted and lubricated seas-l HP-204 6/13/77 Drain valve leakag;e by seat Replaced drain valve
~~> NOTE: HP - High Pressure Injection Svste:L i~~
PR - RB Purge System - 5.1-1
I ' DUKE PDVER C O M P41 Y --- - - - - - - - - - - - - -- -- - - O
~~E ONE! Ntit.E AR 51&T104 .~~
i INTEGRL TE D- LE AT R ATE-TES T 0 7/ 31/7 7 ----- m 29.5-P5f G- VI RI FI C A TION T E S T- -- -- - T RE N D R E P0t i ,6- F 4 01-I NI T-[A1;Ct)-/-415-6 i LA 1 A-5E T IIM E PRE 55URE TEMPER ATUR E ABS LP ABS La 3 4 C DEF 10* Nf E- L I 41 T S j (P51AI (FI M E A SUR ED C A LC UL A TE0
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e ..s I 4 44 1110 n 3:15A 84.5677 0.15995 , 3.157a7 0.15395 0.19995
, 5 *t3CA 44.1090 84.5766 0.1944' O.19620 0;18446 0.14444 6-- 3:45A- '44.1070 84.5743 ,0.22741 0.2n19 0 22741 0.72741 9
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0.34105 a.31662 0.39672 g *., .A 3 9 .4t3CA 44.1000. -_84.5653 _ 3.33616 0.35230 0.24330 0.11202.
.., I C. _ 4845A 44.0990. _84.5672 0.33032, 0.35606 0.21a00 0.19265 . =
11 5:00A 44.0970 84.5648 '0.31005 0.360as 0.22201 0.39ane
! 9 12 _ 581!A .44.0960 84.5716 . $ .312 7 7 0.36425 0.21341- .3.69612 13 S r 3'C A 44.0940 __ . 84.5710_ 3.32245.. _.0.36883 .D.23346 0.41444..
14 5t45A 44.0930 84.5645 0.27975 0.35991 . G.17110_ ' O. inn 19 8 15 1 6t00A 44.C920 .84.5740 0.28745 0.35449 .3 17462.. ...... 0.6no77 . . _ . , . . . . j 16 6 t15 A __ 44.C9 CO ' 84.5744 0 29808 0.35240. D.13620 .. D .49 9 9 7 . 11 6 830 A 44.C89n' 84.574n n.29276 :L.34917 n.15167 0.4034s
J _ 18 6

45A 44.C878 84.5686 0.29 5 8 L 3.34693 0.1S540 _n.40422 i if f
11 7t0CA '44.C86e a4.5710 n.26400 n .33 634 S.1493Q 0.37870 3 2n 44.034n m 267*8 58-15A _'_ a 4.Q 28 3 a3117e alins n.13m i > 21 1T.3 C A 4.4.0830 84.5725. 0.24345 0.32154 0.13021 0.1667a. CA a . l0 Table 4.2-2 . ! -) . o.
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ML19312C807

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ML19312C807

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