ML20210S023
| ML20210S023 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 11/30/1986 |
| From: | Paulewicz F, Stoehr R, Roxanne Summers GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20210R959 | List: |
| References | |
| NUDOCS 8702170549 | |
| Download: ML20210S023 (97) | |
Text
{{#Wiki_filter:--_ -- THREE MILE ISLAND NUCLEAR STATION UNIT 1 tuclear . REACTOR CONTAINMENT BUILDING INTEGRATED LEAK RATE TEST NOVEMBER 1986 PREPARED BY: h h. a ZAk7 F. W. Paulewicz - IL ect(iorf GPUN j fi PREPARED BY: ; k27N R. G. Stoehr - LLRT Section GPUN REVIEWED BY: L 2
. L. Sumers /
REVIEWED BY: b7 R. O. Barley U GPUN APPROVED BY: _ h8 a!7 f87 J. J.foVtz I GPUN D ADD- 5 0 89 p PDR
TABLE OF CONTENTS 1986 ILRT REPORT
1.0 INTRODUCTION
2.0 GENERAL AND TECHNICAL DATA 3.0 ACCEPTANCE CRITERIA 4.0 TEST INSTRUMENTATION 5.0 TEST PROCEDURE 6.0 METHODS OF ANALYSIS 7.0 DISCUSSION OF RESULTS 8.0 TYPE B AND C LEAKAGE RATE HISTORIES
9.0 REFERENCES
APPENDICES A. ILRT INSTRUMENTATION SCHEMATIC B. AVERAGE TEMPERATURE AND MAXIMUM TEMPERATURE DIFFERENTIAL TABLE C.
SUMMARY
OF CHANGES, EXCEPTIONS AND DEFICIENCIES D. REACTOR BUILDING PRESSURE DURING THE ILRT E. REACTOR BUILDING TEMPERATURE DURING THE ILRT F. REACTOR BUILDING MASS DURING THE ILRT G. 1985 LOCAL LEAK RATE TEST REPORT H. 1986 LOCAL LEAK RATE TEST REPORT I. 1987 LOCAL LEAK RATE TEST REPORT J. TMI-1 LLRT CORRECTIVE ACTION
SUMMARY
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1.0 INTRODUCTION
; The Three Mile Island Nuclear Station Unit i reactor containment building ,
was subjected to a periodic Integrated Leak Rate Test during the period I from November 5, 1986 to November 9, 1986. The purpose of this test was I to demonstrate the acceptability of the building leakage rate at the l calculated design basis accident pressure of 50.6 psig (P a). The l allowable leakage is defined by the Design Basis Accident, applied in the Safety Analysis, in accordance with site exposure guidelines specified by 10 CFR 100. For Three Mlle Island Nuclear Station Unit 1, the maximum allowable "As Found" integrated leakage rate at the design basis accident pressure is 0.10 percent by weight per day (L ). Testing was performed in accordance with the requirements of TMI-1 Tech. Spec. 4.4.1.1, 10 CFR 50, Appendix J, ANSI N45.4-1972, ANSI /ANS 56.8-1981 (as applicable) and the procedural requirements as stated in GPU Nuclear Corporation Three Mile Island Nuclear Station Unit 1 Surveillance Procedure 1303-6.1. This procedure was recommended for approval by the Three Mile Island Nuclear Station, Unit 1, Plant Review Group and approved by the Operations and Maintenance Director TMI 1, prior to the commencement of the test. A summary of the Temporary Change Notices, Exceptions and Deficiencies taken to this procedure is included as Appendix C. All testing was performed by GPU Nuclear Corporation with the technical assistance of Gilbert Associates, Inc. and Volumetrics, Inc. Procedural and calculational methods were witnessed by Nuclear Regulatory Commission personnel and monitored by the GPU Nuclear Corporation site Quality Assurance staff. The last series of Local Leak Rate Testing (App. J, Type C) was completed about seven months prior to the ILRT. The cambined local leakage rate at that time was less than 15 percent of the maximum allowable leakage rate (L.), well below the allowable value of 60 percent, at 50.6 psig prior to the commencement of the integrated leak rate test. < The "As Found" ILRT commenced at 1615 on November 6, 1986, following a satisfactory 4 hour stabilization period. The total temperature change during this period was approximately 0.1 0F. The leak rate was calculated by utilizing a computer and computer program provided by GAI. After approximately 18 hours of data, it was apparent that the "As Found" leak rate was on the order of 0.1 W%/ day by the mass point method. Teams were formed in an attempt to determine the cause of the unacceptable leakage. The source was found to be leakage through the Penetration Pressurization (PP) System line connecting to the AH-V1A/lB (Reactor Building Purge Exhaust) penetration. AH-VIA is the outer valve and AH-VlB is the inner valve. Check valves PP-V101 and PP-V102 were found to be leaking excessively. After these valves were isolated, the interspace between the purge exhaust valves quickly pressurized and the seats on AH-VlA began leaking. The seats were then tightened in place , until no visible leakage was found via a soap bubble test. The Plant Review Group reviewed the "As Found" data and determined that this failure was a reportable event as defined by Tech. Specs, and the NRC was notified per LER No.86-013. 1.0 0111U
After isolating the PP-V101/102 valves and correcting the AH-V1A leakage, increasta OTSG Secondary Pressure from 20 to 45 psig, and assuring that no other significant leakage paths existed, a second ILRT
- was started. After a satisfactory four hour stabilization period, the "As Left" ILRT started at 0130 on November 8, 1986.
The calculated "As Left" leakage rate based on the Total Time method of analysis, for a period of 24 hours was found to be 0.04231 percent by weight per day at 50.6 psig. The leakage rate at the upper bound of the 95 percent confidence. interval was 0.06682 percent by weight per day which is below the allowable leakage rate of 0.075 percent by weight per day at 50.6 psig. The low integrated leakage rate from the ! reactor containment building, less than 75% of the allowable leakage I rate, provides assurance that the reactor containment building is capable to perform its intended safety function. The Total Time results are reported as the results of record. Since the Industrial Cooler System was in operation during the integrated leak rate test, addition of the local leakage rate of the system isolation valves, RB-V2A and RB-V7, (880 scem and 0 seca respectively),to the measured integrated leakage rate must be considered. The combined local leakage rate of these isolation valves is 0.0005 percent by weight per day. The addition of these values changes the Total Time calculated leak rate at the upper bound of the 95% confidence level to 0.06732 WE/ day. Check valves PP-V101-V102 will be repaired and retested along with purge valves AH-V1A/1B prior to plant startup to assure that the 0.6 L criteria from 10CFR50 Appendix J is met for local leak rate testing. The GAI computer program utilized daring the ILRT had the capability of performing ILRT calculations using either the Total Time or Mass Point Methods. Calculations using the TNI-1 data during and after the ILRT showed that the Total Time Nethod results could be grossly affected by i minor changes in the test start times. The Nass Point Method did not l eWhibit these distortions and GPUN considers the Mass Point Method to be more indicative of actual Reactor Building Leakage. The Mass Point "As Left" results were: Leakage Rate: 0.03169 WE/ day f at 95% confidence Level: 0.03428 WE/ day The supplemental Total Time instrumentation verification difference at P wag 0.00963 which is within the 25 percent of Larequirement of ' 10 CFR 50, Appendix J, Section III A.3(b). 2.0 GENERAL AND TECHNICAL DATA , i 2.1 GENERAL DATA !' Owner: General Public Utilities Nuclear Corporation l l l 2.0 0111U
Docket No.: 50-289 Location: Three Mile Island, near the East Shore of the Susquehanna River in Dauphin County, Pennsylvania.
- Containment
Description:
Reinforced concrete structure composed of cylindrical walls (prestressed with a post-tensioning tendon system in vertical and horizontal directions), with a flat foundation mat (conventional reinforcing) and a shallow done roof (prestressed utilizing a three-way post tensioning tendon system). The inside surface is lined with a 3/8" thick carbon steel liner. Date Test Completed: November 9, 1986 2.2 TECHNICAL DATA Containment Net Free Volume: 2 x 106 cubic feet Design Pressure: 55 psig Design Temperature: 2810F Calculated Accident Peak Pressure: 50.6 psig Calculated Accident Peak Temperature: 2810F 3.0 ACCEPTANCE CRITERIA Acceptance criteria established prior to the test and as specified by TMI-1 Tech. Spec. 4.4.1.1., 10 CFR 50, Appendix J. ANSI N45.4-1972 and ANSI N56.8-1981 are as follows:
- a. The measured leakage rate (Lam) at the calculated design basis accident pressure of 50.6 psig (Pa) shall be less than 75 percent of the maximum allowable leakage rate (L.), specified as 0.10 percent by weight of the reactor building atmosphere per day at the upper bound of the 95 percent confidence level. The acceptance criteria is determined as follows:
La = 0.10%/ day 0.75L = 0.075%/ day
- b. The test instrumentation shall be verified by means of a supplemental test. Agreement between the containment leakage measured during the Type A test and the containment leakage determined during the supplemental test shall be within 25 percent of L .
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The Total Time Method calculation was used for the As Left test acceptance criteria because it is recommended in ANSI N45.4-1972. ANSI Standard N56.8-1981 recommends the Mass Point Method, but this standard has not been accepted by the NRC. The GAI computer program provided output using both methods of calculation. 4.0 TEST INSTRUMENTATION 4.1
SUMMARY
OF INSTRUMENTS Test instruments employed are described in the following subsections: 4.1.1 Temperature Indicating System Resistance Temperature Detectors Quantity 24 Manufacturer Yellow Spring Instr. Type YSI Model, 4150-1/4-6-3-6-138-AW-Gl/2-QR (platinum) Range, OF 60-120 Accuracy, OF 10.1 Sensitivity, OF O.01 4.1.2 Dewpoint Indicating System Dewcell Elements Quantity 10 Manufacturer Foxboro Type BD154WB, Lithium Chloride Range, OF 40-100 Accuracy, OF 12.0 Sensitivity, OF 10.1 4.1.3 Pressure Monitoring System Precision Pressure Gauges Quantity 2 Manufacturer Texas Instruments (modified by Volumetrics to interface with ILRT System) 4.0 0111U
Type Model 145.02 Range, psia 0-100 l Accuracy 10.015% of indicated pressure Sensitivity, psi 10.001 NOTE: Since the pressure output was in term of absolute pressure (psia), no instrumentation correction for barometric pressure was required. 4.1.4 Supplemental Test Flow Monitorina System Flowmeter Quantity 2 Manufacturer Sierra Type Model 14636 Range, scfm 0.0 - 13.0 Accuracy, 1% full scale Sensitivity 10.5% of full scale 4.1.5 Outputs from the aforementioned sensors (with the exception of the output from the mass flowmeters) is forwarded to the Data Acquisition System (DAS) for conversion, display and forwarding to the data printer. The installed DAS unit was a Model A-100, manufactured by Volumetrics. The DAS unit has the capability of monitoring over 100 channels. For
~the ILRT, the channels were utilized as follows:
- a. Precision Pressure Gauge - 2 channels
- b. RTDs - 24 channels (Channels 1 to 24)
- c. Dewcells - 10 channels (Channels 30 to 39)
- Output from the DAS went to both a record printer and a utility printer.
4.1.6 Sensor input conditioning cards, precision pressure gauges, DAS unit, and other accessories were purchased as a rack mounted unit from Volumetrics by GPUN in March, 1984. Details of the unit and equipment s;ecifications are available onsite for review. ' 0111U 5.0
4.1.7 After the DAS unit converted the signal input into the desired parameter of temperature or pressure, these values were printed out on a utility printer for manual entry into the GAI computer. The computer program performed the following functions:
- a. Pressure - corrected the two inputs by the applicable calibration equation and averaged the two inputs.
- b. Temperature - The data from the 24 RTD's was summed and the average was taken. No weighting factors were used.
- c. Deweells - The data from the 10 dewcells was summed and the average was taken. The average dewpoint temperature was then converted to partial pressure of water vapor ucing the Keenan and Keyes Steam Table Equation.
4.1.8 The accuracy of the DAS unit with respect to the different monitored parameters is given below:
- a. Pressure - direct transfer of the number of counts from the precision pressure gauges to the printer for manual entry into the computer,
- b. Dewpoint accuracy: 12.0 0F.
Temperature: 10.1 0F (60 F 0 range). 0 to 120 F c.
- d. Repeatability - For automatic DAS:
Dry bulb temp: i .01 0F Dewpoint temp: i .01 0 F Abs. press: i .001 PSI 4.1.9 All operable RTDs and dewcells were assigned equal weighting factors. This is because: 4
- a. There are very few cubicles inside the Reactor Building.
- b. There is free communication between all levels of the building and also between the cubicles and the Reactor Building.
- c. The air inside the Reactor Building is continually recirculated by the installed ventilation system.
- d. Almost all of the equipment in the Reactor Building, with the exception of the aforementioned recirculating fans and required instrumentation, was deenergized during the test. This eliminated any heat producing equipment in the building which could cause local hot spots.
- e. No stratification has ever been observed during an ILRT at TMI-1.
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4.2 SCHEMATIC ARRANGMENT The arrangement of the four measuring systems summarized in Section 4.1 is depicted in Appendix A. The arrangement of temperature sensors can be grouped into five levels as follows: Level Elevation Sensors 1 287 feet TE-655R TE-655S TE-655T TE-655U TE-655V 1 2 314 feet TE-655M to TE-655N 321 feet . TE-6550 TE-655P TE-655Q 3 352 feet TE-655A TE-655G TE-655I TE-655K 4 365 feet TE-655D to TE-655J 405 feet TE-655L TE-655W TE-655I TE-655B 5 437 feet TE-655C TE-655E TE-655H TE-655F The average Reactor Building temperature varied by only 0.272 0F over the 24 hour As Left test. Of the 24 RTDs utilized, RTD 0 (314' Elev.) indicated consistently low and RTD's N and W (314' and 365' Elev.) indicated consistently high. The average difference was 2.05 0F. These results are summarized in Appendix B. This analysis demonstrates that there was no large regional temperature variation in the Reactor Building and also that no large temperature fluctuations occurred during this ILRT. Small fluctuations, as discussed in section 5.2.2. were noted. Operation of the three Reactor Building recirculation units provided satisfactory temperature equalization throughout the building. 7.0 0111U
. . ~ _ . .
k 4.3 CALIBRATION CHECKS ~ Temperature, dowpoint, pressure and flow measuring systems were checked for calibration before the test in accordance with GPU Nuclear Corporation Procedure 1430-Y-23, as recommended by ANSI N56.8-1981. The results of the calibration checks are on file at Three Mile Island Nuclear Station Unit 1. The supplemental test at 50.6 psis confirmed the instrumentation acceptability. 4.4 INSTRUMENTATION PERFORMANCE The twenty-four temperature sensors, ten dewcells, two precision. l pressure gauges, flowmeters, and readout equipment performed satisfactorily throughout the integrated leakage rate test. ,No sensor or readout equipment malfunctions occurred during performance of the test. 4.5 INSTRUMENTATION SELECTION GUIDE VALUE Justification of instrumentation selection was accomplished, using manufacturer's sensitivity and repeatability tolerances stated in Section 4.1, by computing the instrumentation selection guide (ISG) value. Utilizing the methods, techniques and assumptions in Appendix G l to ANSI NS6.8-1981, the ISG was computed for the absolute method as follows:
- a. Conditions La = 0.1w%/ day P = 65 psia
~T = 81.250 F = 540.94 0R dry bulb (typical)
Tdp = 67.2 0F dewpoint (typical) t = 24 hours (test duration in hours)
- b. Total Absolute Pressure: op Sensor sensitivity error (E): 10.001% of full scale Measurement system error (c): t0.002% of full scale
+ 1/
ep= (EP)2 (,p)2-- no.ofsensor)1/2 ep=1 0.001)2 + (0.002) 1/2 g/ (2) 1/2 e p=1 0.0016 psia 8.0 0111U
s
- c. Water Vapor Pressure: epy Sensor sensitivity error (E): 10.10F Measurement system error (c), excluding sensor: i0.10 F At a dowpoint temperature of 67.2 0 F, the equivalent water vapor pressure change (as determined from the steam tables) is 0.01149 psia / 0F.
Epy = 10.10 F (0.01149 psia /0F) Epy = 10.001149 psia spy = 10.10F (0.01149 psia /0F) t e py = 10.001149 psia ey=tp (Epy)2 + (epy)2- 1/2 no. of sensors 1/2
= _
ey=ip (0.001149)2 + (o,001149)2 1/ (to) 1/2 e py = 1 0.0005139 psia
- d. Temperature: eT No. of Sensors: 24 Sensor sensitivity error (Ey): 10.01 0F = 10.010 R Measurement System Error (cT): 0.020F = 10.020R 1/2 (no. of sensors) 1/2 eT = (ET 2) + (cT)2 -
(.01)2 + (0.02)2 1/2 (24) 1/2 eT = eT = 0.00456440R
- e. Instrumentation Selection Guide (ISG)
ISG = 2400 +2 2+ 2e 2 1/2 1/2 ISG = i 2400 2 0.0016)2 +2 0.000514}2+ 20.00456}2 24 hrs. 65 / 65 / 540.94/ ISG = 10.00385%/ day The ISG does not exceed 0.25 L. (0.025%/ day) and it is therefore concluded that the instrumentation selected was acceptable for use in determining the reactor containment integrated leakage rate. 9.0 0111U i
- . a . 4.6 SUPPLEMgNTAL VERIFICATION 4.6.1 Superimposed Test In addition to the calibration checks described in Section 4.3, test instrumentation operation was verified by an approximately 6 hour flow test subsequent to the completion of the 24 hour leakage rate test. This test consisted of imposing a known calibrated leakage rate on the reactor containment building. After the flow rate was established, it was not altered for the duration of the test. During the supplemental test, the calculated leakage rate was L e=L.+Lo y where, Le= calculated composite leakage rate consisting of the reactor building leakage rate plus the imposed leakage rate Lo= imposed leakage rate L.=y leakage rate of the reactor building during the supplemental test phase Rearranging the above equation, L.=y Le-La The reactor containment building leakage during the supplemental test can be calculated by subtracting the known superimposed leakage rate from the calculated composite leakage rate. The containment building leakage rate during the supplemental test (Ly .) was then compared to the calculated reactor containment building leakage rate during the preceding 24 hour test (Lam) to determine instrumentation acceptability. Instrumentation is considered acceptable if the agreement between the two building leakage rates is within 25 percent of the maximum allowable leakge rate (L ). 5.0 TEST PROCEDURE 5.1 PREREQUISITES Prior to commencement of reactor containment building pressurization, the following basic prerequisites were satisfied.
- a. Proper operation of all test instrumentation was verified.
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,. , - _ , y - _ . - - - - . , , - -_ y
/ ;.
- b. .All reactor containment building isolation valves, with the exception of those within the Reactor Building Cooling System and Decay Heat _ System, were closed using the normal mode of operation. All associated system valves were placed in post-accident positions. The Reactor Building Cooling System was in service for temperature control during the test and the Decay Heat System was in service to maintain the plant in a safe and stable condition during the test.
- c. Equipment within the reactor containment building, subject to damage, was protected from external differential pressures,
- d. Portions of fluid systems which, under post-accident conditions become' extensions of the containment boundary, were drained and vented.
- e. The Penetration Pressurization System was depressurized.
Manometers were installed at penetration pressurization manifolds to provide means for detection of leakage into the system.
- f. Manometers or pressure gauges were installed on the purge valve interspaces and access hatch interspaces to provide means for
~
detection of leakage into such systems.
- g. Local leakage rate testing of selected containment isolation valves and penetrations was concluded (AH-V1A/lB/lC/1D, Equipment Hatch Flange, BS-V37A/B/C/D related instrumentation tubing). All of these were found by the cognizant engineer to be acceptable without repairs / adjustment, though a minor adjustment was done inadvertently but only on the AH-VID seat. A local leak test was decumented before and after that adjustment.
- h. Potential pressure sources were removed or isolated from the containment.
- i. All accessible liner weld channels'were vented to the containment atmosphere.
J. A general inspection of the accessible interior and exterior areas of the containment was completed with no problems noted.
- k. The Fluid Block System has been partially removed and can no longer pressurize the space between containment isolation valves.
The entire system will eventually be removed. The valves in this 3 system were placed in the following configuration: (1) Manual valves were placed in the postion required by the present normal valve lineup.
-(2) Any automatic valves, which were still operable, were placed in their original post-accident condition.
(3) The non-seismic sections of the system were vented. i 11.0 0111U
6 5.2 TEST PERFORMANCE' i 5.2.1 PRESSURIZATION PHASE TMI-1 was shutdown on October 31, 1986. This was the first time that the ILRT was performed at the beginning instead of the end of the outage. Pressurization of the reactor containment was started at about 0045 on November 5,1986. A pressurization rate' of approximately 2.5 psig/hr. was maintained using four 1500 cfm diesel driven air - compressors which were rented for this purpose. Buildling temperature and pressure were monitored periodically and the amperage required by the recirculating fans (AH-Eli, 1B, and IC) was monitored every two hours. When containment pressure reached 12 psig at about 0600 on November 5, 1986, pressurization was secured to. allow for~a containment internal inspection. Six GPUN supervisors entered the Reactor Building to inspect for any obvious damage due to the pressurization. The inspection was completed by 0730 with no anomalies observed. Pressurization was restarted at 0737. 4 During pressurization from ambient to the 50.6 psig pressure level, the following observations were made: i
- s. The pressure connection from the precision pressure gauges to the containment pressure root valve was found to be disconnected. The connection was made and then leak checked when this was discovered.
- b. Manometers on several of the penetration pressurization manifolds i indicated increasing header pressure due to valve leakage. The manifolds were:
3
- 1. Manifold F - Electrical Penetrations.
- 2. Manifold M - Leak Rate System.
- 3. Manifold 0 - Fuel Transfer Tube Flanges.
- c. Pressurizer level decreased slowly over the duration of the ILRT.
Concurrently, a gradual-increase was noted in the Borated Water Storage Tank level. This has been noted during previous ILRT's i and is due to leakage through DH-VSA/B and DH-V14A/B. These } valves were repaired during the 6R Outage.
- d. The Reactor Building Purge Exhaust (AH-VIA/B) interspace volume pressure began increasing shortly af ter pressurization began.
Final pressurization was secured at 0410 on November 6,1986, with Reactor Building pressure at 50.8 psig. The required four hour stabilization period began at 0430. During this stabilization period,
- Reactor Building temperature decreased slowly due to temperature i variations of the Industrial Cooler System. Ambient temperature was
! decreasing and this resulted in minor variations in both air inlet and I ' Industrial Cooler cooling temperature. The air inside the Reactor I Building was also not in temperature equilibrium. The Reactor Building 12.0 0111U
was repressurized twice due to this temperature drop. The stabilization period was commenced again at 1029 on November 6,1986, with the first 24 hour ILRT commencing at 1615. Reactor Building temperature was then 84.625 0 F with a pressure of 65.323 psia. 5.2.2 "AS FOUND" INTEGRATED LEAK RATE TEST At 1615 on November 6, 1986, the "As Found" Leak Rate Test commenced. Data was collected and entered into the computer every 15 minutes. The following sets of data were collected and entered:
- a. Pressure indicated by the two precision pressure instruments.
- b. The outputs from twenty-four RTD's were recorded.
- c. The ten dewpoint values were recorded.
The use of vapor pressure (Pwy), average temperature (T) and the total pressure (P T) is described in more detail in Section 6.1. All original data is on file at Three Mile Island Nuclear Station Unit 1. The plot of average pressure, temperature and weight of air was maintained for each fifteen minute reading. Periodically, the calculated leak rate values were obtained. A graphic display of leak rate, temperature, pressure, etc., vs. time was also available. This graphics display was very useful in monitoring and trending the inputs into the computer. The Reactor Building average temperature was dropping slowly over the ; ILRT. Two abrupt increases in Reactor Building temperatures were noted ) due to automatic operation of pumps in the Industrial Cooler System. ! The Control Room was notified when these events occurred, and the appropriate corrective action was taken. It became apparent within several hours after starting the test that an unacceptable leakage out of the Reactor Building was occurring. An investigation of all potential leakage paths was begun. The leakage was eventually traced to the Reactor Building Purge Exhaust Valves (AH-VIA/1B). Rotameters were installed off of Manifold J of the Penetration Pressurization System and PP-T-1B, the pressurization tank for the purge valve interspace in an attempt to quantify this leakage. The total leakage through these rotameters was approximately 33,000 scem. No leakage was found on the outboard valve, AH-VlA. Therefore this leakage was determined to be the total leakage from AH-V1B, the inboard isolation valve. This leakage remained steady until about 2400 on November 6,1986, when it was observed to increase to about 37,000 seem or 0.0217 W%/ day. The ILRT was continued for about another twelve hours until it became apparent that the calculated leakage was approximately 0.1 w%/ day. Searches were made in an attempt to locate additional leakage points. No other significant leakage Paths were noted. l 13.0 0111U
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A At1430onNovhaber7,theTestDirectordeclaredthe"AsFound" test to be a' failure'with the integrated leakage at about 0.1 w%/ day by the Mass Point method. ,The Total Time method indicated a slightly higher leakage but its level of. confidence was extremely poor. The Plant Review Group concurred with this evaluation and declared the ILRT a failure. The Site NRC inspector.was notified shortly thereafter. 1 After the evaluation and NRC notification, the following immediate actions wore taken: 4
- s. TCN 1-86-0176 was processed to isolate the Penetration, i Pressurization System check valves in-the Purge Exhaust-penetration. - This TCN also allowed OTSG secondary side pressure.
to be increased from 20 psis to 45 psig to ensure that minimal , leakage would occur into the OTSG Secondary Side. Leakage from the containment into the OTSG's has been noticed during previous , ILRT's.
, b. Continued to search for additional leakage paths. None were discovered.
- c. After isolating the Purge Exhaust penetration, the interspace pressure rose rapidly. , Seat leakage was then noted on the AH-V1A seating' surface. The seats were tightened and the seat leakage
- was reduced to essentially zero.
4 Since AH-VIA is downstream of AH-V1B, the previously quantified seat leakage through AH-V1B, about 37,000 scem, was the most that could pass through.the AH-V1A valve and quantified the leakage prior to recommencing the testing.
- 5.2.3 "AS LEFT" INTEGRATED LEAK RATE TEST
< The Test Director verified that all prerequisites were still satisfied and that the "As Left" ILRT could commence. The Operations Department and GPUN QA concurred with this evaluation. The Reactor Building was repressurized and another 4 hour stabilization began at 2130 on November 7. The ILRT was started at 0130 on November 8 after t satisfactory completion of the stabilization period.
No significant procedural, technical or operational problems were o'aserved during the "As Left" test. This test was terminated on November 9 at 0130. The test results were: t Calculated Leakage Leakage at 95% U.C.L. Total Time 0.04231 WE/ day 0.06682 WE/ day Mass Point 0.03169 W%/ day 0.03428 W1/ day 3 Plots of containment temperature, pressure and mass are included in Attachments D E, and F. ) These results were reviewed with the TMI-l Shift Supervisor and the PRG I Vice-Chairman soon thereafter and this part of the test was declared a j ' success. 14.0 0111U 't
)
i
k 5.2.4 SUPPLEMENTAL LEAKAGE RATE TEST , After the 24 hour Integrated Leak Rate Test data was obtained and evaluated, the leakage rate found to be acceptable, and a release permit had been obtained, a known leak rate was imposed at 0315 on November 9, 1986, on the reactor containment building through a calibrated flowmeter for a period of approximately 6 hours. During this time, temperature, pressure, and vapor pressure were monitored'as described above. The calculated leak rate during this period was 0.1437 d/ day. The average superimposed leak for the 6 hour period was 5.58 SCFM. This value equates to an integrated leak rate of .09174 d/$ay. If this value is added to the calculated leak rate _of 0.04231 d/ day, the leak rate is 0.13405 W/ day. This is approximately .00963 d / day difference when compared to the calculated superimposed leak rate of 0.14367 d/ day, and, therefore, meets the requirements to be within 25% of L. (0.025) d/ day. 5.2.5 .DEPRESSURIZATION PHASE
. After all required data was obtained and evaluted.,the supplemental test results were found to be acceptable. Permission from the Rad Con t Department and Control Room was obtained, and depressurization of the reactor containment building was started. A post test inspection of the building revealed no unusual findings. The depressurization rate was about 5 psig/hr.
4 6.0 METHODS OF ANALYSIS 6.1 CENERAL DISCUSSION ' The Absolute Method of leakage rate determination was employed during testing at the 50.6 psig pressure level. The Gilbert Associates, Inc. l ILRT computer code calculates the percent per day leakage rate for the Mass Point and Total Time methods. 6.1.1 MASS POINT ANALYSIS The Mass Point method of computing leakage rates uses the following ideal gas law equation to calculate the weight of air inside I containment for each 15 minute interval. : y ,144 PV = E RT T Where: W = Mass of air inside containment, Ibn K = 144 V/R = 5.3983 x 10 6 lba OR - in.2
- lbf I
- i 15.0 0111U f
l t
, , - , - -. --.-.,-,.-.-..-,n_, ,.,,nn.n-..nc,, ,-,...,,_-,,,-,,,,.,,-,-,_,,,..,.,---..,,c.-., , - _ , , , - - -,
P = Partial pressure of air, psia T = Average internal containment temperature, OR V = 2 x 106 ft3 R = 53.35 lbf - ft lba OR The partial pressure of air, P, is calculated as follows: P=PT-Pwy Where, PT= True corrected pressure by converting two pressure gauge readings and averaging, psia. Pwy = Partial pressure of water vapor determined by ave' raging the ten dewpoint temperatures and converting to partial pressure of water vapor, psia. The average internal containment temperature, T, is calculated as follows: T = Average of 24 RTD's + 459.69 0R The weight of air is plotted versus time for the ILRT test and for the supplemental test. The Gilbert Associates, Inc. ILRT computer code fits the locus of these points to a straight line using a linear least squares fit. The equation of the linear least squares fit line is of the form W = A t + B, where A is the slope in ihm per hour and B is the initial weight at time zero. The least squares parameters are calculated as follows: A = N (Ett W) i _ (Eti) (IWi ) S xx l B = (It i2) (IW i ) _ (It i ) (Iti W) i S xx . Where: Sx x = S (Iti 2) - (Et i)2 The weight percent leakage per day can then be determined from the ! following equation-l Lam = -2400 A ; B where the negative sign is used since A is a negative slope to express , the leakage rate as a positive quantity. I l i 16.0 0111U , ?
6.1.2 TOTAL TIME ANALYSIS The. Total Time method utilizes the following equation to determine the leakage rate of the reactor containment building: L = 2400
~
1T1 P2 t T2 Pi
- Where
L = Measured leak rate in weight percent per day' t = Time interval, in hours, between measurements T,T2= 1 Average internal containment temperature, OR, at ' the beginning and the end of the test interval respectively. P,P2= 1 Average containment pressure (corrected for water vapor pressure) at the beginning and end of the test interval respectively. The individual leakage rates are then plotted against time for the duration of the 24 hour test. The Gilbert Associates, Inc. ILRT computer code fits the locus of these points to a straight line using a linear least squares fit. The equation is of the form L = Lo+Lt i where Lt is the slope in percent per hour and Lo is the initial _ I leakage rate at time zero. The least squares parameters are calculated as follows: g , Et 2 - Et - Et EL S XX E L1= N i 1- i i S XX Where: ] S xx=N(It)2 i - (It i)2 F 6.2 STATISTICAL EVALUATION 6.2.1 GENERAL Af ter performing the least squares fit, the ILRT computer code i . calculates the following statistical parameters: ! 1. Limits of the 95% confidence interval for the mass point leakage rate (Cg). 4
- 2. Limits of the 95% confidence interval for the total time leakage j rate (C L).
! 17.0 0111U
These statistical parameters are then used to determine that the measured leakage rate plus the 95 UCL meet the acceptance criteria. 6.2.2 MASS POINT CONFIDENCE The upper 95% confidence limit for the mass point leakage rate is calculated as follows: Cg = 2400 t95 (SA/B) Where: Cg = Upper 95% confidence limit t95= Student's t distribution with N-2 degrees of freedom SA = Standard deviation of the slope of the least squares fit line B = Intercept of the least squares fit line The standard deviation of the slope of the least squares fit line (SA ) is calculated as follows: SA" S (N) 2 (Itg 2 - (Et g) Where: S = Common standard deviation of the weights from the least squares fit line N = Number of data points tt = Time interval of the ith data point The common standard deviation (S) is defined by: S= I(W4 - W)2 1/2 N-2 Where: Wi = Observed mass of air W = Least squares calculated mass of air The ILRT computer code calculates an upper 95% confidence leakage rate as follows: UCL = Lam + 2400 t95 (SA/B) This UCL value is then used to determine that the measured leakage rate at the upper 95% confidence limit meets the acceptance criteria. 18.0 0111U
6.2.3 TOTAL TIME CONFIDEWCE The 95% confidence limit for the Total Time leakage rate is calculated as follows: 1/2 Cn = t95 S 1 + (t - t) Where: (i ~ '} -- t - Total Time interval t = Iti n t1 = Time interval for each data point n = Number of individual Total Time leakage rates A L = Least squares calculated Total Time leakage rate t95 = Student's t distribution with N-2 degrees of freedom S = Common standard deviation of leakage rates from the least squares fit-line 1/2 3, 'I(L - L) 2 - N-2 -- 7.0 DISCUSSION OF RESULTS 7.1 RESULTS AT P. The method used in calculating the total time leakage rate is defined in Section 6.0. The result of this calculation is a leakage rate using absolute values of 0.04231 %/ day. The 95 percent confidence limit associated with this leakage rate is 0.02451 percent per day. Thus, the leakage rate at the upper bound of the 95 percent confidence interval becomes Lam = 0.04231 %/ day Lam (at 95% Confidence Limit) = Lam + 0.02451 = 0.06682 %/ day l The measured leakage rate at the upper bound of the 95 percent confidence level is below the acceptance criteria of 0.075 percent per day (0.75 L.). Therefore, reactor containment building leakage at ; the calculated design basis accident pressure (P a) of 50.6 psig is ; considered to be acceptable. l l I 19.0 0111U
1 i 7.2 SUPPLEMENTAL TEST RESULTS After conclusion of the 24 hour test at 50.6 psig, a mass flowmeter was placed in service and a flow rate of 5.58 SCFM was established. This flow rate is equivalent to a leakage rate of 0.09174 percent per day. After the flow was established, it was not altered for the duration of the supplemental test. The measured composite leakage rate (Le) using absolute values during the supplemental test was calculated to be 0.1437 percent per day using the total time method of analysis. The building leakage rate during the supplemental test is then determined as follows: L'=L y -L a Ly' = 0.1437 - 0.09174 %/ day Ly' = 0.05194 %/ day { Comparing this leakage rate with the building leakage rate measured ^ during the 24 hour test yields the following: LAN - Lv ' = 0.04231 - 0.05194 = 0.0963 LA 0.1 The building leakage rates agree with 9.63 percent of La which is below the acceptance criteria of 25 percent of L . Therefore, the acceptability of the test instrumentation is considered i to have been verified. 8.0 TYPE B AND C LEAKAGE RATE HISTORIES Refer to Appendices G, H, and I for the report on Type B and C testing I performed since the previous Type A test.
9.0 REFERENCES
- 1. SP 1303-6.1, " Reactor Building Integrated Leak Rate Test", GPUNC, Surveillance Procedure, Rev. 21.
- 2. ANSI M45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Reactors", American Nuclear Society, (March 16, 1972).
- 3. Steam Tables, American Society of Mechanical Engineers, (1967).
I
- 4. MP 1430-Y-23 " Reactor Building Integrated Leak Rate Test Instrument Calibrations", GPUNC Maintenance Procedure, Rev. 5.
- 5. ANSI N56.8-1981, N274, " Containment System Leakage Testing Requirements", American Nuclear Society, (February 19, 1981).
- 6. 10CFR50 Appendix J.
- 7. TMI-1 Tech. Spec. 4.4.1.1.
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APPENDIX B TEMPERATURE VARIATIONS DURING THE "AS LEFT" ILRT (ALL VALUES IN F) DATE TIME AVG. TEMP. HIGH TEMP. LOW TEMP. (HIGH-LOW) 11/08/86 0130 81.329 82.704 80.374 2.330 0230 81.315 82.545 80.389 2.156 0330 8'. 260 82.539 80.354 2.185 0430 81.219 82.571 80.300 2.271 0530 -81.185 82.386 80.278 2.108 0630 81.135 82.698 80.233 2.465 0730 81.088 82.315 80.206 2.109 0830 81.047 82.435 80.165 2.270 0930 81.039 82.276 80.183 2.093 1030 81.049 82.405 80.224 2.181 1130 81.107 82.258 80.360 1.898 1230 81.160 82.244 80.444 1.800 1330 81.146 82.199 80.444 1.755 i 1430 81,198 82.263 80.508 1.755 4 4 1530 81.204 82.269 80.545 1.724 1630 81.202 82.356 80.520 1.836 e 1730 81.183 82.543 80.503 2. 04 0 1830 81.100 82.209 80.418 1.791 1930 81.070 82.193 80.371 1.822 2030 81.068 82.655 80.387 .2.268 4 2130 81.021 82.385 80.336 2. 049 2230 81.016 82.486' 80.299 2.187 2330' 80.952 82.524 80.256 2.268 11/09/86 0030 80.976 82.377 80.339 2.038 0130 81.057 82.351 80.477 1.874 Additional Information:
- 1) The average reactor building temperatuge, using the above data, was 81.25 F. The average temperature decreased by 0.272 F during the 24 hour test.
- 2) 'RTD W typically indicated the highest temperature, but RTD N was occasionally slightly higher. The periods when RTD N was the highest indicated temperature 4
are indicated by *. RTD 0 indicated consistently low.
- 3) The High-Low average is 2.051 F. This compares favorably with the 1984 ILRT where the High-Low average was 2.10 F.
22.0 i'
1
<(
4. APPENDIX C
SUMMARY
OF CHANGES, EXCEPTIONS AND DEFICIENCIES TO SP 1303-6.1,, REACTOR BUILDING ILRT i _ DOCUMENT DATE PURPOSE COMMENT
! TCN 1-86-0170 11/04/86 To provide an alternate means of Processed at the request of providing makeup to'the RCS during TMI-l Operations Department.
the ILRT. i TCN 1-86-0171 11/05/86 Provided additional guidance for Resolved Exception 1.
- placement of RTD strings in the i containment dome area.
l. Exception - 1 11/05/86 Identified areas where RTD place- Analyzed and resolved in i ment did not agree with the refer- TCN 1-86-0171. j erced drawings. Exception - 2 11/06/86 Identified step in procedure which References to Data Sheet 2C will referenced Data Sheet 2C. This data be deleted during next procedure 4 sheet was deleted during the previous revision.
; procedure revisita.
, TCN 1-86-0176 11/07/86 Provided additional guidance for Resolved problem areas identified i' OTSG Secondary Side Pressure, leak- during "As Found" ILRT. age through Penetration Pressurization i Valves, and operation of the Industrial
, Cooler System.
Exception - 3 11/07/86 Provided additional signoff sheets for. Administrative requirement. j repressurization of Reactor Building.
- Deficiency - 4 11/07/86 Identified "As Found" ILRT as a failed Administrative requirement.
test. Deficiency - 5 11/09/86 Extended duration of Superimposed Leak Flow remained constant over j Rate Test to compensate for missed period of missed t2adings. No 4 readings, effect on results. Exception - 6 11/17/86 Maintained selected equipment in place Testing performed after completion to facilitate equipment testing. of ILRT. 1 23.0 4
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4' e - . ...m., n > u.x . 4 O o 1 APPENDIX G 4 THREE MILE ISLAND UNIT 1 1 1985 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT' 1 i ' SF 1303-11.18
, . =
b l l l f t l ,l ' l F 27.0 0111u i l
INDEX - 1985 R. B. LOCAL LEAK RATE TESTING REPORT
- 1. PURPOSE
- 2.
SUMMARY
OF WORK ACCOMPLISHED 2.1 Valve Testing / Repairs 2.2 Access Hatches 2.3 Penetration Pressurization
- 3. METHODS OF TESTING 3.1 Valves 3.2 Access Hatches 3.3 Penetration Pressurization
- 4. TEST EQUIPMENT USED 4.1 Valves 4.2 Access Hatches 4.3 Penetration Pressurization
- 5.
SUMMARY
AND INTERPRETATION OF DATA 5.1 Valves 5.2 Access Hatches 5.3 Penetration Pressurization
- 6. ERROR ANALYSIS 6.1 Valves 6.2 Access Hatches 6.3 Penetration Pressurization
- 7. REFERENCES
- 8. ATTACHMENTS 8.1 Results Evaluation Procedure / Repair Criteria 8.2 Tabulation of Individual Test Repairs 28.0 0111U
REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT 1985 REFUELING FREQUENCY-
- 1. PURPOSE 1.1 To provide analysis to the Nuclear Regulatory Commission on the periodic type B and type C leakage tests performed on the Three Mile Island Unit 1 Reactor Building.
This is in accordance with " Reactor Containment Leakage Testing for Water Cooled Power Reactors". -Appendix J, Part 50, Title 10 code of Federal Regulations which required the contents of this summary report to become part of the Type A test report along with the details of any other type B and type C testing performed since the previous type A test (also required per technical specification 4.4.1.1.8). No complete round of Leak Rate Testing was perfoceed in 1985. Local Leak Rate Testing is refueling frequency rather than annual.
- 2.
SUMMARY
OF WORK ACCOMPLISHED 2.1 Valve Testing / Repairs A small number of Appendix J Type B and C leak tests were performed as scheduled in TMI Unit 1 Technical Specification 4.4.1 and as noted on the table in Attachment No. 2. Repairs were initiated on the following components due to higher than desirable leakage.
- 1. AH-VIA/B & AH-VIC/D - Adjusted seats to regain acceptable leakage 2.2 Access Hatch Testing / Repairs Door Seals SP 1303-11.25 (Ref. 7.6) 2.2.1 Door seal leak tests are performed at the 72 hour frequency as required by Technical Specification 4.4.1.2.5 while hot functional testing was being performed. All of the seal tests satisfied the surveillance procedure administrative acceptance criteria.
2.2.2 Overall Hatch Test SP 1303-11.18 (Ref. 7.2) Semiannual integrated type leak tests were performed on each access hatch in 1985 as required by Technical Specification 4.4.1.2.5. All of the tests satisfied the surveillance procedure administrative criteria, 4 29.0 0111U
2.3 Penetration Pressurization SP 1303-11.24 (Ref. 7.5) Quarterly readings are recorded from the flow rotameters which supply air pressure or nitrogen pressure to reactor building mechanical and electrical penetrations as required by Technical Specification 4.4.1.2.5. No penetration leakage problems were noted although flow meter malfunctions required meter repair, and occasional tubing leaks in the air supply system were found and eliminated. 1 30.0 0111U
- 3. METHODS OF TESTING 3.1 Valve Test Methods Testing was performed by use of TMI Unit 1 surveillance procedure SP 1303-11.18 Reactor Building Local Leak Rate Tecting. This procedure gives detailed guidance on the test equipment and methods to be used for each penetration / valve. The following general philosophy is contained in the surveillance procedure.
. 3.1.1 Use air or nitrogen at a pressure differential across the valve greater than P. (50.6 - calculated accident pressure). 55 psig nitrogen was normally used. 4 3.1.2 Assure that the pressure is exerted in the accident test direction unless it can be demonstrated that pressurizing in the opposite direction is as conservative. Butterfly valves AH-V1B/lc were tested in the reverse direction. 3.1.3 Assure that the test volume is drained of liquid so that air or nitrogen test pressure is against valve seats. 3.1.4 Assure that the test verifies valve packing integrity in those cases where the packing would be an R. B. leakage boundary.
- 3.1.5 Assure adequate time period for stabilization of test i
conditions. 3.1.6 Assure test equipment is calibrated and used in a manner consistent with the data accuracy desired (weekly meter standardization was performed during the test program to verify meters accurate within i 4% full scale (Ref. 7.1). 3.1.7 Assure valves to be tested are closed by the normal method prior to testing. 3.1.8 Document as-found conditions (prior to adjustments / repairs) and as-left conditions. 3.1.9 Record test instrument scale readings prior to doing any data
- corrections.
3.1.10 Assure that system drains and vents which could serve as I containment isolation valves, are closed and capped and ! tagged af ter completion of the test program. j 3.2 Access Hatch Test Methods f 3.2.1 Door Seal Leak Tests-Method Door seal leak tests were performed by use of SP 1303-11.25 (Ref. 7.6). This procedure gives detailed guidance on the j test equipment and methods to be used. I f ! 31.0 0111U l
The door seal tests are performed by pressurizing the interspace between the double seals on each door with metered air at the manufacturers recommended test pressure of 10 psig. After stabilization the air rotameter indicates the rate of air' input' required to maintain the test pressure , 3.2.2 Overall Hatch Leak Test -- Semi-annual overall hatch laak' testing was performed by use of TMI Unit 1 Surveillance Procedure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This procedure gives detailed guidance on the test-equipment and methods to be used. The overall integrated leak test verifies the integrity of all of the following barriers:
- a. Hatch shell/ welds,
- b. Rubber door seals,
- c. Teflon operating shaft packing,
- d. Bulkhead electrical penetrations,
- e. Penetration pressurization check valves,
- f. Emergency air flange and associated'"0" rings on outer bulkhead, 4
J
- s. Bulkhead equalizing ball valves and associated
, mounting flanges /"0" rings.
i The overall leak test is performed by pressurizing the hatch to greater than calculated accident pressure and observing the rate of pressure drop on a high accuracy (Heise) pressure sage. Pressure corrections are made by reference to a barometer. Minimum test' duration is 4 hours after a 1 hour stabilization period. 3.3 Penetration Pressurization - Method Quarterly readings were taken on the flow rotameters which are permanently installed in the penetration pressurization system. These readings represent the air / nitrogen makeup rate required to , maintain approximately 60 psig in mechanical penetrations and 30 ! psig in electrical penetrations. High meter readings have occasionally occurred but these have been attributed to leaks in the compression fittings in the penetration pressurization system or to malfunctioning (stuck) rotameters. Testing was per plant surveillance procedure SP 1303-11.24. r i 5 32.0 0111U
- 4. TEST EQUIPMENT USED 4.1 Valve Test Equipment (See Enclosure 1)
- a. Rotameters - Sets of 3 Mfge. - Brooks Inst. Co.
Model - 1114 Full View Ranges: Float Mat'l. Tube No. Range Pyrex R-2-15D 8-1.120 SCCM Sapphire R-2-15C 100-12.200 SCCM Carboloy R-6-15B 1,000-142,000 SCCM Accuracy i 2% full scale industrial accuracy
- b. Temperature Indicators (as follows or similar)
Mfgr. - Ashcroft Model - EH or AH / 3" or 5" bial Range - 300 -1300 F Accuracy - 1 20 F
- c. Pressure Indicators (as follows or similar)
Mfgr. - Ashcroft Model - 1279 1/2" Dial Range 60 or 0-100 psig Accuracy - 1 2 psig
- d. Pressure Regulator (as follows or similar)
Mfge. - Union Carbide Corp. Model - UPG 3-75-580 Range 100 psi output / 0-3000 psl input 33.0 0111U
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- e. Calibration Rotameters (Set of 2).
Mfge. - Brooks Inst. Co. j l Models - 1110-05K2B1249, 1110-08K2B1206 i Ranges 16,000 SCCM, 3,600-234,000 SCCM Repeatability - i 1/4% of instantaneous Accuracy - i 1% instantaneous
- f. Flow rate Calibrator Mfge. - Brooks Inst. Co.
Model - 1056A - Range O to 2,400 SCCM Accuracy - 0.2% of indicated volume 4.2 Access Hatch Test Equipment
- a. Precision Pressure Gage (as follows or similar)
Mfgr. _Heise Model - CM Range - 0.60 psig Resolution - 0.25 pcis Accuracy - 0.1%
- b. Barometer (as follows or similar) l Mfgr. - Pennwalt i
, Model - FA185260A I f Range - 10.8 - 15.5 psia l Resolution - 0.005 psia l Accuracy - 0.1% l l l l , 35.0 0111U l l
._ ~ _ _ _ , _ . . _ . _ __ _ _ .,. _ . _ ._ ._
-4.3 Penetration Pressurization Test Equipinent
- a. Flow Rotameters - (Permanent System Equipment)
Mfge. - Brooks Inst. Co. Model - 1114 Range 10 SCFH at 60 psia air Accuracy - t 2% Industrial accuracy 36.0 0111U 9
q a
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- 5. ,SUtel&RY AND INTERPRETATION OF DATA 5.1 Valve Test Results i
See tabulation of individual results in Attachment #2 for the few valve tests scheduled during 1985 (quarterly / semi-annual / maintenance retests). 5.2 Access Hatch Test Results i 5.2.1 'Overall Hatch Leakage - SP 1303-11.18 (Ref. 7.2) - See the computer tabulation of 1985 leak rates in Attachment No. 2. The leakages were considered to be satisfactory. 5.2.2 Door Seal Leakage - SP 1303-11.25 (Ref. 7.6) None of the door seal leak tests exceeded the 3 SCFM administrative leakage limit. Typically, the leakage was less than 1 SCFH. 5.3 Penetration Pressurization (PP) Leakage - SP 1303-11.24 (Ref. 7.5) Leakage Rates - SCFH Mechanical Electrical Date 03/15/85 ~2.8 0.0 06/15/85 1.0 0.0 09/11/85 45.0 0.0 > 12/12/85 _020 0.6 No penetration leakage problems were identified. Occasional meter problems were found and repaired and occasional leaks of tubing / pipe fittings in the PP system were located and eliminated. There is no technical specification limit on penetration pressurization system leakage. The system leakage is maintained as low as practical. 1 l 37.0 01110 1
5
- 6. ERROR ANALYSIS n'
6.1 Valve Testina Errors (For purge valves see Section 6.2) The flow meters used in the field have normal industrial accuracies of i 21 full scale lLn the 10-100% (15-150 mm) scale range. Prior to use, um versus seem graphs were developed for the meters by 10 point calibrations using high accuracy (i 1% instantaneous) lab rotameters. During the leak test program, weekly 3 point standardizations were performed on the field rotameters to verify continued accuracy. The acceptance criteria for these Je standardizations was a variance of no more than 4% from the calibration graphs. If meters were repaired or the 3 point standardization exceeded the inaccuracy limit a new 10 point calibration was performed. Scale readings on the leak rate procedure (SP 1303-11.18) data sheets were evaluated and corrected , using the methods in Attachment 1. Conservative bias was i introduced into the results by assuming 15 mm (10% of scale) as the minimum scale. 6.2 Access Hatch and Purge Valve Testing Errors The measured pressure drops were corrected by adding the minimum scale increment of the gage used for both the helse sage and the barometer. This conservatively corrected for the resolution and repeatability errors. Gages used were recently calibrated. A minimum one hour temperature / pressure stabilization period was used prior to each pressure drop test. The access hatches and purge valves are not instrumented to allow temperature corrections. 6.3 penetration Pressurization Testing Errors These test results are used for information only and do not count toward the total leakage limit for Technical Specification conformance. The meters, installed permanently in the system, have i 2% full scale industrial accuracy. 38.0 01110
F~ f
- 7. REFERENCES ,
7.1 1430-Y-22 Standardization of Flow Rotameters 7.2 SP 1303-11.18 Reactor Building Local Leak Rate Testing (Rev's. 28 through 33 w/TCN No. 1-85-0202/3). 7.3 Three Mile Island Unit 1 Technical Specification 4.4.1 7.4 TMI Surveillance File (for Data sheets) 7.5 SP 1303-11.24 R. B. Local Leakage Penetration Pressurization (Rev. 5) 7.6 SP 1303-11.25 R. B. Local Leakage Access Hatch Door Seals (Rev. 11). 39.0 0111U
l 4 i ATTACHMENTS 40.0 0111U
( ATTACID(ENT 1 RESULTS EVALUATION PROCEDURE (SP 1302,-11.18 Enclosuro 9) 41.0 01110
Attachment 1 R. B. LOCAL LEAN RATE TESTING RESULTS EVALUATION The vent rotameter reading will be used if it can be demonstrated by the test data that all significant CIV leakage is being accounted for. If CIV packing, fluid block check valve, or gasket leakage was evident the supply rotameter results will be used unless this non-seat leakage was measured reliably and documented. FOR USE OF SUPPLY ROTAMETER DATA: FOR USE OF VENT ROTAMETER DATA: Proc edure : Procedure: a) Record supply meter reading in (1) a) Record vent meter reading in (1) below#. Also identify the meter below.M used by tube # in (8) below and the metering pressure in (9), b) Record downstream verification meter reading in (2) below. Also b) Convert meter units in SCCM units identify the respective meters used using latest lah meter calibration in (8) below and the metering curve. Enter in (3) below. pressure in (9). c) Correct results for temperature. c) Convert meter units to SCCM units Enter supply temperature in (4) using latest lab meter calibration below, curve. Enter in (3) below. Calculate and enter in (7) below, d) Correct results for temperature. Enter vent temperature (OF) in (4) below, then Calculate and enter in (5) below. e) If measurements of any other significant leakage paths (fluid M If meter scale reading was less block check valve, packing) are than 15 mm (minimum scale) use being claimed enter corrected flow 15 mm in calculations. (SCCM) in (6) below. (MM) (SCCM) 530 , ( + ) convert ( + )X + 460 = SCCM (1) (2) (3) (4) (5)
+ SCCM (6) (Identify meters used) (6) @ = CIV Leakage SCCM (9) (Meter Pressures) (7) 42.0 0111U
6 h s ATTACHMENT 2 DATA 1985 TYPE C REACTOR BUILDING i.EAK RATE TESTING 43.0 0111U r
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1985 1985 1985 1985 1985 1985 1985 1985 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM) NO TAG DESC OPER SIZE ASFOUND ASLEFT COMTS85 DATE c** ******** ******** ****** *** ********** ********* ******* ******** 1 AH-VIA/B BFLY P/MO 48 306 ' 306 LOW 3/1/85 2 2ND 48 .001 429 HIGH 6/3/85 3 3RD 48 424 424 LOW 9/13/85 4 4TH 48 97584 332 FAILED 12/17/85 5 AH-V1C/D BFLY M0/P 48 1967 1947 OK 2/22/85 6 2ND 48 235000 1599 FAILED 6/7/85 7 3RD 48 1599 1599 8 OK 9/13/85 4TH 48 1580 1580 OK 12/14/85 9 10 11 CA-Vi GLOBE MO 1 .01 .01 12 CA-V2 GATE P t .61 .01 13 CA-V3 GLOSE MO 1 .41 .01 14 CA-V4A GLOBE MO 1 69 49 OK 07/10/85 15 CA-V4B GLOBE MO i .01 .01 16 CA-V5A GATE P 1 .01 .01 17 CA-V5B GATE P 1 .01 .01 18 CA-V13 GLOBE NO 1 48 68 OK 07/12/85 19 CA-V189 GATE P 2 .01 .01 20 CA-Vi92 LFT CHN N/A 2 .01 .01 21 22 23 CF-V2A GL0bE MO 1 .01 .01 24 CF-V2B GLOBE NO 1 .01 .01 25 CF-V12A LFT CHK N/A 1 .01 .01 26 CF-V12B LFT CHK N/A 1 .41 .01 27 CF-Vt9A GATE P 1 .01 .01 28 CF-V19B GATE P 1 .01 .01 29 CF-V20A GATE P , 1 .01 .01 30 CF-V20B GATE - P 1 .01 .01 31 CH-Vi BALL P 1 .61 .01 32 CM-V2 BALL P 1 .61 .01 33 CM-V3 BALL P 1 .01 .01 34 CM-V4 BALL P 1 .61 62 9/24/85 35 DH-V44 GLOBE HW 2 .01 .41 = 36 DH=V69 STOP CHK HW 2 .91 .01 37 38 39 FTTEAST FLANGE N/A 30 .01 .01 40 FTTWEST FLANGE N/A 30 .61 .01 41 HM=V1A GLOBE S .5 .et .01 l 42 HM=V1B GLOBE S .5 .et .61 ! 43 HM-V2A GLOBE S .5 .01 .01 l 44 HM-V2B GLOBE S .5 .01 .91 1 45 HM-V3A GLOBE S .5 .et .01 46 HM-V3B GLOBE S .5 .01 .61 l 47 HM-V4A GLOBE 5 .5 .01 .01 48 HM-V4B GLOBE S .5 .01 .91 49 HP-Vi GATE HW 6 .01 .01 50 HP-V6 GATE HW 6 .01 .01 51 HR-V2A/B GLOBE HW 2 .01 .01 52 HR-V4A/B GLOBE HW 2 .01 .01 53 HRV22A/B GLOBE S 2 .01 .01 54 HR-V23A GLOBE S 2 .01 .01 53 HR V23D GLOBE S 2 .01 .01 56 44.0
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1985 1985 1985 1985 1985 1985 1985 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM) NO TAG DESC OPER - SIZE ASFOUND ASLEFT COMTS85 DATE cm* **mame** ******** mane ** *** ********** *enemmene ***me** ******** 57 58 IA-V6/20 GLOBE HW 2 .01 .01 59 IC-V2 CATE MO 6 .01 .01 60 IC-V3 GATE P 6 .01 .01 61 IC-V4 GATE P 6 .01 .01 62 IC-V6 GATE P 3 .01 .01 63 IC=V16 CHECK N/A 4 .01 .01 64 IC-V10 CHECK N/A 6 .01 .01 - 65 LR-V1/10 GATE HW 6 .01 .01 66 LR-V4 GLOBE HW .75 .01 .01 67 LR-V5 GLOBE HW 2 .01 .01 68 LR-V6 GLOBE HW 2 .01 .01 69 LR-V49 GATE HW 6 .01 .01 70 71 72 MU-V2A GLOBE NO 2.5 .01 .01 73 MU-V2B GLOBE MO 2.5 .01 .01 74 MU-V3 GATE P 2.5 .01 .01 75 MU-V18 GATE P 2.5 .01 .61 76 MU-V20 GATE P 4 .01 .01 77 MU-V25 GLOBE MO 4 .01 .01 78 MU-V26 GATE P 6 .01 .01 79 MU-Vit6 PIST CHK N/A 1.5 .01 .01 80 81 NI-V26 GLOBE HW 1 .01 .01 82 NI-V27 GLOBE HW i .01 .01 83 NS-V4 GATE MO 1.5 .61 .01 84 NS-Vil CHECK N/A 8 .01 .01 85 NS-V15 GATE MO 8 .01 .01 86 NS-V35 GATE NO 8 .01 .01 87 88 09 PENET104 BLK FLG N/A 2 55 55 OK 06/05/85 90 PENET105 BLK FLG N/A 10 .01 .01 91 PENET106 BLK FLG N/A 4 .61 .01 92 PENET210 BLK FLG N/A 2 55 55 OK 06/05/85 93 PENET211 BLK FLG N/A 2 55 55 OK 06/05/85 94 PENET241 BLK FLG N/A 18 .41 .01 95 PP101/02 LFT CHK N/A 1/2 .01 .01 96 PP133/34 LFT CHK N/A 1/2 .01 .01 97 98 99 100 101 RB-V2A GATE MO 8 .01 .01 102 RB-V7 GATE MO 8 .01 .01 103 SA-V2/3 GLOBE HW 2 .01 .01 104 SF-V23 GATE HW 8 .01 .01 105 WDG-V3/4 GL/GA M0/50L 2 .01 .01 106 WDL-V303 GLOBE MO 4 .01 .01 107 WDL-V304 GATE D 4 .01 .01 100 WDL-V534 GATE P 0 .01 .01 109 WDL-V535 GATE P 8 .01 .01 110 111 45.0
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1985 1985 1985 1985 1985 1985 1985 1985 RESULTS GIVEN IN STD CUBIC CENTIMETERS PER MINUTE (SCCM) NO TAG DESC SIZE, own OPER ASFOUND
******** ******** ****** *** ********** ASLEFT COMTS85 DATE 112 EQPFLG FLANGE ********* ******* ********
N/A 216 .91 113 PERACCES MISC. N/A .01 114 96 462 462 2ND MISC N/A LOW 05/26/85 96 1957 1957 115 EMEACCES MISC N/A 96 2546 11/30/85 116 2ND MISC 2546 OK 04/07/85 N/A 96 2525 2525 117 OK 11/11/85 118 119 MINPATH 120 MAXPATH .01 .01 NO LLRT NO LLRT 121 ACC CRIT .01 .01 NO LLRT NO LLRT 194846 104846 46.0
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING Following is the terminology used in the previous computer data:
- 1) .1 - (Alone) or any other number other than zero in the first decimal place means test scheduled.
- 2) .01 - (Alone) means no data available for the year or that the test was delayed. (e.g. valve not installed yet or not in previous testing scope.)
- 3) .001 - (Or any number other than zero in the third decimal place) after a leak rate (i.e. 59500.001) means actual leak rate was greater than measured / recorded value.
- 4) AsFound - leak rate (SCCM) in the As-Found condition before any repairs or adjustments.
- 5) Astoft - The leak rate (SCCM) after any adjustments / repairs.
- 6) Dates - Date of the last acceptable test results for the item.
- 7) Desc - Description of the valve or penetration.
- 8) Oper - Type of valve operator (actuator).
- 9) Notest - The Tech Spec scope did not require this valve to be tested during the respective year.
- 10) Novalve -_This valve was installed during a later refueling outage.
- 11) Comments : Cognizant Engineer comments about the results:
A) Failed - Exceeded the plant established leakage rate limit from SP 1303-11.18 Enclosure 10 which made repair / adjustment necessary. B) High/ Low - Subjective judgement of cognizant engineer. Represents the results with respect to the leakage which the type of leakage barrier (e.g. gate valve, globe valve, check valve, flanse, etc.) is considered to be capable of without extraordinary repair / adjustment C) OK - No problems with leakage. D) Other - E.G. newvalvo, novalve, notest, repacked seatwork, stembent, etc. (self-oxplanatory). E) NO LLRT - No full found of LLRT was required, data obtained was either quarterly / semi-annual or maintenance related testing - not refueling frequency.
- 12) Size - The nominal pipe oire for the leakage barrier.
47.0 0111U
M APPENDIX H THREE MILE ISLAND UNIT 1 $ 1986 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT i l ! SP 1303-11.18 r i t ! 02/12/86 - 05/24/86 i 1 i i l l I l 48.0 0111U 1
INDEI - 1986 R. B. LOCAL LEAK RATE TESTING REPORT
- 1. PURPOSE
- 2. SUIGIARY OF WORK ACCOMPLISHED 2.1 Valve Testing / Repairs 2.2 Access Hatches 2.3 Penetration Pressurization
- 3. METHODS OF TESTING 3.1 Valves 3.2 Access Hatches 3.3 Penetration Pressurization
- 4. TEST EQUIPMENT USED 4.1 Valves 4.2 Access Hatches 4.3 Penetration Pressurization
- 5. SUlstARY AND INTERPRETATION OF DATA 5.1 Valves 5.2 Access Hatches 5.3 Penetration Pressurization
- 6. ERROR ANALYSIS 6.1 Valves 6.2 Access Hatches 6.3 Penetration Pressurization
- 7. REFERENCES
- 8. ATTACHKENTS 8.1 Results Evaluation Procedure / Repair Criteria 8.2 Tabulation of Individual Test Repairs 49.0 0111U
- -- . ~ . . . .
REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT 1986 REFUELING FREQUENCY
- 1. PURPOSE 1.1 To provide analysis to the Nuclear Regulatory Commission on the ninth periodic type B and type C leakage tests performed on the Three Mile Island Unit 1 Reactor Building.
l' This is in accordance with " Reactor Containment Leakage Testing for Water Cooled Power Reactors". Appendix J. Part 50, Title 10 Code of Federal Regulations which required the contents of this summary reporc to become part of the Type A test report along with the details of.any other type B and type C testing performed since the previous type A test (also required per technical specification 4.4.1.1.8). TMI Unit I restarted October 3, 1985, after a prolonged shut-down which began March 28, 1979 (THI Unit II accident). For this round of leak rate testing, the plant was in a 30 day maintenance outage which began March 22, 1986 and ended April 23, 1986.
- 2. SUlstARY OF WORK ACCOMPLISHED 2.1 Valve Testina/ Repairs Appendix J Type B and C leak tests were performed on the components j as listed in THI Unit 1 Technical Specification 4.4.1. In addition the following components were leak tested though not yet listed in the Technical Specification. A Tech. Spec. change was pending to add these:
4
- 1. HK-VIA/B, 2A/B, 3A/B, 4A/B - New System
- 2. NI-V26 Repairs were initiated on the following components due to higher than desirable leakage.
- 1. CA-V5A - Sten packing leakage repair.
- 2. RB-V7 - Stem packing leakage repair.
i 3. IC-V3 - Extensive seat / wedge work, i { 4. CA-V189 - Lubricated stem - retest. ' j 2.2 Access Hatch Testina/ Repairs i l, i 2.2.1 Door Seals SP 1303-11.25 (Ref. 7.6) ! Door seal leak tests were performed as required by Technical
- Specification 4.4.1.2.5. All of the seal tests satisfied the surveillance procedure administrative acceptance criteria.
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IQ Y Overall Hatch Test SP 1303-11.18 (Ref. 7.2) 2.2.2
~
Semiannual integrated type leak tests were performed on each access hatch in 1986 as required by Technical Specification 4.4.1.2.5. All of the tests satisfied the surveillance procedure administrative criteria. 2.3 Penetration Pressurization SP 1303-11.24 (Ref. 7.5) Quarterly readings were recorded from the flow rotameters which supply air pressure or nitrogen pressure to reactor building mechanical and electrical penetrations as required by Technical Specification 4.4.1.2.5. No penetration leakage problems were noted although flow meter malfunctions required meter repair, and occasional tubing leaks in the air supply system were found and eliminated. i J .w l l l i I l l 51.0 0111U 4 1 4
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- 3. MgTMODS OF TESTING 3.1 Valve Test Methods Testing was performed by use of TMI Unit 1 surveillance procedure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This procedure gives detailed guidance on the test equipment and methods to be used 1 for each penetration / valve. The following general philosophy is l
+. contained in the surveillance procedure.
3.1.1 Use air or nitrogen at a pressure differential across the valve greater than Pa (50.6 - calculated accident' pressure). 55 psig nitrogen was normally used. 3.1.2 Assure that the pressure is exerted in the accident test direction unless it can be demonstrated that pressurizing in the opposite direction is as conservative. Butterfly valves AH-V1B/lC, and globe valves WDG-V14, DH-V64, SA-V3, and IA-V20 were tested in the reverse direction. 3.1.3 Assure that the test volume is drained of liquid so that air or nitrogen test pressure is against valve seats. 3.1.4 Assure that the test verifies-valve packing integrity in l those cases where the packing would be an R. B. leakage ' i boundary. 3.1.5 Assure adequate time period for stabilization of test conditions. 3.1.6 Assure test equipment is calibrated and used in a manner consistent with the data accuracy desired (weekly meter standardization was performed during the test program to verify meters accurate within i 4% full scale, MP 1430-Y-22 Ref. 7.1). 3.1.7 Assure valves to be tested are closed by the normal method prior to testing. 3.1.8 Document as-found conditions (prior to adjustments / repairs) and as-left conditions. 3.1.9 Record test instrument scale readings prior to doing any data corrections. 3.1.10 Assure that system drains and vents which could serve as containment isolation valves, are closed capped, and tagged after completion of the test program. A training program prior to the refueling outage was performed to help assure that the above philosophy was understood by the personnel involved in the testing. 52.0 0111U m.
3.2 Access Hatch Test Methods 3.2.1 Door Seal Leak Tests-Method Door seal leak tests were performed by use of SP 1303-11.25 (Ref. 7.6). This procedure gives detailed guidance on the test equipment and methods to be used. The door seal tests are performed by pressurizing the interspace between the double seals on each door with metered air at the manufacturees recommended test pressure of 10 psig. After stabilization the air rotameter indicates the rate of air input required to maintain the test pressure. 3.2.2 Overall Hatch Leak Test -- Semi-annual overall hatch leak testing was performed by use of TMI Unit 1 Surveillance Procedure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This procedure gives detailed guidance on the test equipment and methods to be used. The overall integrated leak test verifies the integrity of all of the following barriers:
- a. Hatch shell/ welds,
- b. Rubber door seals,
- c. Teflon operating shaft packing,
- d. Bulkhead electrical penetrations,
- e. Penetration pressurization check valves,
- f. Emergency air flange and associated "O" rings on outer bulkhead,
- s. Bulkhead equalizing ball valves and ascociated mounting flanges /"0" rings.
l The overall leak test is performed by pressurizing the hatch to greater than calculated accident pressure and observing j the rate of pressure drop on a high accuracy (Heise) pressure sage. Pressure corrections are made by reference to a barometer. Minimum test duration is 4 hours after a 1 hour stabilization period. I S3.0 0111U
3.3 Penetration Pressurization - Methad Quarterly readings were taken on the flow rotameters which are persenently installed in the penetration pressurization system. These readings represent the air / nitrogen makeup rate required to maintain approximately 60 psig in mechanical penetrations and 30 pois in electrical penetrations. High meter readings have occasionally occurred but these have been attributed to leaks in the compression fittings in the penetration pressurization system or to malfunctioning (stuck) rotameters. Testing was per plant surveillance procedure SP 1303-11.24 (Ref. 7.5). 54.0 0111U
- 4. TEST EQUIPMENT USED 4.1 Valve Test Eeuimment (See Flaure 1)
- a. Rotameters - Sets of 3 Mfgr. - Brooks Inst. Co.
Model - 1114 Full View ' Ranges: Float Mat'l. Tube No. Ranae Pyrex R-2-15D 8-1.120 SCCM Sapphire R-2-15C 100-12.200 SCCM Carboloy R-6-15B 1,000-142,000 SCCM Accuracy i 2% full scale industrial accuracy
- b. Temperature Indicators (as follows or similar)
Mfge. - Ashcroft 4 Model - EH or AH / 3" or 5" Dial Range - 300 -1300 F Accuracy - t 24
- c. Pressure Indicators (as follows or similar)
Mfge. - Ashcroft Model - 1279 1/2" Dial Range 60 or 0-100 psig Accuracy - i 2 psig
- d. Pressure Regulator (as follows or simil.r)
Mfge. - Union Carbide Corp. Model - UPG 3-75-580 Range 100 psi output / 0-3000 psi input i 55.0 0111U l l
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- e. Calibration Rotameters (Set of 2)
Mfgr. - Brooks Inst. Co. Models - 1110-05K2B1249, 1110-08K2B1206 Ranges 16,000 SCCM, 3,600-234,000 SCCM Repeatability - t 1/4% of instantaneous Accuracy - t 1% instantaneous
- f. Flow rate Calibrator Mfge. - Brooks Inst. Co.
Model - 1056A Range O to 2,400 SCCM Accuracy - i 0.2% of indicated volume 4.2 Access Hatch Test Equipment
- a. Precision Pressure Gage (as follows or similar)
Mfge. - Heise Model - CM Range - 0.60 psig Resolution - 0.25 psig Accuracy - 0.1%
- b. Barometer (as follows or similar)
Mfge. - Pennwalt Model - FA185260A Range - 10.8 - 15.5 psia Resolution - 0.005 psia Accuracy - 0.1% 57.0 0111U i t
4.3 Penetration Pressurization Test Eauipment
- a. Flow Rotameters - (Permanent System Equipment)
Mfgr. - Brooks Inst. Co. Model - 1114 Range 10 SCFH at 60 psia air Accuracy - t 2% Industrial accuracy l 58.0 0111U
- 5. SUIBIARY AND INTERPRETATION OF DATA 5.1 Valve Test Results i
As-Found/As-Left Leakage - Also see tabulation of individual results in Attachment #2. Total Leakane Tech. Spec. Limit % Tech. Spec. Limit As-Found MAIPATH 97.845 SCCM 104.846 SCCM 93% As-Left MAIPATH 28,213 SCCM 104,846 SCCM 27% NOTE: The total shown above is cumulative by penetration and not the total of all valve leakages. i.e., only the highest valve leakage on each penetration is counted. This number is labeled as "MAXPATH" on the tabulation of results in Attachment 2. EXAMPLE: Penetration XYZ has three containment isolation valves inside the reactor building in parallel and one outside. The leakage from the three valves inside totals 500 SCCM and the outside valve is 1000 SCCM. The penetration 4 leakage is counted as 1,000 SCCM not 1,500 SCCM. 5.2 Access Hatch Test Results 5.2.1 Overall Hatch Leakage - SP 1303-11.18 (Ref. 7.2) - See the computer tabulation of 1986 leak rates in Attachment #2. 5.2.2 Door Seal Leakage - SP 1303-11.25 (Ref. 7.6) None of the door seal leak tests exceeded the 3 SCFH administrative leakage 10mit. Typically, the leakage was less than 1 SCFM. 5.3 Penetration Pressurization (PP) Leakage - SP 1303-11.24 (Ref. 7.5) ; s .w e Leakane Rates - SCFH l i Mechanical Electrical Date -- 03/22/86 2.0 0.5 06/12/86 20.0 0.0 ! 09/11/86 0.0 0.0 12/12/86 21.5 . 0.0 No penetration leakage probleem were identified. Occasional meter problems were found and repaired and occasional leaks of tubing / pipe fittings in the PP system were located and eliminated.' ) There is no technical specification limit on penetration pressurization system leakage. The system leakage is maintained as low as practical. 59.0 0111U
l
- 6. ERBOR ANALYSIS 6.1 Valve Testina Errors (For purge valves see Section 6.2)
The flow meters used in the field have normal industrial accuracies of i 2% full scale in the 10-100% (15-150 mm) scale range. Prior to use, em versus seem graphs were developed for the meters by 10 point calibrations.using high accuracy ( 1% instantaneous) lab rotameters. During the leak test program, weekly 3 point standardizations were performed on the field rotameters to verify continued accuracy. The acceptance criteria for these standardizations was a variance of no more than 4% from the calibration graphs. If meters were repaired or the 3 point standardization exceeded the incccuracy, limit a new 10 point calibration was performed. Scale readings on the leak rate procedure (SP 1303-11.18) data sheets were evaluated and corrected using the methods in Attachment 1. Conservative bias was introduced into the results by assuming 15 mm (10% of scale) as the minimum scale. Approximately half of the test results actually showed a minimum scale reading. More involved error corrections were not considered meaningful based on the acceptable total leakage as-found and the low total leakage as-left. 6.2 Access Hatch and Purze Valve Testina Errors The measured pressure drops were corrected by adding the minimum scale increment of the gage used for both the helse sage and the barometer. This conservatively corrected for the resolution and repeatability errors. Gages used were recently calibrated. A minimum one hour temperature / pressure stabilization period was used prior to each pressure drop test. The access hatches and purge valves are not instrumented to allow temperature corrections. 6.3 Penetration Pressurization Testina Errors These test results are used for information only and do not count toward the total leakage limit for Technical Specification conformance. The meters, installed permanently in the system, have i 2% full scale industrial accuracy. ( 60.0 0111U
- 7. REFERENCES 7.1 1430-Y-22 Standardization of Flow Rotameters (Rev. 4) 7.2 SP 1303-11.18 Reactor Building Local Leak Rate Testing (Rev. 35) 7.3 Three Mile Island Unit 1 Technical Specification 4.4.1 7.4 TMI Surveillance File (for Data sheets) 7.5 SP 1303-11.24 R. B. Local Leakage Penetration Pressu'rization (Rev. 6) 7.6 SP 1303-11.25 R. B. Local Leakage Access Hatch Door Seals (Rev. 9).
61.0 0111U
ATTACHMENTS 62.0 0111U
ATTACHMENT 1 RESULTS EVALUATION PROCEDURE (SP 1303-11.18 Enclosure 9) , 63.0 0111U
? Attachment 1 R. 8. LOCAL LEAK RATE TESTING RESULTS EVALUATION The vent rotameter reading will be used if it can be demor.strated by the test data that all significant CIV leakage is being accounted for. If CIV packing, fluid block check valve, or gasket leakage was evident the supply rotameter results will be used unless this non-seat leakage was measured reliably and documented. FOR USE OF SUPPLY ROTAMETER DATA: FOR USE OF VENT ROTAMETER DATA: Procedure: Procedure: a) Record sucolv meter reading in (1) a) Record vent meter reading in (1) below*. Also identify the meter below.e used by tube # in (8) below and the metering pressure in (9). b) Record downstream verification meter reading in (2) below. Also b) Convert meter units in SCCM units identify the respective meters used using latest lab meter calibration in (8) below and the metering curve. Enter in (3) below, pressure in (9). c) Correct results for temperature. c) Convert meter units to SCCM units Enter supply temperature in (4) using latest lab meter calibration below. curve. Enter in (3) below. Calculate and enter in (7) below. d) Correct results for temperature. Enter vent temperature (OF) in (4) below, then Calculate and enter in (5) below. e) If measurements of any other significant leakage paths (fluid
- If meter scale reading was less block check valve, packing) are than 15 mm (minimum scale) use being claimed enter corrected flow 15 mm in calculations. (SCCM) in (6) below.
(HM) (SCCM) 530 . ( + )C**'***( + )x + 460 = SCCM (1) (2) (3) (4) (5)
+ SCCM (8) (Identify meters used) (6)
= CIV Leakage SCCM
-(9) (;teter Pressures) (7)
- 64. 0 0111U
ATTACHMENT 2 DATA 1986 TYPE C REACTOR BUILDING LEAK RATE TESTING 65.0 0111U
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1986 1986 1986 1986 1986 1986 1986 1986 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SJCM) NU FAG DESC OPER SIZE ASFOUND ASLEFT COMTS86 DATE o** ******** unnamenn mammum *** **memmen* **mune*** **ummune ******** 1 AH-VIA/B BFLY P/MO 48 897 2028 OK 4/15/86 2 2ND 48 18433 234 HIGH 6/14/86 3 3RD 48 780 780 OK 9/25/86 4 4TH 48 168000.001 .01 B-NOLEAK 11/4/86 5 AH-V1C/D BFLY N0/P 48 2067 3995 OK 4/18/86 6 2ND 48 1540 1560 OK 6/14/86 7 3RD 48 1229 1229 OK 9/17/86 8 4TH 48 4600.001 1989 C(1989 11/4/86 9 10 11 CA-V1 GLOBE MO 1 de 63 OK 4/13/86 12 CA-V2 GATE P 1 453 397 OK 4/8/86 13 CA-V3 GLOBE MO 1 60 59 OK 4/11/86 14 CA-V4A GLOBE MO l' 41 43 OM 4/12/86 15 CA-V4B GLOBE MO 1 , 41 63 OK 4/12/86 16 CA-V5A GATE P 1 663 1783 OK 3/30/84 17 CA-V5B GATE . P 1 536 536 OK 3/31/86 18 CA-V13 GLOBE MO 1 60 59 OK 4/11/86 19 CA-V189 GATE P 2 2266 1802 HIGH 4/1/86 20 CA-V192 LFT CHK N/A 2 60 60 OK 4/1/86 21 22 23 CF-V2A GLOBE MO 1 45 45 OK 3/23/86 24 CF-V2B GLOBE MO i 45 45 OK 3/24/86 25 CF-V12A LFT CHK N/A 1 219 219 OK 3/29/86 26 CF-V12B LFT CHK N/A i 441 441 OK 3/29/86 27 CF-V19A GATE P 1 314 314 DK 3/30/86 28 CF-V19B GATE P 1 140 140 OK 3/30/86 29 CF-V20A GATE P 1 45 45 OK 3/23/86 30 CF-V20B GATE P 1 45 45 OK 3/24/86 31 CM-V1 BALL P i 57 57 OK 2/12/86 32 CM-V2 BALL P 1 57 57 OK 2/12/86 33 CM-V3 BALL P 1 57 57 OK 2/12/86 34 CM-V4 BALL P 1 80 80 OK 2/12/86 35 DH-V64 GLOBE HW 2 167 167 OK 4/5/86 36 DH=V69 STOP CHK HW 2 64 64 OK 4/4/86 37 38 39 FTTEAST FLANGE N/A 30 139 139 UK 4/2/86 40 FTTWEST FLANGE N/A 30 68 68 OK 4/3/86 41 HM=VIA GLOBE S .5 45 45 OK 3/23/86 42 HM=VfB GLOBE S .5 45 45 OK 3/23/86 43 HM-V2A GLOBE S .5 45 45 OK 3/23/86 44 HM-V2B GLOBE 3 .5 45 45 OK 3/23/86 45 HM-V3A GLOBE S .5 45 45 OK 3/23/86 46 HM-V3B GLOBE S .5 45 45 OK , 3/23/86 47 HM-V4A GLOBE S .5 45 45 OK 3/23/86 48 HM-V4B GLOBE I .5 45 45 OK 3/23/86 49 HP-Vi GATE HW 6 63 63 OK 4/2/86 50 HP-V6 GATE HW 6 94 . 94 OK 4/2/86 51 HR-V2A/B GLOBE HW 2 63 ' 63 OK 4/19/06 l 52 HR-V4A/B GLOBE HW 2 281 63 OK 4/19/86 53 HRV22A/B GLOBE S 2 63 63 OK 4/19/86 54 HR-V23A GLOBE S 2 63 63 OK 4/19/86 55 HR-V23B GLOBE S 2 63 63 OK 4/19/86 56 I 66.0
I LOCAL LEAK RATE IEST RESULIS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1986 1986 1986 1986 1986 1986 1986 1986 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM) NO TAG DESC OPER SIZE ASFOUND ASLEFT on* COMTS86 DATE
**u***** unamanne .**me** **# ######### ********* ******** ********
57 58 IA-V6/20 GLOBE HW 2 49 86 OK 3/24/86 59 IC-V2 GATE NO 6 227 227 OK 3/24/86 de IC-V3 GATE P 4 62977 3103 FAILED 61 .IC-V4 3/27/86 GATE P 6 1209 1209 OK 3/26/86 62 IC-V6 GATE P 3 45 45 OX 3/26/86 63 IC=V16 CHECK N/A 4 242 242 OK 44 3/27/96 IC-V18 CHECK N/A 4 45 45 OK 3/27/86 65 LR-V1/10 GATE HW 6 123 123 OK 3/26/86 64 LR-V4 GLOBE HW .75 44 44 OK 4/1/86 67 LR-V5 GLOBE HW 2 45 45 OK 3/22/86 48 LR-V4 GLOBE HW 2 , 45 45 OK 3/22/84 69 LR-V49 GATE HW 6 138 138 DK 4/1/86 70 0 0 71 0 0 72 MU-V2A GLOBE MO 2.5 58 44 OK 4/10/86 73 MU-V2B GLOBE NO 2.5 58 64 OK 4/10/86 74 MU-V3 GATE P 2.5 251 251 OK 3/29/86 75 MU-V18 GATE P 2.5 430 430 OK 4/2/86 76 MU-V20 GATE P 4 43 63 OK 3/29/86 77 MU-V25 GLOBE NO 4 127 449 OK 4/13/86 78 MU-V26 GATE P 6 60 60 OK 3/29/86 79 MU-Vite PIST CHN N/A 1.5 2702 2702 HIGH 3/29/86 80 0 0 81 NI-V26 GLOBE HW 1 di 61- OK 4/5/86 82 NI-V27 GLOBE HW i 61 41 OK 4/5/86 83 NS-V4 GATE NO 1.5 44 44 LOW 3/27/86 84 NS-Vit CHECK N/A 8 1625 1625 OK 3/28/86 85 NS-V15 GATE MO 8 44 44 LOW 3/28/96 86 NS-V35 GATE NO 8 404 404 OK 3/27/86 87 0 0 88 0 0 89 PENET194 BLK FLG N/A 2 59 59 OK 4/19/86 99 PENET105 BLK FLG N/A 10 43 63 OK 4/3/84 91 PENET106 BLK FLG N/A 4 63 63 OK 4/3/86 92 PENET210 BLK FLG N/A 2 44 44 OK 4/19/86 I 93 PENET211 BLK FLG N/A 2 46 46 OK 4/19/86 94 PENET241 BLK FLG N/A 18 59 59 OK 4/10/86 95 PP191/02 LFT CHN N/A 1/2 .61 .61 NOTEST 96 PP133/34 LFT CHN N/A 1/2 .01 .01 NOTEST i 97 98 99 100 101 RB-V2A GATE NO 8 1447 1447 OK 4/6/86 102 RB-V7 GATE MO 8 14013 392 PACKING 4/7/84 103 SA-V2/3 GLOBE HW 2 46 44 OK 4/19/86 104 SF-V23 GATE HW 8 34 34 OK 3/31/86 105 WDG-V3/4 GL/GA M0/SQL 2 330 330 OK 3/30/06 106 WDL-V303 GLOBE MO 4 108 108 OK 3/30/06 107 WDL-V304 GATE D 4 60 60 OK 3/30/86 , 1G8 WDL-V534 GATE P 8 63 63 LOW 4/3/84 109 WDL-V535 GATE P 8 63 63 LOW 4/3/86 , 110 til 67.0
I LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING l 1986 1986 1986 1986 1986 1986 1986 1986 l RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM) l l NO TAG DESC DPER IIZE ASFOUND ASLEFT COMTS86 DATE com ammamene mammmmme mamene een mammeaume naammaans mammmmme mammanum 112 EQPFLG FLANGE N/A 216 46 44 OK 4/18/86 113 PERACCES MISC. N/A 96 462 462 LOW 5/24/84 114 2ND MISC N/A 96 .1 .1 OUTAGE 115 EMEACCES MISC N/A 96 1753 1753 OK 5/16/86 116 2ND MISC N/A 96 .1 .1 OUTAGE 117 118 119 MINPATH 24625.02 14054.02 OK 120 MAXPATH* 97845.02 28213.02 HIGH 121 ACC CRIT 194846 194846
- The semiannual and quarterly tests performed nearest to the round of refueling frequency testing are added into the "MAXPATH" calculation. For valves in series the valve with highest leakage is used in the "MAXPATH" calculation.
For valves in parallel or where the inner and outer valves can not be tested separately, the total leakage is used. 68.0
LOCAL LEAK RATE TEST RESULTS l i THREE MILE ISLAND UNIT 1 REACTOR BUILDING l Following is the terminology used in the previous computer data:
- 1) .1 - (Alone) or any other number other than zero in the first decimal place means test scheduled.
- 2) .01 - (Alone) means no data available for the year or that the test was delayed. (e.g. valve not installed yet or not in previous testing scope.)
- 3) .001 - (or any number other than zero in the third decimal place) after a leak rate (i.e. 59500.001) means actual leak rate was greater than measured / recorded value.
- 4) AsFound - leak rate (SCCM) in the As-Found condition before any repairs or adjustments.
- 5) AsLeft - The leak rate (SCCM) after any adjustments / repairs.
- 6) Datas - Date of the last acceptable test results for the item.
- 7) Desc - Description of the valve or penetration.
l 8) Oper - Type of valve operator (actuator).
- 9) Notest - The Tech Spec scope did not require this valve to be tested during the respective year.
- 10) Novalve - This valve was installed during a later refueling outage. -
- 11) Comments - Cognizant Engineer comments about the results:
A) Failed - Exceeded the plant established leakage rate limit from SP 1303-11.18 Enclosure 10 which made repair /adjusLaent necessary. B) High/ Low - Subjective judgement of cognizant engineer. Represents the results with respect to the leakage which the type of leakage barrier (e.g. gate valve, globe valve, check valve, flange, etc.) is considered to be capable of without extraordinary repair / adjustment C) OK - No problems with leakage, j D) Other - E.G. newvalve, novalve, notest, repacked seatwork, stembent, etc. (self-explanatory). E) B-NOLEAK - B valve did not leak as indicated by soap bubble l testing. All leakage was across AH-V1A; no 'ASLT' data due to the 6R Outage. C<1989 - indicates that the leakage across the C valve was less j than 1989 SCCM. AH-VlD seats alone were adjusted. 1 j 69.0 0111U
NOTEST - PP-V101/102/133/134 were added to testing scope for 6R Outage (1987). OUTAGE - Testing not performed due to 6R Outage.
- 12) Size - The nominal pipe size for the leakage barrier.
( 70.0 0111U
APPENDIE I THREE MILE ISLAND UNIT 1 1987 REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT SP 1303-11.18 11/4/86 to end of current outage v 71.0 0111U
I'! DEI - 1987 R. B. LOCAL LEAK RATE TESTING REPORT 1 I
- 1. PURPOSE
- 2.
SUMMARY
OF WORK ACCOMPLISHED l 2.1 Valve Testing / Repairs 2.2 Access Hatches 2.3 Penetration Pressurization
- 3. METHODS OF TESTING 3.1 Valves 3.2 Access Hatches 3.3 Penetration Pressurization
- 4. TEST EQUIPMENT USED 4.1 Valves 4.2 Access Hatches 4.3 Penetration Pressurization
- 5.
SUMMARY
AND INTERPRETATION OF DATA 5.1 Valves 5.2 Access Hatches 5.3 Penetration Pressurization
- 6. ERROR ANALYSIS 6.1 Valves 6.2 Access Hatches 6.3 Penetration Pressurization
- 7. REFERENCES
- 8. ATTACHMENTS 8.1 Results Evaluation Procedure / Repair Criteria 8.2 Tabulation of Individual Test Repairs 72.0 0111U
REACTOR BUILDING LOCAL LEAK RATE TESTING REPORT 1987 REFUELING FREQUENCY
- 1. PURPOSE 1.1 To provide analysis to the Nuclear Regulatory Commission on the tenth periodic type B and type C leakage tests performed on the Three Mile Island Unit 1 Reactor Building.
This is in accordance with " Reactor Containment Leakage Testing for Water Cooled Power Reactors". Appendix J. Part 50, Title 10 Code of Faderal Regulations which required the contents of this summary report to become part of the Type A test report along with the details of any other type B and type C testing performed since the previous type A test (also required per technical specification 4.4.1.1.8). TMI Unit I restarted October 3, 1985 after a prolonged shutdown which began March 28, 1979 (TMI Unit II accident). For this round of leak rate testing the plant was shutdown for the 6R Refueling Outage. Testing began on November 2, 1986 and was still in progress at the time this report was submitted.
- 2.
SUMMARY
OF WORK ACCOMPLISHED 2.1 Valve Testina/ Repairs Appendix J Type B and C leak tests were performed on the components as listed in TMI Unit 1 Technical Specification 4.4.1. In addition the following components were leak tested though not yet listed in the Technical Specification.
- 1. HM-VIA/B, 2A/B, 3A/B, 4A/B - New System
- 2. NI-V26.
- 3. PP-V101/102/133/134 Repairs were initiated on the following components due to higher than desirable leakage.
- 1. IC-V3 replaced with a new valve - upgrade
- 2. IC-V4 replaced with a new valve - upgrade
- 3. AH-VIA - Seat leakage 4
- 4. AH-VID - Seat leakage 73.0 Oll1U
- 5. Penet 211 - Test connection cap leak - replaced improper fitting (double / tubing connector as-found rather than the single connector desired. Resulted in pipe threads mating with tubing threads with resultant thread leakage).
- 6. NU-V2A - Packing leakage 2.2 Access Hatch Testinz/ Repairs 2.2.1 Door Seals SP 1303-11.25 (Ref. 7.6)
Door seal leak tests are performed as required by Technical Specification 4.4.1.2.5. Testing of door seals has been postponed until the end of the 6R Outage - containment integrity is currently not required. 2.2.2 Overall Hatch Test (Ref. 7.5) Semiannual integrated type leak tests were not performed because the plant was in cold shutdown and containment integrity was not required. These tests will be performed when the 6R Outage is complete. 2.3 Penetration Pressurization SP 1303-11.24 (Ref. 7.5) Quarterly readings are recorded from the flow rotameters which supply air pressure or nitrogen pressure to reactor building mechanical and electrical penetrations as required by Technical Specification 4.4.1.2.5. No penetration pressurization tests have been performed for 1987 as of yet, the initial quarterly test is scheduled for March 12, 1987. i l l l i l 74.0 0111U t
- 3. METHODS OF TESTING 3.1 Valve Test Methods Testing was performed by use of TMI Unit 1 surveillance procedure SP 1303-11.18 Reactor Building Local Leak Rate Testing. This procedure gives detailed guidance on the test equipment and methods to be used for each penetration / valve. The following general philosophy is contained in the surveillance procedure.
3.1.1 Use air or nitrogen at a pressure differential across the valve greater than Pa (50.6 - calculated accident pressure). 55 psig nitrogen was normally used. , 3.1.2 Assure that the pressure is exerted in the accident test direction unless it can be demonstrated that pressurizing in the opposite direction is as conservative. Butterfly valves AH-V1B/1C, and globe valves WDG-V4, DH-V64, SA-V3, and IA-V20 were tested in the reverse direction. 3.1.3 Assure that the test volume is drained of liquid so that air or nitrogen test pressure is against valve seats. 3.1.4 Assure that the test verifles valve packing integrity in those cases Where the packing would be an R. B. leakage boundary. 3.1.5 Assure adequate time period for stabilization of test conditions. 3.1.6 Assure test equipment is calibrated and used in a manner consistent with the data accuracy' desired (weekly meter standardization was performed during the test program to i verify meters accurate within i 4% full scale (Ref. 7.1).
- 3.1.7 Assure valves to be tested are closed by the normal method
! prior to testing. I 3.1.8 Document as-found conditions (prior to adjustments / repairs) and as-left conditions. , 3.1.9 Record test instrument scale readings prior to doing any data corrections. 3.1.10 Assure that system drains and vents which could serve as l containment isolation valves, are closed and capped and tagged after completion of the test program. J A training program prior to the refueling outage was performed to help assure that the above philosophy was understood by the personnel involved in the testing. 75.0 0111U
4 3.2 ' Access Hatch Test Methods
^
i , 3.2.1 Door Seal Leak Tests-Method- j Door seal leak tests were performed by use of SP 1303-11.25 (Ref. 7.6). This procedure gives detsiled guidance on the l test equipment and methods to be used. The door seal tests are performed by pressurizing the interspace between the double seals on each door with metered air at the manufacturers recomunended test pressure of 10 - psig. After stabilization the air rotameter indicates the rate of air input required to maintain the test pressure. 3.2.2 Overall Hatch Leak Test -- Semi-annual overall hatch leak testing was performed by use of TMI Unit 1 Surveillance Procedure SP 1303-11.18 Reactor Building Local'Lesk Rate Testing. This procedure gives detailed guidance on the test
. equipment and methods to be used. The overall integrated leak test, verifies the integrity of all of the following
- . barriers
- I
- a. Hatch shell/ welds, i
\
- l b. Rubber door seals,
- c. Teflon operating shaft packing,-
- d. Bulkhead electrical penetrations,
- e. Penetration pressurization check valves,
- f. Emergency air flange and associated "O" rings on outer bulkhead,
- s. Bulkhead equalizing ball valves and associated
- mounting flanges /"0" rings.
The overall leak test is performed by pressurizing the hatch to greater than calculated accident pressure and observing j the rate of pressure drop on a high accuracy (Heise) pressure j gage. Pressure corrections are made by reference to a barometer. i Minimum test duration is 4 hours after a 1 hour stabilization ., period. 1 1 76.0 0111U
3.3 Penetration Pressurization - Method Quarterly readings were taken on the flow rotameters which are permanently installed in the penetration pressurization system. ' These readings represent the air / nitrogen makeup rate required to maintain approximately 60 psig in mechanical penetrations and 30 psig in electrical penetrations. High meter readings have occasionally occurred but these have been attributed to leaks in the compression fittings in the penetration pressurization system or to malfunctioning (stuck) rotameters. Testing was per plant surveillance procedure SP 1303-11.24 (Ref. 7.5). i 4 77.0 0111U
1 l l 4 l
- 4. TEST EQUIPMENT USED 4.1 Valve Test Equipment (See Finure 1)
- a. Rotameters - Sets of 3 Mfge. - Brooks Inst. Co.
Model - 1114 Full View Ranges: Float Mat'l. Tube No. Ranae Pyrex R-2-15D 8-1.120 SCCM Sapphire R-2-15C 100-12.200 SCCM Carboloy R-6-15B 1,000-142,000 SCCM Accuracy i 2% full scale industrial accuracy
- b. Temperature Indicators (as follows or similar)
Mfgr. - Ashcroft Model - EH or AH / 3" or 5" Dial Range - 30 0 -130 0F Accuracy - i 20 F
- c. Pressure Indicators (as follows or similar)
Mfge. - Ashcroft Model - 1279 1/2" Dial Range 60 or 0-100 psig Accuracy - t 2 psig ,
- d. Pressure Regulator (as follows or similar)
Mfge. - Union Carbide Corp. Model - UPG 3-75-580 Range 100 psi output / 0-3000 psi input 78.0 0111U
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- e. Calibration Rotameters (Set of 2)
Mfgr. - Brooks Inst. Co. Models - 1110-05K2B1249, 1110-08K2B1206 Ranges 16,000 SCCM, 3,600-234,000 SCCM Repeatability - i 1/4% of instantaneous Accuracy - 1% instantaneous
- f. Flow rate Calibrator Mfgr. - Brooks Inst. Co.
Model - 1056A Range O to 2,400 SCCM Accuracy - 0.2% of indicated volume 4.2 Access Hatch Test Equipment I
- a. Precision Pressure Cage (as folicws or similar)
Mfge. - Heise Model - CM Range - 0.60 psig Resolution - 0.25 psig Accuracy - 0.1%
- b. Barometer (as follows or similar)
Mfgr. - Pennwalt Model - FA185260A l [ Range - 10.8 - 15.5 psia Resolution - 0.00S psia l l Accuracy - 0.1% i i l 80.0 0111U
o 4.3 Penetration Pressurization Test Rouipment
- a. Flow Rotameters - (Permanent System Equipment)
Mfgr. - Brooks Inst. Co. Medel - 1114 Range 10 SCFH at 60 psia air Accuracy - t 2% Industrial accuracy 81.0 0111U
.- = - ,_.
- 5. SUlstARY AND INTERPRETATION OF DATA 5.1 Valve Test Results As-Found/As-Left Leakage to this date - Also see tabulation of i individual results in Attachment #2.
Total Leakate Tech. Spec. Limit % Tech. Spec. Limit As-Found MAIPATH >156,224 SCCM 104,846 SCCM >100% As-Left MAIPATH 13,981 SCCM 104,846 SCCM <20% NOTE: The total shown above is " maximum pathway" and not the total of all valve leakages. i.e., only the highest valve leakage on each penetration is counted. This number is labeled as "MAXPATH" on the tabulation of results in Attachment 2. EIAMPLE: Penetration KYZ has three containment isolation valves
..i" inside the reactor building in parallel and one outside. The leakage from the three inside totals 500 SCCM and the outside valve is 1000 SCCM. The ,
penetration leakage is counted as 1,000 SCCM not 1,500 SCCM. 5.2 Access Hatch Test Results 5.2.1 overall Hatch Leakage - SP 1303-11.18 (Ref. 7.2) - Testing of the Personnel Access and Emergency hatch has been postponed until after the 6R Outage - containment integrity is currently not required 5.2.2 Door Seal Leakage - SP 1303-11.25 (Ref. 7.6) Testing of the door seals has been posponed until the end of the 6R Outage - containment integrity is currently not required. 5.3 Penetration Pressurization (PP) Leakage - SP 1303-11.24 (Ref. 7.5) Leakaze Rates - SCFH Mechanical Electrical l Date 12/12/86 21.5 0.0 First Quarter '87 Not Performed
- Not performed *
- Not yet performed at the time this report was submitted.
There is no technical specification limit on penetration l pressurization system leakage. The system leakage is maintained as l low as practical. l 82.0 0111U l I
- 6. ERROR ANALYSIS 6.1 Valve Testina Errors (For purge valves see Section 6.2)
The flow meters used in the field have normal industrial accuracies of i 2% full scale in the 10-100% (15-150 mm) scale range. Prior to use, mm versus seem graphs were developed for the meters by 10 point calibrations using high accuracy (i 1% instantaneous) lab rotameters. During the leak test program, weekly 3 point standardizations were performed on the field rotameters to verify continued accuracy. The acceptance criteria for these standardizations was a variance of no more than 4% from the , calibration graphs. If meters were repaired or the 3 point standardization exceeded the inaccuracy limit, a new 10 point calibration was performed. Scale readings on the leak rate procedure (SP 1303-11.18) data sheets were evaluated and corrected using the methods in Attachment 1. Conservative bias was introduced into the results by assuming 15 mm (10% of scale) as the minimum scale. Approximately half of the test results actually showed a minimum scale reading. More involved error corrections were not considered meaningful based on the acceptable total leakage as-found and the low total leakage as-left. 6.2 Access Hatch and Purae Valve Testina Errors I The measured pressure drops were corrected by adding the minimum scale increment of the sage used for both the heise sage and the barometer. This conservatively corrected for the resolution and repeatability errors. Gages used were recently calibrated. A minimum one hour temperature / pressure stabilization period was used prior to each pressure drop test. The access hatches and purge valves are not instrumented to allow temperature corrections. 6.3 Penetration Pressurization Testina Errors These test results are used for inforos. tion only and do not count
.toward the total leakage limit for Technical Specification conformance. The meters, installed permanengly in the system, have i 2% full scale industrial accuracy.
I l 83.0 Oll1U
- 7. REFERENCES 7.1 1430-Y-22 Standardization of Flow Rotameters (Rev. 4) 7.2 SP 1303-11.18 Reactor Building Local Leak Rate Testing (Rev. 38) w/TCN 1-86-0179 (RB-V2A/7 Testing).
7.3 Three Mile Island Unit 1 Technical Specification 4.4.1 7.4 THI Surveillance File (for Data sheets) 7.5 SP 1303-11.24 R. B. Local Leakage Penetration Pressurization (Rev. 6) 7.6 SP 1303-11.25 R. B. Local Leakage Access Hatch Door Seals (Rev. 11). S4.0 0111U
j? ATTACHMENTS 85.0 0111U
ATTACHMENT 1 RESULTS EVALUATION PROCEDURE (SP 1303-11.18 Enclosure 9) 86.0 0111U
Attachment 1 R. B. LOCAL LEAK RATE TESTING RESULTS EVALUATION The vent rotameter reading will be used if it can be demonstrated by the test data that all significant CIV leakage is being accounted for. If CIV packing, fluid block check valve, or gasket leakage was evident the supply rotameter results will be used unless this non-seat leakage was measured reliably and documented. I FOR USE OF SUPPLY ROTAMETER DATA: FOR USE OF VENT ROTAMETER DATA: Procedure : Procedure : a) Record supply meter reading in (1) a) Record vent meter reading in (1) below#. Also identify the meter below.* used by tube # in (8) below and the metering pressure in (9). b) Record downstream verification meter reading in (2) below. Also b) Convert meter units in SCCM units identify the respective meters used using latest lab meter calibration in (8) below and the metering curve. Enter in (3) below, pressure in (9). c) Correct results for temperature. c) Convert meter units to SCCM units Enter supply temperature in (4) using latest lab meter calibration below. curve. Enter in (3) below. Calculate and enter in (7) below. d) Correct results for temperature. Enter vent temperature (OF) in (4) below. then Calculate and enter in (5) below. e) If measurements of any other significant leakage paths (fluid
- If meter scale reading was less block check valve, packing) are than 15 mm (minimum scale) use being claimed enter corrected flow 15 mm in calculations. (SCCM) in (6) below.
(MM) (SCCM) 530 , ( + )C "'***( + )X + 460 = SCCM (1) (2) (3) (4) (5)
+ SCCM (8) (Identify meters used) (6) @ = CIV Leakage SCCM (9) (Meter Pressures) (7) 87.0 0111U
i I 1 l ATTACHMENT 2 DATA 1987 TYPE C REACTOR BUILDING LEAK RATE TESTING 88.0 0111U
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING f 1987 1987 1987 1987 1987 1987 1987 1987 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM) NO TAG DESC OPER SIZE ASFOUND ASLEFT CONTS87 DATE ce* ******** ******** ****** *** ********** ********* ******** ******** i 1 AH-VIA/B BFLY P/MO 48 .01 .61 , 2 2ND 48 .61 .01 l 3 3RD 48 .61 .01 l 4 4TH 48 .01 .01 1 5 AH-V1C/D BFLY M0/P 48 .61 .01 6 2ND 48 .01 .01 7 3RD 48 .61 .01 8 4TH 48 .01 .01 l 9 10 11 CA-V1 GLOBE MO 1 71 71 OK 11/29/06 12 CA-V2 GATE P 1 1132 1132 HIGH 11/29/86 13 CA-V3 GLOBE MO 1 71 71 OK 11/29/86 ! 14 CA-V4A GLOBE MO 1 71 71 OK 12/3/86 15 CA-V4B GLOBE MO 1 98 98 OK 12/3/86 16 CA-V5A GATE P 1 4407 .01 HIGH 17 CA-V5B GATE P 1 4047 .01 HIGH 18 CA-Vt3 GLOBE MO 1 71 71 OK 11/29/86 19 CA-V189 GATE P 2 1389 1389 HIGH 12/1/86 - i 20 CA-V192 LFT CHK N/A 2 71 71 OK 12/1/86 21 22 23 CF-V2A GLOBE MO 1 40 40 OK 11/21/86 24 CF-V2B GLOBE MO 1 38 38 OK 11/22/86 25 CF-V12A LFT CHK N/A i 13 13 OK 11/25/86 26 CF-V12B LFT CHK N/A i 88 88 OK 11/25/86 27 CF-V19A GATE P 1 1306 1306 HIGH 11/26/86 28 CF-V19B GATE P 1 32 32 OK 11/25/86 29 CF-V20A GATE P 1 367 367 OK 11/22/06 30 CF-V20B GATE P 1 38 38 OK 11/22/86 31 CM-Vi BALL P 1 75 75 OK 12/9/86 32 CM-V2 BALL P 1 319 319 HIGH 12/9/86 33 CM-V3 BALL P 1 75 75 OK 12/9/86 34 CM-V4 BALL P 1 75 75 OK 12/9/86 35 DH-V64 GLOBE HW 2 39 39 OK 1/6/87 36 DH=V69 STOP CHK HW 2 46 46 OK 1/6/87 37 38 39 FTTEAST FLANGE N/A 30 145 145 OK 12/9/86 40 FTTWEST FLANGE N/A 30 52 52 OK 12/9/86 41 HM=V1A GLOBE S .5 75 75 OK 12/6/86 42 HM=ViB GLOBE S .5 75 75 OK 12/6/06 43 HM-V2A GLOBE S .5 75 75 OK 12/6/86 44 HM-V2B GLOBE S .5 75 75 OK 12/6/86 45 HM-V3A GLOBE S .5 75 75 OK 12/6/86 46 HM-V3B GLOBE S .5 75 75 OK 12/6/86 47 HM-V44 GLOBE S .5 75 75 OK 12/6/86 48 HM-V4B GLOBE S .5 75 75 OK 12/6/86 4 49 HP-Vi GATE HW 6 75 75 OK 12/9/86 50 HP-V6 GATE HW 6 75 75 OK 12/9/86 51 HR-V2A/B GLOBE HW 2 66 66 OK 12/10/86 52 HR-V4A/B GLOBE HW 2 273 76 OK 12/10/86 53 HRV22A/B GLOBE S 2 76 76 OK 12/10/86 54 HR-V23A GLOBE S 2 76 76 OK 12/10/86 55 HR-V23B GLOBE S 2 76 76 OK 12/10/86 56 89.0
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1987 1987 1987 1987 1987 1987 1987 1987 RESULTS CIVEN IN STD. CUBIC CENTIMETERS PER MINUTE (SCCM) NO TAG DESC OPER SIZE ASFOUND ASLEFT COMTS87 ,DATE com ******** ******** ****** *** ********** ********* ******** ******** 57 58 IA-V6/20 GLOBE HW 2 52 52 OK 12/5/86 59 IC-V2 GATE MO 6 190 190 OK 11/12/86 40 IC-V3 GATE P & 18500 66 NEWVALVE 2/3/87 61 IC-V4 GATE P 4 71 42 NEWVALVE 1/24/87 62 IC-V6 GATE P 3 47 47 OK 11/17/86 63 IC=V16 CHECK N/A 4 141 141 OK 11/17/86 44 IC-V18 CHECK N/A 6 71 71 OK 11/11/86 65 LR-V1/10 GATE HW 6 1904 1904 OK 12/5/86 66 LR-V4 GLODE HW .75 52 52 OK 12/5/86 67 LR-V5 GLOBE HW 2 52 52 OK 12/5/86 68 LR-V6 GLOBE HW 2 52 52 OK 12/5/86 69 LR-V49 GATE HW 6 76 76 OK 12/5/86 70 71 72 MU-V2A GLOBE MO 2.5 52875 .01 PACKING 73 MU-V2B GLOBE NO 2.5 72 72 OK 74 MU-V3 GATE P 2.5 .01 .01 75 MU-V18 GATE P 2.5 330 330 OK 12/8/86 76 MU-V20 GATE P 4 66 66 OK 11/15/86 77 MU-V25 GLOBE MO 4 88 88 OK 11/26/86 78 MU-V26 GATE P 6 88 88 OK 11/26/86 79 MU-V116 PIST CHK N/A 1.5 1319 1319 OK 11/15/06 80 81 NI-V26 GLOBE HW i 71 71 OK 12/2/86 82 NI-V27 GLOBE HW i 71 71 OK 12/2/06 83 NS-V4 GATE . MO 1.5 621 621 OK 11/20/86 84 NS-V11 CHECK N/A 8 1810 1810 OK 11/19/86 85 NS-V15 GATE MD 8 1042 1042 OK 11/19/86 86 NS-V35 GATE NO 8 7607 495 HIGH 12/19/86 87 88 89 PENET104 BLK FLG N/A 2 71 71 OK 11/4/86 90 PENET105 BLK FLG N/A 10 45 45 OK 12/13/86 91 PENETiO6 BLK FLG N/A 4 45 45 OK 12/13/86 l 92 PENET210 BLK FLG N/A 2 45 45 OK 12/13/06 I 93 PENET211 BLK FLG N/A 2 1J452 71 CAP LEAK 11/4/86 l 94 PENET241 BLK FLG N/A 18 92 45 OK 12/12/86 i 95 PP101/02 LFT CHN N/A 1/2 37000.001 .01 FAILED l 96 PP133/34 LFT CHK N/A 1/2 .01 .01 97 98 99 l 100 101 RB-V2A GATE MO 8 880 880 OK 11/18/86 l 102 RB-V7 GATE NO 8 0 'O 11/22/86 103 SA-V2/3 GLOBE HW 2 .01 .01 104 SF-V23 GATE HW 8 42 42 OK 12/4/86 l 105 WDG-V3/4 GL/GA N0/ SOL 2 82 82 OK 12/2/06 106 WDL-V303 GLOBE MO 4 25 25 OK 11/24/86 107 WDL-V304 GATE D 4 25 25 OK 11/24/86 l 108 WDL-V534 GATE P 8 71 71 OK 12/11/86 109 WDL-V535 GATE P 8 71 71 OK 12/11/86 110 111 90.0
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING 1987 1987 1987 1987 1987 1987 1987 1987 RESULTS GIVEN IN STD. CUBIC CENTIMETERS PER MINL'TE (SCCM) l NO TAG DESC OPER IIZE ASFOUND AKLEFT COMTS87 DATE cco ***emmee nameuseu ****** www ****mummen amannewee ** men *** ******** 112 EQPFLG FLANGE N/A 216 243 77 OK 1/2/87 113 PERACCES MISC. N/A 96 .01 .01 114 2ND MISC N/A 96 .01 .01 115 EMEACCES MISC N/A 94 .01 .01 116 2ND MISC N/A 96 .61 .01 117 118 119 MINPATH 61478.071 5360.1 FAILED 120 MAXPATH
- 156224.061 13981.09 FAILED 121 ACC CRIT 194847 104847
- The semiannual and quarterly tests performed nearest to the round of refueling frequency testing are added into the "MAXPATII" calculation. For valves in series the valve with highest leakage is used in the "MAXPATII" calculation.
For valves in parallel or where the inner and outer valves can not be tested separately, the total leakage is used. 91.0
LOCAL LEAK RATE TEST RESULTS THREE MILE ISLAND UNIT 1 REACTOR BUILDING Following is the terminology used in the previous computer data:
- 1) .1 - (Alone) or any other number other than zero in the first decimal place means test scheduled.
- 2) .01 - (Alone) means no data available for the year or that the test was delayed. (e.g. valve not installed yet or not in previous testing scope.)
- 3) .001 - (Or any number other than zero in the third decimal place) after a leak rate'(i.e. 59500.001) means actual leak rate was greater than measured / recorded value.
- 4) AsFound - leak rate (SCCM) in the As-Found condition before any repairs or adjustments.
- 5) AsLef t - The leak rate (SCCM) after any adjustments / repairs.
- 6) Dates - Date of the last acceptable test results for the item.
- 7) Desc - Description of the valve or penetration.
- 8) Oper - Type of valve operator (actuator).
- 9) Notest - The Tech spec scope did not require this valve to be tested during the respective year.
- 10) Novalve - This valve was installed during a later refueling outage.
- 11) Comments - Cognizant Engineer comments about the results:
A) Failed - Exceeded the plant eslablished leakage rate limit from SP 1303-11.18 Enclosure 10 which made repair / adjustment necessary. B) High/ Low - Subjective judgement of cognizant engineer. Represents the results with respect to the leakage which the type of leakage barrier (e.g. gate valve, globe valve, check valve, flange, etc.) is considered to be capable of without extraordinary repair / adjustment C) OK - No problems with leakage. D) Other - E.G. newvalve, novalve, notest, repacked seatwork, stembent, etc. (self-explanatory). E) CAPLEAK - see para. 2.1.5 PACKING - see para. 2.1.6. i
- 12) Size - The nominal pipe size for the leakage barrier.
i i l 92.0 0111U i
.l APPENDIX J THREE MILE ISLAND UNIT 1 THI-1 LLRT CORRECTIVE ACTION SUlstARY I
t 1 l l l l l i l l 93.0 0111U i
PURPOSE The purpose of this section is to provide a summary of the corrective action plan for upgrading Three Mile Island Unit I containment isolation provisions. Trending of the past twelve years of local leak rate testing results has served to indicate which containment isolation valves (CIV's) have been most prone to excessive leakage. The corrective action plan, which began in 1981, is intended to resolve the repetitive leakage problems and, thereby, improve the reliability of the containment. I. REPLACEMENT OF CONTAINMENT ISOLATION VALVES WITH HISTORY OF HIGH "AS-FOUND" LEAKAGE 1981 RB-V7, formerly a solid wedge gate valve, has been replaced with a gate valve having a wedge with a supplemental resilient wedge insert. The valve seats have performed with no leakage problems since then. 1983 WDG-V4 a solid wedge gate valve was replaced with a solenoid open/ spring closed globe valve. 1987 IC-V3 and IC-V4, formerly solid wedge gate valves, have been replaced with plastic lined plug valves. We expect these to be much more reliable for the application. II. CONTAINMENT ISOLATION VALVE MODIFICATION I 1986 Actions have been initiated to procure modified valve internals, e.g. split disc wedges, from the Anchor / Darling Valve Company for the future modification of the following solid wedge gate valves: CA-V2/5A/5B/189 IC-V2 Many of the necessary measurements were taken during the 6R Outage. ! We expect these modified internals to improve the leak tightness of the valves. The modifications would be done as required in future outages after the new internals are received. 1983 Seats with teflon seat inserts were installed in HP-V1/6 and LR-V49. FUTURE Modifications to the following check valves may also be accomplished based on the success with those valves listed above: CF-V12A/12B MU-V116 PP-V101/102/133/134 III. INCREASED CIV TESTING 1985 In response to a request from the Nuclear Regulatory Commission, the reactor building purge isolation valves, AH-V1A/1B and AH-VlC/lD, were placed on a quarterly testing frequency; these valves had previously been tested annually. The increased testing frequency improves operability of these valves. 94.0 0111U
l i I l 1987 The Penetration Pressurization check valves, PP-V101/102/133/134, associated with the purge interspaces have been added to the LLRT program. Required Interval - Refueling. IV. CHESTERTON VALVE PACKING PROGRAM 1986 The Chesterton Packing Program has been initiated in an attempt to eliminate valve packing leakage, thereby, reducing valve maintenance, radiation exposure and Laproving valve operability. The program, which is being implomented on various plant valves, including containment isolation valves, consists of the following:
- installing graphitic packing materials
- reducing packing box depth by installing spacer material, e.g. carbon bushings
- eliminating leak-off provisions
- live-loading of packing glands on motor operated valves (where possible)
We expect these packing modifications to improve the leak tightness of these valves, y V. PROCEDURAL IMPROVEMENTS onnoinz Various surveillance and maintenance procedures have been updated and revised to provide guidance for inspecting and repairing purge valves, such as Surveillance Procedure 1301-8.3 " Reactor Building Purge Valve Seat Inspection"; Required Interval - Refueling. Methods for shimming seats have been determined and implemented to greatly improve adjustability and seat reliability of the purge valves. I i l 95.0 01110 l}}