ML17305B278
| ML17305B278 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 12/27/1990 |
| From: | Conway W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 161-03672-WFC-J, 161-3672-WFC-J, NUDOCS 9101020254 | |
| Download: ML17305B278 (30) | |
Text
ACCELERATED DIST BUTION DEMONST ION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9101020254 DOC.DATE: 90/12/27 NOTARIZED: NO DOCKET FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 AUTH.NAME AUTHOR AFFILIATION CONWAY,W.F.
Arizona Public Service Co. (formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards response to NRC request for addi info re application for amend to License NPF-41,extending interval for steam generator eddy current exam.
DISTRIBUTION CODE:
AOOID COPIES RECEIVED:LTR
(
ENCL
/
SIZE: 2+
TITLE: OR Submittal: General Distribution NOTES:STANDARDIZED PLANT 05000528 RECIPIENT ID CODE/NAME PD5 LA TRAMMELL,C INTERNAL: ACRS NRR/DET/ESGB NRR/DST 8E2 NRR/DST/SICB 7E NUDOCS-ABSTRACT OGC/HDS1 RES/DSIR/EIB EXTERNAL: NRC PDR NOTES:
COPIES LTTR ENCL 1
1 2
2 6
6 1
1 1
1 1
1 1
1 1
0 1
1 1
1 1
1 RECIPIENT ID CODE/NAME PD5 PD TRAMMELL,C.
NRR/DET/ECMB 9H NRR/DOEA/OTSB11 NRR/DST/SELB 8D NRR/DST/SRXB 8E
- 8~M REG FILE 01 NSIC COPIES LTTR ENCL 1
1 2
2 1
1 1
1 1
1 1
1 1
0 1
1 1
1 NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 27 ENCL 25
Arizona Public Service Company P.O. BOX 53999
~
PHOENIX, ARIZONA85072<999 WILLIAMF. CONWAY EXECUTIVEVICEPREEI DENT NUCLEAR 161-03672-MFC-JRP December 27, 1990 Docket No.
STN 50-528 U.
S. Nuclear Regulatory Commission Attention:
Document Control Desk Mail Station Pl-37 Washington, D.C.
20555
Reference:
Letter from W. F.
dated November 14, 1990
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Unit 1 Technical Specification Amendment Extension of Interval for SG Eddy Current Examination
Response
to Request for Additional Information File: 90-F-005-419.05; 90-056-026 This letter is being provided in response to a request for additional information from Mr.
C.
M. Trammell of the NRC concerning APS'equest for a Technical Specification amendment for an extension of the interval for Steam Generator Eddy Current examination for Unit l.
Based on this request, the following information is presented in response to the NRC request:
910i020254 90i227 PDR ADOCK 05000528 PDR (II I(
~
~
1 672-WFC-JRP December 27, 1990 U.
S. Nuclear Regulatory Commission Document Control Desk Page Two MECHANICAL WEAR The most recent inservice inspection of the Unit 1 steam generator tubes was completed in July 1989, during the second refueling outage.
With but a few exceptions, the degraded/defective tube indications noted during eddy current inspection are the result of tube to support wear.
The'ear indications can be divided into the following categories:
f 1.
2.
3.
Cold Leg Corner Wear Central Cavity Batwing Wear Eggcrate and Vertical Support Wear COLD LEG CORNER WEAR Tubes in the cold leg corner region of the System 80 steam generators are subject to radial cross flow of sufficient magnitude to cause the tubes to vibrate within the eggcrate supports.
This unique condition is isolated to the small number of tubes near the intersection of the tube lane (Figures 1 & 2) and the downcomer cold leg side recirculation entrance window. It has been APS'ractice since the discovery of this wear mechanism to plug all tubes in this region of the steam generator with any measurable
- wear, which essentially eliminates any potential for this wear mechanism causing a tube leak during the next operating cycle.
Accordingly, a total of four tubes were identified during the Unit 1 second outage to have measurable wear indications (24% to 54% through wall) were plugged and staked.
Staking is the process of inserting a one-half inch diameter stainless steel cable or rod into the tube above the plug to prevent potential collapse or severance of the tubes.
CENTRAL CAVITY BATWING WEAR Tubes in the central cavity of the steam generator tube bundle are susceptible to wear resulting from interaction with the batwing supports, diagonally oriented tube spacer strips similar in nature to the anti-vibration bars in Westinghouse steam generators.
It has been APS'ractice since the discovery of this wear mechanism to plug all tubes in this region of the steam generator with any measurable
- wear, which essentially eliminates any potentials for this wear mechanism causing a tube leak during the next operating cycle.
Accordingly, a
total of three tubes were identified during the Unit 1 second refueling outage to have measurable wear indications (34% to 36% through wall) were plugged and staked.
161-03672-WFC-JRP December 27, 1990 U.
S. Nuclear Regulatory Commission Document Control Desk Page Three EGGCRATE AND VERTICAL SUPPORT WEAR Random tube wear has been observed to occur at the various horizontal eggcrate supports and,vertical supports.
For these wear indications, a plugging criteria which would prevent any indication from growing to 40% through-wall or greater during the next operating cycle was developed..
Wear projections were performed assuming a constant volumetric wear rate for a given tube over the next operating cycle and conservatively projecting a
500 day operating cycle until the next refueling outage.
Next using a geometrical volume to depth relationship for flat wear scars, it was determined that a 30% through-wall indication at the second refueling outage woul'd grow to a maximum 40% through wall (assuming constant volumetric wear rate) 'during the next operating cycle.
Accordingly, all tubes with wear indications of this nature, 30% through wall or greater were plugged.
The constant volumetric wear rate assumption is appropriate.
The wear mechanism is consistent with the classical Archard's equation for wear which states that the volumetric wear rate is directly proportional to the normal force between the surfaces and a material-dependent wear coefficient.
Since there is no mechanism by which the normal force between the tubes and the horizontal eggcrate and vertical supports will increase, the'"volumetric wear rate willnot increase.
The tube wear that occurs in the steam generators is due to secondary side fluid flow causing the tubes to vibrate against the supports.
Since the requested time extension to the inspection interval corresponds to the time Unit 1 was shutdown (i.e., with no secondary side fluid flow) during the extended outage of 1989, there is no actual increase in the amount of time in which wear is actively driven by the secondary side fluid flow.
The 500 operating days, conservatively assumed in the wear projections upon which the second refueling outage plugging criteria was developed, is not impacted by the time extension.
FOREIGN OBJECTS Unit 1 has had only one instance in which the presence of a significant foreign object in the steam generators has been detected.
During a January 1987 mid-cycle inspection, the presence of an approximately 12 inch long bar-like object (Figure 3) wedged between the Row 1 tubes and the economizer divider plate just above the flow distribution plate was discovered.
Low frequency eddy current data (20KHZ Absolute) indicated the apparent presence of a foreign object on the outside of the tube for nine successive Row 1 tubes.
Subsequent laboratory mock-up tests produced similar eddy current signals for a bar of ferritic stainless steel.
A fiber-optic inspection of the ID of one of the affected tubes revealed the presence of a slight straight-edged indentation above and opposite to a
smaller partial circumferential indication.
Based on the above, APS concluded that a steel bar-like object had become wedged between the tubes and the economizer divider plate, forming an impression of itself on one side of the tube and a reactive dent on the opposite side of the tube caused by the reactive force at the top edge of the drilled holes in the flow distribution plate.
The
QO
-03672-WFC-JRP December 27, 1990 U.
S. Nuclear Regulatory Commission Document Control Desk Page Four eddy current data revealed no wear or loss of tube wall at the location of the object.
A means by which an object of this size could have been introduced into this region of the steam generator during maintenance or operation could not be identified.
Therefore, it was concluded that the object was most likely introduced during fabrication of the steam generators.
Due to the location of the object, an inspection from the secondary side was not possible.
The steam generator flow distribution analysis predicted flow velocities of less than 1.5 ft./sec. in the primarily vertically upward direction, which is quite low.
Since the foreign object is large, wedged in place and in a relatively low flow velocity flow field, there is no potential for migration causing further damage to other tubes.
Therefore, the affected tubes were plugged and the object left in place.
Inspection of the surrounding unplugged tubes during the most recent second refueling outage inspection revealed no additional damage caused by the continued presence of the foreign object.
In addition to the above instance, two tubes with minor wear defects caused by foreign objects were noted during the 100% second refueling outage inspection.
In both cases, review of the eddy current data indicates the objects were present during previous inspections, had remained in place, and had now begun to initiate wear on the tubes.
In both cases, the objects are located in the interior of the bundle, indicating they were introduced during fabrication and have not migrated.
The two tubes were plugged and staked and adjacent tubes will be inspected during subsequent inspections to verify the objects continue to remain in their present locations.
As a general practice at PVNGS, 100% of the tubes around the outer periphery and along the tube lane have been eddy-current inspected, using a multi-frequency eddy-current probe.
The low frequency eddy current data provides for the detection of objects on the outside diameter of the tubes.
The outer periphery and tube lane are the locations where industry experience has indicated that damaging loose parts are more likely to accumulate.
To date, no loose parts have been detected in these locations at Palo Verde.
- Recently, APS has instituted computerized data screening.
In this process, the data from all inspected tubes is passed through the computer which automatically flags abnormalities in the low frequency data.
A detailed analysis can then be performed to positively identify the presence of any foreign object.
The affects of the objects presence can then be evaluated.
PVNGS utilizes a Foreign Material Exclusion (FME) process for preventing the introduction of loose parts into the steam generators.
This process requires all parts and tools utilized in the course of maintenance activities to be logged in and out to ensure all material is accounted for.
FME covers are used to seal system openings whenever work is not in process to prevent inadvertent introduction of material into the opened system.
The process is performed by the Quality Control organization.
It should be noted that the ADVs and MSIVs, on which maintenance and modification activities were performed during the Unit
~0 g
61-03672-NFC-JRP December 27, 1990 U.
S. Nuclear Regulatory Commission Document Control Desk Page Five 1 refueling outage, are physically located outside of the containment building downstream of the steam generators.
Therefore, in the very unlikely event that materials were introduced into the system during these activities, the potential for entering the steam generators is remote.
As in the case of tube vibration and wear, tube wear from foreign objects is due to secondary side fluid flow causing the object to vibrate against the tubes.
Since the requested time extension to the inspection interval corresponds to the time Unit 1 was shutdown (i.e., with no secondary side fluid flow) during the extended outage of 1989, there is no actual increase in the amount of time in which wear is actively driven by the secondary side fluid flow.
The actual operating time on the Unit 1 steam generators from the date of the last eddy current examination (July 1989) to the next scheduled refueling outage (February 1992; total of 32 calendar months) is expected to be approximately 19 months (June 1990 to February 1992).
The 19 months is within the Technical Specification surveillance requirement intent of not less than 12 months nor more than 24 calendar months.
In response to an additional question from Mr. Trammell regarding the length of time the Unit 1 steam generators were without the nitrogen overpressure, the main steam isolation valve modification began shortly after the steam generators entered wet lay up conditions on April 15, 1989.
The nitrogen overpressure was vented off on approximately May 3,
- 1989, and was re-introduced on February 28, 1990 (10 months).
During this period, chemistry control was maintained in accordance with plant procedures which dictates strict adherence to prescribed lay-up practices and specifications.
CHEMISTRY Chemistry has reviewed the operating data for the steam generators for 1986-1988 in response to questions by NRR.
Tables 1 through 5 show the dates and highest values reached for that date for the out of specification conditions in Mode 1.
In addition, the chemistry data was reviewed for the worst case condition for the chemistry parameters for the specified time frame.
This condition occurred at the end of June of 1987 due to a major failure in the circulation water system piping.
The review was limited to Mode 1 data as that is the operating conditions under which the steam generators are being fed from the condensate system and that is when anion throw from the condensate polishers would be seen.
The listed out of specifications are limited to anionic impurities and do not include out of specification sodium values as NRR's question was addressing anion throw from the condensate polishers
~
'1 f
w Nuclear Regulatory Commission Document Control Desk Page Six 161-03672-WFC-JRP December 27, 1990 Eddy-current testing performed in 1989 did not indicate any damage to the steam generators that was attributable to corrosion.
Eddy-current maps (Figures 4
through 9) show the sludge accumulation on the S/G tube sheet to be minor.
With continued improved chemistry in the steam generators, propagation of any corrosive type of attack will take longer due to the lower concentration of impurities present in the bulk water to contribute to the corrosive environment.
The second question asked by NRR pertained to steam generator chemical additions for the period of time wi.th no nitrogen overpressure (May 3, 1989 to February 1990).
They are as follows:
Steam Generator ¹1 7 additions for Hydrazine 2 additions for Ammonia Steam Generator ¹2 3 additions for Hydrazine 3 additions for Ammonia Steam generator water levels were maintained above the tube bundle during this time frame and chemical additions were made to compensate for makeup water added to the steam generators or increase the hydrazine concentration within the operating band.
In response to one final question from the staff, as stated in UFSAR Section 3.11.2.2, copper is used as a pressure boundary material only in the containment fan cooler coils since its corrosion rate in the spray solution is acceptably low.
If you should have any questions, please contact Michael E. Powell of my staff at (602) 340-4981.
Sincerely, WFC/JRP/pmm CC:
C.
M.
J.
B.
D. H.
A.
C.
A.
HE Trammell Martin Coe Gehr Gutterman
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ECONOMY,'ll ZER Vl.AH VII'Pl
I l
DO'llNCOMER PARTITION l L
SU<PORT L"=VEL'OSITION HOT SIDE MAIN FLOW SrlROUD COLO SiDE RECIRCULATING WATER ENTRANCE WINDOW 9 IN.
16 IN.
7 IN.
ECOtlOMIZER DIVIDER PLATE SUP.
LEV.
3 POS ECONOMIZER FLOW SHROUD CORNER TUBE SYSTEM 80 STEAtl GENERATOR COLD SIDE RECIRCULATING FLL'ID ENTRANCE REGIOtl FXGURE 2
DIVIDER PLATE OBJECT PRIMARY INDENTATION (CAUSED BY OBJECT) l I
I REACTIVE DENT (CAUSED BY TOP EDGE OF FLOYD D ISTR IB'JT ION PLATE)
( AS V I E'>YED BY BORESCOPE EXAMINATION)
FLOiU DISTRIBUTION PLATE ROW 1
TUBES UNIT 1
FOREIGN 08JECT FlGURH 3
w TABLE 1 1986 STEAM GENERATOR DATA MODE 1 STEAM GENERATOR ¹1 DATE 1/08/86 1/09/86 1/20/86 1/2I/86 1/23/86 1/24/86 1/30/86 2/01/86 2/02/86 2/03/86 2/08/86 2/09/86 2/13/86 2/14/86 2/21/86 2/22/86 2/23/86 2/24/86 2/25/86 6/01/86 6/03/86 6/08/86 6/17/86 6/25/86 7/11/86 7/12/86 12/21/86 Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Chloride Sulfate Chloride Sulfate Chloride Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Chloride Sulfate Chloride Chloride Chloride Chloride Chloride Chloride Chloride Sulfate Chloride Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Chloride Chloride
<0.020
<0.015 PARAMETER LIMIT (ppm)
MAX. VALUE (ppm) 0.046 0.022 0 '39 0.071 0.043 0.017 0.025 0.026 0.017 0.074 0.119 0.029 0.061 0.034 0.068 0.065 0.076 0.123 0.049 0.037 0.027 0.025 0.033 0.136 0.022 0.026 0.036 0.042 0.024 0.050 0.050 0.024 0.064 0.091 0.025 0.047 0.025 0.021 0.023 I
Unit 1 spent approximately 237 days in Mode 1
of specification for anionic impurities on 27 specification for anions on 30 days.
1'n 1986.
Stea'm generator ¹1 was out days.
Steam generator ¹2 was out of
Oy 4
.T ff K
I L
TABLE 2 1986 STEAM GENERATOR DATA MODE 1 STEAM GENERATOR P2 DATE PARAMETER LIMIT (ppm)
MAX. VALUE (ppm) 1/08/86 1/09/86 1/20/86 1/21/86 1/24/86 1/30/86 2/01/86 2/02/86 2/03/86 2/08/86 2/09/86 2/13/86 2/14/86 2/23/86 2/24/86 2/25/86 5/24/86 5/31/86 6/03/86 6/06/86 6/08/86 6/19/86 6/20/86 6/21/86 6/25/86 7/11/86
, '8/06/86 8/31/86 9/10/86 10/02/86 Chloride
<0.020 Sulfate
<0.015 Chloride Sulfate Chloride Sulfate Chloride Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride Sulfate Chloride chloride Chloride Sulfate Chloride Sulfate Chloride Chloride Chloride Sulfate Sulfate Sulfate Chloride Sulfate Chloride Sulfate Chloride
Sulfate Sulfate Sulfate Chloride 0.180 0.030 0.082 0.234 0.064 0.023 0.025 0.074 0.116 0.059 0.028 0.051 0.026 0.057 0.026 0.070 0.085 0.074 0.029 0.036 0.020 0.027 0
F 170 0.026 0.016 0.058 0.032 0.025 0.050 0.030 0.021 0.050 0.050 0.050 0.018 0.023 0.017 0.024 0.016 0.037 0.058 0.022 0.049 0.044 0.016 0.080
TABLE 3 1987 STEAM GENERATOR DATA MODE 1 STEAM GENERATOR ¹1 DATE PARAMETER LIMIT (ppm)
Max.
VALUE (ppm) 3/10/87 3/11/87 3/12/87 3/27/87 8/3/87 8/5/87 8/7/87 8/8/87 8/16/87 8/21/87 Chloride Sulfate Chloride Sulfate Sulfate Chloride Chloride Chloride Chloride
'hloride Chloride Chloride
<0.020
<0.015'.110 0.077 0.071 0.053 0.020 0.026 0.022 0.025 0.036 0.071 0.029 0.023 STEAM GENERATOR ¹2 DATE PARAMETER LIMIT(ppm)
Max.
VALUE (ppm) 3/7/87 3/8/87 3/9/87 3/10/87 3/11/87 3/12/87 3/25/87 8/5/87 8/7/87 8/8/87 Chloride Sulfate Chloride Sulfate Sulfate Chloride Sulfate Chloride Sulfate Sulfate Chloride Chloride Chloride Chloride
<0 '20
<0.015 0.030 0.038 0.031 0.034 0.023 0.164 0.186 0.052 0.054 0.020 0.027 0.021 0.038 0.041 Unit 1 spent approximately 191 days in Mode 1 during 1987.
¹1 and ¹2 were outside of the Mode 1 operating bands on 10 of the 191 days.
There were 3 condenser tube failures during 1987 that probably contributed to the out of specifications on these days but the exact dates of the tube failures (with the exception of the June event) were not able to be established.
TABLE 4 1987 STEAM GENERATOR DATA WORST CASE CONDITIONS (1986-1988)
On'une 29, 1987 Unit 1 experienced a major transient in the circulating water system.
This transient caused a plant trip and a major ingress of circulating water into the condenser hotwells.
The following is a listing of the chemistry associated with this transient. All of these values are for Modes 3 through 5.
STEAM GENERATOR ¹1 DATE PARAMETER MODE LIMIT(ppm)
Max.
VALUE 6/29/87 6/30/87 7/1/87
'/2/87 7/3/87 7/4/87 7/5/87 Chloride Sodium Chloride "Sodium Chloride Sodium Chloride Chloride Sodium Chloride Sodium Sulfate Sodium
<0.10
<0.10
<0.10
<0.10
<0.10 5
<1.00 0.550 0.600 1.900 1.700 1.100 1.000 1.100 1.200 1.200 1.300 1.100 1.500 0.012 STEAM GENERATOR ¹2 DATE 6/29/87 6/30/87 7/1/87 7/2/87 7/3/87 7/4/87 7/5/87 PARAMETER Chloride Sodium Chloride Sodium Chloride Sodium Chloride Sodium Chloride Sodium Sulfate Chloride Sodium Sulfate Sodium MODE LIMIT (ppm)
<0.10
<0.10
<0.10
<0.10
<0.10 5
<1.00 Max.
VALUE 0.350 0.500 1.000 1.000 0.600 0.800 0.700 0.800 0.960 0.840 1.140 0.880 0.860 1.160 0.440 As indicated above the steam generators were out of specification for anion and cation species from 6/29 to 7/5.
Subsequent return to power did show limited problems with chloride return which was cleaned up using S/G blowdown. Subsequent eddy current, testing showed no damage that was corrosion related.
~0 TABLE 5 1988 STEAM GENERATOR DATA MODE 1 ea.
STEAM GENERATOR ¹1 DATE 3/11/88 3/13/88 3/23/88 3/30/88 PARAMETER LIMIT (ppm)
<0.020 Chloride Chloride Chloride Max.
VALUE (ppm) 0.034 0.064 0.024 0.022 STEAM GENERATOR ¹2 DATE 3/11/88 3/13/88 3/22/88 8/27/88 PARAMETER LIMIT(ppm)
<0.020 Chloride Chloride Sulfate
<0.015 Max.
VALUE (ppm) 0.030 0.063 0.032 0.023 Unit 1 spent approximately 239 days in Mode 1 during 1988.
Steam generators 1 and 2 were outside of the Mode 1 operating bands for 4 of the 239 days.
The out of specifications were attributed to returning the unit to power in early March, and to operation with the bypass valve around the condensate polishers open at the end of March.
The August value for sulfate appears to be an incorrect result as the cation conductivity does not support the anion data.
STEAN GENERATOR:
1 DATE:
11/02/87 LOCATION: ALL CRITERIA:
SLG PLUGGED 3i SLG 10/8?,
ARIZONA PUBLIC SEAVICE CO.,
PALO VEBOE, UNIT 1
~TUDES=.gyTH SLUDGE. DEPOSITS 261 PROC:
73TI"9RCOi REY 1
TIHE:
16: 06 STAY 160 160 150 gmp 140 iao 130 130 120 120 1'0 1P 1CO 1CO 90 60 7P jo 60 60 50 50 40 40 30 20 20 10 i0 20 30 40 50 C0 70:0
'30 "00 I"0 i20 5:-0 1+0
'=:
i~0 "70
<<:0
STEAM GENERATOR:
2 DATE:
11/04/87 LOCATION: ALL CRITERIA:
SLG PLUGGED 40 SLG 10/8?,
ARIZONA'IA PUBLIC SERVICE CO.,
PALO VEBOE, UNIT i PROC:
73TI-9RCQi REV 1
TIME:
15: 19 STAY 4y 0
'60 qa 140 14 1
Q 120
~
0
'00 90 CQ 7n 60
=0 4Q 40
=0
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=0 an
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STEAH GENERATOR: ii 06/09, ARIZONA PUBLIC SERYICE CO.,
PALO VEROE, UNIT I DATE 07/ie/ea TIHE:
11: 00: 57 LOCATION:
TSH TO TSHt 15.0 CRITERIA:
SLG PLUGGED 0
STAYS UNIT 1
S/G 1
TUBE SHEET-HOT LEG ENO OF CYCLE 2 0.0 in - 0.2 in 0
0.3 in - 0.4 in 2
0."
'... - O.g in 88 0.7 in - 0.8 in 267
> 0.8 in 140 ieO rA 14~ - -
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06/09, AHIZOHA PUBLIC SEHVICE CO.,
PALO VEHOE, UNIT i DnTB 07/19/69 TINE:
11: 43: 56 Locnriotl:
Tsc To Tsci 15.0 cnrTGnrn:
sLo PLUGG60 0
srnvs UNIT 1
S/G 1
'TUBE SHEET COLO LEG ENO OF CYCLE 2 0.0 in 0.2 in 0
0.3 in - 0.4 in 2
9.5
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0.7 in - 0.8 in 194
> 0.8 in 67
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STEAM GEIIERATOA:
12 06/09, ARIZONA PUBLIC SERVICE CO.,
PALO VEROE, UNIT 1
DATE! 07/19/89 TIME:
13: 43'4 LOCATION:
TSC TD TSC+ 15. 0 CRITERIA:
SLG PLUGGED 0
STAYS UNIT 1 S/G 2
TUBE SHEET'OLD LEG END OF CYCLE 2 0.0 in 0.2 in 1
0.3 in 0.4 in 0
0.5 in - 0.6 in 58 0.7 in - 0.8 in 67
> 08 in 39 CBp
!no I30
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Rp 10 ip ip 10 30 10 10 c?0 30 40 50 60 70 80 90 100 110 120 13t) 140 150 160 170 180
STEAM GENEBATOB:
12 06/09, ARIZONA PUBLIC SERVICE CO.,
PALO VERDE, UHIT i E
OATE: 07/19/89 TIME:
13: 27: 56
. LOCATION:
TSH TO TSH+ 15.0 CBITEAIA: SLG PLUGGEO 0
STAYS UNIT 1
S/G 2
TUBE SHEET HOT LEG ENO OF CYCLE 2 0.0 in 0.2 in 0
0.3 in - 0.4 in i4 0.5 in - 0.6 in 104
- 0. 7 in 0. 8 in i40
> 0.8 in 58 0
.:n.
30
~
~ in.
PQ '00 l0 20 10-i0 7.0 30 40 50 60 70 80 80 100 110 120 130 140 160 150 170 180
I