ML20085N287
| ML20085N287 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/27/1995 |
| From: | Burski R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-95-03, GL-95-3, W3F1-95-0095, W3F1-95-95, NUDOCS 9506300134 | |
| Download: ML20085N287 (14) | |
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June 27, 1995 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 l
NRC Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes" Gentlemen:
In accordance with NRC Generic Letter 95-03, Entergy Operations, Inc.,
Waterford 3, has evaluated recent operating experience with respect to the detection and sizing of circumferential indications to determine the applicability to Waterford 3.
This submittal provides: (1) a safety assessment justifying continued operation and (2) the ste m generator inspection plans for the seventh refueling outage, currently scheduled for September 1995. This response is submitted under oath of affirmation.
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,. NRC Seneric Letter 95-03, "Circumferential Cracking of' Steam.
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Generator Tubes"
.W3F1-95-0095.
Page'2:-
' June l27, 1995 i
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- Please ' contact me' or Robert J. Murillo at (504) 739-6715 should there be l
any questions regarding this response.
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.Very truly yours,
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R.F. Burski Director
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Nuclear Safety j
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Attachments I
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L.J. Callan, NRC Region IV
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C.P. Patel, NRC-NRR R.B. McGehee l
N.S. Reynolds 3
NRC Resident Inspectors Office i
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.l UNITED STATES OF-AMERICA l
NUCLEAR REGULATORY COMMISSION l
In the matter _of
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'Entergy Operations, Incorporated
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Docket No. 50-382 l
Waterford 3 Steam Electric Station
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AFFIDAVIT-R.F. Burski, being duly sworn, hereby deposes and says that he is Director, Nuclear Safety - Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign and file with the Nuclear-Regulatory Commission the attached Waterford 3 response to Generic Letter 95-03; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
6?d/LJ R.F. Burski Director, Nuclear Safety - Waterford 3 STATE OF LOUISIANA
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) ss PARISH OF ST. CHARLES
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Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this 7~4" day of d o us-
, 1995.
_ Notary Public My Commission expires W' N N
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l WATERFORD 3 SAFETY ASSESSNENT JUSTIFYING CONTINUED OPERATION PURSUANT TO GENERIC LETTER 95-03 p
f INTRODUCTION This document is the Waterford 3' Safety Assessment justifying continued-operation pursuant to Generic Letter 95-03. The safety assessment justifying continued operation.is predicated on the following technical bases:
RECENT OPERATING EXPERIENCE APPLICABLE TO CIRCUMFERENTIAL CRACKING The Waterford 3 steam generators were desianed and fabricated by Combustion Engineering, Incorporated. They are designated as Model 3410 steam generators, which refers to the design thermal output of the unit.
The 3410's each contain 9,350 high temperature mill annealed Inconel-600 tubes which were explosively expanded the full depth of the tubesheet. Design operating inlet temperature is 611 F.
Actual operating inlet temperature has been reduced to 604 F.
Secondary chemistry since start-up has met or exceeded EPRI Secondary Water Chemistry Guidelines.
1)
SCOPE OF S/G TOP OF THE TUBESHEET INSPECTIONS Waterford 3 has conducted extensive and comprehensive Steam Generator (S/G) eddy current inspections utilizing primary and independent secondary data analysts.
Eddy current inspections to date have been conducted during each refueling outage. During the spring 1991, fourth refueling, Waterford 3 examined 259 tubes on the hot leg, top of the tubesheet, in S/G #2 with a single coil motorized rotating pancake coil (MRPC) probe. These examinations were in addition to the full-length bobbin coil tests performed on a 3% sample size in each steam generator.
In the fall of 1992, fifth refueling, Waterford 3 used a 3-coil MRPC probe for a selected tube inspection of 300 tubes in each S/G on the hot leg transition area focused in the sludge pile region and a 21% full length bobbin coil sample inspection for each S/G.
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1The }992 MRPC examinations, performed at the top of the hot leg were
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conducted in accordance with recommendations published in ABB-CE Information Bulletin 92-02, "Circumferential Cracking in Steam Generator Tubing," dated August 21, 1992.
During the Spring, 1994 sixth refueling outage,-Waterford 3 performed a comprehensive examination of both S/Gs. The examinations included, for each steam generator, a full-length bobbin coil examination of 100% of the inservice tubes (18,023 tubes), three-coil MRPC examination of 65% of the j
tubes (100% of sludge pile) at the top of the tubesheet on the hot-leg side, and flexible MRPC examination of 300 tubes in the upper bundle region i
potentially subject to steam drying. To date, Waterford 3 has no evidence to indicate that any circumferential tube cracking has been incurred in either steam generator. This inspection is more fully described in Special Report 95-001-00 submitted March 21,1995.
2)
SUSCEPTIBILITY TO CRACKING Waterford 3 S/G tubes have some inherent susceptibility to cracking due to l
the original design and manufacturing process for all Combustion Engineering S/Gs; nonetheless, mitigating conditions are present which reduce Waterford 3's susceptibility.
l The tube to tubesheet explansion process qualified for Combustion Engineering recirculating S/Gs' utilized a detonator cord positioned axially along the tube centerline encapsulated by a polyethylene sheath held within the tube. The explosive force generated upon detonation was transferred into the tube radially. The Inconel-600 high temperature mill annealed tubing without thermal stress relieving inherently remains in a condition with high levels of residual stresses due to the aforementioned process. Waterford 3's susceptibility is based upon original design and fabrication processes employed for all CE S/Gs.
Mitigating conditions to the development of circumferential cracking in the explanded region are related to: T-hot reduction; minimal sludge deposition; removal of copper from the secondary, and molar ratio control. Waterford 3 has reduced T-hot t
and has remained aggressive in maintaining or exceeding EPRI S/G Secondary Water Chemistry Guidelines.
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, NON-DESTRUCTIVE EXAMINATION METHODS (PROBES. INSTRUMENTS AND HARDWARE 1 The Waterford 3 eddy. current testing program optimized the test methods to minimize electrical noise and signal interference to maximize flaw 1
sensitivity.-
During the most recent Waterford 3 refueling outage (RF #6), both low-loss cables and low-noise slip rings were utilized for the MRPC and bobbin coil testing. The larger 0.115 inch diameter pancake coil was utilized for the i
MRPC tests to enhance flaw ~detectability and reduce the signal-to-noise ratio. The site specific Waterford 3 Steam Generator Eddy Current Data j
Analysis Guidelines, ISI-I-001, contains information relating to noise reduction and signal interference that was used to ensure that flaw indications in the transition region were' identified by the eddy current analysts.
4)
WATERFORD 3 PLANT SPECIFIC FACTORS THAT COULD AFFECT SENSITIVITY i
Waterford 3 does not have plant specific factors present which could have adversely affected flaw detectability.
The Waterford 3 ISI-I-001 guideline contains information relating to the anticipation of potential sources of interfering signals.
Eddy current testing utilizing frequency modulation.provides the analyst with a means to identify interfering signals such as those caused by dents, copper deposits, tube bowing, bulges, etc., and to understand the effects of these signals in relation to the overall process of flaw detection and analysis.
Waterford 3 to date has not experienced secondary deposition of copper, magnetite, or denting which affect flaw detectability. Waterford 3's S/Gs' utilize an eggcrate tube support system, rather than a tube support plate system which is more susceptible to denting.
Waterford 3 steam generator secondary chemistry has been All' Volatile Treatment (AVT) and full flow condensate polishing since start-up. Sludge f
lancing results from Refuel #6 evidenced relative low sludge loadings with 1
47 lbs. wet removed from S/G #1 and 40 lbs. wet removed from S/G #2.
Bobbin coil data verified minimal sludge deposition utilizing 100 kHz absolute as a screening frequency which was utilized to produce sludge maps identifying relatively small deposits of sludge. Ammonium chloride has been used since July 1993 to maintain molar ratio control within 4
specifications. Waterford 3 has increased the secondary feedwater pH to t
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,further'reduceirontransport. Waterford 3 lowered T-hot from 611 F to 604 F in Cycle 6 (Refuel #5).
Temperature is an important factor in the mitigation of Primary Water Stress Corrosion Cracking (PWSCC) and Outside Diameter Stress Corrosion Cracking (0DSCC). Waterford 3 also removed the last source of copper from the secondary plant by replacing the copper / nickel MSR bundles with stainless steel.
5)
EQUIPMENT SET-UP FOR TOP OF THE TUBESHEET MRPC TECHNIOUE The equipment set-up, technique, and qualification provide a high assurance of flaw detectability and-thru-wall characterization.
During Refuel #6, March 1994, Waterford 3 used the following techniques for the top of the tubesheet hot leg 3-coil MRPC examination. Data was acquired on the push to minimize geometry discontinuities.
MIZ-30 Data Acquisition System Zetec 0.600" MRFC Pull speed 0.4" ips e
Zetec 3-coil MRPC 0.115 inch pancake coil 600 RPM 800 samples /sec.
e 600 kHz data evaluation / indication measurement 400 kHz optimum frequency for plotting and 100 kHz data e
evaluation terrain maps 400/100 kHz mix to suppress support and OD deposit signals e
200 kHz mix suppression and secondary plotting e
100 kHz data evaluation and component of mix to suppress e
supports and deposit signals Currently 3-coil MRPC technology is qualified to Appendix H of the EPRI "PWR Steam Generator Examination Guidelines" for the detection of the following:
PWSCC; ODSCC; expansion transitions; dented structure freespan; and non-dented structures. The qualification requirement technique requires a minimum probability of detection (P0D) of 80% with a 90%
confidence level for flaws greater than or equal to 60% thru-wall depth on a suitable specimen as defined by Appendix H.
The fact that Waterford 3 examined 100% of both steam generator hot leg sludge pile explansion regions with a probe of equal or greater probability of detection than the 0.080 inch pancake coil with No Detectable Degradation (NDD) ensures tube i
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,stru[cturalintegrity.. Tube structural integrity related to circumferential cracking is assured, per Reg. Guide 1.121, for defects with less than 79%
average thru-wall. Supporting documentation related to tube structural integrity for those plants that have experienced circumferential cracking has been verified through in-situ pressure testing results in Table 1.
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ANALYSIS OF DATA (PLANT SPECIFIC GUIDELINES) i Waterford 3 took a pro-active approach for data acquisition and analysis guidelines used for potential tube crack indication areas.
Waterford 3's sixth refueling MRPC testing was performed on 65% (100% of the sludge pile region) of the tubes at the top of the tubesheet on the hot-leg side of each steam generator. This testing also utilized C-Scans I
(Terrain Plots) for each of these tubes.
This data provided further information relating to the potential presence of tube crack indications.
The MRPC examinations at a miniraum scanned a given point 2.88 times based upon the following parameters:- coil diameter (field spread); pull speed j
i (inches per second); and RPMs (sample rate). This additional' scan coverage further minimizes the possibility of not identifying a crack indication.
No crack indications were present for the Waterford 3 S/G tubes.
The Waterford 3 ISI-I-001 guidelines also provides the eddy current testing analyst with information relating to unique plant circumstances (e.g.,
dents, copper deposits) which may necessitate special test procedures found not to be necessary at other similarly designed steam generators. This type information would be included in the Operational Experience section of the guide. Other utility plant operational experience pertaining to eddy current testing is also contained in this section of the Waterford 3 i
guidelines.
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CONFORMANCE TO EPRI NP-6201 PWR STEAM GENERATOR EXAMINATION SUIDELINES The Waterford 3 eddy current testing program is generally in conformance with EPRI NP-6201 guidelines and has demonstrated proactive initiatives related to S/G integrity.
. An NRC inspection was conducted at Waterford 3 during the February / March 1994 timeframe (Report No. 50-382/94-06). The regional initiative consisted of an inspection of the history and material condition of steam generator tubing and an assessment of the effectiveness of our programs in detection and analysis of degraded tubing, repair of defects and correction of conditions contributing to tube degradation.
The NRC reviewed various Waterford 3 procedures / guides and concluded that the Waterford 3 eddy current examination program criteria were found to be generally consistent with EPRI NP-6201 Rev. 3.
The NRC also concluded that the scope of the Waterford 3 eddy current examinations has been significantly increased during the last two refueling outages, with the examinations performed in 1994 during Refueling Outage #6 considered by the inspectors to be very comprehensive. The NRC also stated the inclusion of a sample of tubes for MRPC examination in the upper bundle region was considered commendable and indicative of strong management support for steam generator integrity initiatives 8)
DATA ANALYSIS TRAINING (PERFORMANCE DEMONSTRATION PROGRAM)
The data analysts training requirements are well documented and comprehensive.
Waterford 3's Design Engineering Steam Generator Eddy Current Data Analysis Guidelines 1S1-1-001 Rev. 2 contains the qualification requirements, including training, which must be met by primary and secondary data analysts responsible for evaluating the steam generator tube eddy current test results at Waterford 3.
The training requirements include successful completion of a written examination and a performance demonstration examination specific to Waterford 3.
The performance demonstration examination must be completed by all analysts at each outage, unless the data analyst has completed the site specific training and examinations within the previous twelve months. The N0ECP-252 Rev. 3 (Steam Generator Eddy Current Inservice Testing) procedure also elaborates on the eddy current testing analyst testing requirements.
The Procurement / Programs department is in the process of revising the 1S1-1-001 guide to incorporate a more detailed definition of the training requirements in the program for eddy current data analysts. During their inspection, the NRC pointed out that the eddy current examination program criteria were found to be generally consistent with EPRI NP-6201.
The NRC identified a need to address the lack of definition of training requirements in the program for analysts (e.g., minimum classroom requirements for initial training and the scope of re-training necessary when an analyst fails a written I
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examinations').;ThdISI-I-001 guide'willbeupdated-prior-toRefuel#7to-l
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- include minimum classroom requirements'and retraining.
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WATERFORD 3 i
STEAM GENERATOR INSPECTION PLAN FOR FALL 1995 REFUEL #7 t
INTRODUCTION Waterford 3's steam generator inspection plan for Refuel #7, Fall of 1995, is in a accordance with EPRI NP-6201 Rev. 3 "PWR Steam Generator Examination Guidelines." The following inspection plan includes full length bobbin coil and hot leg top of.the tubesheet:
1)
S/G INSPECTION PROGRAM FOR REFUEL #7 A 20% full length bobbin coil examination of both S/Gs with a 500 tube per S/G augmented inspection is planned for Refuel #7, The method of sampling is based upon a planned random selected sample of every tube within a line varying 5 to 7 lines over for each consecutive starting point. This sampling method ensures adequate coverage of the entire S/G bundle.
Expansion criteria will be based upon the degradation mechanism identified.
To date,the only active damage mechanism identified is wear in the upper bundle hot and cold leg anti-vibration bars (Batwings) and vertical supports.
Per Technical Specifications, degraded tubes with indications of 20% through 39% thru wall shall be re-examined at subsequent refuelings.
Waterford 3 will continue to monitor' degraded tubes with wear indications to evaluate percent thru-wall growth.
A 20% top of the tubesheet hot leg inspection scope is planned for the expansion transition area of both S/Gs focused in the sludge pile region.
Additionally, Waterford 3 will include a 500 tube per S/G augmented top of the tubesheet hot leg scope.
Waterford 3 will utilize probes capable of reliably detecting circumferential1y oriented indications which are qualified to Appendix H of EPRI NP-6201, "PWR Steam Generator Examination Guidelines," Rev. 3.
The expansion criteria requires 100% inspection of both S/Gs if one circumferential crack is identified during the inspections.
The requirements for data analysts will be QDA certification and experience with identifying CE S/G degradation mechanisms specific to PWSCC and ODSCC in the transition region.
Performance demonstration training and tecting will be conducted to ensure damage mechanisms associated with Maine Yankee, l
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.,CalvkrtCliffs' Unit #2andANOUnit#2areidentified. The Waterford 3 Steam Generator' Data Analysis Guidelines will be updated.to include these concerns on circumferential cracking. - The Waterford 3 site specific 1
examination will require a minimal score of 80% in order.for the-analysts to be qualified to perform data analysis.
i REfBEED NRC Generic Letter 95-03:
"Circumferential~ Cracking of Steam Generator '
Tubes" EPRI NP-6201, "PWR Steam Generator Examination Guidelines," Rev. 3 NEI Letter dated June 1,1995, "NRC Generic Letter 95-03; Circumferential Cracking of Steam Generator Tubes."
NRC Inspection Report 50-382/94-06 Waterford 3 Steam Generator Integrity Review February / March 1994 Waterford 3 Special Report 95-001-00 dated March 21, 1995 Letter No.
l W3F1-95-0028 EPRI Technical Advisory Group, " Appendix H Qualification of Plus Point, Cecco, and Other Coils" dated May 25, 1995 EPRI TR-102134 " Secondary Water Chemistry Guidelines," Rev. 3.
ABB Combustion Engineering, "CE0G Experience Summary Regarding Circumferential Cracking 'of Steam Generator Tubes,". dated May 25, 1995 j
Letter No. CE0G-95-296 ABB-CE Info Bulletin 92-02, "Circumferential Cracking in Steam Generators" NRC IN 94-88, " Inservice Inspection Deficiencies Result In Severely Degraded Steam Generator Tubes"
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i TABLE 1 IN-SITU PRESStlRE TEST RESULTS Page 1 of 2 Bad M1 Test Tube MRPC Data In-Situ Pressure Test Data Leak Rate Test Data Comments Qgatt R1 Max Depth Length Avg Depth 3.1P Target Pressure Reg. Guide Leak Presstre*
Leak Rate
%TW (degrees)
%TW Pressure
- Achieved Compliance (psi) fp)
Calvert Cliffs 22 495 3G'28 49 180 24.5 4200 5400 7000 Yes No (Unit 2) 22 495 46/32 83 360 83 4200 5400 6400 Yes No l
22 4/95 7N42 82 360 82 4200 5400 7057 Yes No l
22 495 95/51 91 360 91 4200 5400 6700 Yes No 22 495 91/49 92 360 92 4200 5400 7075 Yes No 22 495 88/52 81 360 81 _
- 4200 5400 7050 Yes No ANO A
1G92 6N48 87 239 57 '.
4050 4700 4700 Yes No (5)
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A 594 24132 90 205 51.3 4050 4700 4600 Yes Yes 2000 0.05*
(6)
A 5/94 32/126 97 360 97 4050 4700 3600 Inconclusive Yes 2000 0.15" (6)
A 594 48/50 79 217 47.7 4050 4700 4700 Yes Yes 4700 0 04*
A 1/95 23/135 81 304 68 4050 4450 4450 Yes No A
1/95 77/97 51 241 34 4050 4450 4450 Yes No A
1/95 2 #134 97 143 39 4050 4450 4450 Yes No Palo Verde 12 1G94 SN75 100
302 83.9 3540 3900 3900 Yes No (5)
@"" II 12 1G94 SN73 100
285 79 2 3540 3900 3900 Yes No (5) 12 1G94 SN71 100*
251 69.7 3540 3900 3900 Yes No (5) 12 10/94 35t10 100*
197 54.7 3540 3900 3900 Yes Yes 3000 0.011 (5) 12 1G94 48/79 70*
200 38.9 3540 3900 3900 Yes No (5)
N9es:
(1)
Target Pressure is three Times Normal Operating Differential Pressure (3.P) with plant specific correcten factors applied.
(2)
Pressure at which leakage was measured (3)
Leak rate rrr asured following 3.iP pressure test (4)
Depth measured by UT (5)
Full Tube Pressure Test (6)
Pump Capacity Limited Max Test Pressure (7)
Leak rate at initial NOP test (8)
Leak rate at Peak Accident Pressure (9)
Leak rate at FJOP following Peak Accident Pressure Test (10)
Re-tested with Bladder Tool (11)
Unplugged and then Teste1
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TABLE 1 (cont'd)
IN-SITU PRESSURE TEST RESULTS
- Page 2 of 2 Eleut 3 61 Test Tube MRPC Data h-Stu Pressure Test Date Leek Rate Test Data Comments Dalt 81 Max Depth Length Avg Depth 3.5P Target Pressure Reg. Guide Leek Pressure
Leek Rate
%TW (degrees)
%TW Pressure
- Achieved Compliance (pe.)
(mm)
Maine 2
&94' 30/131 94
-300 94 4350 4800 4800 Yes No -
Yankee 2
8/94 40/124 92 360 92 4350 4800 4800 Yes No 2
8/94 49/126 95 132 35 4350 4800 4800 Yes
- No 2
8/94 81/72 92 181 47 435' 4800 4800 Yes No 1
8/94 62/27 97 360 97 435l 300 4800 Yes No 1
8/94 66/33 360 94 4350 4800 4800-
~Yes No 1
8/94 106/63 st 320 88 4350 4800 4800 Yes No m
Yes 1450'"
0.001'O (6) 2 8/94 49/122 94 360 94 4350 4800 4000
_ _Ye_s_.._1_45_0* _--_ _0._19_2"_ _
3/95 4350 4800 5700 Yes (10,11) 2
&94 89/46 94 262 69 4350 4000 2700-m Yes 1450*
0.01'"
(6)
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_ _Ye_s__ _1_45_0* _. 0 28_*__
3/95 4350 4800 Yes (10,11) 1 8/94 70/55 89 360 89 4350 4800 2900_
m Yes 1450*
O.001 (6)-
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_ _Ye.s_ - 1450*_- _ _0._31_3*__
3/95 4350 4800 0000 Yes (10,11) 2 3/95 91/46 83 360 83 4350 5700 6475 Yes No 1
3/95 72/65 96 358 95 4350 5700 8500 Yes No 1
3/95 92/73 97 311 84 4350 5700 6400 Yes No 1
3/95 55/102
- 99 247 69 4150 5700 7000 Yes-No -
Notes: (1)
Target Pressure is three Tirnes Normal Opereting Drfferential Pressure (33P) with plant specific correcten factors applied (2)
Proseure at which leakage was measured (3)
Leek rate measured foRomng 3aP pressure test (4)
Depth m-ed by UT (5)
Fun Tube Pressure Test (6)
Pump CapacRy Limited Max Test Pressure (7)
Leek rate at initial NOP test "
(8)
Leek rate at Peek Accident Pressure i
(9)
Leek rate at NOP foNowing Peak Accident Pressure Test (10)
Re-tested with Bledder Tool (11) '
Unplugged and then rested
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