ML20081C031
| ML20081C031 | |
| Person / Time | |
|---|---|
| Site: | Ginna, 05000000 |
| Issue date: | 02/28/1976 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML20081B916 | List: |
| References | |
| FOIA-83-296 NUDOCS 8310310076 | |
| Download: ML20081C031 (120) | |
Text
{{#Wiki_filter:. _ _. _ _ _ . . _ _ _ . _ _ _ _ . _- . . ._ _ _____ . J ' R. E. GINNA NUCLEAR POWER STATION 1: . 1
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REACTOR CONTAINMENT BUILDING ,
- c. . n.. . - 1 l
l INTEGRATED LEAK RATE TEST - , a n ! l
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FEBRUARY,1976 - I Cf L -
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_-[. ROCHESTER GAS & ELECTRIC 4 l CORPORATION
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c0 8310310076 830706 PDR FOIA l CONNERB3-296 PDR !
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; ,_ TABLE OF CONTENTS
( SECTION PAGE l
1.0 INTRODUCTION
1 2.0 ACCEPTANCE CRITERIA AND CONCLUSIONS 2 3.0 TEST INSTRUMENTATION 4 3.1 CALIBRATIONS 5 3.1.1 Texas Instrument Precision Gauges 5 3.1.2 Foxboro Dewpoint Recorder 6 3.1.3 Leeds and Northrup Temperature Recorder 7 3.1.4 Supplemental Test Rotometer 9 3.2 ERROR ANALYSIS 11 4.0 LEAKAGE TEST 16 4.1 PREREQUISITES 16 4.1.1 Administration 16 4.1.2 Test System 17 4.1.3 Containment Isolation 17 4.1.4 Containment Penetration Line
/' Venting 17
(_O 4.1.4.1 4.1.4.2 Venting Outside Containment Venting Inside Containment 17 17 4.1.4.3 Isolation Valves 18 4.1.5 Containment Inspection 19 4.2 TEST PERFORMANCE 20 4.3 TEST RESULTS 24 4.3.1 Calculated Leakage Rates 24 4.3.2 Preferred Calculational Methods 24 4.3.3 Best Estimate Leakages 25 5.0 SUPPLEMENTAL TEST 27 5.1 TEST PERFORMANCE 27 5.2 TEST RESULTS 27 6.O TYPE "B" AND "C" LEAKAGE RATE HISTORIES 29 6.1 DISCUSSION OF LEAKAGE HISTORY 29 6.2 PENETRATION MODIFICATION DURING REPORTING PERIOD 33 Lv
-i- 4
1.0 INTRODUCTION
s s
) A reactor containment building integrated leak rate test !
%J was performed in February 1976 on the Ginna Station reactor containment building. The objective of the test was to confirm that leakage from the containment did not exceed 75 percent of the maximum allowable leakage rate based on the limiting conditions of 0.2 percent per day mass leakage at 60 psig or 0.153 percent per day at 35 psig. A leak-age margin of 25 percent from allowable is maintained to allow for nominal deterioration over a three-year period.
An initial attempt to meet the integrated leak rate test acceptance criteria failed and is the subject of a separate
, s s accompanying report. This report discusses the success-
\
ful integrated leak rate test performed after repairs were made to leakage paths identified during the failed test. A successful reduced pressure test was performed at 35 psig and the leakage rate was determined to be 0.044 percent per day. A supplemental test was performed by imposing a known leakage rate on the reactor containment building. The re-sults of this test established a containment leakage rate i which was within 0.25 Lt of the leakage determined in the 24 hour test and thereby confirmed the validity of that test. The testing was performed by Rochester Gas and Electric [ ) Operations Department and all data have been reduced by l' RG&E personnel.
- 2.0 ACCEPTANCE CRITERIA AND CONCLUSIONS j i
The maximum allowable leakage rate (La) at 60 psig (Pa) has been established in the Technical Specifications as 0.2 percent per day. The maximum allowable leakage rate for reduced pressure testing (Lt) at presscre Pt is deter-l/2 minedbytherelationLt=(Pt\j p (La), and is 0.153 percent per day. This is in accordance with the present Ginna Station Technical Specifications and a Request for Exemption from Appendix J to 10 CFR Part 50 which was submitted to the Nuclear Regulatory Commission in October 1975. 1 l l The acceptance' criterion for the Type A test was set at 75 '
~
percent of the maximum allowable leakage, or 0.115 percent per day for testing at 35 psig. The acceptance criterion for the supplemental test was established as a containment leakage rate which differed from the Type A test leakage rate by no more than 0.25 Lt, or 0.038 percent per day for testing at 35 psig. These acceptance values are summarized below: parameter value Pa GO psig Pt 235 psig La 0.2%/ day Lt 0.153%/ day 0.75 Lt 0.115%/ day supplemental test variance 0.038%/ day The leakage rate for the reactor ccntainment b0ilding was [Os determined by two methods. The leakage determined by
l applying a linear least squares fit to the weight of con-tainment air at each hour of the test (Mass Plot Method)
'N was 0.044 percent per day. The leakage determined by ap-plying the Total Time Method of ANSI N45.4-1972 was 0.038 percent per day.
Leakage rates for the supplemental test were also calcu-lated by the same two methods. The leakage rates for the containment building, including the imposed known leakage, were 0.089 and 0.050 percent per day respectively for the ANSI N45.4 Total Time Method and the Mass Plot Method. The imposed, known leakage rate was 0.037 percent per day. Thus the calculated reactor containment building leakage rates using the Total Time and Mass Plot Methods were 0.052
~~ and 0.013 percent per day respectively. Both of these val- '- \_ / ues differ by less than .25 Lt from the 24 hour test values.
Therefore it is concluded that (1) leakage through the primary reactor containment and systems and components penetrating the primary containment does not exceed the allowable leakage rate specified in the Technical Speci-fications and (2) the accuracy of the Type A test has been verified by an acceptable supplemental test.
l
3.0 TEST INSTRUMENTATION
- l. The test instrumentation utilized for the leakage portion of this test was as follows:
I( Os i 2 Texas Instrument Precision Pressure Gauges
; Model 144 Range 0-100.000 psia Accuracy 1 0.015 percent of reading 1 Foxboro Dewpoint Recorder Model 9435 TM Range 2'to 142*F ,
Accuracy 1 0.35'F over range Sensors 6 1 Leeds and Northrup Temperature Recorder e Model G Range 40'to 130*F i
- Accuracy 1 0.l*F over range Sensors' 2.3 1 Wallace and Tiernan Flow Meter Model 1856 Range O to 7 scfm at 1100F i Accuracy i 1% of full scale 4
The sensor locations were the same as those used in pre-vious tests. These locations are shown on Figure 1 of the Preoperational Integrated Leak Rate Test of the Reactor Con-tainment Building Report, dated November 7, 1969. l (' No test instrumentation failed or was recalibrated through-out the course of this test.
j 3.1 CALIBRATIONS 3.1.1 l
! Texas Instrument Precision Pressure Gauges f- s ss Only two means exist to calibrate these precision gauges:
- a. Maintain a 3 micron vacuum in the quartz capsule and thereby zero the instrument.
- b. Relate the instrument readings to the local atmospheric which is simultaneously detected by a mercurial baro-meter.
1 It was decided that if each instrument's barometric ratio (Texan Instrument Reading (counts) divided by mercurial barometer (mmHg)) approached that determined in past tests, , the instruments would still be in calibration as shown below. 1969 T* 1972 1976
- s Instrument Ratio Ratio Ratio PI-3A 0.01964 0.01977 0.01927 PI-3B 0.01904 0.01899 0.01921 It is the repeatability of these instruments, however, which is important in determining leakage rates because the leakages are dependent essentially upon differences in pressure. Repeatability is demonstrated by comparing the differences between the pressure gauge readings during the test as shown below, where PI-3A and PI-3B are the two Texas Instrument Precision Gauges.
Time of Test (Hrs) PI-3A minus PI-3B (psi) 0 0.0296 'I C4 1 2 0.0303 0.0292 3 0.0291
r Time of Test (Hrs) PI-3A minus PI-3B (psi) 4 0.0301 5 0.0291 6 0.0291 7 0.0292 8 0.0292 9 0.0292 10 0.0290 11 0.0294 12 0.0285 13 0.0285 14 0.0286 15 0.0282 16 0.0282 17 0.0272 18 0.0272 19 0.0271 20 0.0270 21 0.0279 22 0.0277 23 0.0277 24 0.0318 0.7175 The average of the differences is 0.0287 psi and the re-peatability, or largest variation from the average, is 0.0031 psi. 3.1.2 Foxboro Dewpoint Recorder Each sensor was saturated with the chloride solution as required for proper operation. The Dewcel's RTD-Ohmic value for a specific Dewcel temperature was measured at the recorder. The same ohmic I value was introduced into the recorder and the recorder's observed value was then subtracted from the actual value, as noted below, to yield A: a 1
BEFORE TEST Dewcel Actual Observed Dewcel Temp. Dewpoint, OF Dewpoint, OF A 0F l 1 85.68 F 25.480F 260F .52 2 85.24 0 F 25.170F 25.50F .33 3 87.030F 26.320F 26.20F +.12 4 84.45 0 F 24.620F 24.9 0 F .28 5 84.17 0 F 24.420 F 24.70F .28 6 84.56 0 F 24.680 F 250 F .31
-1.60 ,
AFTER TEST 1 101.140F 36.40F 37.lOF .70 . 2 102.940F 37.76 0 F 38.1 0 F .34 3 1050 F 39.40 F 39.90F .50
4 103.380F 38.lOF 38.40 F .30 , \
5 103.050F 37.830 F 38.00F .17 6 104.08 0 F 38.66 0 F 39.2 0 F .54
-G The average variation between the before and after cali-brations is 0.16 0F.
3.1.3 Leeds and Northrup Temperature Recorder The ohmic value for each RTD was determined by measuring the lead resistances. With the use of the RTD's ohms ver-sus temperature curve, the actual temperature was noted. The RTD-ohmic value was then introduced into the recorder, and the recorder's observed value was then subtracted from the actual temperature (curve value) as noted below:
i i Thermohm Before Test a O F Af ter Test A 'F l 1 -
.380F -
.620F 2 -
.230 F -
.370 F
- 3 -
.070F -
. 210F 5
4 + .37 F 0
+ .21 F i 5 - .010 F -
.130F 6 -
.250F -
.690 F 7 + .010 F -
.37 F 8 -
.10 F -
.30 F 9 -
.170 F -
.330 F 4
10 -
.01 F -
.420 F 11 -
.0150 F -
.290 F 1
12 -
.030 F - .33 0 F
13 .060F -
.19 F 14 -
.10 F -
.20 F j 15 -
.180F -
.30 F
- 16 -
.020 F -
.3 F j 17 .050 F + .47 F 18 -
.07U F -
.30 F 19 .10 F -
.520F
] I 20 -
.240 F .18 F 21 -
.14 F -
.290F 22 .220 F + .040 F 23 0.000 F -
.470 F 24 -
.140 F +1.87 0 F
- SUM -2.205 -4.22 l
}
4
i The average observed difference in calibration is 0.084 0F. I In addition to the above measurements, a calibrated ther- , mometer was placed inside the containment at RTD location
; 17. The thermometer reading was 44.350 F when the recorder printout was 44.50F, thus verifying the RTD calibration curve.
- 3.1.4 Supplemental Test Rotometer A two point calibration was made against a Rockwell 175 dry gas test meter. The rotometer was tested as de- '
scribed in ASTM D3195 " Standard Recommended Practices for Rotometer Calibration". 7_ Comparison results are as follows:
\- Rockwell Test Meter Wallace & Tiernan Rotometer 2.12 SCFM 2.075 SCFM (30% scale) 1.053 SCFM 1.037 SCFM (15% scale)
The rotometer range at calibration condition (78.2*F) is i 6.914 SCFM. The flow setting of the rotometer was established at 12 percent of scale for the supplemental test. Using the spot calibration data at 15 percent of scale indicates
~
that the rotometer accuracy is + 0.23 percent of full scale. The Rockwell rotometer was calibrated by the RG&E gas meter shop using a bell prover procedure which is , l
)
Y 1
- l j
I approved by the Public Service Commission of the State of New York. The meter factor for this meter was deter-mined to be 1.001. 4 2 l I i l J r ! I t i l 1 1 l 3 l l 1 I I 1 i t : 1 ! (. _ _ _ _ . . . _ . .._.
l 3.2 ERROR ANALYSIS The leakage rate in percent per day based on an interval of measurement of 24 hours duration is:
~ ~
P24Tn L = 100 1P Tg Percent per day where: Pg = partial pressure of air at start P24 = partial pressure of air at finish T o = R. Building mean ambient temperature at start, *R T24 = R. Building mean ambient temperature at finish, 'R The change or uncertainty interval in L due to uncertainties r- - in the measured variables is given by: g , 7 BL Tp ),I BLTpp+I BL T T f 3L T T T" 2 1/2 3P q + [oP 24 4 3 o S ,3T 0 ,3T24 2k. . where T is the standard error for each variable. The error in L after differentiation is: o "P 2 (Pg T o *P 24 *T P 24 To e T 1/2
*L = 100 Pg T 2 T T2 24 ) (p O 24 ) [o 24) (P )
where: ep = 1p
*T 24 24 "Tg "T Tg 24 24
,_ _. _ - _ _ _ _. - - . - _ _ - _ _ - . - - _ _ _ ._ ~_- ._ _ _ . - _ . - _ . .
The error assigned to pressure determinations results from N two sources, the precision pressure instruments and the Dewcels, and may be expressed as ep = e pg +e pd where: e gg = error induced by the two precision gauges e pd = err r induced by the six Dewcels Using the manufacturer's specifications for accuracy of 0.015% of reading the error induced by two precision pres-sure gauges is ( 00015) (49. 350 psi) = 10.0052 psi W Because the repeatability, defined in Section 3.1.1, is 0.0031 psi and is less than the value found using the man-ufacturer's standard, e pg will be assumed to be 0.0052 psi. Using manufacturer's specifications, the error induced by the Dewcels is 10.350 F = 10.14 F V The average variation between the before and after test calibrations was shown to be 0.160 F in Section 3.1.2 and thus the latter value will be used. From steam tables the pressure equivalent to 0.16 F at a dewpoint of 33 F, epd' is 0.0058 psi. Therefore, the induced pressure error e is a p U - l e = Y(0.0052 psi) 2 + (0.0058 psi) 2 P , e = 0.0078 psi P Because both Po and P24 are less than the value used to calculate epg, the value found for ep is a conservative representation for both e p and e p . Thus ep =e p =e p = 0.0078 psi. The error in temperature determination, e , resulting from T 24 sensors with a manufacturer's specification of 0.1 F accuracy, is 0.1 F = 10.0204 YE r l\ However, as noted in Section 3.1.3, the average observed difference in calibration before and after the test was 0.0840 F and thus this value will be conservatively as-sumed for e and will be used for e and e
- T Tg Substitution of e , e and the observed values of the T p pressures and temperatures into equation (1) yields:
e = 0.032%/ day 1
r" i i The error involved in the determination of the weight of air inside the reactor building at any point in tiiIe may' be derived from: W= .- where: ' W = weight of air inside reactor building, LB M' K = constant = V/R = 2.6182 x 106 in2 oR P = partial pressure of air inside reactor building, psia T = mean reactor building temperature, R
~ ' '
The error in W based on the second law of propagation is a
=~f 12" 1/2 p72 +
e = BW T (BW T W
]Tp j yy Tj
+f-KP eTi 2" 1/2 2
= xe .
,i T p/ T 2 Using representative values of 49.35 psia for'P and 508.5 R for T and the values of e T and e p used previously yields:
eg= 58.1 lb The weight error e g is shown as an error band around the least squares fit of the weights on Figure 1-at the end of Section 4.3. The consistency of the data and the conser-vatism of the error calculations are apparent from the graph. Only the last data point falls outside the error band. The discrepancy between this point and the' rest of the data set resulted from a relatively large change in the
-14 '-
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,E service water flow to the containment coolers as explained i in Section 4.2.
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4.0 LEAKAGE TEST 4.1 PREREQUISITES O*N 4.1.1 Administration All actions were performed in accordance with a written procedure approved by the Plant Operations Review Commit-tee. No leak repairs were performed prior to the start of the first test. Repairs were made after the first test failed to meet the acceptance criteria. For both tests the plant was ir. cold shutdown status with the pressurizer level being monitored. All required valves and breaker alignments were properly tagged. The Reactor Coolant Sys-tem and reactor building radioactivity levels had been re . duced as low as practicable prior to testing to permit e N depressurization of the building contents directly to the
'-s atmosphere. The pressurized reactor building atmosphere
- was sampled and recorded prior to release.
1 Each accumulator was vented to the building atmosphere. Equipment was protected from the external differential pressure as required. 1 1 Automatic temperature control was defeated to permit manual building temperature control. The pressurizer was vented to the reactor building atmo- l i sphere. 1~ ,)
4.1.2 Test System i All test instrumentation was calibrated before pressuriza- l tion as stated in Section 3.1. The compressors were checked l
) to ensure no oil carryover was taking place and the test
%d pressure sensing lines were leak tested.
4.1.3 Containment Isolation All automatic containment isolation valves were closed by a manual containment isolation signal without any pre-liminary exercising or adjustment. All lines penetrating the containment were vented in ..cordance with the princi-ples stated in Section 4.1.4, existing Technical Specifica-tions, a Request for Exemption submitted to the Nuclear Regulatory Commission in October, 1975, and a technical specification change proposed to the Nuclear Regulatory Commission in January, 1976.
~ :y) 4.1.4 Containment Penetration Line Venting 4.1.4.1 Venting Outside Containment Lines which penetrate the containment and which are open to the containment atmosphere as in 4.1.4.2, were vented to the atmosphere outside of the containment. Where piping configurations outside containment exist such that the fluid in fluid carrying lines did not drain to expose the isolation valves to the atmosphere by opening existing vents and drains, the fluid was left in the lines.
4.1.4.2 Venting Inside Containment Portions of the fluid systems that are part of the reactor coolant pressure boundary, are open directly to the s_ s containment atmosphere under post-accident conditions, and
)
l l become an extention of the boundary of the containment ( were opened or vented to the containment atmosphere prior to and during the test. Portions of closed systems inside containment that penetrate containment and that also pass in-side the crimary shield wall near the postulated broken leg, and which are postulated to rupture as a result of a loss of coolant accident consistent with the containment integrity analysis of Section 14.3.4 of the FSAR, were vented to con-tainment atmosphere. Where check valves or piping con-figurations exist between the primary shield wall and the containment penetration or in places where damage to the
~
piping system.is not postulated to occur as a result of a LOCA such that fluid seals are formed as a result of normal r{'- operation and containment isolation, the fluid was left undisturbed. That is, those portions of systems not postulated to rup-ture as the result of a LOCA were not drained unless they drained unaided to the postulated breaks in the systems. 4.1.4.3 Isolation Valves Where two isolation valves exist in a single line which are'either check valves, or valves capable of automatic closure, or a combination thereof, no attempt was made to vent to atmosphere from a point between the valves. b
< v
4.1.5 Containment Inspection A visual examination of the accessible interior and ex-c terior surfaces of the containment structure was performed prior to testing to uncover any evidence of structural de-terioration which would affect either the containment struc-ture integrity or leak-tightness. There was no evidence of structural deterioration; however, the containment building did exhibit separation of numerous stainless steel liner sections which cover the liner insulation. The insulation cover serves only to protect the insulation and is not the leaktight containment boundary. - 9
,t t N l
l ____..__..,.__.--,J
l 4.2 TEST PERFORMANCE [ h On February 7, 1976 preparation began for performance of the Integrated Leak Rate Test. The building was pres-surized at a rate of approximately 5 psi / hour. At 14 psig, the building was entered for a second inspection. Representatives from the instrument and control, electri-trical, pipe, and machine shops and from Operations and Test and Results inspected their respective equipment and the general containment appearance but found nothing ab-normal. Pressurization continued until the building pressure was slightly greater than 35 psig. s During pressurization and the subsequent stabilization period, test personnel periodically monitored the leak tightness of isolation valves as could be witnessed by air or water exiting from line vents in the adjacent buildings. In addition, certain closed volumes formed between the in-terior and exterior containment boundaries were montiored for pressure buildup. The following volumes were moni-tored. (a) mechanical penetrations supplied by the various air i manifolds (b) electrical penetrations supplied by the various nitrogen manifolds (c) purge exhaust volume
(d) purge supply volume (e) personnel hatch personnel access lock (f) equipment hatch personnel access lock ( ) Pressure gauge readings were taken prior to the start of pressurization and subsequently at two hour intervals. Test personnel also periodically checked vented lines to identify any isolation boundaries which were not tightly sealed. Leakage was identified from the purge exhaust volume; check valve 1713, nitrogen supply to the Reactor Coolant Drain Tank; and motor-operated valves 813 and 814, inlet and outlet lines for reactor support cooling. Ap-proximately seven hours after reaching 35 psig it was de-termined that leakage from the containment was in excess of the acceptance criterion and the test was terminated. ( V[ } Test personnel then inspected all open line vents and moni-tored all penetrations for additional areas of leakage. Nc further leakage paths were- discovered, other than those given above. This aborted test is discussed in more detail in a separate accompanying report. Repairs were satisfactorily completed on February 10, 1976. The leakage from check valve #1713 was reduced to zero and leakage from the pur<,e exhaust volume was re-duced to 447 cc/ min (60 psig conditions). , b) .g i L/ Preparations were then begun to repressurize the Reactor Containment Building. 35 psig was reached by 1300 hours
I on February 10. The building contents were stabilized for approximately six hours before the zero hour data ( } readings were collected for the type A retest. Every hour, readings were taken from the two precision pres-sure gauges, 6 Dewcels and 24 temperature sensors. Con-tainment temperature was controlled by regulating service water flow to the containment recirculation units. The test proceeded uneventfully until the last few hours of the 24 hour period when difficulty was experienced with 4 the service water flow to the containment recirculation units. Each of the four units has an individual throttle valve in its discharge line. Flow from all four r[~Ng units then combines in an 8-inch line and passes through a
' 4j butterfly valve normally controlled by containment temp-
, erature. For the purpose of this test the 8-inch valve was positioned with the output from a hand-controlled air regulator. The air regulator was apparently faulty and its output pressure slowly drifted up until it closed the 8-inch valve far enough to cause it to assume flow control of the cooling water. The 8-inch valve was apparently interfering with the flow rate during the last 3 to 4 hours, but the interference was not recognized by test personnel. The condition became apparent when an attempt was made near the end of the test to increase the service water flow by
.O a
_ _ , - . . . _ . _ __ _, _ _ ~
opening the 4 recirculation unit discharge valves. No i l flow increase was possible and it was then that the nearly closed 8-inch butterfly valve was discovered. In the pro-i cess of repositioning the 5 valves, flow to recirculation units increased from a nominal 160 gpm each to 500 to 700 4 gpm each. The resulting instability in -he containment atmosphere caused the last hour of data to be of ques-tionable accuracy. Plots of mean containment temperature and pressure are shown on Figure 2. h f^ i i i i d I 1 i j ( i l l
4.3 TEST RESULTS 4.3.1 Calculated Leakage Rates During the integrated leak rate test the leakage rate
'h was calculated after each hour's data using the Total Time N
Method of ANSI N45.4 - 1972. The leakage rates found using this method are shown on the data table at the end of this section. In addition, the leakage rate was calculated periodically by determining the linear least squares fit of the containment air weights which were calculated each hour. Using the Total Time Method, the average (least squares, 1 non time dependent) leakage rate for the 24 hour period i of the test is 0.038 percent per day. A~ linear least squares fit t'o'the containment weights (Mass Plot Method) r for the same period yields a leakage rate of 0.044 percent per day. Both of these values are less than the accept-ance criterion of 0.115 percent per day. Plots of mean containment temperature and pressure and con-tainment air weight are given at the end of this section on Figures 1 and 2. Data taken during the test and hourly a Total Time Method leakage rates are given on Table 4.1. 4.3.2 Preferred Calculational Methods The Mass Plot Method is preferable to the Total Time , Method because it eliminates the dependence upon a single data point, the first point, which is used for each leak-age rate calculation in the Total Time Method. Deviation . of this single set of data from the true ambient condi-tions can cause all of the subsequent leakage rates to vary from the unbiased estimate as determined by the Mass 4 t
- - - . , . , . - ,.,,.,.,-.,.e,.. , . - . - - - , -
Plot Method. The latter method places equal dependence j (A upon all the data points. For this particular test it is seen on Figure 1 that the time 0 weight calculated from the data is slightly less than the weight intercept value found using the Mass Plot Method. Because the temperatures and pressures for time zero yield a weight less than the inter-cept, all subsequent Total Time Method leakages will be less than the unbiased estimate. Changing only this one set of data to yield a time zero weight greater than the intercept will significantly affect the Total Time Method leakage rates even though all of the other data remain unchanged. ,_ Thus, Rochester Gas and Electric accepts 0.044 percent per \- day as the 24 hour test reactor containment building leak-age rate even though our commitment to calculate leakage rates using ANSI N45.4 - 1972 methods yields a smaller value. 4.3.3 Best Estimate Leakages The leakage rates given in Section 4.3.2 are based upon 24 hours of data because of our present technical specification requirement to extend the length of the test to that time. However, because of the difficulty with service water flow to the containment recirculation units and the varying con-tainment temperature, the last set of data is believed
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i to be invalid. Omitting this data yields least square j leakage rates of 0.036 percent per day for both the Mass Plot and Total Time methods. The fact that these values are the same is coincidence. : l 1 l l l l l l l t I t , I s I
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% FIGURE 2 24 HOUR INTEGRATED LEAK ItATE TEST i'
REACTOR BUILDING MEAN TEMPERATURE AND PRESSURE VS TIME
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TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS [d TIME /IIOUR PI-3A, Counts 1900/0 50.5810 2000/1 50.5910 2100/2 50.5930 2200/3 50.5980 PI-3B, Counts 48.9870 48.9960 48.9990 49.0040 Avg. Press., Psia (PT) 49.4419 49.4513 49.4538 49.4588 Dewpoint Temp., *F 1 31.0 30.5 30.5 30.5 2 34.5 34.5 34.0 34.0 3 35.0 35.0 35.0 34.5 4 34.0 34.0 34.0 33.5 5 34.0 34.0 34.0 34.0 6 34.0 34.0 34.0 33.5 Avg. Dewpoint Temp., 33.6 33.5 33.4 33.2
*F, (Note 1)
Water Vapor Press., Psia 0.0945 0.0941 0.0937' O.0930 Drybulb Temp., *F 1 48.4 48.5 48.5 48.5 2 48.4 48.6 48.6 48.7 3 49.9 50.1 50.1 50.1
- c. 4 50.4 49.7 49.8 49.8 5 47.4 47.5 47.5 47.6
\- 6 48.0 48.0 48.3 48.2 7 48.4 48.4 48.5 48.5 8 48.4 48.4 48.5 48.5 9 49.0 49.0 49.2 49.2 10 49.5 49.6 49.6 49.7 11 48.4 48.5 48.5 48.5 12 48.5 48.7 48.6 48.7 13 48.6 48.8 48.9 48.7 14 48.1 48.3 48.2 48.3 15 48.6 48.7 48.8 48.9 16 47.9 47.9 48.0 48.0 17 48.1 48.4 48.3 48.4 18 48.0 48.1 48.2 48.2 19 48.0 48.1 48.2 48.1 20 48.0 48.0 48.1 48.1 21 48.4 48.4 48.4 48.6 22 47.9 48.2 48.2 48.2 23 47.6 47.7 47.8 48.3 24 47.9 48.1 48.2 48.3 Avg. Temp., *R (T) (Note 2) 508.0 508.1 508.2 508.2 Weight of Air, LES. 254,333 254,334 254,299 254,328 Leakage Rate, Total N/A -0.004 +0.164 +0.017 Time Method, %/ day Sheet 1
TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS
-TIME / HOUR 2300/4 2400/5 0100/6 0200/7 PI-3A, Counts 50.5990 50.5990 50.5980 50.5960 PI-3B, Counts 49.0040 49.0050 49.0040 49.0020 Avg. Press., Psia (PT) 49.4593 49.4598 49.4588 49.4567 Dewpoint Temp., *F 1 30.5 30.5 30.2 30.5 2 34.0 33.5 33.5 33.5 3 34.5 34.5 34.5 34.5 4 33.5 33.0 34.0 33.5 5 33.5 33.5 33.5 33.5 6 34.0 33.5 33.0 33.5 Avg. Dewpoint Temp., 33.2 32.9 32.9 33.0
'F, (Note 1)
Water Vapor Press., Psia 0.0930 0.0919 0.0919 0.0922 Drybulb Temp., 'F 1 48.6 48.6 48.6 48.5 2- 48.8 48.7 48.8 48.6 3 50.1 50.1 50.2 50.0 4 49.7
/ g 5 49.8 49.8 49.6 47.5 47.5 47.5 47.4 y ,) 6 48.2 48.2 48.3 48.2 7 48.6 48.6 8
48.6 48.6 48.6 48.6 48.7 48.6 9 49.2 49.2 49.3 49.2 10 49.7 49.7 11 49.7 49.8 48.5 48.6 48.6 48.6 12 48.8 48.7 48.8 48.3 13 48.9 48.8 49.0 48.9 14 48.4 48.4 48.4 48.4 15 48.9 48.9 49.0 49.0 16 48.1 48.1 48.1 48.2 17 48.4 48.4 48.4 48.5 18 48.3 19 48.3 48.3 48.4 48.1 48.2 48.2 48.4 20 48.1 48.2 48.2 48.2 21 48.6 48.6 48.6 48.6 22 48.2 48.3 48.3 48.2 1 23 47.8 47.8 47.9 47.9 24 48.3 48.3 48.4 48.3 Avg. Temp., 'R (T) (Note 2) 508.2 508.2 508.3 508.2 Weight of Air, LBS. 254,331 254,339 254,284 254,321 i [ )LeakageRate, Total +0.006 -0.010 +0.078 +0.016 I k / Tine :lethod, 7,/ day Sheet 2
\
TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS TIIIE/IlOUR 0300/8 0400/9 0500/10 0600/11 PI-3A, Counts 50.5960 50.5960 50.5870 50.5880 PI-3B, Counts 49.0020 48.9920 48.9930 48.9940 Avg. Press., Psia (PT) 49.4567 49.4517 49.4478 49.4488 Dewpoint Temp., *F 1 30.5 30.5 30.5 30.0 2 33.5 33.5 33.0 32.5 3 34.0 34.0 34.0 33.5 4 33.0 33.0 32.5 32.0 5 33.0 33.5 32.5 32.5 6 33.5 33.0 33.0 33.0 Avg. Dewpoint Temp., 32.7 32.7 32.4 32.1
'F, (Note 1)
Water vapor Press., Psia 0.0911 0.0911 0.0900 0.0889 Drybulb Temp., *F 1 48.6 49.0 48.9 48.9 2 48.8 49.0 49.0 49.0 3 50.1 49.8 49.9 49.9 4 49.7 49.4 49.4 49.5 7- 5 47.5 47.4 47.3 47.4 6 48.3 48.5 48.5 48.5 7 48.7 48.7 48.8 48.8 8 48.7 48.6 48.6 48.6 9 49.4 49.2 49.2 49.1 10 49.8 49.5 49.6 49.6 11 48.6 48.6 48.6 48.6 12 48.8 48.8 48.8 48.9 13 48.9 48.7 48.8 48.8 14 48.4 48.3 48.4 48.4 15 49.0 48.9 48.9 48.9 i 16 48.2 48.1 48.2 48,1 17 48.5 48.4 48.4 48.5 l 18 48.4 48.3 48.3 48.3 19 48.3 48.3 48.3 48.3 20 48.3 48.1 48.2 48.2 21 48.6 48.5 48.6 48.6 22 48.3 48.2 48.3 48.2 23 47.9 47.9 47.9 47.9 24 48.4 48.3 48.3 48.3 Avg. Temp., R (T) (Note 2) 508.3 508.2 508.2 508.2 Height of Air, LBS. 254,277 254,301 254,287 254,298 Leakage Rate, Total +0.066 +0.034 +0.044 +0.031 (" Tine 1:ethod, t/ day Sheet 3
TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AUD RESULTS TIME / HOUR 0700/12 0800/13 0900/14 1000/15 PI-3A, Counts 50.5860 50.5850 50.5820 50.5910 PI-3B, Counts 48.9930 48.9920 48.9890 48.9980 Avg. Press., Psia (PT) 49.4474 49.4464 49.4434 49.4523 Dewpoint Temp., *F 1 30.0 30.0 30.0 30.1 2 33.0 33.5 33.0 31.0 3 33.5 34.0 34.0 34.0 4 33.0 33.0 33.0 32.1 5 33.0 33.0 33.0 32.1 6 33.0 33.0 33.0 32.1 Avg. Dewpoint Temp., 32.4 32.6 32.5 31.7
*F, (Note 1)
Water Vapor Press., Psia 0.0900 0.0908 0.0904 0.0873 Drybulb Temp., *F 1 48.8 48.9 48.8 48.9 2 48.9 48.9 48.8 49.0 3 49.9 49.9 49.8 49.8 4 49.3 49.3 49.3 49.5 5 47.3 47.3 47.2 47.3
N 6 48.4 48.5 48.4 48.5
-( 7 8
48.8 48.6 48.7 48.6 48.7 48.5 48.7 48.6 9 49.1 49.1 49.0 49.3 10 49.6 49.6 49.5 49.0 11 48.6 48.6 48.5 48.6 12 48.8 48.8 48.7 48.8 13 48.8 48.8 48.6 48.8 14 48.4 48.4 48.3 48.4 15 48.9 48.4 48.7 48.8 16 48.1 48.1 48.0 48.2 17 48.5 48.4 48.4 48.5 18 48.4 48.4 48.3 48.3 19 48.2 48.3 48.2 48.3 20 48.2 48.2 48.2 48.1 21 48.6 48.6 48.5 48.6 22 48.2 48.3 48.2 48.4 23 47.9 47.9 47.8 48.0 24 48.3 48.3 48.3 48.5 Avg. Temp., *R (T) (Note 2) 508.2 503.2 508.1 508.2 Weight of Air, LDS. 254,285 254,276 254,312 254,324
' ~
Leakage Date, Total +0.038 +0.042 +0.014 +0.006 l ime tiethod, %/ day 1 %) Sheet 4
TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS TICE/ HOUR 1100/16 1200/17 1300/18 1400/19 PI-3A, Counts 50.5930 50.5910 50.5920 50.5960 PI-3B, Counts 49.0000 48.9990 49.0000 49.0040 Avg. Press., Psia (PT) 49.4543 49.4528 49.4538 49.4578 Dewpoint Temp., 'F 1 30.0 30.0 30.5 30.0 2 33.0 33.0 33.5 33.0 3 33.1 34.0 34.0 34.0 4 33.0 33.0 33.0 33.0 5 33.0 33.0 33.0 32.0 6 33.0 33.0 33.0 33.0 Avg. Dewpoint Temp., 32.3 32.5 32.7 32.3 , *F, (Note 1) Water Vapor Press., Psia 0.0897 0.0904 0.0911 0.0897 . Drybulb Temp., *F 1. 48.8 49.2 49.0 49.2 2 49.0 49.2 49.1 49.3 3 50.0 49.7 49.7 49.7 4 49.5 49.4 49.4 49.4
-)
- \- \ )
5 6 47.3 48.5 47.4 48.6 47.4 48.6 47.4 47.6 7 48.7 49.0 48.8 48.8 8 48.6 48.7 48.6 48.6 9 49.3 49.0 49.1 49.0 10 49.6 49.5 49.5 49.5 11 48.7 48.5 43.6 48.6 12 48.8 49.0 48.8 49.0 13 48.8 48.7 48.8 48.8 14 48.4 48.4 48.4 48.5 15 49.0 49.0 48.8 49.0 16 48.2 48.4 48.3 48.3 17 48.6 48.5 48.5 48.5 18 48.5 48.5 48.5 48.5 19 48.4 48.4 48.4 48.4 20 48.3 48.3 48.3 48.6 21 48.7 48.6 48.5 48.5 22 48.4 48.4 48.4 48.4 23 48.0 48.0 47.9 48.0 24 48.3 48.4 48.4 48.4 Avg. Temp., *R (T) (Note 2) 508.3 500.3 508.3 508.3 Weight of Air, LDS. 254,272 254,261 254,262 254,290 r ' Leakage Rate, Total +0.036 +0.040 +0.037 +0.022 1 x Time Method, %/ day Sheet 5
\
TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS TIME / HOUR 1500/20 1600/21 1700/22 1800/23 PI-3A, Cocnts 50.6020 50.6060 50.6090 50.6100 PI-3B, Counts 49.0100 49.0130 49.0160 49.0170 Avg. Press., Psia (P ) 49.4637 49.4672 49.4702 49.4712 T Dewpoint Temp., *F 1 30.5 30.0 30.5 30.0 2 33.0 33.0 33.5 32.5 3 34.0 34.0 34.0 34.0 4 33.0 32.5 33.0 32.5 5 33.0 33.0 32.5 32.5 6 33.0 32.5 32.5 32.5 Avg. Dewpoint Temp., 32.6 32.3 32.5 32.2
*F, (Note 1)
Water Vapor Press., Psia 0.0908 0.0897 0.0904 0.0893 Drybulb Temp., *F 1 49.4 49.4 49.6 49.6 2 49.4 49.4 49.6 49.6 3 49.9 49.9 50.0 49.9 4 49.5 49.6 49.6 49.7 5 47.5 47.6 47.8 47.9 r" 6 48.7 48.8 49.0 49.1 (m , 7 49.0 49.0 49.1 49.3 8 48.8 48.8 48.8 48.8 9 49.2 49.3 49.4 49.3 10 49.6 49.6 49.6 49.6 11 48.7 48.8 48.8 48.9 12 49.0 49.2 49.2 49.2 13 49.0 49.0 49.0 49.0 14 48.5 48.6 48.6 48.6 15 49.0 49.2 49.2 49.2 16 48.4 48.4 48.5 48.5 17 48.7 48.8 48.7 48.7 18 48.6 48.6 48.7 48.6 19 48.5 48.5 48.5 48.5 20 48.5 48.5 48.4 48.5 21 48.6 48.8 48.8 48.7 22 48.5 48.6 48.5 48.6 23 48.2 48.2 48.4 48.3 24 48.5 48.5 48.6 48.6 Avg. Temp., *R (T) (Note 2) 508.4 508.5 508.5 508.6 Weight of Air, LDS. 254,265 254,238 254,250 254,211
' Leakage Rate, Total +0.032 +0.043 +0.036 +0.050 .- Time Method, %/ day Sheet 6
TABLE 4.1 GINNA STATION FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS TI:iE/l: CUR 1900/24
\
'N - PI-3A, Counts 50.5780 PI-3B, Counts 48.9820 Avg. Press., Psia (PT) 49.4379 Dewpoint Temp., *F 1 30.0 2 32.5 3 33.5 4 32.5 5 32.0 6 32.0 Avg. Dewpoint Temp., 31.9
*F, (Note 1)
NOTE 1: For calibration Water Vapor Press., Psia 0.0881 purposes 1.l*F is
- subtracted from Drybulb Temp., *F 1 48.7 the sum.
2 49,o 3 49.6 4 49.4 _ rp) 5 6 47,4 Note 2: For calibration purposes 2.2*F is
'(/
7 49.2 49.2 subtracted from 8 48.8 the sum. 9 49.2 10 49.6 11 49.2 12 49.3 13 49.1 14 48.6 15 49.1 16 48.5 17 48.6 18 48.6 19 48.5 20 48.4 21 48.5 22 48.4 23 48.0 24 48.4 Avg. Temp., *R (T) (Note 2) 508.4 Weight of Air, LBS. 254,146
+0.074 lf OLeakageRate, Time Method, %/Total day Sheet 7
5.0 SUPPLEMENTAL TEST 5.1 TEST PERFORMANCE The supplemental test was started approximately six hours after completion of the 24 hour integrated test. This time lapse was used to establish a controlled leakage rate but primarily the time was required to stabilize the ambient conditions within the containment after the diffi-culties encountered with the containment recirculation units. A controlled leakage rate from the containment was estab-lished through calibrated rotometer SN 1856. The rotometer flow was 12 percent of scale or 0.037 weight percent per day at test conditions. Twelve sets of hourly data were
,_ s
) taken after containment conditions were reestablished.
These data appear in tabular form at the end of this sec-tion. Plots of mean containment temperature and pres-sure are shown on Figure 3. The weight of containment air is plotted on Figure 4. 5.2 TEST RESULTS Leakage rates were calculated using both the Mass Plot and Total Time methods. The leakage rates were 0.050 and 0.089 percent per. day respectively. Reducing these values by the amount of the controlled leak rate yields containment building leakage rates of 0.013 percent per day by the Mass Plot Method and 0.052 percent per day by 'the Total / Time Method. Both results are within 0.25 Lt (0.038 per- l cent / day) of the leakage rates found during the 24 hour
- -- ._. . , - . - . - .. ..- . . ~ . . . -.
2 test. Therefore it is concluded that the accuracy of the { 24 hour test has been established. The supplemental test results are summarized in the table below where: L = leakage rate determined by 24 hour test S = total leakage rate found in supplemental test in-cluding known leak rate K = imposed, known leak rate l (S-K)-Ll is the measured difference between the 24 hour test and supplemental test results. 0.25 Lt is the acceptance criterion variance between the
- 24 hour and supplemental test results.
LEAKAGE RATES, Percent / day r' Calculational s_ Method L S K S-K l(S-K) -L l 0.25 Lt Total Time 0.038 0.089 0.037 0.052 0.014 0.038 i Mass Plot 0.044 0.050 0.037 0.013 0.031 0.038
- (.' -]
i FIGURE 3 SUPPLEMENTAL TEST REI.CTOR BUILDING MEAN TEMPERATURE AND PRESSURE VS TIME
.. g , 1 .--. _ . _: ,-_ ._}=u
. - = ._,,-..w.; === v.n -- - - - -
- .=__--._. s._
==ta= 4 .m u= -- -=-+-- --- + - -
"-- =~ --
- =;: =-:-~ - , =;=g:=._
~ - " - - '=--' " - ~~~
49 35 =
- n=4:== 4.__ -+- ._,_.
; ,. _. ~ h.-
m . v._.pg _. . . . _..'.-
~~" ' ' . " . ' .
' " ' " - ~- ' ~ ~ -~~ ~ ' ~ '3 ~~~~
4 9 . 3 4 (<,'_..,._.._,
._ ..L.
'r:
~ ""~ " ~~"~~~f- ...~~~
REACTOR BUILDING 49.33 ~- ~d*^ '-'~ MEAN AIR M- '
'~~$ ~~ ^y^ ~ ~ ~_
PRESSURE 49.32
- N~ (U--~m-- ;
- w_
M - (PSIA) ~.-----. -- v. 49.31 -, .- _w
-:*- ~ a_=
49.30 *~
+
49.29 . . - . . _ _ 49.28 - - " . - _.+ --
,. y. . _ -
f p..- -.+ . J 0 -- - " - " - - ' - -
+-l..---
n4.- - . .
. i
- L --
48.9 _- 48.8 - - - -- - -
.,_u--_
~
REACTOR BUILDING ~ ~ ~ ' MEAN TEMPERATURE 48.7 (DEGREES . .
- - + - " + --+ W FAHRENHEIT) 48.6 :[ C
~
+
~ '~-
4 8 5(^v "._ #c
~~~~
=.= w--
- * "U
- 4 . . , ._ ~
~ - - .= - - u
~^-
._^ _- ".--
^t
^
48.4 - ._- f_
.- _- .,,- .- ~_== _._
'";- 4'+t4g++ = -
48 3 =~n1;===du=--r-*-- -w
~~
" . " ~."i--- - D- "--- ~
_ ce
- . = : =mn
--- - A. . - . -.:...: :nx2- - - - n nt2 72- _:b==
. :=u. tr
=;-- _ rr. :. +. .g .
z -*- i ,y. +-
-a
- .- a:
....E=....:u:==nt.w
~ .; = = -
. = . . _u=-
/:.u- m:nm -t**-- ---~ ---
- :- -~; -- ....
N ===~,.4.
.1;*'. gt 4 ; *.;r++2 ' ' - ' + + + - t.,. .- n 1 -'+'n Np*- t , . .; ___.-*r **-t _ ; _
m r.an"r '*.;j.n.n=1:n-:+ -*4u . 1;*t".1*' . u ;j:=_.nd
- r. : nun ;nt--- --
tt-- -- . . tn. --t+ - t . nt'.=;
. ! i 'i __
7--*:- :=q= 7**+*n1=Pi ---++?---4 -- ? - 1 nj .*-+-:=.'3 n
- 0 [*; .1 *.O_ -
0 1 2 3 4 5 6 7 8 9 10 11 TIME (HOURS)
FIGURE 4
/m SUPPLEMENTAL TEST
/
, REACTOR CONTAINMENT BUILDING AIR WEIGHT t
v/ VS TIME
.-w,
_. u
~
'~~ ; . m - - -"~ ~ ~ ' ~
254,260 ~ ~~+~
~. -.
254,240 ._,__. 4_. . _ i w. ._ -
- ,. .~ . . ._._._
~ ~
254,220 ' N + a %
^
++9'* ~,ut, 254,200 .. "O ;-
~ ~ ~ "
e _ :-; 254'180 'h. =
-~ - "" -
' ~
'~-+"-
v,.w4.-u
. , ,- ..a - .
.. m --
254,160
-3%*
'" A
~w
~~~ _
~ ;,_. _ _
REACTOR BUILDING *i __- AIR WEIGHT '~ n
.m - . ,
254,~140 'x- - (LBS) .35Cgg_ -- . 254,120 ' ' ' ~ ~ '
" ~'
/
/n \
w' ~
- -:m+
- a. :
.v- .:
. -j-- .
[j _
~
( ; -~. -- x \s .) 254,100 %- .. / _~ ~~ lLEAST SQUARES FIT k .. / ' 254,080 jSLOPE = 5.27 LBS/HR..ORj 7' -
!0.050 PERCENT / DAY k- - - * " -"- " -
su.. . _,m 254,060 f F
- !=::i "
c 254'040 .-. , 58.1 LBS. ERROR BARS h
+FROM ERROR ANALYSIS v--
254,020 _ - .
. -- - a .
~
254,000 "~~
*~ ~~
253,980 g
* * "i-#^"'
253'960 =2:: :: w
~~~ ~ ~'n4'?
+ td. 4++ +4t +++t I. u. __i .i.a.,a 1
**:!*,;g++-+t~;; ;;
+-
-++r;;7+;+*,;- + j; ,4 4;
+
r;;.gt; *" ha -. n ;+y jgu .,n-- -.
, .+. .
4.gx i NT'N*. I {.D ,, **
- 4'**" * * * * ' M*
"" ' ~ ~ " " ~ ^ -',. . 444
~~~'^"'"
0-
, 0 1 2 3 4 5 6 7 8 9 10 11
,/ ,ss
,/ \ I s < TIME (HOURS)
TABLE 5.1 GINNA STATION FEBRUARY 1976 SUPPLEMENTAL TEST - LEAKAGE RATE DATA AND RESULTS
% TIME / HOUR 0200/0 0300/1 0400/2 0500/3 PI-3A, Counts 50.5650 50.5700 50.5690 50.5590 PI-3D, Counts 48.9730 48.9790 48.9770 48.9670 Avg. Press.,. Psia (PT) 49.4270 49.4324 49.4310 49.4211 Dewpoint Temp., *F 1 30.0 30.0 30.0 29.5 2 32.5 32.5 32.5 32.0 3 33.0 33.0 33.0 33.0 4 32.0 32.0 32.0 32.0 5 32.0 32.0 32.0 32.0 6 32.0 32.0 32.0 32.0 Avg. Dewpoint Temp., 31.7 31.7 31.7 31.6
*F, (Note 1)
Water Vapor Press., Psia 0.0873 0.0873 0.0873 0.0870 Drybulb Temp., *F 1 48.7 48.8 48.9 48.7 2 49.0 49.0 49.1 49.0 3 49.6 49.7 49.7 49.6 4 49.2 49.3 49.3 49.1 5 47.3 47.3 47.3 47.2 7(, g) 6 48.3 48.4 48.4 48.3
-L/ 7 48.8 48.8 48.8 48.7 8 48.6 48.7 48.7 48.6 9 49.0 49.1 49.2 49.1 10 49.5 49.5 49.5 49.4 11 48.6 48.6 48.6 48.6 12 48.8 48.9 48.8 48.8 13 48.8 48.9 48.8 48.9 14 48.4 48.5 48.4 48.4 15 48.9 49.0 48.9 48.8 16 48.2 48.2 48.2 48.2 17 48.5 48.5 48.5 48.4 18 48.4 48.4 48.5 48.4 19 48.3 48.4 48.3 48.3 20 48.3 48.4 48.3 48.3 21 48.5 48.6 48.6 48.6 22 48.3 48.4 48.4 48.2 23 47.9 47.9 47.9 47.9 24 48.4 48.5 48.4 48.4 Avg. Temp., *R (T) (Note 2) 508.2 508.3 508.3 508.2 !
Ucight of Air, LDS. 254,194 254,.171 254,164 254,165 1 l Leakage Rate, Total -0.142 +0.034 +0.045 +0.033 !
/(s Time Method, %/ day i Sheet 1
. A
TABLE 5.1 GINNA STATION FEBRUARY 1976 SUPPLEMENTAL TEST - LEAKAGE RATE DATA AND RESULTS TI:iE/ HOUR 0600/4 0700/5 0800/6 0900/7 h jPI-3A, Counts 50.5510 50.5480 50.5450 50.5380 PI-3B, Counts 48.9570 48.9550 48.9520 48.9450 Avg. Press., Psia (PT) 49.4126 49.4102 49.4072 49.4002 Dewpoint Temp., *F 1 29.5 29.5 29.9 29.9 2 32.5 32.5 32.0 32.0 3 33.0 33.0 33.0 31.0 4 32.0 32.0 32.0 32.0 5 32.5 32.0 32.0 32.0 6 32.0 32.0 32.0 32.0 Avg. Dewpoint Tenp., 31.7 31.7 31.6 31.3
*F, (Note 1) 0.0873 0.0873 0.0870 0.0858 Water Vapor Press., Psia Drybulb Temp., *F 1 48.7 48.6 48.8 48.8 2 48.8 48.8 48.8 48.8 3 49.5 49.6 49.5 49.4 4 49.0 49.1 49.1 49.0 5 47.1 47.2 47.2 47.3
/" 6 48.2 48.2 48.4 48.4 s_ 7 48.6 48.6 48.7 49.6 8 48.5 48.5 48.5 48.5 9 49.0 49.0 48.8 48.8 10 49.4 49.3 49.3 49.4 11 48.4 48.4 48.4 48.4 12 48.7 48.7 48.7 48.7 13 48.7 48.6 48.5 48.6 14 48.3 48.2 48.7 48.2 15 48.8 48.8 48.0 48.6 16 48.0 48.0 48.1 48.0 17 48.3 48.3 48.3 48.3
-- 18 48.3 48.2 48.2 48.2 19 48.2 48.1 48.3 48.1 20 48.2 48.1 48.3 48.0 21 48.5 48.4 48.4 48.2 22 48.2 48.2 48.3 48.2 23 47.8 47.8 47.8 48.0 24 48.2 48.2 48.2 48.1 Avg. Temp., R (T) (Note 2) 508.1 508.1 508.1 508.1 Weight of Air, LES. 254,169 254,157 254,143 254,113 c , Leakage Rate, Total +0.017 +0.034 +0.048 +0.077
^, ! ' ime ::ethod, %/ day
%,J ?'
Sheet 2
TABLE 5.1 GINNA STATION i FEBRUARY 1976 SUPPLEMENTAL TEST - LEAKAGE RATE DATA AND RESULTS 1000/8 1100/9 1200/10 1300/11 C,) TIME / HOUR PI-3A, Counts 50.5330 50.5450 50.5470 50.5480 PI-3B, Counts 48.9410 48.9530 48.9550 48.9560 Avg. Press., Psia (PT) 49.3958 49.4077 49.4097 49.4107 Dewpoint Temp., *F 1 29.9 29.9 29.5 29.9 2 32.0 32.0 32.5 32.0 3 33.0 33.0 33.0 32.2 4 32.0 31.9 32.0 31.9 5 32.0 32.0 32.0 32.0 6 32.0 32.0 32.0 32.0 Avg. Dewpoint Temp., 31.6 31.6 31.7 31.5 "F, (Note 1) Water vapor Press., Psia 0.0870 0.0870 0.0873 0.0866 Drybulb Temp., *F 1 48.7 49.0 49.0 49.0 2 48.8 49.1 49.2 49.1 3 49.2 49.4 49.4 49.5 4 48.9 49.2 49.1 49.1 5 47.2 47.2 47.3 47.3 6 48.3 48.5 48.5 48.5 N( 7 8 48.6 48.4 48.7 48.4 48.8 48.8 48.5 48.5 9 48.~8 48.9 48.8 48.9 10 49.3 49.3 49.2 49.4 11 48.4 48.4 48.4 48.5 12 48.6 48.8 48.7 48.8 13 48.5 48.6 48.7 48.7 14 48.1 48.2 48.2 48.3 15 48.6 48.7 16 48.7 '48.8 47.9 48.0 48.0 48.0 17 48.2 48.4 48.4 48.4 18 48.2 48.3 48.4 48.3 19 48.0 48.2 48.2 48.4 20 48.0 48.0 48.1 48.3 21 48.3 48.3 48.4 48.5 22 48.0 48.1 48.2 48.3 23 47.6 47.8 47.9' 47.9 24 48.0 48.2 48.2 48.3 Avg. Temp., *R (T) (Note 2) 508.0 508.1 508.1 508.2 ! Weight of Air, LBS. 254,134 254,146 2'4,155 5 '254,113
' Leakage Rate, Total +0.046 +0.031 +0.021 +0.051 ,, jTime Method, t/ day ,
f NOTE 1: For calibration purposes 1.l*F is subtracted from the sum. NOTE 2: For calibration purposes 2.2*F is subtracted from the sum. l i Sheet 3 - I s -
.,m_.-_ ,-
l i 1 6.0 TYPE "B" AND "C" LEAKAGE RATE HISTORIES
) 6.1 DISCUSSION OF LEAKAGE HISTORY During 1973 there was no refueling shutdown. Type "C" tests were performed for all isolation valves during Plant outages for various maintenance purposes. Type "B" tests for 1973 were performed during the month of March. For the other years, 1974, 1975 and 1976, all Type "C" testing was performed during each reactor shutdown for refueling. Class "B" testing was performed l at yearly intervals and in general coincided with the re-fueling shutdown interval. -
All local type "B" and "C" tests were performed at 60 psig. The acceptance criterion for Type "B" and "C" tests, as per Plant Technical Specification 4.4.2.2, is that total leakage from all penetrations and isolation valves shall not exceed .45 La. In the event that .45 , , La is exceeded, when at power, repairs are to be initiated i immediately. If conformance to .45 La cannot be demon-l started within 48 hours', the reactor shall be shutdown and depressurized and remain so until local leakage meets the l acceptance criterion (.45 La is equivalent to 17,200 cc/ min.). i Tabulation of the leakage rate results for the various 4
! /
testing periods' indicate that the limiting criterion was exceeded during the years 1974'iand'1976. At all times l ! when local leakage rates were discovered to be in excess j
, s
- - _ _ - - - ~ - , _ _ . _._.___ _ ___ _ _ _ , _ . _ . _ .
i i 4 of the allowable limit the reactor was already in the de- ! 3 ( pressurized state. Subsequent Plant startup was not per-
- mitted until repairs had been accomplished and leakage rates reduced to meet the .45 La acceptance criterion.
The following components have been the major contributors to excessive local leakage rates:
- 1. Containment purge supply 48 inch dampers.
- 2. Containment purge exhaust 48 inch dampers.
- 3. Personnel hatch personnel access lock.
It is believed that due to the 30 to 50 Fahrenheit degrees decrease in containment building ambient temperature from operating to cold shutdown status, thermal shrinkage of the' 48 inch dampers and/or "O" ring rubber seats occurs which re-duces the tightness of the seating interface and creates leak-age paths from the purge supply and exhaust volumes. Excessive leakage from the personnel hatch personnel ac-cess lock has occurred during extended plant outages when the repeated door openings resulting from the heavy personnel trafzic entering and exiting containment cause the handwheel shaft packing gland nuts to back off. This relaxes the shaft packing and permits a leakage path to develop The equipment hatch personnel access lock, al-though somewhat larger in volume, is of the same basic design, and has exhibited satisfactory leakage rates throughout this reporting period. Door openings of this l l l
, - . 1
lock also. occur more frequently during a plant shutdown,
\
however, to a lesser degree than the personnel hatch per-sonnel access lock. Maximum known local leakage at the time of any Plant start-up during this reporting interval has been .064 La. An evaluation of the leakage results indicates that the maximum leakage during any power operating period was .39 La. This occurred during the year 1975 and is based upon , the known leakage at plant start-up after refueling (5-10-75) and leakages which apparently developed during . subsequent Plant operation, but which were not discovered until 1976. The .39 La value excludes leakage from the I purge supply and exhaust valves which has been observed j only at cold shutdown and is believed to result from changes in temperature between power operation and cold , shutdown. Major contributors to the .39 La leakage rate were (1) leakage from personnel hatch personnel access lock on January 16, 1976 (corrected on January 17, 1976) and (2) leakage from check valve 1713 in the Reactor Coolant Drain Tank Nitrogen Supply Line on February 8, 1976 (corrected February 9,1976) . Leakage from check valve 1713 was discovered when the nitrogen supply was blocked and the line vented for the integrated leak rate test. O
n I i 1 I l , Identification of penetrations and manifolds and their l leakage history since the Integrated Leak Rate Test in 1972 is found in Tables 6.1-1 through 6.1-6. i a l f l ) i i
\
l I t } O l I i ! s t i f ( l i l l 1 < i i l I , l
- I l
i 1 1 l l l
.c. . - . - -
6.2 PENETRATION MODIFICATION DURING REPORTING PERIOD Spare electrical penetration sleeve AE 12 on manifold 3 was fitted with a Westinghouse type SP 504 electrical penetration assembly to accommodate wiring for a tele-vision monitoring system inside containment. Installa-tion and initial testing of this completed penetration was governed by System Modification procedure SM 75-2.1 and was completed on April 8, 1975. Mechanical penetration sleeves 119 and 123 on manifold F i l were fitted with bellows-type penetration assemblies to accommodate the Standby Auxiliary Feedwater System piping to the steam generators inside the containment building. ( ) Penetration sleeve 119 was a capped spare. Penetration sleeve 123 already contained the Reactor Coolant Drain Tank gas analyzer line. The new penetration assembly . provided for continuing service of this line. Installa-tion and initial testing of these completed penetrations was governed by System Modification Procedure SM 75-5.28 and was completed on March 31, 1976. O v ( 1 ! TABLE 6.1-1 ! TYPE B PENETRATION MANIFOLD DESIGNATION AND LEAKAGE RATES i t NOTE: Where applicable, significant leakage comments follow test data listing j for manifold and/or penetrations. MANIFOLD A Total Volume 21.88 ft.3 Penetrations Included: 312, 309, 332, 310, 317, 318, 321, 322, 306, 316, 319, 304 and 313 Date 'lested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 0 1/29/74 17 2/5/75 0 2/28/76 0 MANIFOLD B Total Volume 25.24 ft.3 Penetrations Included: 323, 324, 325, 326, 336, 320, 315, 307, 305, 311, 308, 303 and 301 Date Tested: Leakage Rate cc/ min. (at 60 psig) s 3/19/73 0 1/29/74 50 2/5/75 0 3/1/76 0 MANIFOLD C Total Volume 45.92 f t. Penetrations Included: 300, 401 and 403 Date Tested: Leakage' Rate cc/ min. (at 60 psig) 3/19/73 5 7/24/73 14 8/7/73 0 4/29/74 0 6/18/74 8 2/5/75 0 3/1/76 0 Sheet 1 6
,, . , , , - - - , - - - - v, -,,_ , , -,,...v..7-.., , , -
TABLE 6.1-1 l TYPE B PENETRATION MANIFOLD DESIGNATION AND LEAKAGE RATES MANIFOLD D Total Volume 636 ft. 4 Penetrations Included: Personnel Hatch Personnel Access Lock Date Tested: Leakage Rate ) cc/ min (at psig) 1 3/30/73 959 5/9/73 86 l l 6/22/73 86 ' l 6/28/73 29 j 8/2/73 0 9/13/73 122 ! 3/24/74 9) 5/3/74 1082(3) 663 I 9/5/74 244 l
' ~10/4/74 963(4) 11/20/74 515 5/1/75 6283(5) 5/8/75 306 9/16/75 208 1/16/76 3633(6) 1/17/76 284
.3/31/76 26,814(7) 4 4/1/76 64 i
Comment 1- '
)r i Test on 3/30/73 indicated a leakage rate of 959 cc/ min. from the between door l
volume of personnel access lock. At this time no communicating leakage path ! through the secondary boundary (outside door) could be detected, hence, all t leakage was judged to be passing through the containment side primary boundary r
'into containment. The lock was repressurized on 5/9/73. Containment entry was effected through the equipment hatch personne.1 lock. Leakage was discovered at ,
the inner door lower handwheel shaft packing gland. After tightening the pack-ing gland nut, a satisfactory retest indic,ated a leakage rate of 86 cc/ min.
' I Comment 2: '
i The plant was down for refueling at the time of this leakage discovery. No corrective action was undertaken at this time. However, two leakage paths were identified, one path through the outside (secondary boundary) equalizing valve and the other path at the inner door (primary containment side boundary)-lower handwheel shaft packing gland. Total leakage from the personnel lock at this , time was 1082 cc/ min. A visual assessment (soap solution) allocated approxi-mately 200 cc/ min. to the equalizing valve and the remainder to the shaft pack-ing gland. Comment 3: (
. Repair attempts were made on the equalizing valve on 5/1/74 accor' ding to proce-dure EM-115.
- }of663cc/ min. Testing of the lock after repair of valve indicated a. leakage rate 1('
Slight leakage still existed through the equalizing valve. Although the plant had returned to power operation no further attempts were made j at this time to reduce the observed leakage rate. Known leakage from all type B i and C tests was less than 10% of Technical Specification allowable limit. Sheet 2
. . - _ _ - - - . . . . . - _ -- . . . - . . - . . . _ - _ -,- - . - . ., . . . . . . . , , , ----.-.,n.. -- _
I l l TABLE 6.1-1 ) ['~'s t TYPE B PENETRATION MANIFOLD DESIGNATION AND LEAKAGE RATES l Comment 4: 1 This test was performed after attempts were made to install a new equalizing ! valve in place of the previously mentioned leaking one. The valves, however, were not conveniently interchangeable, hence, the old valve was returned to service. The original valve is still leaking, however, its' leakage rate is much less than that detected initially. Apparently the valve is seating better and the repeated lock depressurization through the valve has dislodged any loose debris. Other than the slight leakage path through the equalizing valve no other communicating leakage paths to the outside building atmosphere were detec-ted. At this time, no further actions were attempted to reduce leakage from the lock volume. Comment 5: , The Plant was shut down for the 1975 refueling outage at the time of this leakage discovery. Leakage is primarily from the outer door upper handwheel shaft pack-ing gland. The packing gland nut was found to be quite loose.'Upon retightening the gland nut and retesting the lock on 5/8/75, a leakage rate indication of 306
'/ cc/ min, was obtsined.
s_. Comment 6: The Plant, a few days prior to this test, had returned to power operation after having been shutdown for Steam Generator Tube Repairs. Including the indicated
- leakage from this volume, the known total leakage of Class B and C tests was well below allowable Technical Specification limits (approximately 30% of limit).
Again, the outer door upper shaft handwheel packing gland nut was found to be partially backed off. The packing gland nut was retightened, also a new grease fitting was installed on the inner door lower handwheel shaft packing gland. Leakage had previously been identified at the grease fitting threads. Retesting on 1/17/76 indicated a remaining leakage rate of 284 cc/ min. from the lock volume. Comment 7: The Plant was shut down for refueling at the time this leakage was discovered. Corrective actions were undertaken immediately. The outer door upper handwheel packing gland nut was found to have backed off completely. The nut was threaded back into the gland and securely tightened, a retest on 4/1/76 indicated a re-maining leakage rate of 64 cc/ min. from the lock volume. Whenever excessive leakages have been experienced from the access lock, it is a g result of the packing gland nuts loosening or turning free from the packing
/
- \(
I [/ gland, the high leakage rates can be keyed into plant outages where, due to ex-tensive maintenance work, heavy personnel traffic into and out of containment subject the access lock doors to many cycles of operation. Sheet 3
k l TABLE 6.1-1 TYPE B PENETRATION MANIFOLD DESIGNATION AND LEAKAGE RATES d The inaccessibility of these packing gland nuts discourages a routine tighten-ing after extended plant outages. Presently space is limited for adjustment and virtually precludes the use of any other but a chain wrench, which at best 1 only permits limited rotation with each bite. At the next dismantling, for overhaul, of the door assemblies, new nuts will be installed which will permit
, the use of a spanner wrench. Secondly, keepers will be placed to prevent the nuts from backing off.
MANIFOLD E Total Volume 37.67 ft.3 Penetrations Included: 202, 203, 209, 201, 204, 210, 206, 205, and 207 Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/20/73 0 1/29/74 0 2/7/75 0 3/1/76 0 MANIFOLD F Total Volume 7.4 ft.3 Penetrations Included: 119, 129, 123, 133
'O Date Tested: Leakage Rate s cc/ min. (at 60 psig) 3/19/73 .35 1/29/74 .12 2/6/75 0 3/1/76 6.5 MANIFOLD G Total Volume 15.25 ft.
Penetrations Included: 120, 121, 124, 125, 126, 127, 128, 130, 131, and 132 Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 28 1/29/74 1.33 2/6/75 0 3/1/76 .77 MANIFOLD H Total Volume 29.31 ft.3 Penetrations Included: 402, 404 Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 0 7/24/73 8.4 4/1/74 9.87 4
') 4/18/75 2.47 j' %) 6/18/75 )
5.14 ' 3/1/76 0 Sheet 4 l l l
TABLE 6.1-1 TYPE B PENETRATION MANIFOLD DESIGNATION AND LEAKAGE RATES 1 MANIFOLD I Total Volume 1065 ft. Penetrations Included: Equipment Hatch Personnel Access Lock Date Tested: Leakage Rate cc/ min (at 60 psig) 4/2/73 166 5/11/73 84 8/3/73 0 1/5/74 151 5/8/74 24 11/14/74 275 5/1/75 171 9/18/75 212 ) 1/15/76 269 3/31/76 254 MANIFOLD J Total Volume 22.22 ft. Penetrations Included: 102, 111, 101, 113, 118, 105, 109, 99, 103 and 107 Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 0
,- 1/29/74 0 2/6/75 0 3/1/76 0 MANIFOLD K Total Volume 20.88 ft.
Penetrations Included: 104, 106, 108, 110, 112, 100 and 140
- Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 0 1/29/74 0 2/6/75 0 3/1/76 0 MANIFOLD L Total Volume 3 ft.3 Penetrations Included: 29, 141, 142, 143 Date Tested: Leakage Rate cc/ min. (at 60 psig)
, 7/28/73 0 2/16/74 0 3/28/75 0 3/1/76 0 FUEL TRANSFER TUBE Date Tested: Leakage Rate (O s 7/28/73 4/1/74 4/30/75 cc/ min. (at 60 psig) 0 0 0 3/1/76 0 sheet S
TABLE 6.1-1 TYPE B PENETIV. TION MMIFOLD DESIGNATION AND LEAKAGE RATES d MANIFOLD: ELECTRICAL 1 Total Volume 102 ft. 3 Penetrations Included: CE-1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22, 23, 24, 25, 27, 29, 30, 31, 32, 33, 34 Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 7.44 1/29/74 1086 (1) 3/7/74 160 4/24/74 85 4/29/74 2.6 3/13/75 180 (2) l 3/4/76 156 (3) l Comment 1: At the time of leakage discovery, the plant was shut down for refueling. This mani-fold supplies nitrogen cover gas to 32 individual penetrations. Each penetration is connected to the manifold by 1/4 inch tubing with in-line shutoff valve for isola-tion of the penetration from the manifold. By selective isolation, leakage was traced to penetrations #10 and #21. Approximately 160 cc/ min. of the total leakage
~
was allocated to Penetration #10 and the remainder, approximately 927 cc/ min., as l leakage from Penetration #21. Penetration #10 is a completely wired spare with the pass-through wiring being terminated at junction boxes inside and outside of the containment building. Ini-tial leakage from this penetration was from the penetration end inside containment, at the peripheral epoxy seal between the penetration tube and penetration end spacer. I l i Repair attempts were made by inducing a slight vacuum in the penetration free volume and then applying an epoxy slurry to the outside surface where leakage was noted. Upon completion of the first repair, visual indication showed that the inside con-tainment.end was effectively sealed, however, leakage subsequently developed from the penetration end outside of containment. During repair efforts on the outboard end it was noted that leakage would continually shift to different areas, i.e, peripheral seal or individual wire holes. A final retest on 4/29/74 indicated a leakage rate from this penetration of 2.6 cc/ min. Penetration #21 passes through the 480 volt power leads for the IC containment recirculation fan unit. Leakage from this penetration was inside containment at the Hy-press connection on Terminal T3. This metal connection exhibited four (4) surface stress cracks, however leakage was observed from only one of these cracks. Repair of this installation was effected on 3/7/74 by placing an overlay of high tin content, sof t solder completely around the metal connector. After repair, a
's 60 psig pressure decay test indicated zero pressure loss over a four-hour period.
Sheet 6
1 1 TABLE 6.1-1 , TYPE B PENETRATION MANIFOLD DESIGNATION AND LEAKAGE RATES Comment 2: ) 1 During the Class B test for 1975, leakage again (180 cc/ min.)was noted from the manifold. All leakage was attributed to Penetration #10. A 60 psig pressure decay test of the manifold, with Penetration #10 isolated, indicated zero pressure loss ever a four-hour interval. Repair attempts were again made on Penetration #10, however, leakage could not be materially reduced. Comment 3: Results of a 60 psig pressure decay test on 3/4/76 indicated a leakage rate of 156 cc/ min from the manifold, all traceable to the outboard end of Penetration #10. The manifold exhibited zero pressure loss over a fifteen (15) hour period with Penetration #10 isolated. Further effort will be made to repair this leaking penetration. Again, a vacuum will be induced in the penetration, but instead of applying the relatively high viscous epoxy slurry to the points of leakage, a silicone varnish will be applied to facilitate penetration into the fissures of the old epoxy seal. MANIFOLD: ELECTRICAL 2 Total Volume 15 ft.3
\ Penetrations Included: BE-1, 2, 3,4
>s Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/19/73 0 1/29/74 -7 3/15/75 7 3/1/76 0 MANIFOLD: ELECTRICAL 3 Total Volume 36 ft.3 Penetrations Included: AE-1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13 and 14 Date Tested: Leakage Rate cc/ min. (at 60 psig) 3/21/73 11.3 1/29/74 11.3 3/13/75 11.25 3/1/76 .34 m U Sheet 7
N. %
\
TABLE 6.1-2 DESCRIPTION OF ISOLATION VALVES TYPE "C" LINE PRIMARY $ECONDARY TEST LINE SERVICE SIZE PENT. ISOLATION ISOLATION NO. (inches) NO. BOUNDARY BOUNDARY 1 PRT. to Gas Analyzer 5/8 120 A0V 539 2 "2 Supply to PRT. 3/4 121 CK 528 3 Make Up Water to PRT. 2 121 CK 529 A0V 508 4 Residual Heat from RCS. 10 140 MOV 701 SA C.V. Sump Suction to RHRS 8 141 MOV 851A MOV 850A, 1813A 5B C.V. Sump Suction to RHRS 8 142 MOV 851B MOV 850B, 1813B 6 Residual Heat to RCS. 10 111 MOV 720 ! 7 Letdown from RCS. 2 112 A0V 371 8 Charging to RCS. 2 100 CK 370 9A RCP Seal Water 2 106 CK 304A 9B RCP Seal Water 2 110 CK 304B 10 Alternate Charging 2 102 CK 383B 11 RCP Seal Rt. and Ex. Letdown 3 108 MOV 313 12A PRZ. Steam. Space Sample 3/8 207 A0V 966A 1 12B PRZ. Liquid Space Sample 3/8 206 A0V 966B 12C RCS Sample -- Loop B 3/8 205 A0V 966C 13A "A" SG. Sample 3/8 206 A0V 5735 13B "B" SG. Sample 3/8 207 A0V 5736 Sheet 1
TABLE 6.1-2 DESCRIPTION OF ISOLATION VALVES TYPE C" LINE . PRIMARY SECONDARY LINE SERVICE SIZE PENT. ISOLATION ISOLATION TEST (Inches) NO. BOUNDARY BOUNDARY NO. 14 C. V. Air Sample (inlet) 1 305 CK 1599 A0V 1598 15 C. V. Air Sample (outlet) 1 305 A0V 1597 16A "A" SG. Blowdown 2 321 A0V 5738 l 16B "B" SG. Blowdown 2 322 A0V 5737 17A C.V. Pressure Trans. 3/8 121 PT. 945, 946 17B C.V. Pressure Trans. 3/8 203 PT. 947, 948 17C C.V. Pressure Trans. 3/8 332 PT. 944. 949, 950 18A "A" Cont. Spray Header 6 105 CK 862A, LCM 864A 18B "B" Cont. Spray Header 6 109 CK 862B, LCM 864B 3 3 110 CK 889A, 889B 19 Safety Injection CK 870A, 870B l 3 3 113 LCM 879 19 Safety Injection 20 RCDT Gas Header 1 129 A0V 1787, CK 1713 A0V 1786 21 RCDT Gas Analyzer 3/8 123 A0V 1789 RCDT Discharge 4 143 A0V 1721 A0V 1003A, A0V 1003B 22 23 Sump "A" Discharge 3 107 A0V 1728 A0V 1723 { 24 Reactor Support Cooling 6 130 MOV 813 Inlet and Outlet 6 131 MOV 814
. Sheet 2 l _ _ _ _ _ _ _ _ _ _ _ _ _ _
-~ -
V TABLE 6.1-2 DESCRIPTION OF ISOLATION VALVES TYPE "C" LINE PRIMARY SECONDARY TEST LINE SERVICE SIZE PENT. ISOLATION ISOLATION NO. (Inches) NO. BOUNDARY BOUNDARY 26 ACS to "A" RCP 4 127 CK 750A 27 ACS to "B" RCP 4 128 CK 750B 28 ACS from "A" RCP 4 126 MOV 759A 4 29 ACS from "B" RCP 4 125 MOV 759B 30 ACS Excess Letdown 2 124 CK 743 Hx. (Supply and Return) 2 124 A0V 745 32 Instrument Air 2 310 CK 5393 A0V 5392 33 Service Air 2 310 LCM 7227 LCM 7141 34 Depress. at Power 6 132 A0V 7970 A0V 7971 35 Purge Supply 48 204 A0V 5870 AOV 5869 36 Purge Exhaust 48 300 A0V 5878 A0V 5879 5 37 "A" Reactor Comp. Cooling 2 201/209 Unit Cooling coil 38 "B" Reactor Comp. Cooling 5 2 209/201 Unit cooling coil 39 Demineralized Water 2 324 LCM 5024 LCM 5021 40 Auxiliary Steam Supply and 2 301 LCM 6151 LCM 6165 Condensate Return 1 303 LCM 6175 LCM 6152 42 Leakage Test Depress. 6 313 Flange MOV 7444 43 Leakage Test Supply 6 317 Flange MOV 7443 Sheet 3
\
\ .
~ 'd 4
TABLE 6.1-2 DESCRIPTION OF ISOLATION VALVES TYPE "C" _____ LINE PRIMARY SECONDARY TEST LINE SERVICE SIZE PENT. ISOLATION ISOLATION NO. (Inches) NO. BOUNDARY BOUNDARY d 44 Leakage Test. Depress. 6 309 Flange MOV 7445 45 Leakage Test 1/2 332 Tubing Cap LCM 7448 45 Instrumentation (3 Lines) 1/2 332 On Each Line 7452, 7456 46 Nitrogen to Accumulator 1 120 A0V 846 48 Dead Weight Tester 1/8 318 LCM 549A LCM 549B 49 Const. Fire Service 2 103 LCM LCM l 50A Post Accident Air 1 305 LCM #1 LCM #4 Sample (3 Lines) 1 305 LCM #2, 3 LCM #5, 6 ! SOB Post Accident Air 1 203 LCM #1, #2 LCM #3, 4 i Sample (2 Lines) 1 203 ! 50C Post Accident Air 1 124 LCM #1, 2 LCM #3, 4 ! 51A "A" H Recomb. Pilot & 3/4 304 S.O. IV3A LCM 1076A ' Main H Supply 2 304 S.O. IV5A LCM 1084A ! 51B "B" H Recomb. Pilot & 3/4 202 S.O. IV5B LCM 1084B i l Main H Supply 2 202 S.O. IV3B LCM 1076B 1 51C "A" & "B" H Recomb. 2 210 S.O. IV2A, LCM S.O. IVIA 0xygen Make Up 2 210 S.O. IV2B, 1080 S.O. IV1B l . l l - l Sheet 4 t
+
s'
~_
TABLE 6.1-2 DESCRIPTION OF ISOLATION VALVES TYPE "C" LEGEND: NOTE: 1. Residual Heat Removal System normally in service, testable only when all fuel is removed from R.V. A0V AIR OPERATED VALVE MOV --- MOTOR OPERATED VALVE 2. System is functional during recirculation phase CK --- CHECK VALVE f MCA. Associated valves are not considered as C.V. isolation boundaries. Valves inside contain-LCM --- LOCKED CLOSED MA:eUAL ment are open during operation and receive no l closure signal. M --- MANUAL j S.O. --- SOLENOID OPERATED 3. Functional system during postulated MCA.
- 4. There is no automatic closing valve or check valve
, in return lines of these systems. The remote oper-3 ated MOV or AOV valve is ensured closed for type C test and considered as primary isolation boundary. l
- 5. Associated'11nes and cooling units are completely missile protected. Test was deleted as per PORC.
{
- Test reinstated 1976 for verification of cooling unit integrity.
i i d I t Sher.t 5 i +
-~
TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1973 PLANT OUTAGES PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE
- 1. There was no re-NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. fueling outage 1 539 0 0 in 1973. Class C testing was 2 528 0 0 performed during the following 3 529 0 0 508 0 0 maintenance outages 7/22/73 through 4 701 In Service --
7/31/73 Not Tested 10/21/73 through SA 851A 113* 113 850A 0 0 10/27/73 1813A
- 2. Asterisks after "as 5B -851B 213* 213 85 B 0 f " leakage de-
- 0 !
action was taken at 6 720 In Service -- this time. Not Tested
- 3. Regarding superscript 7 371 0 0 after "as found leak-age rate", refer to 8 370 0 0 explanation of signifi-cant leakage covering 9A 304A 57* 57 test period.. Explana-tion number is same as 9B 304B 29* 29 superscript number.
10 3838 0 0 4. Leakage rates of tests I SA and SB are not in-cluded in total. Sheet 1
/ T TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.)
t TEST PERIOD: 1973 PLANT OUTAGES l . PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY ! AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE ! TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 11 313 0 0 12A 966A 0 0 12B 966B 0 0 I 12C 966C 0 0 l 13A 5735 374* 374 i 13B 5736 85* 85 14 1599 0 0 1598 0 0 ' 0 0 15 1597 , 16A 5738 4* 4 i 16B 5737 0 0
- 17A PT-945 0 0 PT-946 17B PT-947 0 0 i PT-948 17C PT-944 0 0 Sheet 2
, PT-949 PT-950
- n. gg .
t TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1973 PLANT OUTAGES PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. ce/ min. 1 18A 862A 8* 8 864A (water) (water) 18B 862B 11,250( ) 7 864B (water) (water) 19 889A 0 0 l 889B 870A 0 0 870B 879 4 20 1787 0 0 1786 0 0 i 1713 21 1789 0 0 22 1721 0 0 1003A 0 0 l 1003B j 23 1728 0 0 1723 0 0 24 813 0 0 814 0 0 i Sheet 3
n, ( TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE (cc/ min at 60 psig.) TEST PERIOD: 1973 PLANT OUTAGES PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 26 750A 0 0 27 750B 0 0 28 759A 0 0 29 759B 0 0 30 743 0 0 745 0 0 32 5393 3* 3 5392 0 0 33 7227 0 0 7141 0 0 34 7970 0 0 7971 0 0 35 5870 275* 275 5869 36 5878 187* 187 37 Tests 37 & 38 Not Unit Cooling Tested Colls are considered as 38 primary boundary Not Tested Sheet 4
~ "
)
s 1 TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) , TEST PERIOD: 1973 PLANT OUTAGES t PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 39 5024 0 0 5021 0 0
- 40 6151 0 0 6165 0 0 6175 0 0 6152 0 0 42 Flange 0 0 7444 0 0 ,
]. 43 Flange 0 0 7443 0 0 44 Flange 0 0 7445 0 0 , 45 3 Separate 0 0 7448 0 0 ) Lines with 7452 } Tubing Cap 7456 on each
- line 46 846 0 0 48 549A 0 0 549B 0 0 49 LCM #1 0 0 LCM #2 0 0 i
SOA LCM #1 0 0 LCM #4 0 0 LCM #2 LCM #5 LCM #3 . LCM #6 Sheet 5 ( 4
(-\ TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) OUTAGES i TEST PERIOD: PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY
- AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS
, VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE ' TEST OR RATE RATE OR RATE RATE ) NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 50B LCM #1 0 0 LCM #3 'O O i LCM #2 LCM #4 i
- 50C LCM #1 0 0 LCM #3 0 0 LCM #2 LCM #4
- 51A IVS 0 0 LCM #2 0 0 j IV3 LCM #1 j 51B IV5B 0 0 LCM #2 0 0 j IV3B 23* 23 LCM #1 l SIC IV2A 0 0 lVIA 0 0 lV2B IV1B 17* 17 i
! TOTALS 12,292 1,052 17 17 i i 1 NOTE: Tests 5A and 5B not included in totals. 1 { , Sheet 6 i 1 i
i TABLE 6.1-3 ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1973 PLANT OUTAGES , 1 EXPLANATION OF SIGNIFICANT LEAKAGE l Explanation 1 (Test 18B) Check Valve 862B. B Containment Spray Hec 'er (Line size 6 inch). The spray header is maintained with a minimum of 70 gallons of water above check valve flapper; this is assured monthly during the pump surveillance test. At the exhibited leakage rate, with the continuance of 60 psig driving pressure and maintained open, upstream atmospheric vent, the line would have been voided of this minimum fluid inventory in 24 minutes. . Containment spray is a functioning system during the postulated MCA, automatically operating upon containment pressure reaching h 30 psig. When operating, the spray pump delivers RWST fluid through the check valve to the respective spray ring inside
.e g containment.
The check valve was dismantled, internals cleaned and flapper and seat lapped in. A satisfactory retest was performed on 11-7-73
> with an indicated leakage rate of 7 cc/ min (water) remaining.
l O
,1.J Sheet 7 l
1
- - - - - - - - r.v. , u - . . , y. .- y ,._m,,y,.___ y.__, , _..,_ -,__, ,... y_,, e , ___3 ,_ -
m . Y. U TABLE 6.1-4 ISOLATION VALVE LEAKAGE RATE (cc/ min, at 60 psig.) TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THROUGH 4/9 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY + AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKI.GE LEAKAGE TEST OR RATE RATE OR RATE RATE 1. Asterisks ofter "As ' NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. Found" leakage de-1 539 0 0 notes that no repair action was taken at 2 528 0 0 this time. 3 529 0 0 508 0 0 2. Regarding superscript ' after "As Found" leak-4 701 In Service age; refer to explana-Not Tested tion of significant Icakage covering test SA 851A 21* 21 850A 0 0 period. Explanation number is same as ! 1813A superscript number. 5B 851B 204* 204 850B 33* 33 1813B 6 720 In Service Not Tested 7 371 0 0 i 8 370 0 0 9A 304A 146* 146 } 9B 304B 23* 23 10 383B 0 0
- i. 11 313 0 0 l Sheet 1
(*\ N. s i TABLE 6.1-4 g A. ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) l TEST PERIOD: 1974 REFUELING OUTAGE I 1/1 THROUGli 4/9
, PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY I AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE Test 15: Slight lenk-TEST OR RATE RATE OR RATE RATE age at valve stem pack- 7 NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. ing. Tightened packing l 12A 966A 0 0 gland nuts. Leakage reduced to zero. r 1 12B 966B 0 0 1 Test 16A: Leakage at 12C 966C 0 0 valve stem packing. l 2
Tightened packing gland ! 13A 5735 374(1) 0 nuts. Leakage reduced E **'
- s 13B 5736 85(2) 0 ,
14 1599 0 0 1598 0 0
+
1 15 1597 2 0 ; i 16A 5738 30 0 l i l 16B 5737 0 0
- 17A PT-945 0 0 j PT-946 !
17B' PT-947 0 0 ! PT-948 17C PT-944 0 0 l PT-949 . PT-950 Sheet 2 T
i TABLE 6.1-4 e ISOLATION VALVE LEAKAGE RATE (cc/ min, at 60 psig.) TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THROUGH 4/9 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE Test 34: (A0V 7970) ' V s 1ccated in-NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. side containment and 18A 862A 12* 12 is flanged to pipe 864A (water) (water) penetrating containment. , Tightened all flange ^ 18B 862B 0 0 bolts and reduced Icakage 864B to a cc/ min. Secondary isolation valve (outside 19 889A 0 O containment) indicated 889B zero leakage. This line t discharges into auxiliary ! 870A 0 0 building Hepa and charcoal i i 870B filters. Flange bolts to be '- 879 tightened more securely on ' next test with proper . 20 1787 0 0 1786 0 0 wrenches available. 1713 , Test 34 & 35: "As Found" 21 1789 0 0 leakage column reficcts l the highest leakage rate 22 1721 0 0 1003A 0 0 resulting from tests per-1003B formed during the refuel-l ing outage test period. 23 1728 0 0 1723 0 0 24 813 0 0 814 0 0 1-26 1750A 0 0 Sheet 3 I
f
. {
(. TABLE 6.1-4 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THROUGH 4/9 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY
' AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE ) TEST OR RATE RATE OR RATE RATE "As Left" Icakage col-NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. umn reflects the known 1
leakage from last test 27 750B 0 0 performed during the refueling utage test , 28 759A 0 0 period, prior to plant startup. i 29 759B' 4* 4 4 30 743 0 0 ( i 30 745- 0 0 32 5393 0 0 5392 0 0 h 33 7227 3* 3 7141 0 0 i 34 7970 25 8 7971 4 35 5870 78,000(3) 381 5869 200 0 ' -' '36 5878 78,000(4). 161 5879 200 0 d
', ^
37 Tests 3'i & 38' 'Not ---
- Unit Cooling Tested Coils are
! ' considered as
' 38_ primary boundaryNot ---
Tested , Sheet 4
)
I
J J i TABLE 6.1-4 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) - TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THROUGH 4/9 1 i PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY ' AS LEFT COMMENTS AS FOUND AS LEFT AS FOUND VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE j TEST OR RATE RATE OR RATE RATE l NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. . 39 5024 0 0 5021 6* 0 i i 40 6151 0 0 6165 0 0 40' 6175 0 0 6152' 394(5) 0 42 Flange 0 0 7444 0 0 43 ' Flange 0 0 7443 G 0 j 44 Flange 0 0 7445 0 0 j 45 3 Separate 0 0 7448 0 0 . Lines with 7452 j tubing cap 7456 1 on each line. ! 46- 846 0 0 48 549A 0 0 549B 0 0 49 LCM #1 7* 7 LCM #2 0 0 50A LCM #1 0 0 LCM #4 0 0 LCM #2 LCM #5 LCM #3 LCM #6 i 50B ~ LCM #1 0 0 LCM #3 0 0 Sheet 5 3 LCM #2 LCM #4 .
i TABLE 6.1-4 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1974 REFUELING OUTAGE 1/1 DIROUGH 4/4
. PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE No. . LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE ,
l TEST OR RATE RATE OR RATE RATE NO. BOUNDAR) _, _cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 50C LCM t 0 0 LCM #3 0 0 LCM *! O O LCM #4 0 0 51A IV3A 0 0 LCM #2 0 0 IV5A 3* 3 LCM #1 0 0 51B IV5B 0 0 LCM #2 0 0 IV3B 37* 37 LCM #1 0 0 51C IV2A 0 0 1VIA 20* 20 IV2B 0 0 IV1B 50* 50 TOTALS 156,751 785 903 103 NOTE: Tests SA & 5B not included in totals. Sheet 6 l
l l TABLE 6.1-4 1 ISOLATION VALVE LEAKAGE RATE ( TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THRU 4/9 EXPLANATION OF SIGNIFICANT LEAKAGE Explanation 1 (test 13A) A0V 5735 "A" Steam Generator Sample (3/8 inch line). Leakage was through valve plug and seat. Lap-in of plug and seat proved successful. Retest of 4-3-74 indicated zero leakage rate. Explanation 2 (test 13B) A0V 3736 "B" Steam Generator Sample (3/8 inch line). Leakage was through valve plug and seat. Lap-in of plug and seat proved successful. Retest of 4-3-74 indicated zero leakage rate. Explanation 3 (test 35) A0V 5870 Containment purge supply dampers (48 inch). A0V 5869 Excessive leakage was observed from test volurae bounded by dampers 5870 (inside containment) and f
- 5869 (outside containment). A 60 psig pressure decay of volume indicated approximately two psig
'M loss of pressure in one minute or a calculated leakage rate of 78,200 cc/ min.
At the time of leakage discovery, refueling operations had not yet commenced. Leakage past each damper was assessed visually (soap bubble solution). 200 cc/ min was allocated to damper 5869 and the remainder to damper 5870. Virtually, all leakage from this volume was into containment atmosphere. Communicating leak-age between containment atmosphere and outside building atmosphere would be governed by the leakage past the lesser leaking damper. Corrective actions were effected over 2-20-74 and 2-21-74 by vacuum cleaning the test volume and adjusting both inside and outside damper's adjustable "0" ring rubber seats. Retest on 2-21-74 indicated a remaining leakage rate of 1298 cc/ min from test volume; at this point in time it was deemed that leakage had been sufficiently reduced to permit refueling operations to begin. On 4-4-74, additional adjustment was made on the inside containment damper 5870. Leakage was also detected at the flanged section containing damper 5870; all flange bolts were securely tightened. Sheet 7 l
TABLE 6.1-4 ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THRU 4/9 I b g EXPLANATION OF SIGNIFICANT LEAKAGE A retest on 4-5-74 indicated remaining leakage from test volume of 381 cc/ min. The plant PORC Committee decided that for better awareness of the condition of these dampers, testing of this volume will be at more frequent intervals of approximately two to three months. Subsequent test dates and results are as follows: 5-9-74 ------ 256 cc/ min 11-11-74 ------ 325 cc/ min 6-6-74 ------ 145 cc/ min 2-13-75 ------ 258 cc/ min 9-3-74 ------ 131 cc/ min Explanation 4 (test 36) A0V 5878 Containment Purge Exhaust Dampers (48 inch) A0V 5879 Excessive leakage was observed from test volume bounded by dampers 5878 (inside containment) and 5879 (outside containment). A 60 psig pressure decay of volume indicated approximately two psig loss of pressure in one minute. This is essentially
'~ [ ) the same rate of loss as that which was observed
-\ ,_/ for test volume of explanation 3 above. The plant conditions at this time were the same. Visual assessment (soap bubble solution) disclosed that the majority of leakage from this volume was past damper 5878 into containment atmosphere. Blow-by passed i the outside damper 5879 was minimal. Corrective action was the same as those previously discussed in Explanation 3. Retest on 2-21-74 indicated a remaining leakage rate from test volume of 161 cc/ min.
Subsequent test dates and results for this volume are as follows: 4-4-74 ------ 226 cc/ min 5-9-74 ------ 194 cc/ min 6-6-74 ------ 175 cc/ min 9-4-74 ------- 338 cc/ min 11-11-74 ------ 220,794 cc/ min
- Retest 11-11-74 ------ 164 cc/ min 2-13-74 ------ 169 cc/ min
- Plant was at cold shutdown status for Steam Generator tube inspection. During the process of educting gas from the
, Reactor Vessel head vent, the eductor hose had been draped over fe ~s the open inside containment damper 5878. Upon damper closure,
-(N- ) the hose became pinched between damper and "0" ring seat,
/ preventing adequate seating surface make-up. After removal of hose, adjustment was made on the damper's "O" ring adjustable i rubber seat. Retest same day indicated leakage rate from volume of 164 cc/ min. 3 Sheet 8 i
l 1 1 TABLE 6.1-4 ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1974 REFUELING OUTAGE 1/1 THRU 4/9 4 EXPLANATION OF SIGNIFICANT LEAKAGE Explanation 5 (test 40) Manual Condensate return from C.V. space heaters (one inch). Valve This valve is normally locked closed at power opera-6152 tion. Leakage was through ruptured valve diaphragm. This valve is considered as the secondary boundary; l at the time of leakage discovery the primary boundary valve for this line exhibited zero leakage rate. The valve diaphragm was replaced and retesting on 4-1-74 indicated zero leakage. i 1' t j e Sheet 9 l
.i l !
/ 'T TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.)
TEST PERIOD: 1975 REFUELING OUTAGE 3/11 TilROUGH 5/10 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST. OR RATE RATE OR RATE RATE 1. Asterisks after NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. "As Found" Icakage
- 1 539 0 0 denotes that no re-pair action was 2 528 0 0 taken at this time.
a i 3 529 0 0 508 0 0
- 2. Regarding superscript i 4 In service ---
after "As Found" Not Tested leakage, refer to ex-l planation of signifi-i SA 851A 188* 188 850A 0 0 cant leakage covering 1813A test period. Expla-nation number is same 5B 851B 291* 291 850B 10* 10 as superscript number. 1813B 0 0 6 720 In service Not Tested i 7 371 0 0 1 8 370 0 0 j 9A 304A 336* 336 9B 304B 0 0 1 10 383B 0 0 i i Sheet 1
TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE (cc/ min, at 60 psig.) TEST PERIOD: 1975 REFUELING OUTAGF 3/11 THROUGH 5/10 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 11 313 0 0 12A 966A 0 0 12B 966B 0 0 12C 966C 0 0 13A 5735 0 0 13B 5736 0 0 14 1599 0 0 1598 0 0 15 1597 0 0 16A 5738 17* 17 16B 5737 14* 14 17A PT-945 0 0 PT-946 17B .PT-947 0 0-PT-948 17C PT-944 0 0 PT-949 Sheet 2 PT-950 . l
,/ 's TABT,E 6.1-5 ISOLATION VALVE LEAKAGE RATE (cc/ min at 60 psig.)
TEST PERIOD: 1975 REFUELING OUTAGE 3/11 THROUGH 5/10 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 18A 862A 6* 6 864A (water) (water) 18B 862B 9* 9 864B (water) (water) 19 889A 0 0 889B 19 870A 0 0 870B 879 20 1787 0 0 1786 0 0 1713 21 1789 0 0 22 1721 0 0 1003A 0 0 1003B 23 1728 0 0 1723 0 0 24 813 0 0 4
- 24 814 11* 11 Sheet 3 26 750A 0 0 t
e ( N < % N \ , TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERI 0i): 1975 REFUELING OUTAGE 3/11 111 ROUGH 5/10 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY , AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE 1. Asterisks after "As NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min, cc/ min. Found" leakage de-n tes that no repair 27 750B 0 0 action was taken at this time. 28 759A 0 0
- 2. Regarding superscript 29 759B 0 0 after "As Found Leak-age", refer to expla-30 743 108* 108 nation of significant leakage covering test 30 745 0 0 period. Explanation number is same as super-32 5393 19* 19 5392 0 0 script number.
33 7227 0 0 7141 0 0
- 3. Regarding Tests 35 & 36 "As Found hakage" 34 7970 15* 15 7971 0 0 column reflects the high-est leakage rate result-35 5870 774(1} 774 5869 0 0 ing fr::,m tests performed during the refueling 36- 5878 2493(2) 232 5879 0 0 outage test period.
37 Tests 37 & 38 Not Not "As Left Leekage" column Unit Cooling. Tested Tested reflects the known leak-Coils are- age from last test per-considered as formed during the refuel-38 primary boundary Not Not ling outage test period, Tested Tested prior to plant startup. Sheet 4
~ "
s i ! TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1975 REFUELING OUTAGE 3/11 THROUGH 5/10 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY l 1 AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE i TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. , 39 5024 0 0 5021 0 0 40 6151 0 0 6165 0 0 40 6175 0 0 6152 0 0 I 42 Flange 0 0 7444 0 0 43 Flange 0 0 7443 0 0 44 Flange 0 0 7445 0 0 ] I 45 3 Separate 0 0 7448 0 0 Lines with 7452 tubing cap 7456 on each line i 46 846 0 0 i 48 549A 0 0 549B 0 0 i 49 LCM #1 18* 18 LCM #2 0 0 50A LCM #1 0 0 LCM #4 0 0 LCM #2 7* 7 , LCM #5 LCM #3 0 0 LCM #6 Sheet 5 1
O O TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) 1975 REFUELING OUTAGE TEST PERIOD: 3/11 THROUGH 5/10 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. ., 50B LCM #1 0 0 LCM #3 0 0 LCM #2 LCM #4 50C LCM #1 0 0 LC" CJ 0 0 LCM #2 LCM #4 51A IV2A 3* 3 LCM (2 0 0 1VSA 4* 4 #1 SlB IVSB 0 0 LCM #2 0 0 lV3B 27* 27 51C 1V2A 56* 56 1VIA 0 0 1V2B 0 0 1V1B 30* 30 l l TOTALS 3,917 1,655 40 40 1 L NOTE: Tests SA & SB not included in totals. l . Sheet 6
i TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1975 REFUELING OUTAGE 3/11 THRU 5/10 EXPLANATION OF SIGNIFICANT LEAKAGE l Explanation 1 (test 35) A0V 5870 Containment Purge Supply Dampers (48 inch). A0V 5869 Leakage is from test volume bounded by dampers 5870 (inside containment) and 5869 (outside containment). A 60 psig pressure decay of volume indicated a leakage rate of 774 cc/ min. At this time no attempted corrective action was undertaken, nor was blow-by past each damper assessed. For the worst condition, each damper leaking identical amounts, the communicating leakage from containment atmosphere to outside building atmosphere would be 507, of the measured total. Subsequent testing dates and leakage results for this volume are as follows: 4-28-75 ------ 509 cc/ min 6-25-75 ------ 489 cc/ min lr 9-24-75 ------ 195 cc/ min
'g' 12-18-75 ------
460 cc/ min l
- 1-8-76 ------
47,000 cc/ min j Retest 1-9-76 ------ 80 cc/ min !
- The Plant was at cold shutdown status for "B" Steam Generator tube inspection on the date this test was performed. Calculated leakage was excessive (47,000 cc/ min). A visual inspection, while volume was sub-jected to test pressure, indicated approximately equal leakage past each damper.
Corrective actions began immediately; "O" ring seats and damper sealing surfaces were thoroughly cleaned, adjustments were made on each damper's "O" ring adjustable rubber seats. Retesting of the test volume, af ter cleanup and adjustment, still resulted in a leakage rate of 1,500 cc/ min. Further checking of the penetrating assembly disclosed leakage existing at approximately 5 upper flange bolts on the spacer flange section associated with the containment side damper 5870. Flange bolts were securely tightened, at and adjacent to the leaking area. ( Sheet 7
l TABLE 6.1-5 ISOLATION VALVE LEAKAGE RATE
}
V TEST PERIOD: 1975 REFUELING OUTAGE 3/11 THRU 5/10 EXPLANATION OF SIGNIFICANT LEAKAGE i A satisfactory retest was performed on 1-9-76, with results indicating a leakage rate of 80 cc/ min. Explanation 2 (test 36) A0V 5878 Containment Purge Exhaust Dampers (48 inch). A0V 5879 Leakage was from test volume bounded by dampers 5878 (inside containment) and 5879 (outside containment). A 60 psig pressure decay of volume indicated a leakage rate of 2493 cc/ min. Although this leakage rate was not extremely high, it was deemed prudent to attempt a reduction of this rate. Adjustments were made on each damper's "0" ring adjustable rubber seat. A retest on 4-28-75 showed that results were satisfactory with leakage rate r^ remaining from the established test volume of 232 cc/ min. Subsequent testing dates and leakage results for this volume are as follows: 6-25-75 ------ 178 cc/ min 9-24-75 ------ 248 cc/ min 12-23-75 ------ 511 cc/ min
- 1-8-76 ------
65,130 cc/ min Retest 1- 9-76 ------ 325 cc/ min
- Ihe explanation for the occurrence on 1-8-76 is similar to that for the purge supply dampers. Retesting on 1-9-76 indicated a remaining leakage rate from the purge exhaust between damper volume of 325 cc/ min. A visual assessment at this time allocated 100 cc/ min as leakage pst the inside containment damper 5878, and 225 cc/ min leakage past the outside containment damper 5879.
< N Sheet 8
f% i TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THROUGH 4/6 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE 1. Asterisks after "As cc/ min. cc/ min. cc/ min. Found" leakage de-NO. BOUNDARY cc/ min. BOUNDARY notes that no repair 1 539 0 0 action was taken at this time. 2 528 0 0
- 2. Regarding superscript 0 508 0 0 3 529 0 after "As Found" leak-age, refer to explana-4 701 In Service tion of significant Not Tested leakage covering test 15* 15 Period. Explanation 5A 851A 1232(1) 2156 850A number is same as 1813A superscript number.
SB 851B 213* 213 850B 19* 19
- 3. Test 12A: Stem pack-f ing exhibited slight Icak - tightened gland 6 720 In Service nuts and reduced leak-t ested age to zero.
7 371 0 0 f ' 8 370 0 0 4 9A 304A 148* 148 98 304B 0 0 10 383B 78* 78 Sheet 1
s O O , TABLE 6.1-6' ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) l TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THROUGH 4/6 i PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 11 313 0 0 12A 966A 15 0 12B 966B 0 0 12C 966C 0 0 i 13A 5735 0 0 i 13B 5736 22.6* 22.6
- 14 1599 21* 21 1598 0 0 1
15 1597 0 0 16A 5738 23* 23 16B '5737 397(2) 20 17A PT-945 0 0 PT-946 17B PT-947 0 0 4 PT-948 17C PT-944 0 0 Sheet 2 PT-949 PT-950 4
j' . \ J C 'l 1 TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE (cc/ min. at 60 psig.) 1976 REFUELING OUTAGE TEST PERIOD: 1/29 THROUGH 4/6 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE OR RATE RATE OR RATE RATE TEST , NO. BOUNDARY cc/ min. cc / min . BOUNDARY cc/ min. cc/ min. I 18A 862A 0 0
- 864A 18B 862B 7570( -
0 I 864B (water) i 19 889A 0 0 889B 19 879 0 0 ! 870A j 870B 20 1787 0 0 1786 0 0 ! 1713 i 21 1789 0 0 22 1721 0 0 1003A 0 0 1003B i 23 1728 0 0 1723 0 0 { f I 24 813 -0 0 5 24 814 0 0 Sheet 3 ') S 4
() I \ U s/ ' TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE (cc/ min, at 60 psig.) TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THROUGH 4/6 , f l PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS 4 VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE 1. Regarding Tests 35 i NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. & 36 "As Found" 26 750A 0 lea age c lumn re-0 flects the highest 27 750B -53* 53 leakage rate result-
! ing from tests per-28 759A 45* f rmed during the 45 ; refueling outage 29 759B 0 0 test period.
30 743 0 0 2. "As Left" leakage } column reflects the 4 -30 745 11* 11 known leakage from last test performed 32 5393 0 0 5392 0 during the refueling ] 0 outage test period, j 33 7227 0 0 7141 Prior to plant start-0 0 up. 34 7970 0 0 7971 0 0 35 5870 Excessive (4} 425 5869 0 0 36 5878 9801(5) 315 5879 0 0 t 37 Test 37 & 38 0 0 ^ Unit Cooling .
- 38. Coils are 0 0 considered as
- Primary boundary .
Sheet 4 i l 1
.f %>
)
C., J TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE (cc/ min, at 60 psig.) TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THROUGH 4/6 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE Regarding Tests 35 1* TEST. OR RATE RATE OR RATE RATE ,," .. NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min, cc/ min. a o re-flects the highest leakage rate result-ing from tests per-formed during the 39 5024 0 0 5021 0 0 refueling outage l test period. 40 6151 0 0 6165 0 0
- 2. "As Left" laakage 40 6175 0 0 6152 0 0 column reflects the known leakage from 42 Flange 0 0 7444 0 0 last test performed I during the refueling 43 Flange 0 0 7443 16* 16 outage test period, prior to plant start-44 Flange 0 0 7445 0 0 up.
45 3 Separate- 0 0 7448 0 0 Lines with 7452 Tubing Cap 7456 on each line 46 846 9* 9 48 549A 0 0 549B 0 0 49 LCM #1 0 0 LCM #2 0 0 SOA LCM #1 0 0 LCM #4 11'* 11 Sheet 5 LCM #2 LCM #5 0 0 LCM #3 LCM #6 0 0
m _ . . _ _ _ _ _ _ _ _ . _ _ - _ _ ._ _ . _ _ - .
/
)
\ ,
L TABLE 6.1-6 t ISOLATION VALVE LEAKdGE RATE (cc/ min. at 60 psig.) TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THROUGH 4/6 PRIMARY ISOLATION BOUNDARY SECONDARY ISOLATION BOUNDARY
- ) AS FOUND AS LEFT AS FOUND AS LEFT COMMENTS '
VALVE NO. LEAKAGE LEAKAGE VALVE NO. LEAKAGE LEAKAGE TEST OR RATE RATE OR RATE RATE NO. BOUNDARY cc/ min. cc/ min. BOUNDARY cc/ min. cc/ min. 50B LCM #1 0 0 LCM #3 0 0 LCM #2 LCM #4 50C- LCM #1 0 0 LCM #3 0 0 LCM #2 LCM #4 i 51A IV3A 13* 13 LCM #2 0 0 l IV5A 13* 13 LCM #1 i 51B IV5B 39* 39 LCM #2 0 0 1V3B 0 0 LCM #1 51C 1V2A 0 0 IVIA Combined Combined i IV2B IV1B 15 15 l t TOTALS Excessive 1235 76 76 NOTE: Test 5A & 5B not included in totals. Sheet 6 i - )
TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THRU 4/6 EXPLANATION OF SIGNIFICANT LEAKAGE Explanation 1 (test SA) MOV 851A Containment Sump Recirculation (8 inch). Leakage was through the seat and disc interface of valve 851A (valve is double disc design). This valve is located in containment Sump B and is maintained open during plant operation. This line must be functional during the Recirculation Phase of the Postulated MCA (Flooded conditions would exist at this time). The valve does not receive any isolation signal to close; remote closure of 851A would only be necessary in the event of suction line piping failure outside containment or failure of the associated RHR pump's seal. This valve is not considered to be a containment isolation valve as defined by 10 CFR 50, Appendix J, Section II H, items 1 through h, or further defined in Section III A Item Id.
,- f s
(* Cleanup and repair attempts (lapping in of discs) vere unsuccessful. Final leakage past the valve seat and discs was approximately a factor of two greater than that leakage which was initially measured. The valve manufacturer was contacted for advice regarding this valve. As a consequence, new valve discs are on order, and the manufacturer vill fabricate a lapping tool for in place use on the valve seating rings. At the next convenient plant outage, after receipt of parts and tool, further endeavor vill be made to reduce the disc-seat interface leakage. The final leakage rate was determined by a pressure decay of the test volume, bounded by an installed test flange and the closed valve itself. At best, the leakage calculated for this valve is somewhat of an average rate as the volume beyond 851A is closed; buildup of the entrapped pressure has the effect of reducing the indicated leakage rate from the volume under-going tedting.
/^N i
Sheet 7
TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THRU 4/6 O
\
EXPLANATION OF SIGNIFICANT LEAKAGE Explanation 2 (test 16B) A0V 5737 "B" Steam Generator Blowdown (2 inch line). Leakage was through seat and plug. This valve is in the secondary side system and is not in contact with primary system fluid. Isolation capability is provided, however, in the event of Steam Generator tube rupture which would permit primary to secondary fluid communi-cation. A new valve stem and seat was installed in valve body. Retesting of this valve after replacements of parts still indicated excessive through leakage. The valve seat is back velded in the valve body and apparently became varped as a result of this velding process. Steam and seat were then lapped in; a satisfactory retest was performed on 3-31-76 with a remaining leakage rate of 20 cc/ min existing. Explanation 3 (test 18B) Check "B" Containment Spray Header (6 inch line). Valve , r" ' 862B This system was previously explained under explanation
/ 1, test period 1973 plant outages.
After lapping in the valve's flapper and seat, a satis-l factory retest was performed with indicated results being 0 cc/ min leakage. . Explanation h (test 35) A0V 5870 Containment Purge, Supply Dampers (h8 inch). A0V 5869 The volume bounded by dampers 5870 (inside containment) and 5869 (outside containment) was subjected to a 60 psig pressure decay test on 2-5-76. The concerned volume could not be pressurized beyond 10 psig through the 1 inch 60 psig air supply line. This check was to assess the leak tightness of these dampers prior to performing the Containment Integrated Leakage Rate Test (ILRT). The outside damper was observed to be properly closed, and the switch indicating light for t'he inside damper 5870 indicated that this damper was also closed. No attempt was made to ascertain the inability to pressurize the volume at this time. On 2-7-76 pressurization of containment for the ILRT
/ I began. During containment pressure buildup it was noted
- [', . k,_,/
1 that pressure gauge monitoring the_between damper volume van increasing at the same rate as that of containment. Visual inspection (aoap solution) indicated essentially zero leakage past outside damper 5869, hence leakage from containment atmosphere to outside building atmos-phere was minimal through this path. Shoot 8
. -. - _ - , .=.
TABLE 6.1-6 ISOLATION VALVE LEAKAGE RATE
- f TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THRU 4/6 EXPLANATION OF SIGNIFICANT LEAKAGE V The ILRT was aborted after approximately 7 hours into the equalization period, due to the calculated hourly mass weight loss being substantially in excess of maximum allovable to assure acceptable ILRT test results. After depressurizing the containment building leakage evaluation was conducted on the observed leakage paths. Although the purge supply damper system was not contributing to the loss of containment weight mass, a 60 psig pressure decay was performed on this volume to locate the leakage path (s) to containment interior. Whatever difficulty prevented the pressurization of between damper volume on 2-5-76 was not prevalent at this time. 60 psig pressure was readily attained, and calculated results indicated a minimal leakage rate of 83 cc/ min from this volume.
Preparations continued in other areas to permit resumption of the ILRT, which was completed success-fully on 2-11-76. The purge supply volume was again tested on 2-28-76
// prior to commencing refueling operations; containment d / purge was in effect during inspection of the Steam Generators. Upon securing the dampers closed and applying air pressure it again was noted that pressuri-zation beyond 10 psig could not be attained. Inspection inside containment disclosed that da=per 5870 was not completely seating. Repeated operation of this damper shoved very , jerky motion, and in'the opening direction loud noises were emitted from the spring cannister assembly.
Adequacy of damper closure was variable, which accounted for the randomness of establishing a sufficiently tight test boundary. When the spring cannister piston shaft was fully extended, heavy scoring was noted over approximately 7 inches of shaft length closest to the cannister head plate. Corrective action consisted of smoothing the scored shaft, polishing and dry lubricating both the spring cannister piston shaft and power air cylinder piston shaft. Subsequent damper operation was greatly improved, noise level from the spring cannister was reduced and repeatability of closure degree was con-sistent. A satisfactory retest was then performed with
,(p}- calculation indicating an acceptable leakage rate of
(\ / h37 cc/ min from the established test volume. Sheet 9
TABLE 6.1-6 m ISOLATION VALVE LEAKAGE RATE TEST PERIOD: 1976 REFUELING OUTAGE 1/29 THRU 4/76 EXPLANATION OF SIGNIFICANT LEAKAGE The manufacturer (Henry Pratt Company) was contacted to assess the problems. He has in-spected the valve and found it acceptable for continued use and has recommended some addi-tional work to be performed at a convenient time. On 3-31-76, prior to plant heat up, a retest of this volume was performed; resulting calculations indicated an acceptable leakage rate of 425 cc/ min. Explanation 5 (test 36) A0V 5878 Containment Purge Exhaust Dampers (h8 inch). , A0V 5879 The volume bounded by dampers 5878 (inside containment) and 5879 (outside containment) was initially subjected to a 60 psig pressure decay check on 2-5-76. This check was to assess the leak' tightness of the concerned C dampers prior to perforcing the containment ILRT. g '( j Results of calculations indicated a leakage rate from the test volume of h56 cc/ min. No repair attempts were undertaken at this time. Pressurization of containment for the ILRT began on
- 2/7/76. During containment pressure buildup, it was noted that the pressure gauge monitoring the between damper volume was increasing at the same rate. Visual inspection of the accessible outside damper 5879 showed several areas of rather severe leakage around the seating periphery.
As previously explained in Explanation 4 of this testing period, the ILRT was aborted. After depres-surizing containment, a 60 psig pressure decay test was performed on the purge exhaust volume, calcu3ated results indicated a leakage rate of 9801 cc/ min. Approximately 50% of the total leakage was past each damper. Adjustments were performed on each damper's adjustable "0" ring seat and the follow up retest indicated a remaining leakage rate of 4h7 cc/ min from this volame. m Subsequent testing dates and leakage results of this
) volume are as follows:
'(' L) 2-28-76 - 31h ec/ min 3-31-76 315 cc/ min Sheet 10
l l i I t I I i l R. E. GINNA i NUCLEAR POWER STATION i i I ? i i REACTOR CONTAINMENT BUILDING LEAKAGE RATE i TESTS WHICH FAILED TO MEET ACCEPTANCE CRITERIA
; DURING THE PERIOD 1973 TO 1976 i
4 I i 1 i 1 1 5 1 i 1 l l i 4 ROCHESTER GAS & ELECTRIC i CORPORATION i t I l
TABLE OF CONTENTS SECTION PAGE
1.0 INTRODUCTION
1 2.0 ABORTED INTEGRATED LEAK RATE TEST 2 2.1 TEST PERFORMANCE 2 2.2 TEST RESULTS 3 2.3 CONDITIONS CONTRIBUTING TO TEST FAILURE 4 2.3.1 Isolation Valve Leaks 4 2.3.2 Ingassing 5 2.3.3 Conclusions 6 2.3.4 Schedule for Retesting 6 3.0 TYPE B AND C TESTS WHICH FAILED TO MEET ACCEPTANCE CRITERIA 8 3.1 PERSONNEL HATCH PERSONNEL ACCESS LOCK 9 3.2 PURGE SUPPLY AND EXHAUST DAMPERS 10 3.2.1 Leakage During the Refueling Outage of 1974 10 3.2.2 Leakage During Steam Generator Inspection of November, 1974 12 3.2.3 Leakage During Steam Generator Inspection of January, 1976 13 3.2.4 Leakage During the Refueling Outage of 1976 14 l
4.0 CONCLUSION
S 18
l
1.0 INTRODUCTION
i l Included in this report are leakage rate test results from 4 i Type A, B and C tests that failed to meet their respective acceptance criteria since the last integrated leak rate 1 test in October, 1972. Included are the results of those t tests as well as the corrective action which was taken. i 1 1 l l t f i I i i l i i j l l i 'l I I
- 1
i 2.0 ABORTED INTEGRATED LEAK RATE TEST 2.1 TEST PERFORMANCE
\
Ginna Station was shut down January 29, 1976 because of turbine failure and shortly thereafter the shutdown was declared to be the annual maintenance and refueling out-age. Because of the impending Type A containment test, the normal Type B and C leakage tests were not performed so as to test the containment in as close to "as is" condition as possible. 1 Preparation for, and procedures used to perform the inte-grated leak rate test, including the aborted test, at Ginna Station during February, 1976 are outlined in Section
-s s 4.2 of the Reactor Containment Building Integrated Leak t i \[ Rate Test Report. The acceptance criteria, instrumentation and error analysis are also described in that report and apply to the aborted test. No recalibration of instru-mentation was performed between the aborted test and the subsequent successful test.
The containment was pressurized to 35 psig and the record-ing of hourly containment pressures, temperatures and dew-point temperatures was begun. The pressure buildup between certain inner and outer containment isolation boundaries were also periodically monitored as were open vents out-side containment to detect any existing leaks. /O_ 4 _
2.2 TEST RESULTS After seven hours of data collection the decision was made to abort the test based upon the calculated leakage rates for the period and the identified leaks. It was recognized that because this period included little time when con-tainment ambient conditions were stabilized that the cal-culated leakage rate probably was not representative of the true containment leak rate. The least squares leakage rates calculated by the Mass Plot Method and Total Time Method for the seven hours were 0.435 and 1.057 percent per day respectively. The containment temperature was dropping steadily during the first three hours of this period and thus a better estimate of the containment leak rate can be ['
' s_
found by considering only the last four hours of data. These data points give leakage rates of 0.368 and 0.234 percent per day respectively for the Mass Plot and Total Time methods. Allowing only three hours for the contain-ment conditions to stabilize, however, is inadequate'to properly determine a true containment leak rate as explained i in Section 2.3.2. ! Data taken during the aborted test is given in the table at the end of Section 2. Plots of mean containment temp-erature and pressure and containment air weight are given on Figures 1 and 2. () x_-
- s. I l
2.3 CONDITIONS CONTRIBUTING TO TEST FAILURE
\
2.3.1 Isolation Valve Leaks During the seven hour period of the test and during a thorough inspection of the containment boundary after the
~
test, leakage was identified from only the following areas: (a) Purge exhaust volume (b) Check valve #1713 - Nitrogen Supply to the Reactor i Coolant Drain tank (c) Motor operated valves #813 and #814 - Reactor Support Cooling (inlet and outlet lines) After the decision to abort the test was made but while - the reactor containment building was still at the 35 psig level a leakage determination was made at the tubing vent connection upstream of check valve #1713. The leakage was above the highest range test rotometer available, how-ever, from the degree of rotometer inlet valve opening re- , quired to produce 100 percent scale indication it was judged that total flow was approximately three times the maximum range or a total of roughly 23,000 Std cc/ min. Equating this leakage to a 35 psig atmosphere the leakage would bc 6800 cc/ min. l Local testing of the purge exhaust volume (60 psig pres-sure decay) indicated a leakage of 9800 cc/ min. One half of this total volume leakage (the maximum leakage which could communicate between the building atmosphere and the outside
- O
atmosphere) was attributed to each damper. Equating this leakage to a 35 psig test atmosphere gives a communicating building leakage rate of approximately 4000 cc/ min. Local testing at the motor-operated valves indicated zero leakage. The dripping water from the open vents on these lines was attributed to leakage through system valves sep-arating the active component cooling system from the Reactor Support cooler lines. That is, the leakage came from valves separating the operating component cooling sys.em outside containment from the voided line outside containment where leakage was seen. Thus the total identified leakage from containment through r~ 1 s. check valve 1713 and the purge exhaust was approximately 10,800 cc/ min or 0.06 percent per day. The estimate of leakage through check valve 1713 was rather crude so con-fidence in the magnitude of the identified leakage is not great. 2.3.2 Ingassing - The inner surface of the containment building is lined.with i 1 a thermal insulation. Although a liner plate covers the l insulation the plate is not leak tight at many locations. It is believed that several hours are required for the con-tainment air to leak into the pores and small voids in the insulation, or to ingas. Mistaking the ingassing process I C f (s !
of the insulation for leakage will result in a higher ob-served leak rate during the hours that stabilization is ( } taking place. It was also noted that after repairs were made to the leaks identified in the aborted test and a second integrated leak rate test was begun that the leakage rate during that stabilization period was several times higher than that calculated during the subsequent 24 hour test. l 2.3.3 Conclusions ' Little can be established about the true leak rate of the containment during the aborted test. The calculated leak-age rates were in excess of the acceptance criteria. Ex-perience of the test personnel indicated that leaks ob-rQ m served during the test were large enough to result in failure of the test. Therefore, the test was aborted after seven hours when it was reasonably evident that the test would not meet the acceptance criteria, even though a true leakage rate had not been established. 2.3.4 Schedule for Retesting The thorough examination of the containment penetration boundaries revealed no structural deterioration or ab-normalities other than the leaking purge exhaust dampers and check valve 1713. All remaining portions of the con-tainment were found to be in good repair. After the iden-g tified leaks were repaired the containment leakage rate (J
i i i
)
was less than half of the allowable value. Both of the penetrations which exhibited excessive leakage are periodi-cally retested. l Therefore, the next periodic Type A retest is proposed to , I ! be performed in approximately three years in conjunction with the 10 year inservice inspection. ; )
?
4 i i I i l .l ? f s g I } 1 i i-I ' ,4 l i i
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1 l i l
/m. '
FIGURE 1 ( ABORTED TEST C REACTOR CONTAINMENT BUILDING MEAN TEMPERATURE AND PRESSURE VS TIME _
.r ..
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y Q .u.. . .
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0 . 0 1 2 3 4 5 6 7 TIME (HOURS) I s
\ l y ,/
r FIGURE 2 ABORTED TEST REACTOR CONTAINMENT CUILDING AIR WEIGHT VS TIME x ; a :.: nm:- -* r -++ A ':
. t:.~ .:n;n:. r.a.an
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REACTOR BUILDING 254.980 - 4-- ;' . i LEAST SQUARES FIT . : . a __ AIR WEIGHT -
! :4'--
4' == ' la ..: , : SLOPE = 4 6.2 LBS/HR OR - '
+- L p.
j 33 254,960
. -i - :--- -
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- i i i 0 1 2 3 4 5 6 7 l 1
TH:E (HOURS) l i
r TABLE 2.1 i GINNA STATION ) FEBRUARY 1976 LEAKAGE RATE DATA AND RESULTS ; ABORTED TEST ' I \ TIME /1100R 0600/0 0700/1 0800/2 0900/3 PI-3A, Counts 50.8970 50.7727 50.7580 50.7470 PI-3D, Counts 49.1980 49.1740 49.1600 49.1490 Avg. Press., Psia (PT) 49.7011 49.6234 49.6145 49.6037 Dewpoint Temp., 'F 1 31.5 31.5 31.5 31.0 2 34.0 34.0 33.5 33.5 3 34.5 34.5 34.0 34.5 4 33.5 33.0 33.0 33.0 5 33.5 33.0 33.5 33.0 6 33.5 33.5 33.0 33.0 Avg. Dewpoint Temp., 33.2 33.1 32.9 32.8
*F, (Note 1)
Water Vapor Press., Psia 0.0930 0.0926 0.0919 0.0915 - Drybulb Temp., *F 1 50.0 49.9 49.5 49.7 2 50.2 50.0 50.0 49.8 3 50.0 50.0 49.7 49.7
# 4 49.8 49.7 49.5 49.5 5 49.0 48.8 48.5 48.4 \-
6 49.5 49.5 49.4 49.2 7 49.7 49.5 49.5 49.4 . 8 49.2 49.2 49.0 48.9 9 49.6 49.5 49.5 49.3 , 10 50.2 50.1 50.1 49.9 11 49.9 49.6 49.6 49.4 12 49.8 49.7 49.6 49.4 13 49.4 49.1 49.0 48.8 14 48.9 48.0 48.5 48.5 15 49.6 49.5 49.4 49.5 16 48.8 48.7 48.5 48.5 17 49.3 49.1 49.0 48.9 18 49.2 48.9 48.6 48.7 19 48.8 48.7 48.5 48.4 20 48.8 43.7 48.5 48.4 21 -- -- -- -- 22 49.0 48.7 48.5 48.5 23 48.7 48.5 48.4 48.2 24 49.1 48.9 48.6 48.6 Avg. Tenp., *R (T) (Note 2) 509.0 508.8 508.7 508.6 Weight of Air, LBS. 255,175 254,903 254,885 254,882 s s g( Leakage Rate, Total N/.T 2.556 1.362 0.919
\. Time ilethod, t/ day Sheet 1
TABLE 2.1 j GINNA STATION FEBRUARY 1976 LEAKIsGE RATE DATA AND RESULTS ABORTED TEST
-w TIf!E/IIOUR 1000/4 1100/5 1200/6 1300/7
( ) PI-3A, Counts 50.7360 50.7310 50.7243 50.7170 PI-3B, Counts 49.1390 49.1340 49.1280 49.1210 Avg. Press., Psia (P ) 49.5932 49.5883 49.5818 49.5749 T Dewpoint Temp., *F 1 30.5 30.5 30.5 30.5 2 33.0 33.0 33.0 33.0 3 34.0 34.0 34.0 34.0 4 32.5 33.0 32.5 32.5 5 32.5 33.0 32.5 32.5 6 32.5 33.0 32.5 33.0 Avg. Dewpoint Temp., 32.3 32.6 32.3 32.4
*F, (Note 1)
Water Vapor Press., Psia 0.0897 0.0908 0.0897 0.0900 Drybulb Temp., *F 1 49.6 49.6 49.8 49.7 2 49.8 49.8 49.8 49.9 3 4 49.6 49.6 49.7 49.6 49.5 49.5 49.5 49.4 5 48.4 6 48.4 48.4 48.4 48.1 49.2 49.5 49.5 [jN
\ 7 49.4 49.4 49.4 49.4 8 48.7 48.8 48.8 48.7 9 49.2 49.4 49.2 49.2 10 49.9 49.9 49.7 49.7 11 49.5 49.2 49.2 49.2 12 49.4 49.4 49.4 49.5 13 48.9 48.8 48.8 48.8 14 48.5 48.5 48.4 48.5 15 48.2 49.2 49.1 49.1 16 48.5 48.4 17 48.4 40.4 48.7 48.8 48.7 48.7 18 48.6 48.5 48.5 48.6 19 48.5 48.4 48.2 48.4 20 48.4 48.2 48.2 48.3 21 -- -- -- __
22 48.5 48.5 48.5 48.5 23 '48.1 48.2 48.1 48.3 24 48.5 48.5 48.5 48.5 Avg. Temp., *R (T) (Note 2) 508.5 508.6 503.6 508.6 Weight of Air, LBS. 254,887 254,810 254,77S 254,741 Leakage hate, Total 0.67G 0.684 0.621 0.582 Time Method, t/ day V NOTE 1: For calibration purposes 1.l*F is subtracted from the temperature sum. NOTE 2: For calibration purposes 2.2*F is subtracted from the temperature sum. Sheet 2
f 3.0 TYPE B AND C TESTS WHICH FAILED TO MEET ACCEPTANCE CRITERIA N There have been five occasions since the Integrated Leak Rate Test in October, 1972 on which the total leakage from type B and C penetrations has exceeded the technical speci-fication limit of 0.45 La. In each case the leakage has been found during a reactor shutdown and has been attributed to events related to the shutdown. In no case has leakage in excess of the limit been discovered or known during power operation. The excessive leakage pathways identi-fied have been from the personnel hatch personnel access lock and the purge supply and exhaust dampers. After . repeated excessive leakage from the purge supply and ex-haust dampers the test frequency for these components has y
#O been increased.
De. tails of those tests which failed to meet the acceptance criteria and the conditions contributing to those failures are given in the following sections. 1 i s 1 l
r 3.1 PERSONNEL HATCH PERSONNEL ACCESS LOCK ( ) On March 31, 1976 a leakage rate of 26,814 cc/ min (0.70 La) was discovered from the personnel hatch personnel ac-cess lock. The plant was shut down for refueling at the I time this leakage was discovered. Corrective actions were , undertaken immediately. The outer door upper handwheel packing gland nut was found to have backed off completely. The nut was threaded back into the gland and securely
; tightened. A retest on April 1, 1976 indicated a remaining leakage rate of 64 cc/ min from the lock volume.
l Whenever larger than normal leakages have been experienced i i from the access lock, it has been the result of the packing gland nuts loosening or turning free from the packing gland. ( i- The high leakage rates can be keyed to plant outages where, due to extensive maintenance work, heavy personnel traffic i into and out of containment subject the access lock doors to many cycles of operation.
- The inaccessibility of these packing gland nuts dis-l courages a routine tightening after extended plant outages.
- Space is limited for adjustment and virtually precludes j the use of any other but a chain wrench, which at best i
only permits limited rotation with each bite. At the next dismantling for overhaul of the door assemblies, new nuts will be installed which will permit the use of a spanner
)
's wrench. In addition, keepers will be placed to prevent It the nuts from backing off.
f l l l l l 3.2 PURGE SUPPLY AND EXHAUST DAMPERS l On four separate occasions excessive leakage has been (V) identified from one or both of the volumes formed by the 48 inch purge supply and exhaust dampers. Excessive leak-age has not been found except when the plant is in the cold or refueling shutdown condition. Except for one case of mechanical interference, the leakage past the dampers is believed to be caused by the change in temperature resulting from the transition from power operation to cold shutdown. 3.2.1 Leakage During the Refueling Outage of 1974 A 60 psig pressure decay of the volume formed by the purge supply dampers during the 1974 refueling outage indicated
, ,-m approximately a two psig loss of pressure in one minute, f I ~ \ ,/ or a calculated leakage rate of 78,000 cc/ min (2.05 La).
Refueling operations had not yet commenced. Leakage past each damper was assessed visually (soap bubble solution). 200 cc/ min was allocated to the outside damper and the re-mainder to the inside damper. Virtually all leakage from this volume was into containment atmosphere. Communicating leakage between containment atmosphere and outside building atmosphere is governed by the leakage past the lesser leaking damper. Corrective actions included vacuum clean-ing the test volume and adjusting both the inside and the i adjustable "O" ring rubber seats. outside damper's A retest indicated a remaining leakage rate of 1298 cc/ min from the
,"y
( j test volume. Additional adjustment was made on the inside
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containment damper following refueling. Leakage was also
)
I l I detected at the flanged section containing this damper m and all flange bolts were securely tightened. Another retest indicated remaining leakage from this volume of 381 cc/ min. The plant PORC Committee decided that for better awareness of the condition of these dampers, testing of this volume should be at more frequent intervals of approximately two to three months. Subsequent tests, performed in addition to the normal type C tests, and their results are as follows: 5/9/74 -------------- 256 cc/ min 6/6/74 -------------- 145 cc/ min
'~ 9/3/74 --------------
131 cc/ min T 11/11/74-------------- 325 cc/ min s.\% ,/ 2/13/75 -------------- 258 cc/ min 4/28/75 -------------- 509 cc/ min 6/25/75 -------------- 489 cc/ min 9/24/75 -------------- 195 cc/ min 12/18/75-------------- 460 cc/ min 1/8/76 --------------47,000 cc/ min Retest 1/9/76--------- 80 cc/ min During the 1974 refueling outage excessive leakage was also observed from the volume bounded by the purge exhaust dampers. A 60 psig pressure decay of this volume ir.dicated approximately a two psig loss of pressure in one minute. This is the'same rate of loss as that which was observed for the purge supply test volume above. The plant condi-tions at this time were the same. Visual assessment ( (soap bubble solution) disclosed that the majority of leak-age from this volume was into the containment atmosphere. (G Blow-by passed the outside damper was minimal. Corrective actions were the same as those discussed above. A retest following corrective action indicated a remaining leakage rate from the test volume of 161 cc/ min. ) l Subsequent tests, performed in addition to the normal type C tests, and their results are as follows: 4/4/74 ------ 226 cc/ min 5/9/74
~
194 cc/ min 6/6/74 ------ 175 cc/ min 9/4/74 ------ 338 cc/ min - 11/11/74 ------220,794 cc/ min Retest 11/11/74------ 164 cc/ min 2/13/74 ------ 169 cc/ min 6/25/75 ------ 178 cc/ min 9/24/75 248 cc/ min y[f 12/23/75 ------ 511 cc/ min 1/8/76 ------ 65,130 cc/ min Retest 1/9/76 ------ 325 cc/ min 3.2.2 Leakage During Steam Generator Inspection of November, 1974' The plant was in a cold shutdown condition for a steam generator tube inspection. During the process of educting gas from the reactor vessel head vent, the eductor hose was draped over the open purge exhaust damper inside containment. Upon damper closure the hose was pinched between the damper and "O" ring seat, preventing adequate make-up of the seat-ing surface. After removal of the hose, adjustment was made on the damper's "O" ring adjustable rubber seats. A retest the same day indicated a leakage rate from the volume of 164 ['% cc/ min. I l 1 1 3.2.3 Leakage During Steam Generator Inspection of January, 1976 A leakage rate of 47,000 cc/ min (1.23 La) from the purge supply damper volume was identified on January 8, 1976. The plant was in cold shutdown condition for steam gener-ator tube inspection. A visual inspection, while the volume was subjected to test pressure, indicated approxi-mately equal leakage past each damper. Corrective actions began immediately; "O" ring seats and damper sealing surfaces were thoroughly cleaned and adjust-ments were made on each damper's "O" ring adjustable rub-ber seats. Retesting of the test volume, after cleanup and adjustment, still resulted in a leakage rate of 1,500 cc/ min. Further checking of the penetration assembly dis-
- closed leakage existing at approximately 5 upper flange bolts on the spacer flange section associated with the con-tainment side damper. Flange bolts adjacent to the leak-ing area were securely tightened.
A retest of the volume after the corrective actions were taken indicated a leakage rate of 80 cc/ min. Leakage from the purge exhaust damper volume at a rate of 65,130 cc/ min (1.71 La) was also identified on January 8, 1976. Corrective actions similar to those above were per- l formed and a retest of the volume indicated a remaining l l
.V I
l l leakage rate from the purge exhaust damper volume of 325 "N cc/ min. A visual assessment at this time allocated 100 cc/ min as leakage past the inside containment damper and 225 cc/ min leakage past the outside containment damper. 3.2.4 Leakage During the Refueling Outage of 1976 The volume bounded by the containment purge supply was subjected to a 60 psig pressure decay test on February 5, 1976. The concerned volume could not be pressurized beyond 10 psig through the 1 inch 60 psig air supply line. Because of the leakage history of these dampers during cold . shutdowns, this check was performed to assess the leak tightness of these dampers prior to performing the inte-
#[ T grated leakage rate test (ILRT). The outside damper was JQ observed to be properly closed, and the switch indicating light for the inside damper 5870 indicated that this damper, was also closed. No attempt was made to ascertain the in-ability to pressurize the volume at this time.
On February 7, 1976 pressurization of containment for the ILRT began. During containment pressure buildup it was noted that the pressure between the dampers was increasing at the same rate as that of containment. Visual inspec-tion (soap solution) indicated essentially zero leakage past the outside damper. Hence, leakage from containment atmosphere to outside building atmosphere was minimal b) N,,/ through this path. 1 1
( The ILRT was aborted after approximately 7 hours into the
) equalization period because the calculated hourly weight loss was in excess of the allowable.
After depressurizing the containment building, a leakage evaluation was conducted on the observed leakage paths. Although the purge supply damper system was not contribut- [ i ing to the loss of containment air weight, a 60 psig I pressure decay was performed on this volume to locate the leakage path (s) to the containment interior. Whatever difficulty prevented the pressurization of the damper vol-ume on February 5, 1976 was not present at this time. A 60 psig pressure was readily attained, and calculated re-f-- sults indicated a minimal leakage rate of 83 cc/ min from
'^
this volume. - Preparations continued in other areas to permit resump-tion of the ILRT, which was completed successfully on February ll, 1976. The purge supply volume was again tested on February 28, 1976 prior to commencing refueling operations. Contain-ment purge had been in effect during inspection of the steam generators. Upon securing the dampers closed and apply-ing air pressure it again was noted that pressurization beyond 10 psig could not be attained. Inspection inside containment disclosed that the inside damper was not com- { pletely seating. Repeated operation of this damper showed
~15-
very jerky motion, and in the opening direction loud noises were emitted from the spring cannister assen61y. Adequacy of the damper closure was variable, which accounted for the randomness of establishing a tight test boundary. When the spring cannister piston shaft was fully extended, heavy scoring was noted over approximately 7 inches of shaft length closest to the cannister head plate. Correc-tive action consisted of smoothing the scored shaft, polishing and dry lubricating both the spring cannister piston shaft and power air cylinder piston shaft. Subse- - quent damper operation was greatly improved. Noise level from the spring cannister was reduced and repeatability of
/' \
closure was consistent. A satisfactory retest was
.(
then performed with calculations indicating an acceptable J l leakage rate of 437 cc/ min from the test volume. , 4 l } The volume bounded by the purge exhaust dampers was also subjected to a 60 psig pressure decay check on February ] 5, 1976. This check, also made because of the leakage 1 i history of the dampers during cold shutdown, was to assess the leak tightness of the dampers prior to performing the containment ILRT. Results of calculations indicated a
- leakage rate from the test volume of 456 cc/ min. No re-l pair attempts were undertaken at this time.
1,( es \ Pressurization of containment for the ILRT began on February ( \s / 7, 1976. During containment pressure buildup, it was e .n- ,e ---r . . - - ,
f noted that the pressure between the dampers was increasing ( at the same rate as containment. Visual inspection of the accessible outside damper showed several areas of leakage around the seating periphery. I After the ILRT was aborted and the containment was depres-surized, a 60 psig pressure decay test was performed on the purge exhaust volume. Results indicated a leakage rate of 9801 cc/ min. Approximately 50 percent of the total leakage was past each damper. Adjustments were performed on each damper's adjustable i
"O" ring seat and the follow up retest indicated a re-
- maining leakage rate of 447 cc/ min from this volume.
i , i Subsequent testing dates and leakage results of this vol-ume are as follows:
- 2/28/76 --------------
314 cc/ min 3/31/76 -------------- 315 cc/ min e 4 4 ' (. l l j !
T l
4.0 CONCLUSION
S [
] The failure of the Type A integrated leak rate test was the result of leakage through the purge exhaust dampers and check valve 1713. Abnormally high leakage rates through the purge dampers have historically occurred only when the plant is in the cold shutdown condition. The pathway from containment through check valve 1713 leads into a normally closed nitrogen system outside containment which is pressurized to 90 psig. Venting of this system both inside containment and outside containment was per-formed especially for the Type A test. Therefore, it is -
reasonable to assume that the containment leakage under post accident conditions would not have exceeded the allow-I able limit and no undue risk to the health and safety of the public existed. The failures of Type B and C tests have been shown to re-sult from conditions caused by plant outages. No Type B or C tests performed during plant operation have resulted in failure. Repairs of leaks developed during outages have been made promptly and the status of those components which have previously failed to meet the acceptance criteria have been monitored prior to returning the plant to power. Therefore, it is concluded that the reactor containment building has been kept in a good state of repair and will adequately perform its safety function.
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