ML19340A518
| ML19340A518 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/20/1969 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 8008070703 | |
| Download: ML19340A518 (53) | |
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\\l ANNUAL REPORT YEAR 1968
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DRESDEN NUCLEAR POWER STATION 1
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j ANNUAL REPORT OF STATION OPERATION j
FOR THE YEAR 1968 i.
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January 20, 1969 i
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DRESDEN NUCLEAR POWER STATION ANNUAL REPORT I.
INTRODUCTION This seventh annual report is submitted in compliance with paragraph 3.C (2) of the Utilization Facility License DPR-2, as amended, and covers operation of Dresden Nuclear Power Station during the year 1968.
1 II.
SUMMARY
OF OPERATIONS A.
Scope of Operations i
j Dresden Station operated during 1968 with a total of 15 shutdowns.
Included in this total was the fif th partial refueling, inspection 7
and maintenance outage, which extended from February 3, until June 2, 1968. This outage was quite extensive and included items such as: Overhaul of the main turbine-generator; removal of turbine crossover grating; refueling of 96 fuel assemblies and cleaning of all fuel returning to the core for Cycle 6 operation; overhaul of 40 control rod drives; inspection and testing of major primary system welds; and general maintenance and inspections made available by the shutdown.
During the year, additions to and changes in facility design were made by: Addition of a steam sample station on "E" turbine extraction line; modification of the " trip" circuit on the reactor canal crane; addition of a new radiation monitor in the fuel building; installation of a materials test loop in "B" secondary steam generator compartment; revision of the mechanical vacuum pump six inch discharge line; erection of control rod drive thimble support fixtures; modifications to No. I deepwell pump; addition of mechanical stops to the off gas menitor range switch; erection of an on-site 400 foot meteorological j
tower; reactivation of Hansel and Breen Environs Stations; and the j
addition of tie connections between Units #1, #.2, and #3 service air j
and water systems.
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A total of eight shipments, consisting of 184 spent fuel assemblies, was shipped to the Chemical Processing Plant of Nuclear Fuel Services, i
Inc., at West Valley, New York during the period.
B.
Shutdowns The plant was shutdown 15 times during 1968 as shown in Table 1 and Figure 1.
Twelve of these were forced outages, all of which were due to turbine condenser tube, leak repairs. There were three scheduled outages: Two for operator training and one for the major inspection, refueling and maintenance outage.
Three of the 12 forced outages were temporarily extended during the year. The first of these occurred on October 30, for repairs to "B" clean-up demineralizer pump; the second on November 10, for operatar i
training; and the last, on December 26, for a leak check in "A" unloeding heat exchanger oil cooler.
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Load Restrictions The load restrictions imposed during the year are listed in Table 2.
i Restrictions were due to reactor recirculation loop outages, feedwater heater outages, clean-up demineralizer pump outages, incore instrument j
stabilization and calibration, and end of core life physics tests.
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TABLE 1 OPERATING PERFORMANCE 1968 No. Of Off System On System Outege Date Time Date Time Outage llours Reason For Outage 84 1/13/68 2:36 a.m.
1/15/68 5:34 et.m.
50 Hrs. 58 Min. Operator training.
85 2/ 3/68 12: 56 a.m.
6/ 2/68 5: 52 a.m.
2,883 Hrs. 56 Fif th partial refueling, control rod drive overhaul, fuel cleaning, generator overhaul.
86 7/10/68 6:31 p.m.
7/12/68 12:22 a.m.
29 Hrs. 51 Min.
Condenser tube leak repair.
87 7/20/68 1:33 a.m.
7/21/68 6:50 a.m.
29 Hrs. 17 Min.
Condenser tube leak repair.
88 8/24/68 1:42 a.m.
8/26/68 12:27 a.m.
46 lirs. 45 Min.
Condenser tube leak repair.
89 9/ 6/68 11: 18 p.m.
9/ 8/68 1: 49 p.m.
38 Hrs. 31 Min. Operator training.
90 9/20/68 6:27 a.m.
9/22/68 7: 54 a.m.
49 Hrs. 27 Min.
Condenser tube leak repair.
91 10/28/68 1:36 a.m.
10/29/68 9:45 a.m.
32 lirs.
9 Min.
Condenser tube leak repair.
92 10/29/68 12:17 p.m.
11/ 1/68 1: 51 a.m.
61 lirs. 34 Min.
Condenser tube repairs (15 Hrs. 43 Min.),
"B" clean-up pump repair (45 Hrs. 51 Min.).
93 11/ 9/68 1: 38 a.m.
11/10/68 10:39 a.m.
33 Hrs.
1 Min.
Condenser tube repairs (17 Hrs. 7 Min.),
Operator training (15 Hrs. 54 Min.).
94 11/23/68 5:06 a.m.
11/25/68 5:08 a.m.
48 Hrs.
2 Min.
Condenser tube leak repairs.
95 12/ 8/68 9:23 a.m.
12/ 9/68 11:40 a.m.
26 Hrs. 17 Min.
Condenser tube leak repairs.
96 12/16/68 9:08 p.m.
12/18/68 5:47 a.m.
32 Hrs. 39 Min.
Condenser tube leak repairs.
97 12/25/68 8:26 a.m.
12/26/68 8:42 p.m.
36 Hrs. 16 Min.
Condenser tube leak repairs (24 Hrs. 4 Min.),"A" unloading heat exchanger oil cooler leak check (12 Hrs. 12 Min.).
TABLE 1 OPERATING PERFORMANCE 1968 (Continued)
No. Of Off System On System Outage Date Time Date Time Outage Hours Reason For Outage 98 12/28/68 10: 22 p.m.
12/30/68 2:45 a.m.
28 Hrs. 23 Min.
Condenser tube leaks repairs.
TOTAL OUTAGE HOURS FOR YEAR 3,427 Hrs. 5 Min.
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TABLE 2 LOAD RESTRICTIONS FOR 1968 Reduction From Maximum Date Capability of 210 MW Condition January 16 - January 29 25 "B" Secondary Steam Generator Tube Leak January 29 - January 31 10 "A" Drain Cooler Tube Leak j
January 31 - February 3 30 End of Life Core Physics Test June 2 - June 6 100 Reactor Incore Calibration June 6 - June 11 50 Reactor Incore Calibration Y
June 11 - June 15 20 Reactor Incore Calibration June 15 - June 25 18 "E" Primary Heater Tube Leak July 12 - July 13 110 "B" Reactor Clean-up Pump Failure July 13 - July 17 10 "A" Secondary Feedwater Heater Tube Leak i
October 3 - October 16 35 "C" Secondary Steam Generator Flange Leak October 16 - October 21 10 "C" Primary and "A" Secondary Feedwater Heater Tube Leak October 21 - October 29 25 "C" and "D" Primary Heater Tube Leak and "A" Secondary Heater Tube Leak November 1 - November 4 25 "C" and "D" Primary, "A" Secondary Feedwater Heater Tube Leak November 4 - November 9 18 "C" Primary and "A" Secondary Feedwater Heater Tube Leak l
November 10 110 Condensate Domineralizer High Differential Pressure
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TABLE 2 LOAD RESTRICTIONS FOR 1968 (Continued)
Reduction From Maximum Date Capability of 210 MWe Condition November 11 - December 21 18 "C" Primary and "A" Secondary Feedwater Heater Tube Leak December 21 - December 30 45 "B" Secondary Steam Generator Tube Leak Repair December 30 - December 31 60 "B" Secondary Steam Generator Tube Leak Repair and Feedwater Heater Tube Leak Repair P
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FIGURE 1 PL A NT ELECTRIC A L LO ADIN G YEAR 1968 DRESDEN NUCLE AR POWER STATION 220 -
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FEB. ' liar.
' APRIL
!!AY
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' AUG.
' SEPT.
' OCT, 10V.
DEC.
. III. DISCUSSION A.
Operating Experience 1.
Generation The total reactor operating (critical) time during the year was 5,583 hours0.00675 days <br />0.162 hours <br />9.63955e-4 weeks <br />2.218315e-4 months <br /> and the total power for the period was 133,307. 7 KfDt.
The gross electrical generation during the year was 966,792.23 MWHe; net generation was 915,930.00 K41e. As of December 31, 1968, the total gross generation since conmencement of power operation on April 15, 1960, was 8,421,120.27 MWHe.
2.
4 crams a.
At 6:50 p.m.,
on November 9, during operator training criticals, a spurious signal on Channel 3 out-of-core initiated a scram due to the coincidence /non-coincidence switch not being locked in the full coincidence position.
The reactor power was approximately 200 watts with 49 rods and four notches withdrawn prior to the scram.
b.
At 11: 33 p.m., on December 17, the reactor was scrammed by the plant sa fety system from the No. I vacuum trip. The scram occurred while heating the primary system at approximately 500 F per hour, with the reactor power at approximately 20 MJt, as the unit was returning to service following a condenser tube repair outage. The scram was due to the approach of reactor pressure to 200 psig before the No. 1 vacuum trip had reset at tha setpoint of 23" Hg.
c.
At 11: 50 a.m., on December 26, a reactor scram was in itia ted by a spurious signal from Channel 4 out-of-core while down ranging the micro-micro ammeters. The reactor was on a post outage heating rate, at 3760 F and approximately 700 watts, with 42 rods and five notches withdrawn prior to the scram.
3.
Incidents a.
During scram te. sting on February 5, Accumulator No. 17 failed to scram.
Investigation of the system revealed that the plastic core pilot seat in the scram solenoid valve had broken off and blocked the bleer-off, preventing the scram valves from opening.
The seat was repaired and ten additional seats from the 54 scram solenoid valves were picked at random and inspected.
From the conditions observed, no further replacements were necessary. This has been the second seat to malfunction during the history of these valves.
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b.
On September 23, 1968, the post incident system valves became inoperative from the remote location after rain water leaked into the electrical distribution panel.
The leakage originated from an accumulation of rain water on the roof above the panel due to a plugged roof drain.
I Repair of the distribution panel was completed within three hours, load was reduced to 135 MWe and shutdown requirements were evaluated vs. repair time. As a result t
of the occurrence, permanent protection of the existing panel has been completed and all other distribution panels for plant systems, both for safety and operating functions, have been reviewed by station management and found to be
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adequate.
4.
Control Rod Drives a.
Control Rod Drive Ooeration l
1.
While exercising control rod drives on January 5, Drive A-5 inserted to Position 4 during attempts to withdraw it from Position 8.
The drive was moved to the fully inserted position and flushed.
It was then exercised 5
and found to be working properly. The malfunction was attributed to a sticking Barksdale valve.
The Barksdale was taken out of service for inspection and repair and returned to service on January 16 with no further problems being experienced.
2.
On April 20, while the reactor was in the refuel mode, control rod Drive E-5 could not be withdrawn from Position 0.
Investigation led through a series of orifice and drive water regulator setting changes, inspection of the Barkadale and solenoid valves, and checking the conditions of the seats on the Asco valves. Although no apparent reason for the problem was found, the drive operated properly when it was returned to service. It was believed that the malfunction was caused by a blocked solenoid orifice, which was relieved when the solenoids were disassembled. Although the problem would cause the drive to malfunction in this manner, it would not impair the drive's ability to scram.
3.
While attempting to use control rod Drive F-5 as a cocked rod during reactor fuel loading on April 20, the drive failed to withdraw at normal or increased hydraulic pressures.
Investigation of _ the occurrence revealed that a fuel element nosecone 'ad parted frem the fuel assembly during movement and fell into the guide tube opening on the core lower grid, thereby obstructing control rod blade movement.
No further problems were experienced with Drive F-5 af ter the element nosecone was removed from the control cell.
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4.
During daily control rod drive exercises on May 10, control rod E-6 was withdrawn to Position 12.
It then inadvertently drif ted to Position 11.
The malfunction of the drive is attributed e.o a sticking Asco valve, which when slow or reluctant to move, would cause the drive to drif t af ter previous insert or withdraw movements.
Abnormal operations of the drive were cleared when.the Asco's were changed on August 27.
5.
On June 7, during rod withdrawal, the position indicator for Drive H-9 was noticed to jump to Position 12, hold f or a time, then drop back again. The drive was then 4
j incerted with the indicator again jumping to Position 12, this time remaining there. The cause of the malfunction was due to a faulty probe, which af ter replacement on July 11, revealed no further abnormalities.
j 6.
On June 8, control rod Drive H-4 position indicator was noticed to juup off scale (past 12) when the drive w&s inserted from Position 12 to Position 11.
The drive was withdrawn to Position 12 and again inserted to Position 11.
Deflection was observed once more. From the irratic response of the indicator, it was assumed i
that Switch 12 on the drive probe was remaining closed and the resulting combination of Switches 11 and 12 caused the full scale reading. The faulty probe was replaced on July 11.
J 7.
On June 12, with the drive at Position 0 and the green light on, the drive position indicator of H-5 drive was noticed to suddenly jump to Position 3, hold for a t ime, then drop back. Observation resulted in the i
conclusion that Switch 3 on the drive prebe was shorting out on occasion, thus moving the indicator to Position 3.
2 The faulty probe was replaced on July 11.
8.
During daily control rod drive exercises on June 12, the green position indication light for Drive K-5 was observ-ed to remain on at positions other than 0.
The trans-mitter for Drive K-5 was changed, but the green light remained on.
The conclusion therefore, was that the drive probe S. G. switch was probably shorted out.
The probe was changed on July 11, alleviating the problem.
On July 15, when the transmitter was replaced, a relay fuse blew in the drives' control circuit. Because of the abnormal response of this transmitter, and similar responses to a later transmitter, it was concluded that either the probe or probe connector was shorted. The probe was replaced on July 20, af ter which normal operation i
resumed.
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9.
During reactor startup on July 20, prior to becoming critical, control rod drives on Accumulator 21 (E-8, A-7 and C-3) and Accumulator 17 (D-10, D-7 and C-2) drif ted in following previous withdraw movements.
Extensive investigation of piping, valves, pressure settings, etc.
revealed no explicit cause of the malfunction. However, it is believed that the scram inlet valves from both accumulators were not seated properly at the lower tempera-tures and thus caused the drives to insert. When the reactor was at rated temperature and pressure, and follow-ing extensive movements and testing for drift, all six drives on the two accumulators began operating normally, and have been operating properly since the occurrence.
10.
During control rod drive exercises on September 4, control rod Drive D-7 would not unlatch from Position 0 with normal or increased hydraulic pressures. The Asco orifices to Drive D-7 were checked and the insert orifice found fully closed. Normal position of this orifice is completely open.
The system was returned to correct settings and normal operation was continued.
11.
On December 20, with the reactor at normal operating pressures and temperatures, control rod Drive H-7 failed to withdraw from Position O.
Varied pressures and Asco settings resulted in only intermittent movement of the drive. Finally, af ter successfully moving the drive from Position 0 to Position 1, H-7 drif ted on to Position 2 af ter a 4 - 5 minute settling period. The drive was left at this location with no further movement attempted.
Additional exercising and flushing of H-7 will continue until the problem is resolved.
It is believed that a crud buildup is temporarily hung up in the collet piston -
shuttle piston area and is therefore preventing normal unlatching of the collet fingers, b.
Control Rod Drive Tests on February 5, all control rod drives, except SN 1290 (Core Position F-6), were scram and friction tested and timed for normal insertion and withdrawal. Drive 1290 was not scram tested because of its previous history of short buffer. This testing was performed as a routine check of drives to be over-hauled during the spring refueling outage.
Following the overhaul and replacement of 40 drives during March and April,1968, initial scram and timing tests were performed on the 80 drives cn April 14, to insure the operability of each drive prior to loading fuel. The tests wete conducted using
" dummy" fuel assemblies in the fuel cells. These tests were followed by normal timing and latching tests and scram and friction tests on }by 10 and 11, and on October 29 - 31.
The data obtained from all the above tests were satisfactory on all drives.
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i c.
Inspection The technical specifications to Dresden's License DPR-2, as amended by Change Nc. 7, dated April 9,1964, state that during major outagas, "Not less than two control rod drive mechanisms shall be removed, disassembled, and thoroughly inspected at intervals not to exceed 24 months."
In addition to license requirements, drives removed for inspection are selected on the basis of drive test results and malfunctions experienced during operation.
Forty control rod drives were removed and replaced with repaired drives during the fif th partial refueling outage.
l Many of these drives had exhibited long insert times during Cycle V.
Four spare drives, SN 1233 SN 1273, SN 1257 and SN 1243, were overhauled prior to the outage and were tested satisfactorily in early February, 1968, in the drive test i
facility. These drives replaced four drives removed from the reactor, of which two of the replaced drives plus 38 others removed during the outage were overhauled, inspected returned to the reactor, and satisfactorily tested before i
reloadings sta rted. Figure 2 shows the location of the 40 drives removed.
After removal of reactor head, turning vane and fuel, the control blades were removed and index rubes withdraun. The J
drives were removed from core and transported to the 565' elevation.
1.
Visual Inspection j
All parts were visually inspected as closely as radiation
.i levels would permit. A 1/4" plexiglass face shield was i
used to protect against beta emission when viewing parts, i~
All moving parts on roller mounts were actually moved in i
order to verify freedom of movement. Visual inspection of roller mount assemblies was done at about a three foot j
distance or viewed underwater.
i All canne? magnets were dropped into hot water and heated to about 2000 F.
Leakers were found by streams of air i
bubbles being forced f rom small racks.
2.
Flourescent Dve Penetrant Check i
A fluorescent dye penetrant Zyglo examination was made on components of drives. A penetrant (ZL-2), developer ZP-5, and ultraviolet light were used for the inspection.
Af ter ultrasonic cleaning, parts were painted uith penetrant i
and allowed to set for 20 minutes to allow capillary action time to draw penetrant into any cracks.
The dyc was then wiped off with lint free rags removing all dye except tha t in small scratches and cracks.
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The part was dusted li htly with developer in order to j
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draw dye from any cracks. The dye fluoresces under ultraviolet light and cracks will stand out readily against the background.
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3.
Results of Inspection a.
Of the 40 drives removed, only 38 were overhauled i
for lack of spare parts. The remaining two drives will be overhauled at a later date.
i b.
Nitrided guide roller pins were used to replace all chrome placed pins (ILA3448) in the roller mount l
assembly. Practically all of the pins removed exhibited some wear in the form of loss of chrome.
Of the 152 pine. only one had failed and two were severely wcen.
i.11 others had worn an average of six thousandths on the 250 thousandths diameter.
1 c.
All rollers were checked with a go-no-go gage. This gage has an outside diameter of.260".
The maximum tolerance for the hole of a new roller is.259",
thus only one thousandth was acceptable for roller wear. Only sixteen rollers were worn an amount greater than this. No effort was made to determine the actual wear of rejected rollers.
The replacement rollers were nitrided inside the pin hole to minimize wear.
d.
Dye checking revealed chrome cracking or flaking on seven guide plugs very similar to conditions found on inspection of December, 1962. (Sec comments on individual drives for specific details.)
Aside from guide plug chrome plating, no cracking was found in drive components.
e.
All seals were replaced on every drive overhauled.
In the absence of gross seal wear or breakage, the condition of seals causing long insert times is described as " normal wear" in the report.
f.
Eleven drives had scratches on the inside of the shuttle piston (8563398P1) of magnitude sufficient to cause interference with the collet assembly and in some cases, malfunctions of the collet.
Table 3 contains a summary of the 1968 drive i
inspection and overhaal.
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. FIGURg 2 LOCATION OF CONTROL ROD DRIVES OVERHAULED DURI;;C 1968 MAJOR OUTAGE 10 9
8 7
6
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4 3
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B C
D E
F G
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. TI ELt' 3 1968 DatVE INSFECTION PECAF OtR(DtE SEVERLT EXCESSIVE D'IE8FECT10NS SCRATO(E5 D4tACED BENT worn OK WC9N OR DTE AIVE 0)RE DEFICIENCY WORM ROL12R WEAR ON ON CUIDE ON SHUMLE STRAINER fI?CERS ON 3RCf1N DAIVE BBOKEN STUF CHECK FAET$ BEFLACgD IF.W 1 PostTTOM FINS DOLLFRS FLIC 0. D.
PIST0h ID (6558170) CCLtIT ASS'T FI5 TOM SL4LS FISTDN SE4LS S E$t'LTS et 1294 31 DRIFT SIDd I
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OK 1292 A6 S?JCK SIDd X
X X
CK Magnet, collet pistoa singe 1279 K4 STUCE DRIFT 2
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OK Magnet 1254 J9 seTFT $1Dd i
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OK Inden tube 1274 E4 SLAW I
I OK Collet adapter sleece, collet assembly 1262 B4 DRIFT I
I OK Collet assembly 1244 FI
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OK Cuide plug, atralner 1290 F6 SHORT SUFFER I
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I Cuide plus Calde plug, cellet adapter cracked sleeve 1284 E9 SHORT SUFFER I
I OK 1252 E6 SHORT BUFFER I
3 CK Boller mount essembly 1226 DS SHORT BUFFER E
X OK 1259 C6 SHORT BUFFEk I
CK Cuide plug, sognet 1506 M9 SHORT ALTTER 1
4 E
I X
X OK Cuide plug, cellet assembly 1317 F8
$HORT SLTTER E
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CK Magnet 1237 F4 SHORT BLTFER 2
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I X
Detve pisten 6 pistem head 1267 F5 SHORT BUTFER X
OK F11ter (8568985) 1263 FID SIDd E
I CK Magnet 1310 A7 Stad 2
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OK Cuide plaig, (11ter G 565905) 1248 DI Stad 1
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OK 1297 Cl Stbd 1
X X
X Cuide plug 1238 F9 SIDd I
E CK 1284 89 SIDd 1 broken 3
K OK 1316 D6 Stad X
X X
CK 1265 38 Stad E
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CK Strainer 1315 C8 51De I
E X
CE 1231 K7
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I OK 1313 F7 SIDd I
I OK 1293 K1 SIDA E
OK 1284 C7 SIDd I
I OK 1227 C9 SIDd I
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CK Magnet, lades tube 1260 E8 Stad X
X X
OK 1242 D4 SIDd X
X X
OK 1256 D2 SIDd I
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I CK Meseet, strainer, collet
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} 1239 MS SIDd eseembtfee 1
I OR Filter
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} 1270 1109 GIO SIDd E
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Calde pleg H2 Stad I
I OE 1230 K5 Stad 1
I OK 1210 C2 Stad 1
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h * " DRIFT" meana drive fatted to latch
" STUCK" meene drive ccull not be moved "stDd" sneans drive had LonR insert time, le greater than 26 seconde
- %tur; f BUFFt.p" s.eans ur s ve atJ not slow a.
s-of f icientis at e3J a,f w rame s t ro' v (t rasel tire las t t.e buffer less ti..en. I secuem:s)
- ta nddition to parte replaced as noted, all roller ping, drive pleton seste and stop piston seals teere replaced on all overhauled drives W
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i 1 5.
Blade Following Checks During periods of operation, control rods have been verified for blade following on a weekly basis. Control rod sorth tests were also conducted prior to and following the refuelind outage and during each month of operation. During each startup, control rod patterns for criticality have been predicted and all blade following verified, b.
Control Blade Inspection a
j Control blades were inspected in the fuel building storage pool during the refueling outage on April 7 and 8, 1668.
Eight blades were checked with go-no go gages for dimensional variations and five of the eight by underwater TV for signs of visual wear. All blades inspected were found to be in I
acceptable condition. The blades examined and thei.e locatinns in the core are exhibited below:
A B C D E F G 11 J K A B C D E F G 11 J K 4
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7 7
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6 88 6
38 5
5 4
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B B
3 85 al 3
85 B
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40 21 2
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l Control Blades Control Blades Examined By Examined By l
Co-No p Gage Underwater TV i
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Reactor Clean-up Demineralizer System
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"A" clean-up demineralizar loop Itas been out of service since Novembe r, 1966. The loop was taken out of service due to i
apparent tube leaks in the regenerative hea t exchangers.
l During the month of August, 1968, a decontamination project was j
undertaken to reduce the high contamination levels inside of "A" clean-up loop piping and heat exchangers. Prior to the clean-up p roj ect, work had been seriously hampered by the high radiation fields near the demineralizers.
The decontamination project was under the direction of the Dow Chemical Company, Dow Industrial Services, and Accor, Inc.,
l who, with the locp isolated, circulated an acid solvent through the "A" system. The decontamination of the clean-up loop was completed in September. As a result of the cleaning, the contamination levels were reduced to levels low enough to allow access to the loop for inspection and repair work, At the close of the year, this work was still progressing on the repair of "A" clean-up denincralizer loop heat exchangers.
i On July 12, 1963, during a plant startup, "B" cican-up demineralizer pump tripped out of service af ter a control rod drive spud guide roller wedged between the impeller and pump casing. The unit was dorated 100 We for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> with no
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primary clean-up system in service while the pump from "A" clean-up loop was installed in "3" clean-up circuit.
i
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On October 29, during a condenser tube leak repair outage, "B" clean-up demineralizer pump again tripped out of service j
af ter a small piece of metal wedged between the impeller and the pump casing. The pump was repaired end the unit returned j
to service November 1.
7.
Chanties in Facilitv Desien
)
a.
Fuel Buildine: Radiation Monitor 1
1 During the month of January, an Area nadiation Monitor was installed in the fuel building. The nonitor alarms in the l
control room and fuel building when the sensor, located at j
the north end of building adjacent to the storag pool, sees more i
chan 10 mk por hour.
In addition, the monitor trips all power to
(
the overhead crane, thus preventing any further movement.
As was the original Riggs monitor, the new Area Radiation Monitor is tested each day when men are working in the 4
area.
b.
Reactor Canal Crane Circuit Trip Modification A trip circuit was installed on the reactor canal crane n ring the month of March. Any one of threa radiation I
i
, ~,---
~--.c
-n.
1 i
D 9
<b
~
cn en ca ev SN
. e0.
3 eu eu monitors (set to alarm at 10 mR/Hr.) will now trip the reactor canal crane main contactor during refueling operations. Hand operated trip buttons were also placed at the north and south enda and center of the pool area.
c.
Material Test Loop Installation 4
During the fif th partial refueling outage, General Electric 1
Company initiated installation of a materials test loop in j
. "B" reactor recirculation loop.
This installation was completed on May 28, 1968. The purpose of the facility is d
to determine whether corrosion by a reactor water environment can significantly reduce the fatigue life of austenitic stainless steels. A secondary objective is to determine j
whether metal fatigue in a boiling water reactor coolant can change mode of propagation, from transgranular to an j
intergranular mode.
The facility includes 1 1/2 inch supply and return lines
- j from "B" loop _which will supply approximately l GPM of reactor water to an autoclave outside the generator room.
Both lines of' the loop are terminated outside of "3" compartment with a valve and blind flange pending installation of the autoclave later in 1969. The autoclave will contain stainless steel metal samples for strain cycling and associated j
instrumentation and control. The supply line connects to a cross flange installed on the six inch bypass loop decontamination flange, The return line onnects to a second cross flange installed on the. decontamination flange on the generator inlet line. Both lines have motor operated isolation
. valves inside the compartment.
i The two cross flanges house four electrochemical corrosion '
j -
test flanges which were installed during the outage. The test flanges are a continuation of a corrosion rate test
- {
program initiated during-1967.
i i
- d. _ Vacuum Pump Discharge Line 2evision The six inch discharge line from the mechanical vacuu.n pump was rerouted from the 30 inch air ejector holdup to the 24 l
. inch gland exhauster holdup on May 12-15.
The revision restores the mechanical vacuum pucp tie to 'the 24 inch holdup line as it was originally designed into the plant.
During Unit il startup testing in 1960, the piping had been routed to the 30 inch holdup because of backpressure on.the turbine gland seal system. The backpressure problem was climinated by a connection directly into the 24 inch holdup header rather than the original six inch gland exhauster discharge line.
A pre-operat!onal test was made on this new arrangement on May 27.
Test data, taken from various vacuum pump and gland exhauster operational combinations, indicate that no prcblems
{
exist with the revision.
4 i
i I
D{}
(
e.
Control Rod Drive Thimble Supports Support structures, for the reactor control rod drive thimble housings, were installed during May by Shear Connectors, Inc.,
under the supervision of the Commonwealth Edison Mechanical and Structural Department. The supports, which were installed below the 80 control rod drive flanges at the base of the reactor, are designed to prevent drive ejecticn from the vessel in the event of a thimble weld failure, f.
Off Gas Monitor Range Switch A mechanical stop was installed on the of f gas monitor range switch during August. The mechanical stop prevents ranging the monitor above values which correspond to off gas release limits.
g.
Extension of No. 1 Deep Well The No. 1 deep well was modified during the latter part of 1968 to accommodate the additions of Dresden 2 and 3.
The deep well, which is located approximately 400 feet directly south of the station for supply of well water to Dresden Unit No. 1, was extended from the original depth of 788 feet to a new, more abundant supply depth of 1,500 feet.
Digging of the well uas essentially completed on December 6, and capacity and water sampling tests were in progress at the end of the year.
h.
Reactivation of Hansel and Breen Environs Stations On August 25, the Hansel and 3reen Environs Stations were returned to active service. Both stations were returned for operation as they were originally before deactivation in.1967, although both have been modified for use only in direct radiation monitoring of the environs.
- i. Meteorological Survev Tower Erection For determination of climatological conditions at the s ta tion, a site meteorological survey tower was constructed and put into operation during the early part of 1968. The facility consists of a 400 foot tower with four distinct elevations for measurement of wind speed, direction and temperature. The data obtained from these elevations is fed to a computer at the case of the tower for computatien and recording for contra ctor analysis. The data will be used for determination of:
1.
Climatological conditions 2.
Ground level concentrations from stack gas emissions J. _Dresden Units No. 1, 2 and 3 System Connections Connections between operational systems on Units No. 1, 2 and 3 were made possible during 1968 in the following locations.
m,
_ =
i b
b q
_ w o Ju o J uu dJ
. \\.
2 1.
Unit #1 make-up pumps are available to supply Units 2 and 3 heating boilers in the event of loss of the Unit 2 clean demineralized water pumps. The crosstic connects Unit 1 to Unit 2 via a three inch line with manual isolation valves.
A check valve, also installed, prevents the normally higher Unit 2 pressure from supplying Unit 1.
2.
A tie into Unit I clean demineralized water storage tank was made for the Unit 2 and 3 clean demineralized water pumps. The tie is a three inch line with manual isolation valves for normal supply to Units 2 and 3 pumps.
a 3.
A six inch tie into the Unit I well water tank was made for Units 2 and,3 make-up feed pumps.
4.
A crosstie between Units 1 and 2 clean demineralized water header exists on Unit 2 in order to supply demineralized water to Unit 2 should the Unit 2 clean demineralized water -
pumps be out of service. The crosstie is made via a three inch line with manual isolation valves.
5.
Units 1, 2 and 3 instrument-and service air systems can be interconnected.
Service Air is crosstied to Unit 2 by a two inch line.
The cross connection is automatic at 95 psi on Unit 2 by action of a pressure regulating valve between the systems.
The crosstie is to act as a backup for Unit 2 service air in as much as Unit 2 has only one air compressor.
Instrument Air is crosstied between Units I and 2 through a two inch line. The cross connection is automatic via a pressure control valve set at 92 psi.
It is used to supply instrument air to Unit 2 if the Unit 2 air driers were out of service or instrument air pressure dropped below the required setpoint. The crosstie is limited by 1/4 inch orifice which prevents a break in the Unit 2 a
instrument air header from dropping Unit 1 instrument air pressure.
All systems have been tested and are operational. The connections have all been used intermittently.as construction on the two newer units progresses.
k.
Installation of Unit #1 Evacuation Sirens Tuo Unit 1 evacuation alarm sirens were installed during August in the Unit 4 and 3 construction area. One siren was installed on the cribbouse roof in the Unit 2 area and the other at the south-west corner of the main floor of tee Unit 3 turbine build-ing. The control key and control relays for these two alarms are mounted. on a wall south of Unit I generator exciter. A 4,800 volt feed is supplied from CTR 26, bus SA (Unit 1) at 100 AMPS. The sirens are actuated automatically by Unit I reactor sphere pressure or by a control key for testing.
8 e
1 cn en ch'
1.
"A" Clean-up Loop Decontamination Piping Isolation During Augus t, two isolation valves (A0 600 and A0 601) were installed in "A" clean-up domineralizer pump room to facilitate decontamination of the "A" clean-up loop non-regenerative heat exchangers. Both valves are two inch air operated valves and are installed on a spare two inch stainless steel line (5639) to the radwaste area. The valves are controlled from the control room, with each valve having position indicating lights on Bench Panel B-3.
In the event of a scram, both A0 600 and A0 601 will close automatically through the plant safety system relays.
8.
Personnel Radiation Exposure Personnel exposures to radiation were within limits specified in 10 CFR Part 20.
9.
Liquid Poison System The liquid poisen system was operative at all times during the year. The boron poison was sampled on March 29, May 29, May 30, September 21, and October 31.
There were no conditions which would indicate a loss of boron from the solution tank.
Boron concentrations in the reactor water remained low throughout the year.
10.
Radioactive Waste Disposal Release of radioactive liquid waste was accomplished in batch quantities at controlled release flow rates according to establish-ed procedures. The contribution to the activity of dilution water was always maintained within the limits specified in the applicable federal regulations. The average contribution to the unidentified activity in the cater utilized for radioactive liquid waste dilution during the year was calculated to be.189 x 10-7 uc/ml (18.9 uuc/1) compared to an average limit of 1.00 x 10-7 ue/ml (100 uuc/1) for unidentified mixtures containing radium 226 or radium 228 as specified in 10 CFR Part 20.
Solid radioactive wastes were stored on-site pursuant to License DPR-2.
Table 4 shows the content, shipment locations, and dates of radioactive waste shipments made during the year. A total of 11 radioactive waste shipments were made in 1968.
The concentration of noble fission gases in the stack discharge to atmosphere was maintained well within license limits of 700,000 microcuries per second. The average activity release rate for the year while the, plant was operating was approxima tely 12,500 uc/second.
Eight Nuclear Fuel Services, Inc. - - Stanray rail cask shipments, totaling 184 spent fuel assemblies, were made during the year to the Chemical Processing Plant of Nuclear Fuel Services, Inc., in West Valley, New York. Table 5 is a breakdown of these shipments in addition to all shipments made since initiation in June, 1965.
TABLE 4 RADI0 ACTIVE WASTE SHIPMENTS - 1968 Total Activity Content Date Volume (Millicuries)
Location of Shipment Dry Radioactive Waste May 1 286.65 ft3 318.96 Nuclear Engineering Co.;
Sheffield Nuclear Center; Sheffield, Illinois Dry Radioactive Waste May 2 499.80 ft3 142.21 Dry. Radioactive Waste May 2 405.00 ft3 600.51
~
3 16.48 Dry Radioactive Waste May 3 1,116.00 ft Dry Radioactive Uaste July 9 1,374.60 ft3 147.96 7
Dry Radioac'tive Waste July 10 730.50 fc3 105.06 3
Dry Radioactive Waste November 13 1,523.85 ft 141.85 Dry Radioactive Waste November 14 342.00 ft3 57.20 Dry Radioactive Waste November 15 321.00 ft3 58.96 3
251.27
~ Dry Radioactive Waste November 18 1,147.45 ft High Level Uaste August 23 60.00 ft3 207,000.00 Accor, Inc.;
Hawthorne, New York l
TABLE 5 SPENT FUEL SHIPMENT
SUMMARY
Number of Assemblies or Containers Total Shipment Number Date Batch 3
Rail Truck Shipped 1
2 3
4 5
Ra il Truck Date 1
o/11/65 24 0
0 0
0 24 24 2
6/30/65 24 0
0 0
0 24 48 3
7/16/65 24 0
0 0
0 24 72 4
8/ 3/65 24 0
0 0
0 24 96 5
8/16/65 24 0
0 0
0 24 130 6
9/ 2/65 24 0
0 0
0 24 154 7
9/22/65 24 0
0 0
0 24 168 1
8/ 1/66 0
0 4
0 0
4 172 8
8/ 5/66 16 0
6 0
0 22 194 2
8/15/66 0
0 4
0 0
4 198 3
8/24/66 0
0 4
0 0
4 202 4
8/28/66 0
0 4
0 0
4 206 9
8/31/66 0
0 12 8
0 20 226 5
9/ 5/66 0
G 4
0 0
4 230 6
9/12/66 0
0 4
0 0
4 234 7
9/14/66 0
0 4
0 0
4 238 10 9/16/66 0
0 0
20 0
20 258 8
9/19/66 0
0 4
0 0
4 262 9
9/21/66 0
0 4
0 0
4 266 10 9/25/66 0
0 4
0 0
4 270 11 9/26/66 0
0 4
0 0
4 274 12 9/28/66 0
0 4
0 0
4 278 13 10/ 2/66 0
0 4
0 0
4 282 14 10/ 3/66 0
0 4
0 0
4 286 11 10/ 7/66 0
0 0
24 0
24 310 15 10/11/66 0
0 4
0 0
4 314 16 10/12/66 0
0 4
0 0
4 318 17 10/20/66 0
0 4
0 0
4 322 18 10/23/66 0
0 4
C 0
4 326 19 10/26/66 0
0 4
0 0
4 330 12 10/28/66 0
0 3
16 0
19 349 20 10/30/66 0
0 4
0 0
4 353 21 11/ 1/66 0
0 4
0 0
4 357 22 11/ 6/66 0
0 4
0 0
4 361 23 11/ 8/66 0
0 4
0 0
4 365 24 11/10/66 0
0 4
0 0
4 369 25 11/13/66 0
0 4
0 0
4 373 13 11/14/66 0
0 0
23 0
23 396 26 11/15/66 0
0 4
0 0
4 400 27 11/17/66 0
0 4
0 0
4 404 28 11/20/66 0
0 4
0 0
4 408 29 11/27/66 0
0 4
C 0
4 412 30 11/29/66 0
0 4
0 0
4 416
f Number of Assemblies or Containers Total Shipment Number Date Batch
_T_o -
o Rail Truck Shipped 1
2 3
4 5
Rail Truck Date 14 12/ 2/66 0
0 4
19 0
23 439 31 12/ 6/66 0
0 4
0 0
4 443 3:
12/11/66 0
0 4
0 0
4 447 3*
12/15/66 0
0 4
0 0
4 451 34 12/18/66 0
0 4
0 0
4 455 35 12/20/66 0
0 4
0 0
4 459 36 12/27/66 0
0 4
0 0
4 463 37 1/ 3/67 0
0 4
0 0
4 467 38 1/ 5/67 0
0 4
0 0
4 471 39 1/ 8/67 0
0 4
0 0
4 475 40 1/10/67 0
0 4
0 0
4 479 41 1/15/67 0
0 4
0 0
4 483 42 1/22/67 0
0 1
0 0
1 484 15 7/15/68 0
0 0
24 0
24 508 16 8/ 9/68 0
0 0
24 0
24 532 17 8/29/68 0
0 0
21 0
21 553 18 9/19/68 0
0 0
21 0
21 574 19 10/ 9/68 0
0 0
24 0
24 598 20 10/30/68 0
0 0
24 0
24 622 21 11/20/68 0
0 0
24 0
24 646 22 12/11/68 0
0 0
22 0
22 668
l 11.
Partial Refueling and Fuel Inspection a.
Fif th Partial Refueling To facilitate fuel cleaning and drive removals, all fuel assemblies and control blades were removed from the core i
and stored in the fuel building.
All fuel assemblies were removed from the core. 2/18/68 through 3/1/68. All control blades were removed from the core, 2/23/68 through 2/29/68. Forty drives were removed from the core, 3/6/68 through 4/1/68. Repaired 40 drives reinstalled in core, 3/14/68 through 4/2/68. All control blades reinstalled in core, 4/12/68 through 4/15/68.
All fuel assemblies reinstalled in core, 4/19/68 through 5/9/68. Metal specimens (removed on 3/2/68) were reinstalled in the core on 5/9 and 5/10/68.
1.
New Fuel Characteristics Ninety-six Type VI fuel assemblies were purchased from United Nuclear Corporation of Elmsford, New York. This fuel'is similar in nuclear and thermal hydraulic characteristics to the General Electric Type III-F fuel.
Type VI fuel consists of 84 " normal" assemblies and 12 instrumented assemblies. The " normal" assembly consists of a six by six array of fuel rods containing UO2 with one of the rods being a removable poison rod.containing Gd 02 3 mixed ' with UO. The instrumented assemblies consist 2
of a six by six array without a poison rod. An instru-ment tabe is inse ted in place. of the removable poison rod leaving 35 rods centaining UO2 Thirty-two Type VI assemblies have a peripheral orifice and 64 assemblies have a ' central region orifice. The peripheral orifice has a single hole, 1.49 inches in diame ter.
The central orifice has a single hole, 1.81 inches in diameter.
Critical experiements, using Type VI fuel, were performed at Pawling, New York, laboratory of United Nuclear Corp.
A loading to critical was performed first using only -
unpoisoned assemblies and then with only poisoned assemblies.
A combination of poisoned and unpoisoned assemblies were made critical with the water height near the top of the' fuel for the uniformity check. Eleven assemblies were checked for uniformity with a total spread of 7.2c in reactivity. The critical core configurations e
are shown on the following page.
. X X
X X
X X
X X
X X
2 X
X X X X
X X
X X
X X
X 2
X X
X X X
X X
X X
E X
X X
X X X X
X X
X X
X X
X Unpoisoned Critical Poisoned Critical Uniformity Test Array - 14 Assemblies Array - 18 Assemblies Array 1 - Last Assembly Added 2 - Poisoned Assembly 3 - Substitution Location 2.
Fuel Assembiv Cleanine and Testing a.
Fuel Cleaning Fuel assembly inlet orifice cleaning was initiated on March 4.
Ultrasonic cleaning of the inlet portion of the fuel assemblies s:as attempted in an effort to develop a faster and more efficient method of crud removal. After three days of only partial success with ultrasonic cleaning, it was decided that mechanical cicaning by manually brushing should be initiated in order to complete the job in the allotted outage time.
bbnual brushing began on March 6, and was completed on March 22.
Inlet orifices on all assemblies scheduled for return to the reactor for Cycle VI, with the exception of assemblies PF-10, SA-1, and A-465, were manually brushed and the effectiveness of the cleaning determined by flow testing. A total of 365 essemblies were cleaned, including 163 Type III-B 's, 96 Type III-F's and 106 Type V's.
b.
Fuel Assembiv Flow Testing All but four of the assemblies in the reactor during Cycle V were flow tested in the "as found" or " dirty" condition and all were tested again following clean-ing.
Flow testing of the fuel assemblies was accomplished in the flow test fixture constructed during the 1967 refueling outage.
Data obtained by flow tests of one new Type V assembly with an "E" orifice and one with a "D" orifice during the 1967 outage were used as the
'riteria for cleaning ef fectiveness. A comparison of this dats with data obtained from cleaned assemblies during the present oatage is shown in Figure 3.
. Three new Type VI assemblies with Type I orifices and three with Type II orifices were flow tested to obtain a reference point for future cleaning of Type VI fuel. Data obtained are also shown in Figure 3.
The orifice of the assembly was removed, cleaned, replaced and the assembly was flow tested again to see that it had been cleaned sufficiently.
3.
Gadolinia - Alumina Rod Installation a.
Gadolinia Rod Inspection Gadolinia-urania rods were removed from the fuel assemblies DU-102 and LU-56.
They were replaced with gadolinia-alumina rods.
The removal and replacement scheme was as follows:
ASM Exchanne Gadolinia-urania
~
A B C D
E F
DU-56 storage.
6 Gadolinia-alumina rod (S. N. DS013) to E-5.
5
\\g
$> \\\\q Gadolinia-urania 4
44 rod from B-2 to 3
storage.
Gadoliniaturania N/
2 DU-102 rod from E-5 to B-2.
1 Gadolinia-alumina rod (S. N. DSO41) to E-2.
A visual inspection af the spacer contact elevations
-disclosed no fretting or wear.
. FIGURE 3 180,.
FUEL ASS'EMBLY PRESSURE DROP VS. FLOW TYPE "E" ORITICE 160~
P Type VI Assembly W/
Af ter Mechanical Cleaning 140. Type 1 Orifice (Data (All Types) from 3 Assemblies) 120 mg 100..
g 80 Type > Assembly (New) Data Obtained 1967 E
60 40 20' O
0 1
2 3
4 5
6 7
8 9
10 11 10 13 Assembly Pressure Drop - PSI 180 F"EL ASSEE LY PRESSURE DROP VS. FLOW 160 I"iPE "D" ORIFICE 140 v
Type VI Assembly W/ Type 2 120 Orifice (Data from 3 After Mecbanical Cleanint x
Assemblies)
@ 100 (All Types) 3 80 Type V Assembly (New) Data Obtained
.3 1967 w
60 40 20 0
0 1
2 3
4 5
6 7
8 9
10 11 12 13 Assembly Pressure Drop - PSI
=-
. \\
l b.
Fuel Inspection i
1.
Fuel Capscrew Inspection The Type I fuel was inspected for missing capscrews using the underwater T.V. camera. This check was made prior to scheduling their removal from the core. The results agreed with the inspection performed during the 1967 outage with the exception of one assembly.
It appears that an error was made in recording the data during the 1967 inspection.
The results of the inspection are shown below:
TYPE I FUEL CAPSCREW INSPECTION Assembly Core Location Caoscrews Missins:
A-40 6325 B-2 A-lli 7120 A-2, B-2 A-60 7410 B-2, A-3 A-82 5208 A-3, A-4, B-1, B-2 A-97 6905 A-3 A-6L 5703 A-4 A-49 6724 B-4 2.
General Electric Fuel Inspection General Electric Company conducted inspection work on several selected fuel elements during the months of March and April. The inspection employed television, boroscope, ultrasonic, eddy currents and profilometry methods. The following fuel elements were inspected:
Assembly Type Description G-42 III-F A leak exposuro DU-100 V
A thin clad pilot assembly A-465 I
Highest exposure E-lll III-B A lead exposure DU-46 V
A 1 cad exposure G-3 III-F A leaker i
s l
r Assembly Type Description G-103 III-F A leaker G-102 III-F A leaker G-99 III-F A leaker DU-45 V
Tubing manufactured by G. E.
DU-102 V
Zircaloy spacers G-93 III-F Revised rod cleaning process E-120 III-F Springless spacers E-149 III-B Revised autoclave process The results of the inspection showed the assemblies to be in satisfactory mechanical condition for continued operation with the exception of the known lenkers.
- 12. Generator No. 1,0verhaul, Inspection and Testing The Unit.No. 1 main turbine generator was overhauled, inspected and tested during the fifth partial refueling outage. The armature winding of the stator was inspected and appeared to be in excellent condition. The repair work on the armature inclucad anchoring several loose slot wedges and varnishing portions of the stator end turns. Following the repair work, the armature was meggered and a high potential test performed on the armature winding of the stator. All' tests were satisfactory.
The field rotor and winding was inspected and appeared to be in very good condition except for minor migration out of the block support on the rotor core.
The mica insu'ating blocks were replaced and several retaining wedges in the field winding slots were driven back into place and relocked in position. The field winding was meggered and satisfactory high potential tests were performed.
The gene'rator was reassembled and an air leakage test was performed on the generator casing. The leakage was within the allowable limits. Following the air leakage test, the collector rings were machined, ground, and polished. The final stages of the work was then completed and the startup procedures were initiated.
13.
Inspections a.
Reactor Vessel Inspection On April 16, a visual inspection was made of the reactor
4 l vessel internals with all the fuel and blades removed from the core.
Typical areas observed during the inspections included; the hold down lugs and lifting eyes on the upper grid, the "T" handles on the hold down lugs for the bottom core support plate, the eight baton strips which hold the segments of the thermal shield, the poison sparger ring in the diffuser basket and the thimbles and guide tubes in the bottom of the vessel.
From the conditions observed, no problems exist in the integrity of the vessel internals.
b.
Reactor Flange Inspection i
The reactor flange was dye penetrant and ultrasonically inspected by Pittsburgh Testing Laboratories on May 16, 1968. The tests were conducted the same as those initiated during the 1967 refueling outage inspections, using the same calibration standards and methods employed at that time. Testing was performed under the supervision of the Commonwealth Edison Company's Operational Analysis Department and was witnessed by representatives from the Atomic Energy Commission, Travelers Insurance Company, Babcock and Wilcox Company and the State of Illinois Boiler Inspection Division.
The ultrasonic and dye penetrant examinations were performed at the inside diameter and between the vessel stud holes on the cladding on the face of the flange. The dye penetrant inspection was conducted through two quandrants of the flange, whereas the ultrasonic inspection included an area 3600 around the inside surface of the flange and at 15 various locations between the vessel stud holea. No indications of defects were found.
c.
Reactor Flange Stud Bolts In addition to the weld inspections, ultrasonic tests were conducted on the 56 vessel stud bolts by the Commonwealth Edison Operational Analysis Department. No indications of defects were found. Tests were made on April 2, 1968.
d.
Reactor Thimble Weld Testing Thirty-six of 40 control rod drive housing thimble welds were ultrasonically inspected by Pittsburgh Testing Laboratories following removal of the control rod drives for overhaul during the refueling outage. The inspections were conducted using the ultraronic fixture and reference standard constructed for the 1967 refueling outage inspections. Of the 36 thimble ucids ultrasonically inspected, no indications of discontinuities were found.
Ultrasonic testing of thimble J-welds was initiated during the 1967 refueling outage. A summary of the locations of the
D"*D "D N 3 thimbles inspected during the 1967 and 1968 outages, including locations of the four sleeved thimbles not ultrasonically inspected, is shown in Figure 4.
e.
P rimary System Weld Inspections The 1968 primary system stainless steel piping weld inspection on Dresden Unit No. I was completed during the refueling, maintenance and inspection outage, February 3 through June 2, 1968. This intens ified inspection duplicated the ultrasonic testing of welded areas performed in 1967 with an increase in percentage of large plate pipe welds inspected, and a more extensive use of the dye check technique.
Dye penetrant checking revealed a number of minor surface impe rfec tions. None of these were of significance to nuclear safety and none resulted from operation of the plant. These surface imperfections were largely due to undercu. ting and poor weld finishing techniques, and they a
were all removed by fitfug. None of these imperfections were deep enough for deteetion by ultrasonics. A total of 20 dye check indications were fou e and corrected.
Ultrasonic checking revealed a total of 13 indications in the piping in the range of 37. to 10% wall thickness in depth. A defect indication equal to or greater than 37, of the wall thickness and I" in Icngth, detected by ultrasonic techniques, was the basis for repair or replacement. This standard is described in Section III of the ASM2 Boiler & Pressure Vessel Code, Nuclear Vessels, in paragraph N-322.
One weld in a section of 6" pipe
("B" unloading heat exchanger return line) had a defect indication of 107. of pipe thickness which extended approximately 4" in a longitudinal direction on the inside of the pipe. This same 20' section of pipe had three other defects and was replaced with 304 L stainless. The pipe containing the defects was sent to the General Slectric Company for metallurgical examination.
In addition to replacing the 6" unloading heat exchanger discharge line, an 3" valve (MO-100) on the emergency condenser steam supply line was replaced. This was done because of ultrasonic indications noted in the valva during the 1957 refueling outaga inspections.
The replcccment of this valve was made in accordance to the ASLE prescure vessel codes. Radiographic and ultrasonic tests were conducted on the welds following installation. One srall ultrasonic indication was detected. The indication was located a'.
reld 223 on the body of the talve.
It cpproached 37 of unll thickness near the outer surface of tha valve, with very little detectable length to it.
Since tha discontinuity was well within the allouable code tolerance, no repairs uere necessary.
i FIGURE 4 CONTROI. RCD DRIVE HOUSIPG
_THI:BLE h2LD INSPF.CTIO !S 1967 - 1963 to 0
0 +
l 9
--C h
b b
I 8
g
-f-g g
-l-
+
O 7
-l-0 6
O
-l-
+
5 O
-f-4 k
3 g
O O +
I --
O 2
I 1
_L
_L.
i I
A B
C D
E F
G H
J K
Outcr=e=1c117 " rectea. redre tv - n=ren. 1967
-[- Ultrasonically Inspected, March,1968 Sleeved Drive Supports - Not Tested
} [
b l D
D{
The 6" stainless steel bypass loop in "B" secondary steam generator compartment was also replaced, although no indications of defects were found in this line. The replacement was made because of problems encountered previously with 304 stainless steel in the other three loops. The bypass was replaced with 304 L stainless steel and was radiographic and ultrasonically tested prior to and following installation.
No other significant indications of defects were noted in the rensinder of the piping and equipment inspections. A few defects of 10% of well thickness or less were found on the exterior surface of the pipe. These were all repaired by filing, rewelding and dressing the pipe to full well thickness.
Table 6 summarizes the scope of the inspection as compared to the 1967 inspection.
The integrity of the entire primary system was confirmed by a L205 psig hydrostatic test following replacements and repairs to the primary system piping.
Mr. Janes Long-bucco, of Travelers Insurance Company, witnessed the test on the system.
f.
Dnergency Condenser The emergency condenser was inspected during thy by climbing down the vent.
The interior of the condenser was found in satisfactory condition with no measurable amount of deterioration since the last inspection.
g.
Turbine Crossunder/ Extraction Piping Inspection Two windows were cut in the north and south crossunders to inspect for piping erosion during the February - June refueling and maintenance outage.
There appeared to be little or no change in condition of the pipe from the 1967 inspection.
It should be noted that the unit was in operation only about seven months between inspections. A hole drilled in an eroded area showed remaining metal was 11/16" thick. A ten inch square area in a reducer was padded with stainless steel weld.
A window was also cut in the "D" extraction 16 inch piping which was replaced in 1967. No erosion was observed in either the copper bearing (0.3% to 0.5%) steel or the 21/4%
chrome, 17. molybdenum alloy.
The pipe test segment which was installed in 1967 was removed for inspection. None of the six materials in the test segment exhibited any significant erosion.
Extraction valves on "B" and "C" extractions were inspected through windows cut into extraction lines. All valves appeared to be okay.
, TABLE 6 ULTRASONIC INSPECTION OF PRIMARY SYSTEM PIPING WELDS Walded Plate 1968 Weld Previous Weld Total Welds _
j.aspections Inspections #
Field M
Field Shop Field Shop Risers (16")
56 36 14 9
4 0
Downcomers (16")
50 20 13 5
5 0
Suction Piping (22")
38 32 9
9 9
2 Return Piping (18")
32 22 8
6 5
2 Return Piping (22")
4 4
1 0
4 0
All ultrasonic inspections conducted during 1967.
Number of indications (By U.T.) during 1968 = 0.
Seamless (10" and Under)
- 10/67 Total Pipe Length S ize Welds Inspections Inspections Inspection (% of Pipe ength)_
10 "
12 12 10 8"
29 29 29 97 6"
96 96 126 **
94 4"
60 60 60 68 3"
61 61 61 32 2"
34 29 23 65 1%"
15 10 10 All ultrasonic inspect'cus conducted during 1967, except for 1965-66 inspections of the six-inch bypass loops.
Number of indications (By U.T.) during 1968 = 13.
- All 96 velas were inspected during 1967 30 of the 96 were inspected during 1965 and 1966.
. 14. Tests a.
Shutdown Margin Checks Two shutdown margin checks were performed at the end of Cycle V on February 17.
The results, compared to the beginning of Cycle V, are exhibited in Figure S.
The north edge of the reactor has a shutdawn margin in excess of 1.77., which is greater than it was at the beginning of Cycle V.
Reactor shutdown margin checks were again conducted on May 9, 1968, to demonstrate that the refueled core met license requirements with regard to the " stuck rod" criterion and that the margin is in excess of one percent throughout the core. The reactivity worths of the control rods used during the shuulown margin checks are exhibited in Figure 6.
The margins were found to be in excess of 1.2% on the periphery of the core and in excest of 1.3% in the center of the core.
b.
Fuel Sipping and Leaker Detection On February 10, 1968, eight days af ter the reactor was shutdown for refueling and overhaul, a program of sipping individual fuel assemblies in their Cycle V locations within the reactor vessel was begun. The detection and location of defective fuel assemblies took precedence over all activities involving fuel handling at the reactor. The general order of sipping and number sipped are indicated below. The Type I, SA-1 and PF-10 assemblies, not sipped at the canal, were sipped in the fuel building.
SIPPING AT THE CANAL Number in Core Number Sipped Type I 66 59 Type III-B (Installed BOC 3) 94 94 Type III-B (Installed BOC 4) 96 96 Type III-F (Installed BOC 4) 100 100 Type V (Installed BOC 5) 106 74 TOIAL SIPPED 423 Four defective fuel assemblies were identified by sipping at the canal. The location of the defective assemblies is shown in Figure 7.
A summary of the defective assemblies, including exposure is tabulated on the following page. Three of the four assemblies are powdered, bringing
i the total defective powdered assemblies, to date, to five out of ten.
The ten powdered at semblies were installed in the core at the beginning of Cycle IV (BOC 4), May,1965.
SUML\\RY OF DEFECTIVE FUELS (EOC 5)
Element Cycle V Locatica Exnosure (EOC 5)
Type of Construction G-3 6112 10,740 Pellet G-99 7112 11,252 Powder G-102 6718 11,693 Powder G-103 5908 11,439 Powder Seven Type I assemblies and two special General Electric assemblies were later sipped in the fuel building. The Type I's were not sipped at the reactor because each had one or more bale capscrews missing and it was thought inadvisable to handle them at the core. The special assemblies were sipped in the fuel building at General Electric's request. None of the nine assemblies sipped indicated leakage.
c.
Critical Testing of Refueled Core The whole core was critically tested on May 9, 1968. The control rod withdrawal sequence, SC, was used to attain a critical state after the partial withdrawal of the 17th rod.
Criticals attained during this and previous refuelings are summarized in the following tabulation.
INITIAL CRITICAL AND REFUELING DATA SUINARY Fuel Critical Core Fraction Fuel Type Minimum Critica:
Cycle Date Rods Withdrwn Refueled Added Size 1
3/30/60 18 1/2 5/5*
I 28 2
3/ 8/63 36 1/2 2/5 I
II 28 3
6/ 7/64 43 1/2 1/5 III-B 24 4
4/27/65 20 1/2 2/5 III-B - III-F 16 5
2/20/67 30 1/2 1/5 V
24 6
5/ 9/68 17 1/5 VI Unpoisoned 14 (Ins trun ented)
- ' Initial loading - 448 assemblies i
. FIGURE 5 END OF CYCLE V SHUTDOWN MARGIN CHECKS
- 1. 2 57. Ma rg in ABCvL F G' H J K to X
l IX 9
X X
IX e8 10_
4/1/67 gh
]
H 286 71 (BOC 5) 7 l
l 6
4 l
@8 3
l 10 2/17/68 2
l M
1.4 1.63 (EOC 5)
Rod Withdrawn
- 1. 77. Ma rc in ABCD E F'r, H J K 10 f
_lg O
h8 10 4/2/67 8
@l_
l
_l_l
_ _ _l_g M 24 6.7 (BOC 5) 7 l
c l
~-
l_I_l l_ -l- -
5 l
_ _ _l_l_ _ _ _ _
l 4
l_
b8 10 2/17/68 2
M 1.2 1.14 (EOC 5)
I l
@ Rod Withdrawn h = Chamber BOC 5 Beginning of Cycle V
=
Multiplication EOC 5 M
End of Cycle V
=
=
. FICURE 6 l
S11UTDOWN MARGIN CHECKS 5/9/68 1
2 1.37. Margin 1.37. Margin 1
ABCDEFGHJK ABCDEFGHJK 10 x
10 9
)
9 8
X X
S f
7 7
6 6
x j
5 5
x 4
4 3
3 2
2 l
1 M 8 = 1.03 M 8 = +1.00 M10 = 1.00 M10 = 1.00 M A = 1.13 MA=
1.00 M B = 2.5 11 B = 1.00 l
I 3
4 1.27. !!argin 1.27. Margin ABCDEFGHJK ABCDEFGHJK to x
10 Q
9 l'
9 x
x i
8 8
X @
7 7
6 6
X 5
5 4
4 X
3 3
2 2
I I
Indicates Rod M 8 = 1.00 M 8 = 1.04 X
=
M10 = 1.00 Full Out M10 = 1.11 M A = 1.17 M A = 1.06 M B = 1.91 h =
In-Vessel Fission M B = 1.21 Chamber Locations
=
. FIGURE 6A SHUTDOWN MARGIN CHECKS 5/9/68 5
1.27. Margin ABCDEFGHJK 10 Q
9 8
x Indicates Rod Full Out
=
h = In-Vessel Fission Chamber 6
Locations 5
@ = In-vessel Fission Chamber 4
l Locations 3
2 x
x l
(B) lT M 8 = 1.00 M10 = 1.00 M A = 1.55 M B = 2.54 6
1.27. Margin ADCDEFGHJK 10 o
I s@
_L x
7 6
x 5
4 x
2 i
M 8 = 1.00 M10 = 1.11 M A = 1.06 M B = 1.21
, FLCURE 7 DEFECIIVE ASSJ.1mLY LC_q\\ TION _(EOC 51 26 l
25 24 23 10 22 21 9
20 19 8
Q2 18 17 y
16 15 6-14 13 5
u b
12 3
99 1 4 --
10 5.
08
.'03 07 2
06 05 t
04 L
03 02 I
_Q1 k
l A
B C
D E
F G
H J
K 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74
- Powdered Fuel
s d.
Control Rod Calibration Immediately following initial criticality, a control rod calibration was performed on control rod B-7.
The reactivity worth of rod B-7 was determined to be.37.
K.
The integral worth curve and the state conditions are exhibited in Figure 8.
The procedure followed and the basic data obtained during the calibration is exhibited in Table 7.
e.
Temperature Coefficient of Reactivity Measurements The moderator temperature coefficient of reactivity as a function of moderator temperature was measured during the initial startup, May 30 - June 1, and is exhibited in Figure 9.
The rod patterns associated with these measure-ments are exhibited in Figure la The temperature coefficient is negative at all temperatures of interest.
f.
Control Rod Zones of Influence Tests Control rod zones of influence tests were performed June 5
- 1968, on two control rods, E-10 in the periphery, and D-6 in the central region of the core. These tests were performed to demonstrate compliance with AEC regulations, which specify that if two adjacent control rods are withdrawn to demonstrate a flux distortion, this distortion must be detected by at least two incore monitors. The test results showed that there is a response from at least the four incore monitors nearest the withdrawn rod for any single control rod movement.
g.
Primary Steam Drum Safety Valves The five spare primary safety valves were installed on the primary steam drum on February 24, thus replacing five of the safety valves previously in service. All five safety valves were tested for relief pressure on August 14, 1967, and were set at their respective design pressures + 10 psi.
Relief pressures were checked by repeated popping. The valves installed were cleaned, set and leak-checked in the shop facility.
h.
Air Locks All air locks, ventilating valves, and process isolation valves were tested periodically during the year and found to be within the licensed allowable leakage rate,
- i. General Electric Extraction Steam Quality Test During August, General Electric Company installed test equipment and performed tests to detarmine the feasibility r
[
. 1 of using radioactive tracers to calculate flow through a
?
turbine extraction line. The extraction flow tests were conducted on August 19 through August 23, on "A" and "E" extraction lines. The flow was determined by injecting at sodium-24 tracer of known concentration into the extraction line, sampling it down stream, and subsequently, j
conducting a tracer balance from which wet steam quality and heater heat balances could be calculated.
Preliminary results of the tests compare favorably with previous turbine performance tests.
The results show that the method is feasible and General Electric has indicated they will continue work in this area.
1 i
e f
i f
I c
ea%
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IMAGE EVALUATION
\\
TEST TARGET (MT-3) 1.0 l2 a UM
> = m p=2 ly p6 b bU l}2.0 l-l L
!!M l.25 1.4 1.6 6"
=
MICROCOPY RESOLUTION TEST CHART
- ~o%e.
sp is+
b xxxxx e,
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IMAGE EVALUATION NN i
TEST TARGET (MT-3) 1 j
F 'le;2 m
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i 1.0 HH 2n 1.l
[* ll!lI14 I
l.8
==
1.25 1.4 i.6 4
6" MICROCOPY RESOLUTION TEST CHART 4
. s,> %s/
+6 5
9
+
s 4"f,s%'. Q c, a
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s~ o, p's v u;p u
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. FIGURZ 8 CONTROL RCD CALIBRATION Rod Calibrated B-7, 5/30/68 8 Period measurement from chart
.30 O Period measurement from stop watch
.28
.26
- 24 A e c D E F G H 5 N to
.22 e
K g
i
~
X Xl X X1
=
5 X sX XXV'X R
XX XIXX X 1X
.18 e
s 7
X XX'X XIX;XX
=
6 4 XX X
lX S X 2iX e
X'X XIXX IX
=
7
.14 XXX 6X 4 Im s
=
XXXX a
.2
.10
.08 b
Control rod inserted
.06 Control rod pulled
.04 Rods involved in measurement see Figure
.02 (S - 1, 2, 3, 4, 5, 6, 7) f
- i" 0
I I
I I
I I
I t
I 1
2 3
4 5
6 7
8 9
10 11 12 NorrCH POSITION 1
+
4-
, TABLE 7 CONTROL ROD CALIBRATION Rod Calibrated B-7, 5/30/68 Control Rod Position Period Reactivity Addad Stop Stop Ster B-7 J-6 J-4 H-3 G-2 G-4 E-2 D-? thtch Ch a r1: k'a r c h c%-e 0
0 12 12 l'i 12 12 12 12 1
1 12 12 135 176
.04977
.03988 2
1 2
1;!
l
.0 1
3 2
2 13 160 163
.0424
.04256 4
2 0
4 l
5 3
0 4
y 65 92
.08730
.06748 6
3 0
6 7
4 6
y y
280 230
.0.J 7
.03164 8
4 y
0 0
4 9
5 4
173 126
.04047
.05265 10 5
0 0
2 11 8
0 0
2 y
118 101
.05551
.06278 12 8
0 1
0 3
13 12 y
y y
y 1
0 3
SS 0 FIGURE 9 MODERI. TOR TEMPERATURE CGEFFICIENT OF REACTIVITY
+6.0
+5.0
+4.0
+3.0
+2.0
+1.0 0.0 C
-1.0
~
w h
-2.0 o
M
-3.0 32
-4.0
,ee wu
~5.0 5
W 8
-6.0 e
u e
e
-7.0 4
nee
-8.0 Ie*
-9.0
-10.0
-11.0
-12.0
-13.0
-14.0 I
I 1
I t
t 0
100 200 300 400 500 600 Temperature OF
, FIGL'AE 10 Moderator Temperature Coef ficient Tests ABCDEFGHJK APC DEFGHJK 10 to 9
x 9
x 8
x x
x 8
x x
7 7
2 x
s y
6 x
x 6
x x
x 5
x l
x 5
x x
4 4
x X
Y v
i V
x 3
3 x
y x
.x 2
xl 2
x x
v i
l I
l l
5/30/68 5/31/68 90 F 2500F ABCDEFGHJK ABC DEFGHJK 10 10 l
l 9
l v'
9 x
x x
L S
x x
x 8
x x
- (
7 x
x 7
1 x
x l
6 x
x x
6 x
x x
5 x
1 x
5 x
x 4
x x
x x
4 x
x x
xl 3
x y
3 x
x x
2 x
x 2
x l
I l
5/31/68 6/1/68 327 F 380 F 3320F h nods fully uithdraim h Notches withdraim Rods fully inserted
, B.
License DPR-2 Table 8 lists the amendments to our license requested and/or authorized during the year. Pertinent correspondence pertaining to these requests are listed in the Correspondence References.
l
, TABLE 8 SUf4!ARY OF LICENSE AMEND?"2NTS PENDING DURING 1968 Date Request Authorization Request to amend License DPR-2 to permit operation with 96 Type VI fuel assemblies.
(Change No. 14) 9/14/67 4/22/68 Request to amend License DPR-2 to authorize elimination of the cocked control rod during fuel assembly additions and impose a condition requiring withdrawal and reinsertion of a control rod in the vicinity of the core position being refueled before and af ter each fuel assembly is inserted into the core.
(Change No. 15) 1/1//68 5/10/58 Request to amend License DPR-2 to authorize replacement of one of the gadolinia-urania poison fuel rods with unirradiated gadolinia-alumina poison rods in two of the Type V high gadolinia fuel assemblics. (Change No. 16) 5/ 9/68 5/17/68 Request to amend License DPR-2 to permit the installation of the Emergency Core Cooling System.
(Change No. 17) 10/10/6S
1-
,4. I i-
- Correspondence References - 1968
{
(1) Letter to AEC dated January 17, 1968, subritting a'dditional information j
and analysis pertinent to Ciange No. 14.
i j
(2) ' Letter to Commonwealth Edison Company dated April 22,196S, authorizing j
Change No. 14 to the Operating License DPR-2.
(3) Letter to AEC dated January 17,1968, spplying for an amendment to
}
authorize elimination of - the cocked rod during fuel assemble additions, j
(Change No.15) 1 l
(4) Letter to Commonwealth Edisnn Company dated May 10, 1968, authorizing j
Change No.15 to the Opera t.ing License DPR-2.
(5) Letter to AEC dated May 9,1968, requesting authorizatio, to replace l
one gadolinia-urania rod from ecch of two exposed Troc V high.;cdolinia i
assemblics with unirradiated gadolinia'-alumina rods.
(Change No. 16) 1 ij (G Letter to Commonwealth Edison Company dated May 17, 1968, authorizing j
Change No. 16 to the Opera ting License DPR-2.
}
(7) Letter to. Commonwealth Edison Company dated May 29, 196G, regarding
{
non-compliance with Sections D5h and E3 of. Appendix A to the Operating License DPR-2.
I j
(8) Letter to AEC dated June 15, 1968, concerning facts of non-compliance j
with Sections D5b and E3 of Appendix A to the Operating License DPR-2.
4
)
(9) Letter to AEC dated October 10, 1958, requesting amendment to Licence
{
DPR-2 to permit the installation of the Emergency Core Cooling System.
]
(Change No.17) i.
t
,i 1
i I
b i
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