ML20037A890
| ML20037A890 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/26/1971 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 8008070709 | |
| Download: ML20037A890 (27) | |
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UNIT #1 ANNUAL REPORT YEAR, 1970
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a; DRESDEN NUCLEAR POWER STATION CommonweaHh Edison Company EETURN"OBEGU.ATBRYDElihhI$$
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C01210NUEALTH EDISON Col!PA!W DoFSDEN NUCLEAR POWER STATICN UNIT #1 ANNUAL REPORT OF STATION OPERATION I
FOR THE YEAR 1970 i
January 26, 1971
4 TABLE OF CCNTENTS Page I.
Introduction 1
II.
Summary of Operations 1
A.
Scope of Operations 1
B.
Shutdowns 1
C.
Load Restrictions 5
III.
Discussion 5
A.
Operating Experience 5
1.
Genera tion 5
2.
Scrams 5
3.
Incidents 8
4.
Control Rod Drives 12 a.
Control Rod Drive Operation 12 h.
Control Rod Drive Te.ats 13 c.
Control Rod Drive Inspection 13 5.
Control Rod Blades 14 a.
Blade Follersing Chacks 14 6.
Changes in racility Design 14 a.
Nhterial Tes; Loop Installation 14 b.
Experimental Resin Cleaning Device 14 c.
A. D. S. Ins talla tion 14 7.
Persoanci Radia tion Exposure.
16 6.
Liquid Poison System 16 9.
Radioactive Uaste Disposal 16 10.
Inspections 16 a.
'tradiated Fuci Inspection 161 11.
Changes in Opera ting Procedures 21 12.
Tests 21 a.
Sphere Integrity Test Program 21 b.
Primary Steam Drum Safety Valves 21 c.
Temperature Coefficient or Reactivity Check 23 B.
License DpR-2 23 l
-i-t
. DRESDEN NUCLEAR POCER STATION ANNUAL REPORT I.
Introduction This ninth 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 Unit #1 during the year 1970.
II.
Summary of Operatienji A.
Scope of Operations Dresden Unit #1 operated during 1970 base loaded at various loads through-out the cycle, and experienced a total of 11 shutdowns. Fhjor work items during the year included: Overhaul of four control rod drives; tube leak repairs on "A" and "B" secondary steam generators; feedwater heater tube leak repairs; repairs to MO-101, emergency condenser condensate return valve and inspection of a crack in the 4" decontamination stub tube on "A" secondary steam generator.
During the year, additions to and changes in facility design were made by:
Installation of a macerial test loop in "B" secondary steam generator com-partment; installation of Automatic Dispatch System (A.D.S.) equipment; and installation of an experimental resin cleaner.
Two irradiated fuel shipments were made to General Electric Company, Val-lecitos, California, for analysis during the year. Ore shipment consisted of one poison rod and sixteen fuel rods from several different assemblies.
The other consisted of thirty-nine fuel-rods from special assembly SA-1.
The remaining ten fuel rods from this assembly were previously shipped in 1967 following the fourth partial refueling outage..
B.
Shutdowns The plant was shutdown 11 times during 1970 as shown in Table 1 and Figure 1.
Seven.of the shutdowns were forced; three were caused by operating errors, two by loss of condenser vacuum and two by primary system steam leaks. One forced outage was carried over from 1969 and represents a forced extension of _the 1969 refueling outage. There were four_sch,eduled outages; cne for feedwater heater tube leak repair, one for turbine overspeed tests, ene for scram and friction-tests and the last for turbine crocs under piping repair.
The total unit outage time, as a result of the eleven shutdowns, was. 641 hours0.00742 days <br />0.178 hours <br />0.00106 weeks <br />2.439005e-4 months <br /> 25 minutes.
FIGUiG
-1 PLANT ELECTRICAL L O ADING YEAR 1970 DRESDEN NUCLEAR POWER STATION UMIT 21 220.
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HPLE I OPEPATIfC PERFOPl!ANCE - 1970 No. cf Off Svstem On Syste_m, Ou:ame Date Tirr.e Date Time-Outane llours Outase Cause 109 1/1/70 1227 12 hrs. 27 min.
Forced turbine (12 hrs. 27 min.), turbine thrust bearing instrumentation malfunction.
Carry over of the sixth partial refueling outage.
110 1/1/70 2302 1/2/70
_0626 7 hrs. 24 min.
Scheduled Turbine (7 hrs. 24 min.).
Turbine overspeed tests.
111 1/19/70 0613 1/19/70 1615 10 hrs. 2 min.
Forced Turbine (10 hrs. 2 min.1 Reactor scram due to low condenser vacuum.
112 4/16/70 0940 4/26/70 1308 242 hrs.
28 min.
Forced Plant (10 hrs. 20 min.). Instrument mechanic error, scheduled turbine (180 hrs.
00 min.) feedwater heater tube leak repair, and forced reactor (41 hrs. 25 min.) isolation con-denser valve repair. Forced turbine (10 hrs, d,
43 min.) reactor scram due to loss of conden-ser vacuum.
113 5/23/70 0500 5/24/70 1423 33 hrs. 23 min.
Scheduled turbine (17 hrs. 00 min.) feedwater heater and turbine governor maintenance, turbine overspeed tests.
Scheduled reactor (16 hrs. 23 min.), reactor instrumentation flushing.
114 6/19/70 1658 6/20/70 1822 25 hrs. 24 min.
Forced Plant (25 hrs. 24 min.). operator error.
115 7/11/70 0352 7/12/70 0717 27 hrs.
25 min.
Forcad reactor (27 hrs. 25 minl primary system steam leak repair.
2
No. of Off System On System Outage Date Time Da te Time Outane Hours Outage Causes 116' 8/5/7L 1020 8/5/70 1816 7 hrs. 56 min.
Forced Plant (7 hrs. 56 min.1 Operator Error 117 8/28/70 0737 9/1/70 1429 102 hrs.
52 min Forced Turbine'(21 hrs. 28 min.), H.P. Turbine casing Icak. Forced reactor (81 hrs. 24 min.)
emergency condenser valve MO-101 repair.
118 10/8/70 2309 10/12/70 0824 81 hrs. 15 min.
Scheduled reactor (50 hrs. 51 min.), Control Rod Drive Scram and friction testing..
Forced Reactor (30 hrs. 24 min.). Control Rod Prive Accumulator repair.
119 11/5/70 2108 11/9/70 0236 77 hrs.
28 min.
Scheduled turbine (77 hrs. 28 min.), turbine cross under piping steam leak repair 120 11/23/70 2203 11/24/70 1124 13 hrs.
21 min.
Forced Plant (13 hrs. 21 min.). Loss of circula-ting water flow.
Total Outage Time 641 hrs. 25 min.
{-
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Load Restrictions i
The load restrictions imposed during the year are listed in table 2.
Restrictions wera due to incore calibrations,' rod worth checks, feed-water heater tube leaks, secondary steam genera tor tube leaks, circu-i lating water temperature and of f gas activity.
III. Discussion A.
Operating Experience i
1.
Gene ra t ion The total reactor operating (critical) time during.the year was 8262.15
+
hours and the total power for the period was 198,835.0 MJDt.
The gross electrical generation during the year was 1,4'2 8, 32 6.34 ' Mm e"; net gener-a tion was l',360,438.30 Eme. As of December 31, 1970, the total gross.
generation since cor.nencement of power operation on April 15, 1960, uns
. 10,723,232. 77 MJHe.
J 2.
Scrams On January 19, at 6:13 a.m., a reactor scram was initiated by o.
low' condenser vacuum.
The unit uas operating at 170 MJe.
"C" condensate. demineralizer was being readied for service; ard following established procedure, the demineralizer was blown down to radwaste.
tihen the blowdown to radwaste'uas started, the make-up detrand was increased, which caused the tr.ake-up valve to go wide open. This i
caused air to t.e drawn into the condenser through the overflow pipe.
on the condensate storage tank in.the turbine rocm, resulting in loss of vacuum. The reactor scrammed due to low condenser vacuum at 6:13 a.m.
b.
On April 16, at ~ 40 a.m.,
the reactor scrammed.
The un ' t ua s opa ra t ing a t 200 m e.
The instrument mechanics cor?
perfortring calibration on the incores.
4 Incore 103A had b_ en suitched to the " calibrate" raoda and uas la the process of I-cing calibrated. ilhile in cl.ia conrigua tion,'2he l
I. M. also.euitched incore 133A to the " calibrate" mode, miscahing it for incore 104A. This combination of 10JA and 103/. in the "cali-bra;e". node is sufficient to scram the reacter, and did, so, c.
%t 1:25 a.m.,
on.i ril 26, el e reac tor scravried due to lou p
r condensar vacuum.
i The re.acioc ra 4 ic start :fo{. a.d hea tig : cioc to plaa.. o m.op.
'2actat ut.pera t urc. ra s 3 70M'.
The teactor 9eresi.ed.'ua.o Ic vacu o /.ie a he seri ice unam ieci rer.ulctor relief valve stuch opea, resulting in Icsa of steam seal pressure ard turbine seats.
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. TABLE 2 L_OAD RESTRICTIONS FOR 1970 Reduction From Maximum Date Capability of 210 MW Condition 1
January 1-160 Incore Calibration January 2-January 7 125 Incore Calibration January 7 - January 9 80 Incore Calibration January 9 - Janua y 11' 50 Incore Calibration February 1 - February 2 50 Rod Worth Checks February 21-February 27 50 "B" Secondary Steam G merator Tube Leak.
February 23-March 1 50 Rod Worth Checks March 2 - March 6 10 "B" and "C" Feedwa ter IMa ter tu' e leaks.
March 7 - !! arch 13 50 "A". Secondary Steam C' aerator Tube Leak.
March 14-April 3 10 "B" and "C" Feeduater heater rule Leak.
April 3 - April 6 50 Rod Worth Checks April 7 - April 9 10 "B" and "C" Feedwa ter Hea ter Tube Leak.
April 9 50 Rod Worth Checks April 26-May 1 30 Incore Calibration May 1-May 8 18 "D" and "E" primary heater tube laak.
May 18 - May 23 18 "D" & "3" prima r" hea ter tube lea':.
May 24-May 29 IS "D" & "E" primary hea ter tube lea't iia, 29 - !!ay 31 50 nod Uorth Checks
!:ay 31 - June 7 13 "D" & "C" primary laater tuhe Ica';.
June 13 - June 15 50 7od north Chnchr, m-r+
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-1 Reduction from Maxirium Date Capab ility of 210 iff Conditica July 1 - July 24 10 circulating Unter Temperature July 25-July 27 50 Rod Uorth Checks July 27 - July 28 30 Incore Calibration July 28-Augus t 3 10 circulating Uater Temperature August 3 - August 11 50 Off Gas Activity Au;ust ll-August l '+
35 Off Gas Activity August 14 - August 20 25 Off Gas Activity Augus t 29 - Sep tember 1 25 Off Gas Activity September 1 - September 10 35 Off Gas activity September 18 - September 20 50 Rod Worth Checks September 20 - October 8 35 Off Gas Activity October 12 - October 23 35 Off Gas Activity October 23 - Cetober 25 50 Rod North Chec!'cs October 26 - NovenNer 5 35 Off Gas Activity November 9 - November 15 35 Off Gas Activity November 15 - Mcvember 23 60 Off Gas Activity November 24 - December 1 90 Off Gas Activity December 1 - December 11 70 Off Gas Activity December ll-Decen'2er 22 60 Off Gas Activity December 22 - Decesser 31 70 Orf Gas Activity
8-d.
On bby 23, 1970, a t 5:01 a.n.,
the unit was taken out of service for turbine overspeed tests and minor repairs. Two scrams occurred after the turbine was off the line.
The overspeed tests were completed and mi..or adjustments of the governor high speed were in progress. The reactor uns at 1000 psig with one bypass valve open. The drum level was slouly decreasing.
- The operator manually opened the feedwater valve to restore Jrum level, resulting in a increase in ' flow. The addition of relatively cool water caused nigh flux on channels 3 and 6 resulting in a scram at 7:12 a.m.
Follouing the scram the reactor was brought critical and heating was in progress. The reactor scrammed at 8:14 a.m. due to lou conden-ser vacuum.
Investigation as to the cause of the low vacuum revealed that the turbine steam seal regulator was binding, resulting in a loss o r steam seal preasure. Maintenance adjusted the regulator and the vacuum was restored, e.
At 4: 58 p.m., on June 19, 1970, the reactor scrammed because of an operator error.
The Unit was operating at 205 MWe.
Monthly checks of the reactor low level sensors were being conducted by draining the level sensors one at a time. The B reactor safety system was in a tripped condi-tion at the time of the incident as a result of the surveillance test.
The operator attempted to reset the safety relays using the reset c ontrol switch. In error, he turned the safety system FC set switch, which was in close proximity to the reset control switch, from A safety system to back-up power, causing a momentary interuption of power to the A system which resulted in a trip on the A system and a scram, f.
On November 23, at 10:03 p.m., the reactor scranmed on low con-denser vacuum.
The unit was operating at 150 MRe.under normal conditions. Sub-freez-ing temperatures caused icing of the intake canal and traveling screens.
Icing of the screen progressed rapidly to a point where virtually all of the circulating water flow was cut off.
Loss of condenser cooling flow caused the vacuum to decrease rapidly initiating a reactor scram on low cendenser vacuum.
3.
Incidents a.
Diesel Generator cooling water hose leak.
J At 0102, on April 10, 1970, during a diesel generator bi-weekly test a leak was observed in the cooling water pump discharge hese.
Inspection and repair required that the diesel radiator be drained.
Plant shutdown was initiated and load had been reduced 15 INe at which time it was determined that the repair could be completed in less time than required to shut the unit down. A temporary leak
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. repair was completed at 1410 on April 10 and the diesel generator was returned to service.
The cooling water pump discharge hose was replaced during a maintenance outage on April 20, 1970.
b.
Reactor Low Water Level scram switch failure.
During the day shif t on April 17, 1970, with the reactor shutdown for' a minor maintenance outage, the Operating Department ran a safety system check normal to an extended outage of the unit.
Three of the four reactor water low level magnatrol switches, LSL-1,2,& 3 failed to operate properly.
A flushing of the sensor float chambers was done on the middle shif t.the same day, and a recheck of the four level switches took place; all worked properly.
The proper operation of these switches af ter flushing indicated that a crud buildup sufficient to impair the switch operation had taken place. Over the years these switches have been flushed to insure their operability and to reduce radiation levels. During operation due to the design of the float chamber, a small flow of reactor water is present. Over a ten year period it is possible that clearances have been reduced so that a major upset and introducticn of reactor crud might cause a reduction of operability of these switches. The test crew reported that any vibration would have caused switch operation, however, artificial vibration is not part of the test procedure.
On April 15, a full power scram took place which raised the water turbidity and could have in the next two days caused enough crud to settle out in these switches to impair operation. This, of course is conjecture and the condition of these switches daring the previous three months of steady operation cannot be known.
As a result of the April' 17, 1970 failures, a program of increased surveil-lance and flushing was instituted for the low level switches. Two failures were detected under the new program.
On August 28,1970 LSL-3 failed to operate following flushing and draining, while LSL 1,2, and 4 functioned properly. The switch for LSL-3 vas replaced and LSL-3 retested satisfactorily.
Again, on October 10, 1970 during an outage for control rod drive frictien testing, all low level sensors were tested and LSL-2 failed to cperate pro -
perly until adjustments were made by an instrument mechanic.
During the 1971 outage, the station plans to replace the present reacter low level switches with a differential type switch similar to those used on Units 2 and 3.
This type will eliminate the radiation and crud depos-ition problem as rhara is no flow through this type of switch, c.
Failure of UO-101, 2merscacj Condenser condensate valve (Three 'a tlurec during 1970).
1.
Unit 11 uas shutdcun for maintenance on April 16, 1970. On April
- 23, 1970, the safety system check-off sheets were being completed
' n preparation for startup.. The reactor mode switch was in refuel i
with all rods inserted. During the operational check of the emergency condenser, it was found that MO-101, the north emergency condenser condensate valve, failed to pen. The valve was satisfactorily opened i
and closed manually indicating the 125 V DC motor had failed. The motor was removed and disassembled, and inspection indicated signs of overheating. This motor is located in the primary steam drum com-partment and is exposed to elevated temperatures. The-motor on the south emergency. condenser condensate valve, MD-109 uns removed for i
inspection and found to be in good condition. FQ-101 was repaired and both motors were returned to service on April 25, 1970. These valves were last checked on December 27, 1969 during our. normal sur-veillance and found to operate normally.
2.
Unit I was shutdown on August 23, 1970 for miscellaneous piping j
repairs.
During a check prior to startup, MO-101 failed to open. The motor was removed and disassembled and inspection again indicated en overheating condition.
Extensive repairs were required and the motor was sent -to a local shop for rett nding.
i 1
Upon completion of the repairs, the motor was given a final check out at which time it was discovered that the interpole phasf,g was reversed.
This would have resulted in overheating if the motor had been put in operation in this condition. This defect was corrected and the motor was returned to service.
The reactor was placed in a shutdown condition during the period the repairs were in progress. The unit was returned to service on September 1, 1970.
2 3.
A failure or the Unit I north emergency condenser condensate valve MO-101 occured on Novenber 7, 1970. This was the third failure of this valve in 1970.
As in the previous cases the failure was discovered while completing check of f sheets in preparation for startup af ter an cutage fer. minor repairs. A spare motor had been ordered from the manu?ceturer and was installed and tested. The unit was returned to service on November 9.
The station plans to take action to avoid this problem bu extending both valve operators outside of the steam drum compartnent so that the motors vill be in an accessible and cooler location. T! is reiccation of the motor cperators will be done during the 1971 norueling Cuca32.
d.
On November 28, 1970, a primary system leak uns discovered in tec-ondary stean gercrator room.
The leak was found to be a crach in a 5" stainless steel nipple.
Normal reacter enclosure air activity sampling routines derected a rise in airborne activity uhich led to the search for and discove~ or -pe crac Visual enamina tion o f the "A" secondary steam genera' or cepa rt reat en Uc 7-ember 20th detected a leakini; flange nasket en motor operatC "alve 111,
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<.ounstean cuction valve on "A" recircula tion pump.
TI.e repair to
c Tbn e gah,ket uas completed on Decem' er 1st and the locp uas pcensurized rc checi the repair. The flange gasket repair uas succcove). However, bid, sir-borne cctivity rersisted l
i
4 and further investigction at this time located the Icaking nipple.
The area of failure on the 4 inch nipple was subjected to visual and ultrasonic examination in the years 1967, 1953, and 1939, ut.th no evidence of discontinuities or cracking. Following the discovery of the crack on December 1st the nipple was dye-penetrant. tested and ultrasonically examired with negative results except for the aforementioned crach. The primary purpose of ultrasonically examining the nipple af ter discovery of the defect was to improva techniques in locating and mapping, defects ultrasonically. Uhen the nipple is ayallable for detailed visual examination it is our intent to compare the actual size and orientation of the crack with our conclusions reached as a result of the ultrascuic examina-tion.
At present, the A recirculatien loop piping is isolated and out of service. We are proceeding to contact piping contractors to effect a repair en the defective nipple. Our intended repair technique is to cut the nipple approximately 3 inches below the crack and re-weld the flange to the nipple. The repair will be radiographed and subjected to hydrostatic, ultrasonic and dye--
penetrant tests to ensure a satisfactory repair.
4 The piece of nipple containing the crack will be subjected to visual and me, '.lurgical examination to determine the mode of fa ilure. Dut
.g the next refueling outage, currently scheduled for March 1971, we intend to perform ultrasonic 'and dye-penetrant tests on three nipples of similar type located in the other secondary loops in addition to our regularly scheduled piping inspection.
4 e.
On Dec. 11, 19 70, a t 7:50 p.m. "B" secondary steam generator waste collector tank was discharged to the river at 121 pei/L. A total of 1,026 gallons were discharged. This d is cha rge wa s te rm-inated en Dec. 12, 1970 at 5:3 5 a.m.
2 Investigation revealed that the 0-5 gpm flow regulator bypass valve was in the open position allowing.a small flow to bypass the flow meter. Apparently, this. valve was erroneously opened some tice prior to the discharge. This valve is now closed and logged out of service.
The inportance of checking the valve line up when discharging a tank to the river was reemphasized to the operators involved.
In addition we are revising the discharge card to include a valve check sheet.
This check sheet will be completed before discharging of any tank to the river, 4
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_ 4 Control Rod Drives a.
Control Rod Drive Operation.
1
- 1) On January 11, 1970, Control Rod Drf e.-14 was found to drift in from position 4 to positic. 0 while it was being withdrawn to increase power.
Flushing and exercising failed to stop the drifting from positions 3 and 4.
The ASCO orifice on the drive's insert header was closed and the drive held its location for one hour. After the orifice was re-opened, the drive immediately drifted to 0.
Appurently the insert ASCO valve wac sticking open allowing water to the bottom of the drive and causing it to drift in.
The ASCO orifice was closed completely until the teflon seat in the ASCO valve was replaced in /
The orifice was re-opened following repairs to the valve and the drive was returned to service.no further drifting was experienced.
- 2) On January 19, 1970, during the rod withdrawal sequence for criticality, control rod drive G-7 continuously drifted "in" from Position 12.
The Asco orifice on the drive's insert header was elosed in s
an attempt to limit flow to the drive. Whaa the driva continued to drift, the withdrawn drives on Accumulator 10 (G-7 and H-2) were inserted and the fuses pulled on the accumulator to scram and flush the seats on the scram inlet and outlet valves. After the scram valve flushing, no further drifting was experienced.
It appears that crud on the seat of the scram inlet valve was allowing leakage past the valve and causing the drive to drift "in".
No further anomalies have been experienced since the manual scram of Accumulator 10.
3)
On. April 24, 1970, during daily rod exercises, control rod D-5 was observed to drif t "in" from Position 1.
Further rod movements indicated that the drive also drif ted in from all other positions.
The drive was flushed at Postion 12 with a continuous withdraw signal for 15 minutes.
Following the flushing, the drive was observed to hold its position at each notch during insert.
The drive was moved to Ponition 1 and held its position for four minutes.
A small amount of crud apparently had become lodged under the seat of the inlet ASCO valve, allowing water to the bottom of the drive.
The flushing operation removed the particle allowing the valve to operate normally. No further anomalies have been experienced since the flushing of the drive.
4.
Control Rod Drives a.
Control Rod Drive Operation 1)
On January 11, 1970, Control Rod Drive J-14 was found to drift in from position 4 to position 0 while it was being withdrawn to increase power. Flushing and exercising failed to stop the drifting from positions 3 and 4.
The ASCO orifice on the drive's insert header was closed and the drive held its location for one hour. After the orifice was re-opened, the drive immediately drifted to 0.
Apparently the insert ASCO valve was sticking open allowing water to the bottom of the drive and causing it to drift in.
The ASCO orifice was closed completely until the teflon seat in the ASCO valve was replaced in April.
The orifice was re-opened following repairs to the valve and the drive was returned to service,no further drif ting was experice.ced.
2)
On January 19, 1970, during the rod withdrawal sequence for criticality, control rod drive G-7 continuously drifted "in" from Position 12.
The Asco orifice on the drive's insert header was elosed in s
an attempt to limit flow to the drive. When the drive continued to drift, the withdrawn drives on Accumulator 10 (G-7 and H-2) were inserted and the fuses pulled on the accumulator to scram and flush the seats on the scram inlet and outlet valves. After the scram valve flushing, no further drifting was experienced.
It appears that crud on the seat of the scram inlet valve was allowing leakage pas't the valve and causing the drive to drift "in".
No further anomalies have been experienced since the manual scram of Accumulator 10.
3)
On. April 24, 1970, during daily rod exercises, control roo D-5 was observed to drift "in" from Position 1.
Further rod movements indicated that the drive also drif ted in from all other positions.
The drive was flushed at Postion 12 with a continuous withdraw signal for 15 minutes.
Following the flushing, the drive was observed to hold its position at each notch during insert.
The drive was moved to Position 1 and held its position for four minutes.
A small amount of crud apparently had become lodged under the seat of the inlet ASCO valve, allowing water to the bottom of the drive.
The flushing operation removed the particle allowing the valve to operate normally. No further anomalies have been experienced since the flushing of the drive.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ b.
Control Rod Drive Tests On April 17 and 18, 1970, all control rod drives were scram and friction tested and timed for normal insertion and with-drawal. The drives were again tested on October 9 and 10,1970, in compliance with the unit license requiring semi-annual control rod drive testing.
The test results were satisfactory.
c.
Control Rod Drive Inspection l
Four drives (1229, 1291, 1253, and 1235) were removed during the 1969 refueling outage and are to be used as spares to replace drives to be removed during the 1971 refueling outage.
Inspection and overhaul of the drives began early in January of 1970. Two drives (1229,1291) were completely overhauled and inspected and need only to be tested prior to installation.during the next refueling outage. Inspecticn and overhaul of Drive 1253 is essen-tially complete except for dye checking of the spud and installation of new roller pins.
The overhaut was terminated due to a lack of spare parts. Drive 1235 was put aside during disassembly when the guide threads called during removal.
1 Following completion of the overhaul's of drives 1235 and 1253, all four drives will be available as replacements for drives removed during the 1971 refueling outage. All drives bill be tested on the station test facility just prior to installation.
As previously, visual and dye penetrant techniques were utilized for drive inspection. The results of the inspection are as follows.
- 1) Nitrided guide roller pins were used to replace all pins in the roller mount assembly. Practically all of the pins removed exhibited some wear. None of the 12 pins removed had failed. No attempt was made to accurately measure the amount of pin wear experienced.
- 2) 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 of an inch was acceptable for roller wear.
Five of the rollers inspected exhibited an excessive amount of wear, and were replaced with new rollers.
- 3) Dye checking revealed cracking or flaking chrome on one guide plug, one collet, and one unlocking spring. Aside from the chrome plating anomalies, no cracking was found in drive components.
- 4) All seals were replaced on every drive overhauled.
In the absence of gross seal wear or breakage, the condition of seals causing long insert and short withdraw times is described as " normal wear" in the report.
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U 5) Thread damage and unremovable scratches necessitated the re-placement of one shuttle piston, one guide plug, one stop piston, and two outer filters.
- 6) Table 3 summarizes the inspection results.
5.
_ Control Rod Blades a.
Blade Folloring Checks During periods of operation, control rod have been verified -for blade follouing on a weekly basis.
During each startup, control rod patterns for criticality have been predicted and all blade following verified.
6.
Changes in Facilitv Design a.
Material Test Loop Installation The materials test loop was placed in service September 15, 1970. The test loop, which was installed to determine quantitatively the effect of BUR primary coolant on the low-cycle fatigue life of piping steels, was installed in 1968. This loop was described in the 1968 and 1969 annual report's.
Since that time, design work and shakedown testing of the test vessels have been in progress. The dynamic and static vessels along with the control consoles were insta11ed during the year s
as was a section of piping which will facilitate the addition of a _ third test vessel in the future. The wiring to the control room was also completed; providing a trouble alarm and isolation valve position indi-cation in the control room.
Following a final hydrostatic test of the system and a preoperational check out of the loop isolation system, the loop was placed in service. The test loop has been in service since its startup in September with the exception of a shutdown for specimen inspec-tion and plant outages.
b.
Experimental Resin Cleaning Device In Janua ry, 1970, an experimental resin'eleaning device was installed in the condensate demineralizer regeneration room by General Electric.
The unit is designed to remove any insoluble materials which may be fooling the resins. The cleaning is done with ultrasonic vibrators.
The test period was to last one year, and therefore final results are not available. However, preliminary results indicate that the cleaner has a greater than 70% removal efficiency. The effectiveness of the cleaner allows operation with much fewer bed regenerations than normal operation without the cleaner.
c.
A. D. S. Ins _talla t ion In April,1970, a new shaf t and bearing was installed on the A.D. S. linkage at the turbine governor motor to replace the shaf t which had seized follow-ing the 1969 refueling outage. The system remains out of service pending completion of preoperational tests.
Table 3 1970 Spare control Rod Drive Overhaul and Inspection Summary Core SN Drive SN Drive Positio2 Removed Installed Abnormal Symptoms of Drive Removed Inspection Results H-7 1229 1276 Drive Stuck at Position Zero Scratched shuttle piston; Chrome cracked on collet assembly; chrome cracked oa guide plug; deep scratch in outer filter.
E-10 1291 1234 a) Long Insert Time Seals broken on stop piston; b) Short Withdraw Time threads damaged on stop piston, c) Short Buffer Time collet assembly, and guide plug; outer filter damaged e
during removal.
{
J-5 1253 1275 a) Long Insert Time Guide rollers worn; chrome b) Short Withdraw Time on collet spring cracked.
J-7 1235 1296 m) Long Insert Time Guide plug galled during b) -Short Withdraw Time removal, drive put aside for shop work. Will be overhauled at a future date.
7.-
Personnel Radiation Exposure Personnel exposures to radiation during 1970 were within limits specified in 10 CFR Part 20.
8.
Liquid Boron System The liquid boron system was operative at all times during the year.
The boron concentration was determined on April 18 and October 10.
There were no conditions which would indicate a loss of boron from the solu-tion tank. Boron concentrations in the reactor water remained below detectable limits throughout the year.
+
9.
Radioactive Waste Disposal Release of radioactive liquid waste was accomplished in batch quantities at-controlled release flow rates according to established procedures.
The contribution to the activity of dilution water was always maintained within the limits specified in the applicable fed-eral regulations.
i The average contribution to the unidentified activity _in, the water utilized for radioactive liquid waste dilution during the year was calculated to be Od35_x_10' _uC/ml ('L3.5 uuCi/,1) compared to an average limit of 1.00 x 10 uC1/ml (100 uuci/1) for unidentified mixtures containing no Ra-226, Ra-228 or I-129 as specified in 10 CFR Part 20.
Solid radioactive wastes were stored on-site pursuant to AEC License DPR-2.
Table 4 shows the radioactivity content, shipment destina-tions, and dates of radioactive vaste shipments made during the year.
Ten radioactive waste shipments were made in 1970.
The concentration of noble f_iss,1,on gases in the chimney discharge to atmosphere was maintained well within license limits.
The-annual average noble gas release rate from the chimney was approximately 03.948 uCi/sec.
There were two shipments of privately owned. spent fuel in 1970. On March 23, 1970, irradiated fuel from special assembly SA-1 was shipped by motor freight to the General Electric Company, Vallecitos Nuclear Center, Pleasonton, California. On April 27, 1970, 16 fual rods from nine irradiated assemblies and one rod containing Gd 0 -A10 23 2 3 were shipped by motor freight to the ceneral Electric Company Vallecitos Nuclear Center, Pleasonton, California. Table 5 is a breakdown of Commonwealth Edison Company spent fuel shipments for reprocessing since initiation of fuel shipping in June, 1965. There were no C.E.Co.
owned spent fuel shipments in 1970.
10.
Inspections a)
Irradiated Fuel Inspection A post-outage fuel inspection on Unit #1 was conducted by Ceneral Electric Company. 16 fuel rods from nine fuel assemblies were removed and shipped to San Jose for further analysis.
The rods and bundles involved are listed in Table 6.
This fuel shipment took place April 27, 1970.
. TABLE 4 RADWASTE SHIPhTNTS Date Destination Activity (mc.)
Volume (ft.3) 1/ 5/70 Nuclear Engineering Company, 87.4 1230.6 Incorporated, Sheffield, Illinois 1/ 6/70 485.07 1492.05 1/13/70 150.07 1201.20 i
5/ 6/70 136.01 1304.55 5/ 7/70 722.52 994.50 9/11/70 114.79 1100.50 4
9/14/70 241.07 876.20 9/15/70 52.97 1044.90 9/21/70 270.00 189.00 9/23/70 300.00 189.00 i
4 3
9 h
,,,r...
..-.c
..-<c.
-.. - -,. -. - - ~, - - -
.m y
t Table 5 COMMONWEALTH EDISON COMPANY SPENT FtIEL S1tIP: TNT StR2tf.RY thimber of Asser-blics or Containers Total Shipment Nu:-he r Date Bateh To Rail Truck Shinned 1
2 3
4 5
Ra il Truck Date 1
6/11/65 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
S/15/66 0
0 4
0 0
4 198 3
8/24/66 0
0 4
0 0
4 202 4
5/28/66 0
0 4
0 0
4 206 9
8/31/66 0
0 12 8
0 20 225 5
9/ 5/66 0
0 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
'O O
4 0
0 4
322 18 10/23/66 0
0 4
0 0
4 326 19 10/26/66 0
0 4
0 0
4 330 12 10/2C/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/ 6/65 0
0 4
0 0
4 361 23 11/ 6/66 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 I4 0
0 4
40S 29 11/27/66 0
0 4
0 0
4 412 30 11/29/66 0
0 4
0 0
4 416
. TABLE 5 (Cont.)
k T SNd7 SP U
11 Number of Assemblies or Containers Total Shipment Number Date Batch To Rail Truck Shipped 2 3 4
5 6
7 Rail Truck Date 14 12/ 2/66 4
19 0
0 0
23 439 31 12/ 6/66 4
0 0
0 0
4 443 J2 12/11/66 4
0 0
0 0
4 447 33 12/15/66 4
0 0
0 0
4 451 34 12/18/66 4
0 0
0 0
4 455 35 12/20/66 4
0 0
0 0
4 459 36 12/27/66 4
0 0
0 0
4 463 37 1/ 3/67 4
0 0
0 0
4 467 38 1/ 5/67 4
0 0
0 0
4 471 39 1/ 8/67 4'
O O
O O
4 475 40 1/10/67 4
0 0
0 0
4 479 41 1/15/67 4
0 0
0 0
4 483 42 1/22/67 1
0 0
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 23 1/ 3/69 0
0 0
19 0
19 687 24 1/29/69 0
0 24 0
0 24 711 25 2/26/69 8 0 0
3 0
0 11 722
4 4 TABLE 6
-TUEL RODS SHIPPED TO G.E. FOR ANALYSIS 4/27/70 Assembly No.
Number of Rods DU 89 2
DU 82 2
DU 92 3
XE 45 1
E 139 2
G 14 2
DU 36 i
G 20 2*
G 42 2
- One of these was a Gd 0 -A1 07 3 poison rod.
23
_ _ _ _ _ _. 11.
Channes in Operating Procedure Following the failure of three out of four level switches in April,1970, (see Incidents, section III, 3, b.) a temporary surveillance procedure was set up and will be used until the switches can be replaced.
The procedure calls for testing of the switches once a month and flushing of the switches during each outage, but not more than once a month. The procedure will remain in effect until the 1971 refueling outage when the switches will be replaced, i
12.
Tests a.
Sghere Integrity Test Program As part of the sphere integrity test program, the sphere ventilation valves the 16 foot equipment hatch, and the sphere access air locks were leak tested during the year.
Table 7 lists individual and total leakage results founa during test in 19'O.
b.
_P_r ima ry Stean Drum Sa fety Valves Four of the five spare primary steam drum safety valves were overhauled during January,1970. Following the overhaul, the valve " popping" pressures were set and tested using the station nitrogen test facility. The results l
of the tests were all satisfactory. The fif th valve was overhauled and tested satisfactorily in 1969.
l c.
Temp. Coefficient of Reactivity The moderator temperature coefficient of reactivity was checked three times l
during the yerr as the cycle VII exposure increased from 0 to 198,835 MWDt, the latest measurerent being on October 10, 1970.
In figure 2 the temperature l
coe f ficient is plotted versus moderator temperature for the three aforementioned checks.
B. License DPR-2 No amendments to License DPR-2 or changes to the technical specifications were re-quested during 1970. However, supplement A to proposed change #17 (Dresden core cooling system) was submitted on September 17, 1970, t
. TABLE 7 SPHERE LEAKAGE Component
- 7. of Allowable Exhaust valves 2.12 %
Supply valves 5.10 Equipment lock 2 90 Escape lock
.063 Personnel lock 1.66 Frimary steam isolation valves
$.10 Emergency condenser manheads
.053 Transfer tube cover 2 56 16 Foot bolted equipment cover 0.00 TOTAL 48 556 $
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