ML20238A511
| ML20238A511 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom |
| Issue date: | 12/31/1976 |
| From: | Ullrich W PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| References | |
| EFF-76, NUDOCS 8708310105 | |
| Download: ML20238A511 (99) | |
Text
'l (RETS MASTER FILEl
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E Pre-76 PHILADELPHIA ELECTRIC COMPANY' PHILADELPHIA A r
PEACII BnTTOM ATOMIC POW 1:R STATIO:I.
i U:J I T :1 0. 2 A:1D U? LIT NO. 3 ANNUAL OPI:PI. TING RCPORT MO. 1 January 1, 1976 through December 31, 1976 i
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Submitted to j
The United States Nuclear Regulatory Commission Pursuant to racility Operating Licenco :lo.'DPR-44 & 56
.)
8708310105 761231 PDR ADOCK 05000277 R
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PIIILADELPIIIA ELECTRIC COUPN1Y PEACl! BOTTOM ATOMIC POWER STATION.
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UNIT 1:0. 2 and IJNIT NO. 3 l
ANNUAL OPERATING REPORT l
MO. 1 l
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January 1, 1976 through December 31, 1976 l
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l Submitted to The linited States Muclear Reculatory Comaission Pursuant to l
Pacility Operating License No. DPR-44 & 56 i
Preparation Directed by:
W.
T.
Ullrich, Superintendent Peach nottom Atomic Power Station
PBAPS TABLE OF CONTENTS PAGE I
INTRODUCTION 1
I.
OPE RATIONS l
A.
Summary 1
B.
Unit 2 Operations 2
C.
Unit 3 Operations 16 D.
Common Plant Operations 27 E.
Reactor Shutdowns, Power Reductions and Maintenance Summaries 28 II.
MODIFICATIONS 76 III.
PERSONNEL EXPOSURES l
1 l
A.
Personnel Exposure by Job Function 81 D.
Wholobody Exposure 81 l
s C.
Single In-plant Personnel Exposure Due to an Outage which Resulted in 10% of the Annual Limit 81 D.
Single Gaseous Releases, Due to an Outage which Resulted in 10%
of the Annual Limit 81 E.
Liquid Radioactive Release 82
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i.
P3APS TADI,E OF COf!TEUTS 1
_Pt_.i E P.
Isotopic Analysis of Liquid Radioactive. Releases 82 C.
Isotopic Analysis of Gascous Radioactive Effluents 92 Gaseous Radio.=ctive Release Data 82 I.
Solid Radioactive Waste Shipments 82 j
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IP.
IPRADIATED PURL I:X7MINATIONS 90 l
APPENDIX A 3,y l
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PBAPS I
TADLE OF COtITENTS TABLES' PAGF I-A PEACH DOTTOfi UNIT 2 AND 3 3
PERPORt'A!!CE PARAMETEPS I-E-1 UNIT 2 SIIUTDOWNS AtID POWER 29-31 (a) ^thru (c)
REDUCTIONS - January 1, 1976 to December 31, 1976 I-E-2 UNIT 3 SUUTDOWr:S AND POWER 32-36 (a) ' thru (c)
REDUCTIONS - January 1, 1976 to Docenber 31, 1976 I-E-3 PEACII BOTTO11 UtIIT 2 SIGNIFICANT 37-47 l
(a) thru (h) t'AINTENANCE SUf'ItARY - 1976 l
I-U-4 PEACil BOTTO!! UMIT 3 SIGNIFICA!!T 48-54 l
(a) thru (g) t1AINTENANCE SUtil'ARY l
I-E-5 PEACH DOTTet! UNIT 2 t'AINTENANCP 55-66 (a) thru (1)
SUlif'? RY INSTRUtU:NTS & CONTROLS I-E-6 PEACH BOTToft UNIT 3 t'AINTENANCF 67-74 (a) thru (i)
SUt'ttABY INSTRUtfENTS & cot:TROLS III-A STAT!DARD TORitAT FOR REPORTING 83
??Ut1 DER OF PERSONNEL & !!AN-REli POR WrRK & JOB PUtICTION III-B PICORDED ANNUAL UHOLEDODY FOR 84 CALEND7R YEAR 197G III-C LIQUID RADIO?,CTIVE RELEISE DATA 85 III-D ISOTOPIC ANALYSIS OF LIQUID RAD 86 PADIOT.CTIVE RELEASES III-E ISOTOPIC ANALYSIS nr CASEGUS 87 RADIOACTIVE ITTLUENTS III-F GASEDUS HADIDI.CTIVE RELTASP DATI 88 III-C SOLID RADIOACTIVE WASTE SIIIP1T,t:T 89
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PEAPS INTRODUCTION i
This first Annual Operating Report' for Peach Bottom Atomic Power Station is being issued in accordance with Amendments No.
j 12 and 10, issued on November 10, 1975, to License Number DPR-44 and DPR-56 respectively, and is intended to fulfill the annual
]
reporting requirements of those amendments.
Prior to these license amendments, five Semi-Annual Operating Reports were issued in accordance with the licenses in effect at that time.
The material included in this report is intended to meet, but is not necessarily limited to, the Peach Bottom Technical specifications now in effect and the guidance of U.
S.
Nucleas Regulatory Commission Regulatory Guide 1.16, Revision 4, where applicable.
Peach Bottom Atomic Power Station consists of two Boiling Water Nuclear Power Units with each having a licensed capacity of 3293 MWt and is located within the operating territory of the Philadelphia Electric Company.
The facility is owned by, and licensed to, Philadelphia Electric Company, Public Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electric Company.
Philadelphia Electric Company is the factility operator.
I.
OPERATIONS A.
Summary Unit ' experienced 11 forcea outages,. fivi power reductions and one major refueling outage during 1976.
Forced outages were caused by steam leaks in the drywell
( 2), turbine generator /EHC difficulties (5), relief valve malfunctions (2), recombiner difficulties (1), and trouble-sk oting error (1).
The refueling outage duration was approximately three months.
During this outage the feedwater spargers were replaced and an integrated leak rate test was performed.
Unit 3 experienced 17 forced outages, 11 power reductions and two major outages.
Forced outages were 1
PBAPS l
caused by recirculation system vibration (6), primary coolant leaks (2), surveillance testing (2),
recombiner/ vacuum pump difficulties (2), surveillance testing (2), relief valve malfunctions (2), and turbine generator /EHC difficulties (2).
Core plugging in order J
to eliminate LPRM/ channel box interaction was I
accomplished during a one month. outage in' January 1976.
The first refueling outage was started on December 24, 1976 and continued into 1977.
j i
Operating Unit Status Reports for Peach ~ Bottom Atomic Power Station Units 2 and 3 are published on a monthly q
basis by the Nuclear Regulatory Commission, as part of 1
" Operating Unit Status Report" NUREG 0020 Series.
)
i Plant performance data is given in Table I-A, and Peactor Shutdowns, Power Reductions and Maintenance Summaries are tabulated in I-E (a) through (c) and I-E-2 (a) through (e).
i B.
Unit 2 Operations Following the Unit 2 start-up near the end of 1975, Unit
)
2 continued to increase power in accordance with the fuel preconditioning program.
Full load was achieved on January 2.
Load was slowly dropping through January 8, as Xenon built-in and reactivity decreased.
On January 8, a blown fuse in the 125 volt d.c. control system associated with the EHC tripped the turbine and scrammed the reactor.
An investigation that followed identified several grounds on the 24 volt d.c. system, but none on the 125 volt d.c. cystem.
The reactor.was returned to service late on January 8.
Startup was delayed because of the f ailure of the 25B RHR valve on the previous day.
The motor on this valve was replaced prior to startup.
On January 9, the turbine generator was synchronized, but then removed from service when a severe icing condition resulted in breaking of the shear pins on all 3 inner circulating water pump screens.
Following the replacement of the shear pins, the reactor was again increased in power and the unit synchronized.
Only one circulating water pump was operated during the etartup in order to increase the temperature in the discharge pond.
Once a large volume of warmer water was,
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PBAPS accumulated, the cooling towers were placed in service.
Because of the severe weather conditions the-cooling tower fans were not started until January tenth.
The unit continued to increase in. power in accordance with the fuel preconditioning program ac.S was operating at approximately 875 MWe on January 12.
On January 13, a core monitor alarm on the main generator occurred.
The shift responded by immediatley l
reducing load.
Investigation quickly determined that the generator core monitor alarm was false.
Load was recovered to its original value within approximately ten minutes.
Later on January 13, the offgas activity at the air ejector increased significantly.
A second significant increase occurred on the morning of January 14.
Maximum activity levels measured at the air ejector were in the order of 100,000 uCi/sec.
This activity eventually resulted in an increase release to the environment.
on January 15, with the plant operating at essentially full load, maintenance activities were in progress on the 154B valve.
These operations required the 154B valve to be closed.
Surveillance testing on the "A" RHR loop required by this maintenance operation identified an inoperable 25A valve.
This essentially made both RHR loops inoperable.
When this condition was recognized, the 154B valve was immediately opened and the
'B' loop returned to an operable status.
Prior to this occurrence, the four diesel generators and both core spray lorps had been proven operable.
The 25A valve was returned to an operable status at 2:30 p.m. on January 16.
Late on January 17, the pump-out rate from the drywell increased.
Various systems were isolated and valves backseated in an attempt to reduce the unidentified leakage.
These efforts were unsuccessful.
Unit shutdown was initiated at 4:30 a.m. on January 18, and a hot standby condition was achieved by noon of the same day.
A drywell entry succeeded in identifying the leak on the packing of the 54A valve.
This leak was-corrected by backseating the valve.
The unit was returned to power operation with the generator being synchronized at 2:24 p.m.
From January 18 through 23,
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PBAPS plant load was _ increased in accorance with the fuel preconditioning program.
The unit achieved full' power on January 22.-
Of fgas activity decreased to approximately 60,000 uci/sec. at the air ejector.- On January 23, a leak occurred in the -recombiner building which increased the Unit 3 vent stack release rate to approximately 1500 ' uci/sec. - which is approximately 24%
i of MPC.
Total gaseous release increased to approximately 50% of MPC on January 24.
A significant leak on the packing of one of the mechanical compressor discharge' valves in the recombiner was identified and corrected on this date.
This reduced the total release rate to less than 7%.
l On January 25, a chemical upset in.the reactor water l
chemistry following the return of the Reactor Water Cleanup System to service increased the conductivity and l
resulted in a power reduction of approximately 25MWe.
A3 the conductivity decreased, power was slowly returned to its original value.
Unit 2 continued to operate at essentially full power j
through February 6.
Load slowly decreased at 3
approximately 4 MWe per day due to fuel burnup, creating i
I an end of cycle coastdown situation.
On February 6, load was reduced to 475 MWe in order to change contrcl rod sequence.
Power was being increased in accordance with the preconditioning program on February 9.
Late on February 9, load was reduced to approximately 350 MWe in order to repair the RCIC outboard icolation valve MO-16.
Following this repair, power was again increased in accordance with the fuel preconditioning program.
Full 1
load was reached on February 14.
On this date power was reduced to 688 MWe for turbine valve testing.
No problems resulted from the testing and the unit was returned to 1063 MWe by February 16.
Xenon build-in and fuel burnup caused power reduction to 1021 MWe by February 20.
During the weekend of February 21, power was reduced in order to perform additional turbine stop valve and control valve tests.
During this power reduction, the-fifth feedwater heaters were removed from service in l
order to increase reactor power during the end-of-cycle
_ r,_
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PBAPS I
coastdown.
Power was increased to essentially full load by February 26.
Late on February 26, reactor power was reduced by 50 MWe-to permit routine turbine stop valve testing.
Power.was reduced, the stop valves tested and power was returned to its original value within 15 minutes. _ Approximately one hour later reactor power was again reduced, this time by 100 MWe,'the turbine control valves were tested and the power recovered to its original value again within 15 minutes.
During each of these tests, offgas activity at the air ejector showed a significant increase.
Following these tests, the offgas continued to ramp upward at a very slow rate.
This slow increase may actually have started prior to turbine' valve testing.
The increase in offgas activity leveled off on March 2, at approximately 130,000 uCi/sec.
Data indicated that this increased. activity was from fission products being released from elements which had previously failed.
The slow ramp increase was not indicative of additional failures.
On February 26, the drywell on Unit 2 was pressurized to 1 psi and the Standby Gas Treatment System was placed in service to maintain a slight vacuum of 2" to 3" of water in the torus compartment.
This differential pressure i
continues to be maintained in conservative consideration of the Mark I pool swell ef fects.
Gaseous release rate as of March 3 was approximately 19%.
This is a decrease from the 32% total release rate which was experienced on March 2, prior to correcting several leaks in the 2B mechanical compressor.
On March 5, power was reduced to 900 MWe in order to repack the "A" and "C" condensate pump shaft seals thereby reducing the leakage into the Radwaste System..
During the power reduction, the turbine stop. valves were tested full stroke with no apparent increase in offgas ac tivity.
The unit returned to maximum load operation on March 0.
By March 19, load had decreased to approximately 1000MWe gross.
Offgas activity had stabilized at approximately 130,000 uCi/sec.
At approximately 9:00 p.m. on Friday, March 26, load on Unit 2 was reduced preparatory to removing the unit from 6-
PBAPS service for the first refueling outage.
During the-shutdown at approximately 45W power, a special recirculation pump ' trip and restart test was performed in. order to provide data which would indicate the ability to start a rc :1rculation pump following the removal of the 4 inch recirculation pump discharge -valve bypass line.
The test showed that a recirculation pump could be started with the reactor at power without producing excessively high flux spikes.
The teet also showed that a jogging circuit for the pump discharge valve was unnecessary.
Prior to removing the turbine generator from service, various testing was performed on the turbine.
This
' included a turbine overspeed test, as well as a test to determine excessive through leakage of stop valves and y
control valves.
1 Depressurization of the primary coolant system was delayed until 12:30 p.m.,
on Saturday, March 27, in order to minimize the release of gaseous activity from the fuel.
The reactor remained critical until 11:30 p.m.,
in order to control the rate of depressurization.
With the reactor pressure less than 50 psig, the main steam isolation valves were closed and an attempt made to place the 2A RHR heat. exchanger in service.
A leak on this heat exchanger was quickly identified.
This i
leak resulted in an increase in reactor level and conductivity when the heat exchanger was first placed in service.
The heat exchanger was removed from service and isolated.
The delay in placing the RHR system in service permitted the reactor pressure to again increase above 50 lbs., resulting in the automatic isolation of the shutdown cooling suction valves.
In order to depressurize the reactor, the MSIV's were opened.
A l
bypass valve was then opened using the test circuitry.
At this time, there was no vacuum in the condenser and the vacuum breakers were open.. Opening of the bypass valve therefore resulted in a transport of gaseous i
activity from the reactor through the main steam piping, condenser and turbine building to the Unit 2 reactor builiing vent stack.
(LER 2-76-18/1P)
A condenser vacuum was established and the reactor depressurized to - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _. _ _ _.
1 i
1 PEAPS the condenser.
Subsequently the
'D' RHR heat exchanger was placed in service.
1 Once shutdown cooling had been established, the primary coolant was further cooled to less than 2120F.
The reactor was then pressurized using instrument air in order to perform local leak rate tests on the MSIV's.
During this same period, testing of the feedwater check valves was accomplished.
This testing continued through March 30.
Results of these tests indicated that one
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inner MSIV, and all four outer MSIV's had excessive leakage.
All four feedwater check valves also showed unacceptable leakage.
With the completion of the 50 psig testing, reactor pressure vessel head removal operations were started.
The head was removed on April 2.
Removal of the steam separator and dryer was completed by April 3.
Flooding of the dryer separator pool was initiated following replacement of the shielding plugs.
Hydro-lazer operations on the reactor pressure vessel walls was begun in preparation for installation of the feedwater sparger work platform.
This platform was installed and the spargers removed on April 6.
The inspection of the feedwater sparger nozzles showed approximately 1SO indications on the nozzle cladding.
Detailed reports were presented to the NRC, Division of Reactor operations, on May 14 and August 18.
The feedwater sparger work platform was removed and the service platform installed.
An inspection of the jet pump hold-i down nut tack welds was performed using a TV camera.
All welds were intact.
A shutdown margin test was performed on April 9, preliminary to control rod. drive maintenance.
Work then continued on installation of the jet pump plugs as a preliminary requirement for starting removal of the 4 inch bypass piping.
Jet pump plugs were installed in the 'B' jet pumps by late April 15.
The service platform was then removed and the reactor head cavity flooded.
Check-off-lists were completed and fuel movement initiated on April 17.
Fuel relocation and sipping continued from April 17 through May 1.
Fuel sipping identified at least 17 Type II leaking fuel elements.
A pattern of leaking fuel elements was noted which correlated the leaking fuel
_a_
'I 1
PBAPS with control rod manipulations associated with the January 12 return to power.
By May 1, fuel handling had progressed to the point where LPRM replacements could be H
made.
Prior to start of fuel handling, three control rod drives were replaced.
The last of these drives was not latched successfully.
An investigation via TV camera of the control blade associated with this drive unit indicated a latch handle stuck in the up position.
This problem was corrected on May 5 by replacing the control blade with a new blade.
During Unit 2 refueling outage, April 1976, control rod drive serial number 2773 was removed for routine inspection and maintenance.
The CRD collet housing was lignid dye penetrant inspected as recommended by the NSS Suppljer.
An approximate 1" long circumferential crack was found in the collet housing adjacent to the attachment weld to the main housing assembly
(" flange and tube assembly").
It was the opinion of the NSS
)
Supplier that the CRD ct ld be reused if required, but j
that replacement of colle c housing is recommended.
The CRD serial No. 2773 will not be reused, however, until the collet housing can be replaced.
Following the completion of LPRM replacement, the remaining cycle two fuel in the fuel pool was transferred to the core.
Core verification was completed and fuel handling activities terminated on May 7.
Head cavity level was then lowered and hydro-lazer cleaning of the head cavity walls initiated.
This cleaning effort was completed by May 10.
Other major areas of work during the outage include modifications to the torus supports, repairs in the condenser steam space, recoating of the circulating water inlet piping, snubber inspection and rebuilding, inservice inspection of primary coolant welds, reactor vessel stabilizer modification and repair, lapping of various containment valves and a major turbine inspection program.
From May 11 through May 15, the start-up vibration instrumentation in the reactor vessel was removed.
Concurrent with this operation, a significant number of Type I fuel elements were sipped
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PBAPS in order to identify leakers.
This activity was completed on May 12.
In order to support removal of the four inch bypass piping, the jet pump plugs were transferred from the
'B' jet pumps to the
'A' jet pumps.
This work was completed on May 17.
The reactor vessel level was then lowered and the vessel walls decontaminated.
The feedwater sparger work platform was installed and work proceeded to grind out the indications indentified on the feedwater nozzles and install the new feedwater spargers.
Removal of the sparger indications necessitated removal of about 1/16" to 1/8" of base metal in approximately five locations.
Feedwater sparger work activities were completed on May 28.
with the completion of the feedwater sparger work, operations associated with reinstalling the vessel internals were initiated on May 29.
The jet pump plugs were removed and the dryer and separator installed by June 2.
During this operation, the main steam isolation valves were tested.
Seven of the eight MSIV's were found to be acceptable.
The eighth valve was disassembled, reworked, and successfully tested on June 8,
following the reactor vessel hydro.
The reactor vessel head was set on June 2.
Tensioning was completed on June 5.
A hydrostatic test was started on June 7, and completed on June 8.
During the hydro, all control rods were scram tested and all excess flow check valves were checked for operability.
The integrated leak rate test was started early on June 11.
Following pressurization of the containment, several leaks were identified and corrected.
The magnitude of these leaks was sufficient to cause a leak rate in excess of allowable limits.
Following the correction of these leaks, the ILRT was conducted.
Data taken indicated that a negative leak rate or in-leakage was occurring.
A great deal of effort was exerted in trying to identif y the source of in-leakage.
No source was identified.
On June 15, the investigation disclosed that the reactor head was not properly vented to the containment.
Since there was a continuous water transfer from the reactor vessel to the drywell sump, the test data indicated a negative leak rate.
Vent l
PBAPS l
piping external to the containment was installed between instrument lines from the reactor to the containment.
Additional test data was taken which show a successful leak rate.
A validation test following the ILRT was completed on June 16.
The containment was subsequently l
depressurized and the drywell to torus bypass test performed.
This test was successfully completed by 8:15 a.m.,
on June 17.
A detailed report, " Reactor Containment Building Integrated Leakage Rate Test Report" was issued to Docket No. 50-277 for this test.
Followir.g completion of tha ILRT, check-off-lists and surveillance tests required prior to startup were initiated.
The surveillance tests associated with the stroking of the MSIV's indicated that the closing time of the
'c' and
'D' outer isolation valves was too fast.
The f ast operation was caused by leakage of air from the operator through a defective shaft seal into an oil dash pot in the hydraulic circuit.
Dismantling of the operators was required.
The seal was replaced in the 86D valve.
Difficulty in dismantling of the 86C valve resulted in damage to the origiani stem.
A new stem was not available until June 24.
Plant startup was 3
therefore authorized with the
'C' main steam line isolation valves closed.
Unit 2 reactor was made critical on June 22, following i
an extended refueling outage.
Testing indicated that l
sufficient shutdown margin existed with the reload core.
Startup testing and surveillance testing in accordance i
with the Startup Program were successfully completed on June 23 and June 24, resulting in synchronizing the generator at 3:33 a.m.
on June 24, 1976.
" Report of Plant Start-up Following First Refueling Outage June 1976" was issued to Docket No. 50-277.
Power was I
subsequently increased in accordance with the Puel Preconditioning Program.
A power level of 75%, or i
approximately 828 MWe, was reached on June 29.
At 5:03 p.m. on this date, the 71L relief valve spontaneously I
opened, resulting in a primary coolant blowdown from i
operating pressure to approximately 350 psig.
The reactor was scrammed manually following the opening of the relief valve.
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1 PBAPS l
i I
)
k During this shutdown, the 71L relief valve was replaced j
with a rebuilt valve recently removed from another location on Unit 2.
Additional work accomplished during the outage included the replacement of the operator stem j
on the 86C MSIV.
This valve was stroked satisfactorily j
following repair.
1 i
Following the replacement of Relief Valve 71L, Unit 2 j
was made critical at 4:40 p.m.,
on July 2.
The a
generator was synchronized at 2:54 a.m.,
on July 3.
i Load was then increased in accordance with the preconditioning program.
By July 6',
the load on Unit 2
.]
had been increatied to approximately 912 MWe, on this 5
.date, an "a" switch in the 2A condensate pump breaker compartment malfunctioned, causing a runback of both
{
recirculation pumps.
The "a" switch was repaired and power level recovered.
1 Maximum load of approximately 1000 MWe was achieved on July 8.
Power was limited by the preconditioning j
l envelope which existed prior to startup.
On July 9, load was reduced to reposition control rods.
Power was i
again increased and full load achieved on July 14.
On July 18, a 300 MWe load reduction was initiated to l
prevent loss of condenser vacuum when the air ejector malfunctioned.
The air ejector malfunction was traced to low steam pressure supplied to the air ejector.
This condition was corrected and the load again increased in accordance with the preconditioning limitation.
Surveillance testing on July 19, identified electronic component failures in the ramp generator associated with the HPCI speed control.
The HPCI was declared inoperable and the require surveillance testing on other engineered safeguard systems initiated.
Repairs were completed and the HPCI quick-start tested on July 21.
On July 26, the reactor buidling ventilation fans tripped because of a supply damper malfunction.
Loss of ventilation rapidly increased the main steam tunnel temperature.
In order to prevent a scram associated with this problem, the setpoints on the temperature switches were temporarily increased until the ventilation system could be re-established.
Tunnel - - _ - - _
PBAPS temperatures were monitored frequently during this period of time.
Within 15 minutes, the ventilation j
system was re-established and the setpoints restored to j
their original values.
These instruments were then i
recalibrates.
Repairs to the supply dampers have br?n accomplished.
Full load continued through August 5, when a scram occurred.
This scram was caused by a l
f atigue failure of the EHC hydraulic tubing connected to
)
the number 4 control valve.
The tubing failure was immediately repaired and the reactor was made critical on August 6, at 1: 00 a.m.
The generator was synchronized at 7:50 a.m. on the same date.
Power was increased in accordance with the fuel preconditioning limitations.
l On August 9, with the unit operating at 870 MWe, Unit 2 again scrammed at 2:35 p.m.
The scram was caused by low reactor level.
Investigation into the cause of scram indicated that while troubleshooting within the static inverter cubical associated with the non-interruptable a.c.
supply, the neutral conductor was inadvertently opened.
This resulted in a fictitious high level output j
from the reactor level instrumentation which i
subsequently tripped the main turbine and all three j
feedwater pumps.
An immediate restart was authorized.
The reactor was again critical at 8:32 p.m. on August 9, j
and the generator was synchronized at 12:43 a.m. on August 10.
On August 11, with the unit at approximately 600 MWe, the main generator stator cooling pump tripped.
Before the standby pump could come up to cpeed, the pressure in
{
i the system decayed to initiate a load runback resulting l
in tripping of one fo the reactor recirculation pumps.
{
The pump was successfully restarted and the load recovered in accordance with the preconditioning l
program.
Full load was reached on August 15.
Following the tripping of the stator cooling pump, immediate repair operations were initiated.
The failed pump was returned to service on August 16.
Shortly thereafter, the operating stator cooling pump (2B) tripped.
This again resulted in a system pressure decay and the tripping of one reactor recirculation pump.
Load was subsequently reduced and the recirculation pump. _ _ _ - _ - _ _ _ _ _ - _ _ - - - - _ - _ _ _ _ - _ - _ _ _
g-PBAPS successfully restarted.
Power was increased in accordance with the preconditioning program and full load was achieved on August 18.
Later on August 18, Unit 2 again scrammed.
This scram was caused by a 9-second EHC tubing failure in the same line as the original failure which occurred on August 5.
The unit was shutdown and depressurized in order'to permit repair of a recombiner mechanical compressor and to correct l
leaks on several drain cooler manways.
Additionally, an
]
electrical connection at the recirculation pump MG set
.{
i transformer was renewed following its flash-over to i l
g round.
This maintenance work extended the outage
)
through August 20.
The unit was returned to service at j
4:24 p.m. on this date.
Poaer ' level was increased to 4
approximately 780 MWe by August 23, following the normal fuel preconditioning requirements.
Surveillance testing on August 23, indicated a primary I
coolant leak inside the drywell in excess of the Technical Specifications limit.
The unit was subsequently shutdown at 11:23 p.m.,
on this date.
Following purging of the drywell, a packing leak on the
'A' recirculation pump suction valve was identified as I
the cause of the excessive leakage.
Following i
depressurization of the reactor, the packing on this valve was replaced.
Additional maintenance work was performed on various equipment, including correction of a diaphragm leak on the 96A feedwater check valve.
Maintenance work was completed on August 26, and the unit resynchronized at 7:06 a.m.,
on August 27.
Fall load operation was established on September 1.
On September 21, the NSS Supplier informed Philadelphia i
Electric Company of an inconsistency in the FCCS l
calculations submitted to the Nuc3 ear Regulatory Commission as part of the reload / document.
These calculations assumed that the Coi'e Spray and RHR injection valves received an opening ' signal as reactor pressure decreased to 450 psig and that the recirculation pump discharge valves received a closing signal at 250 psig.
In reality, both these circuits were actuated at a pressure setpoint somewhere between 1
350 and 300 psig.
This was reported to the Nuclear l.
Regulatory Commission, and power reduction was initiated i!
as a conservative measure until the full implications of t _ _ _ - _ _ _ - _ _ _ -
'l 1
this error could be evaluated.
Later on the same day, it was determined that the unit ~ould be returned to I
f ull power, with the MAPLHGR limiu "stablished at less d
i p
than 0.98 of the Technical Specifict ion limits, as an j
appropriately conservative measure.
i On September 24, loss of the 2A mechanical recombiner compressor, because of a faulty discharge relief valve, j
resulted in a slow loss of condenser vacuum which eventually resulted in a plant scram.
During the outage that followed, the recombiner relief valves were rebui' c.
The reactor'was taken critical on September 25, and the turbine generator synchronized on September 1
26, at 2:15 a.m.
Power was increased in accordance with I
the fuel preconditioning program and full load achieved on September 30.
on the same date. additional difficulties with the relief valve on the 2A mechpnical recombiner compressor again I
resulted in a decreading vacuum condition on the main condenser.
In orderito recover vacuum while relief _
valve repairs were being made, unit load was reduced to 300 MWe.
The unit returned to full power operation in l
accordance with fuel preconditioning requirements.
Full l
load was achieved on October 3.
On October 9, Unit 2 was reduced in power in order to i
perform a control rod sequence exchange.
Power level l
1 was recovered in accordance with fuel preconditioning requirements uduil full load was achieved on October 16.
On October 17, the unit experienced a scram due to a l
The turbine trip was caused by a loss of l
125V d.c. power supply in the EHC System.
The 125V d.c.
fuse blew during a thrust bearing wear detector surveillance test.
The excessive current was caused by a defective motor in the thrust bearing wear detector l
circuitry.
Scram recovery operations were initiated immediately.
The reactor was again critical at 5:30 a.m. on October 18.
The generator was synchronized at 9:52 a.m. on the same date.
With the unit in service, the thrust bearing wear detector motor was replaced.
Testing of the replacement motor at 2:05 p.m. on the same date again resulted in a thrust bearing turbine trip and reactor scram.
Turbine trip in this case was +
?
I I
PBAPS 1
caused by the improper rotation of the replacement 4
motor.
The reactor was again recovered and criticality established at 9:33 p.m.
The main generator was synchronized at 8:25 a.m. on october 19.
Power was
)
increased in accordance with the fuel preconditioning program with full load being achieved on October 23.
With the completion of maintenance work on November 25, the reactor was taken critical and the generator synchronized at 12:50 p.m. on November 26.
The four i
relief valves which had been maintained or replaced during the outage were successfully surveillance tested at approximately 170 psig reactor pressure prior to synchronizing the unit.
Power was subsequently increased and full load operation achieved on December i
3.
I During ^he month of December 1976, Unit 2 operated at essentially full power except for load changes in response to high Susquahana River differential temperatures as measured between Holtwood and the Maryland-Pennsylvania State line.
Unit 2 was operating at full load at the end of the year.
C.
Unit 3 Operations As of December 31, 1975, Unit 3 reactor was critical in the hot standby condition, while an investigation into the source of the condenser vacuum leak was in progress.
This investigation proved unsuccessful.
On January 1, 1976, a decision was made to start the core plugging outage to eliminate LPRM/ channel box interaction.
Operations associated with gaining access to the top of the core were started on January 2.
Head removal and vessel internal removal progressed satisfactorily.
The reactor head cavity was flooded on January 6, but the water was very cloudy.
Cleanup of the cloudy water conditions delayed fuel movement until Friday, January 9.
Concurrent with fuel handling operations, local leak rate testing and repair of leaking containment isolation valves were in progress.
By January 23, refueling floor operations had progressed to the point where all channels had been inspected and _ _ - _ _ _ _ _ - _ - - _
3 d
.l q
PDKPS
]
thyde which were not acceptable replaced with new l
channels.
Core plugging had-been completed.
All fuel had been returned to the core.
Core verification was complete and rod operability testing was in progress.
During the outage, three control rods were removed and replaced with new blades..The removed blades were j
shipped to. General Electric Cumpany for analysis.
]
By January 30, the steam separator and'dfyer had.been-installed.
The reactor vessel head was installed and the reactor prepared for' hydro..The major work l
associated with repair of containment isola tion valves and leak testing of these valves was completed.
Total i
local leak rate test leakage was less than all.owable.
The reactor vessel hydro was completed on January 30.
Drywell head arid shield plugs were installed by February 1.
check-off-list prior to startup were completed.on February 2.
The reactor was. critical at 1:53 p.m. on the same day, and the unit 1 synchronized by 10:09 a.m.
on February 3.
' Power. was then increased in accordance with the precondi'tloning program and reached approximately 900 MWe on February 9.
The unit continued at this load until, late
- February 13, when power was reduced to as low as 420 MWe for rod sequence exhange and special turbine valve testing.
No problems resulted from the teuting and power was increasing in accordance with the preconditioning program when the Unit scrammed on February 18, due to action of the power-load unbalance instrumentation.
This is believed to have been caused by 4 momentary partial closure of the CIV's but no cause of the failure could be found.
An immediate restart was authorized.
The Unit was critical at 3:00 a.m. on February 19, synchronized at 8:23 a.m.,
and power increased in accordance with the preconditioning program.
The unit was removed from service on February 22 in order to correct. a significant packing leak on a jet pump differential pressure sensing line manual block valve.
Repair of this valve required a depressurization and cooldown of the primary system.
The unit'was returned to power operation on February 23.
On February 26, the dryvell on Unit 3 was pressurized and the -
- as placed in service to i
~17-
1 PBAPS maintain a slight vacuum in the torus compartment.
This differential pressure continues to be maintained in conservative consideration of the Mark I pool swell effects.
Power was increased in accordance with the fuel preconditioning program through February 27, when
)
the unit was removed from service in order to correct a ground in the control circuit for the "B" loop recirculation pump discharge valves.
This ground was identified on February 25, and essentially made one-half of the RHR-LPCI system inoperable.
This outage was
. extended when the disc on the "B" Loop RHR throttleable angle valve 154B was found to be separated from the stem.
Temporary repairs were made and the unit returned to service on March 7.
On March 8, the unit was shutdown due to a valve error and resynchronized the same day.
On the morning of March 13, leakage inside Unit 3 drywell increased rapidly between 8:00 a.m.
and noon to approximately 6.5 gpm.
Shutdown was started at 12:30 p.m. on the same day.
Later on the same date, a drywell entry identified the leak as a cracked 1" pipe on the suction side of the "B" recirculation pump upstream of the pump inlet valve.
The reactor was placed in cold shutdown.
i on the morning of March 14, leakage rate was about 11 gpm with the reactor depressurized.
The crack existed i
at 10 O' clock in the heat affected zone of the socket I
weld made on the end of a half coupling which was welded to the 29" recirculation line.
The leak was repaired by reducing pressure at the point of leakage to a minimum by pulling 25" Hg. on the reactor, removing the cracked pipe, installing a tapered stainless steel plug, seal welding the plug in place, installing a socket weld cap over the plug with a full penetration weld to the existing half coupling, and dye penetrant testing the weld.
Dye-penetrant tests were also performed on the "A" recirculation pipe, half coupling weld to the 28" OD recirculation pipe and the socket weld to the 1 inch pipe.
No defects were identified.
On March 15, an inspection was made of the closure on i
the "B" recirculation pipe with the primary coolant et full operating temperature and pressure.
No leaks were j
noted.
The unit was returned to power on March 15, and i
reached full load by March 19.
After reaching full load 1
I I 1
l l
l l
l l
PBAPS l
i on March 19, the plant continuted to operate at maximum power level through March 20.
On this date, an instability in the 3B recirculation pump voltage regulator caused the recirculation pump to trip.. The 1
operating pump was reduced in speed and load reduced to i
536 MWe.
With the pump tripped, an attempt was made to
)
close the discharge valve MO-53B prior to attempting a restart.
All position indication on the valve was lost during the attempt to close the valve.
An investigation determined that the difficulty was in the limitorque operator.
Since this valve is part of the LPCI logic
~
l circuitry, the unit was shutdown.
A drywell entry was made and two broken wires on the limitorque operator i
repaired.
During this time, the voltage regulator on i
the MG set was also repaired.
The generator was returned to service at 8: 24 a.m., on March 22 and j
increased in power in accordance with the fuel preconditioning limits.
By March 24, reactor load had been increased to 924 MWe, At 11:05 a.m. on this date, the
'B' Target Rock relief valve momentarily opened and reclosed.
Electrical tests were run which determined that the electrical circuitry for this valve was not I
defective.
In order to prevent recurrence, reactor pressure was limited to about 985 psig.
Full load operation was essentially reached on March 26.
On April 13, a reactor scram occurred due to a leaking instrument bypass valve associated with the differential pressure switch which monitors main steam line flow.
The valve block associated with the differential pr. essure switch was replaced.
During the outage, information was received that a recent review of the RHR System logic indicated a deficiency which would result in only 2 of the 4 RHR pumps starting during a loss of offsite power accident simultaneous with a LOCA.
The test circuitry review was conducted in conjunction with a review of the loss of outside power surveillance test.
The logic deficiency was discussed with Philadelphia Electric Company Engineering Department and agreement reached on the deficiency.
The Engineering Department provided a circuitry change.
These changes in circuitry were l
completed by April 14, and a Unit 3 loss of power-LOCA test performed.
This test functionally verified that j
all RHR pumps would operate under such an accident.
l 1 1 j
PBAPS With the completion of the loss of power test following the LPCI modification on Aril 14, the reactor was returned to service.
The reactor was made critical on April 15, and the generator synchronized at 11:30 a.m.,
j on April 16.
Power was increased in accordance with the fuel preconditioning program.
The reactor operated through April 29, when a scram occurred at 9:00 p.m.
The scram was caused by a high
'i flux trip which is believed to have been caused by a rapid flow increase in the recirculating water system.
At the time of scram, an electrical feed to the 3B recirculation pump MG set control circuit was reclosed after being inadvertently opened.
The unit was returned to service, with the genarator being resynchronized at 8:14 a.m.,
on April 30.
On May 1, during an investigation of some difficulty in the level control system of Unit 3, an inoperable discharge check valve on the 3A reactor feedpump was identified.
The 3A feedpump was subsequently removed from service until repairs could be made.
This limited maximum power to approximately 930 MWe.
On May 2, the 3B recirculation pump tripped at approximately 8:00 a.m.
The pump was returned to service at 10:45 a.m. on the same day.
No difficulties in returning the pump to service were experienced.
On May 3, the 3B recirculation pump again tripped.
Investigation again showed no pump difficulty.
However, the pump discharge valve (53B) failed to close.
This required a plant shutdown.
The unit was removed from the bus at 2:19 p.m. on May 4.
A drywell entry was made and several broken wires in the valve operator were repaired or replaced.
The unit was resynchronized at 7:22 a.m. on May 5.
On May 6, the HPCI System was determined to be inoperable because of high steam flow condition which occurred during automatic initiation of the turbine start.
Surveillance testing required by the inoperability of the HPCI System was initiated.
This testing identified an ADS timer which was slightly outside the Technical Specification limits.
A power reduction was initiated.
Repairs to the timer were completed in a short period of time and power level _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
[
PBAPS I
recovered.
HPCI Hydraulic System repairs were completed and operability proven on May 8.
On May 8, the 3B recirculation pump again tripped.
Electrical testing of the recirculation pump motor circuitry identified no deficiencies.
The pump discharge valve closed properly '
but would not open after the recirculation pump was j
restarted.
The plant was removed from service at 9:00 p.m. on May 8, and a drywell entry made.
Inspection inside the drywell indicated a grounded motor lead on the 53B valve, a loose wire on the current transformer circuitry on the 3B recirculation pump motor protective relaying, and a leak in a one inch instrument line between the pump and the discharge valve.
The primary system was subsequently depressurized to make repairs.
During the outage, vibration instrumentation was installed on the ' B' and
'A' recirculation loops in order to investigate the cause of the recurring difficulties in this area.
During the outage which started on May 8, an 1
investigation into the
'B' recirculation loop vibration identified a defective bearing in the 3B recirculation pump motor.
The upper guide bearing in this motor was replaced on May 11.
Repairs were also made to the balance plate in this motor.
Vibration instrumentation was installed on the motor connection box, as well as the pump discharge valve limitorque operator.
i Additional vibration instrumentation was installed on a one inch pipe which connects to the recirculation loop between the
'B' pump and its discharge valve.
A crack near the socket weld in this pipe was repaired during this outage.
The reactor was started up on May 13, and the generator synchronized on May 14.
Power of this unit was limited to 20% because of the inoperabilty of all three cooling towers.
Permission to operate above 20% without cooling towers was received from the NRC via a Technical Specification Change on May 15.
On May 17, readings taken from the vibration instrumentation installed on the recirculation loop indicated excessive vibration on the 3B recirculation motor and discharge valve with pump speed in excess of 52%.
Pump speed was subsequently reduced to get vibration levels down to 1 mil on the pump connection box and approximately 3 mils on the i - _ _ _ _
PBAPS discharge valve motor operator.
This limited plant output to approximately 760 MWe.
On May 21, Unit 3 was shutdown to investigate the cause 1
of recirculation system vibration.
The 3B recirculation
(
pump motor was uncoupled and data taken to permit balancing of the motor.
Following rebalancing the reactor was started up on May 23, and the generator synchronized on May 24.
Approximately four hours later, a water slug from the air ejectors into the recombiner sufficiently reduced recombiner temperature to the point where hydrogen concentrations downstream of the recombiner tripped the mechanical compressors.
Since l
the system could not be returned to service in a short period of time, the plant was removed from service by scramming the reactor.
A restart was attempted later in the day.
At this time, a ground detector in the l
generator field annunciated.
After a significate amount l
of investigation, a decision was made to synchronize the l
l genera tor.
Several hours later, however, a generator l
core monitor alarm annunciated and the turbine was i
manually trippea.
The pressure pulse to the reactor caused a scram on high flux level.
The unit was i
subsequently cooled down, the condenser vacuum broken, the generator hydrogen purged, and the generator field k
partially disassembled.
The field ground was eventually I
located in the generator rotor bridging stud insulating
)
I collar.
Repairs were made; the generator was reassembled and resynchronized on May 31.
During this
(
outage, additional balancing was performed on the i
recirculation pump motor.
Following return to service, l
the unit was increased to full power, while vibration was monitored.
At maximum speed, pump vibration. appears to be 1 to 2 mils and discharge valve operator vibration I
was approximately 2 to 3 mils.
l Unit 3 operated at essentially full power from June 5 through June 18.
On June 18, power was reduced to approximately 400 MWe in order to perform a rod sequence exchange.
During the low power period, the brushes in the recirculation pump MG sets were replaced, a damaged bus duct cooling f an was removed from the bus for repair off aite, and the generator stator liquid cooling system heat exchangers were cleaned.
On June 19, power level was again increased in accordance with fuel.
_m
PBAPS preconditioning requirements.
Full load was reached on June 24, and continued through July 19.
During this period, power slowly decreased due to fuel burn-up.
?ower was limited during thir, period because of the
_; restrictions associated with rod motion imposed by fuel preconditioning requirements.
surveillance testing on July 5, identified electronic problems in the RCIC speed control system.
Repairs were accomplished and the system quick-start tested on July 7.
The necessary surveillance testing associated with the inoperability of the RCIC system was accomplished l
during this period.
On July 9, Unit 3 power was reduced to reposition control rods in order to return the unit to full load operation.
During the recovery from this load drop, a lightning storm disrupted the offsite system sufficiently to cause the loss of the number 3 startup feed.
Loss of this feed at 3:08 a.m. on July 11, resulted in a Unit 3 reactor scram.
During this lightning strike the main stack flow monitoring system was lost for approximately twelve hours and the radioactive effluent monitoring system was lost for about six hours.
Following the re-establishment of both startup feeds and return to service of the stack flow monitoring and sampling system, Unit 3 reactor was made critical et 1:40 a.m. on July 12.
Reactor pressure was increased.
Wfth the reactor power at approximately 9%, the 71B relief valve spontaneously opened at approximately 6:00 a.m., resulting in a reactor blowdown to approximately 200 psig.
Following primary coolant depressurization and reactor shutdown, the
'B8,
'F',
and ' K' relief valve operators were replaced.
A torus inspection was made which indicated that no damage bad occurred inside the torus.
Following replacement of the relief valves, the reactor was returned to critical at 3:51 p.m. on July 14.
Primary coolant pressure was increased during several attempts to restart the eA' recirculation pump.
I During each attempt, the scoop tube arm knuckle joint l
broke.
A decision was then made to bring the reactor i
sub-critical and inspect the internals of the hydraulic coupling.
This inspection identified two ball joints
-?3-
.1 I
1 PBAPS
^
l inside the hydraulic coupling which had excessive i
. clearance.. These ball joints were replaced and the recirculation pump returned to service.
The reactor was again made critical at 8:55 a.m. on July 15, and pressure increased to approximately 400 psig.
A drywell inspection at this time identified a plug leak.
j on the bottom of MO-18 of the RHR shutdown cooling i
suction valve.
The manual block valve upstream of MO-18 was subsequently closed to stop this leakage.
The generator was synchronized at 6:15 p.m. on the same date i
and load increased in accordance with the preconditioning program.
Full load was achieved on July J
19.
On July 20, at 10:50 p.m.,
the 71G relief valve spontaneously opened, requiring a manual scram of the reactor.
Following shutdown and depressurization, the
'G',
'D',
'E',
'H' and ?J' valve operators were replaced.
These valve operators were replaced based on l
high temperatures monitored downstream of these valves i
prior to shutdown and a leak test of the pilot valve I
assemblies performed during the shutdown.
A torus inspection made during this outage again identified no internal damage.
Following maintenance operations on the relief valves and other miscellaneous work, the reactor was returned to service on July 25, and the unit synchronized at 10:55 a.m.
Following synchronizing, power was increased and full load achieved on July 29.
Unit 3 has operated through August 13, at essentially full power.
A rod sequence exchange was initiated on this date requiring a load reduction to approximately 400 MWe.
Following the sequence exchange load was'again increased in accordance with the fuel preconditioning limits.
Full load was again achieved on August 19.
On September 6, control rod 06-23 scrammed in the reactor because of failure of the inlet scram valve to the drive unit.
This resulted in a load decrease of approximately 20 MWe.
The failed diaphragm was repaired.
In order to again withdraw the rod, power was reduced to less than 800 MWe.
At the same time, additional deep control rods were withdrawn one or two notches.
Load was then again increased at the normal j
1 l
PBAPS preconditioning rate of 5 MW/HR, resulting in a return to full load operation on September 9.
i Unit' 3 operated at essentially full load from September J
9 through September 18.
On this date, a high i
unidentified leakage in the Unit 3 drywell required Unit
-f 3 to be removed from service.
The source of this leakage was identified as a leaking packing on the shaft of the ' D' recirculation pump discharge bypass valve (MO-54B).
This valve was repacked, the reactor taken critical and the unit resynchronized by 4:00 a.m. on September 20.
Immediately after synchronizing, difficulty maintaining condenser vacuum was experienced.
This required the removal of the unit from service at 1:00 p.m. on the same date.
The source of vacuum leak was traced to the startup vacuum pump. seals.
The pump was valved out and the unit resynchronized at 8:12 p.m.
on the same date.
Power level was increased in accordance with preconditioning requirements and full load achieved on September 25.
During this power increase, starting on October 7, the air flow through the recombiner system showed a steady increase.
On October 8, difficulties with one of the mechanical recombiner compressors resulted in a partial loss of condenser vacuum.
A load reduction of 200 MWe was taken in order to re-establish vacuum following maintenance on the mechanical compressor.
An I
investigation into the source of leakage did identify a minor leak on the manway in the upper portion of the B1 condenser.
This leak was temproarily corrected.
The unit continued in operation with air flow through the recombiner.
The PORC noted that Unit 3 appeals to be starting the end of cycle coastdown.
It was estimated that two to three megawatts per day would be lost as a result of fuel burnup.
On October 19, the source of the B1 Condenser inleakage was identified as a failed bellows on a steam jet air ejector relief valve.
The problem was immediately corrected by inserting a pipe plug in a vent connection on the relief valve bonnet.
-2b-
i 1
1 PBAPS s
l on October 19, Unit 3 load was significantly reduced in order to remove the 5th heaters from service.
Operation l
in this mode provides for a cooler feedwater temperature which would extend the core life.
Following removal of the 5th heaters, the control rods were repositioned and load increased in conformance with the fuel preconditioning requirements.
A maximum load of 1048 MWe was achieved on October 21.
operation at essentially this load continued until l
October 28, when load was reduced by 60 MWe immediately following operation of the Reactor Water Cleanup System in a vessel-to-vessel mode.
There was no increase in main steam activity or reactor water conductivity noted immediately following this power decrease.
Full load was again reached on October 29.
By November 24, reactor power was reduced to 936 MWe due to fuel burnup.
On this date, a rod sequence exchange was initiated.
This required a power reduction to approximately 450 MWe.
Power was subsequently increased j
to a maximum of 944 MWe by December 5.
Power reduction j
was initiated because of high river differential temperatures starting on December 3.
j Unit 3 operated from December 3 through December 24, at approximately 880 MWe.
Load reductions were required during this period in response to high Susquehanna River differential temperatures, as measured between Holtwood I
and the Maryland-Pennsylvania State Line.
Maximum load was limited by the reactivity available at the end of the first core.
The first refueling outage for Unit 3 was started on December 24.
Shutdown cooling was established on December 25.
Following cooldown, the reactor was l
pressurized to approximately 50 psig, using instrument air in order to do an ILRT on the MSIV's.
This test resulted in the determination of leaks on the A, C, and D inner MSIV's.
Following the valve testing, outage work continued with the drywell head being removed on December 27, and the reactor vessel head being removed on December 30, 1976.
-N.-
_____ _ _ _ _ _ _ m
PBAPS D.
Common Plant Operations Iodine Release
\\
During the second and third quarters of 1976, the plant operated with iodine releases in excess of the 8%
1 average release permissible in the Technical Specifications.
An NRC inspection in October, i
identified this operating parameter as unacceptable to Nuclear Regulatory Commission Licensing.
A meeting was held on site during the last week of October with i
representatives of the NRC Licensing and Region 1.
At f
this meeting, the plant was informed that both unit
.shall be shutdown when the average lodine release for the fourth quarter reaches 8% of the instantaneous limit.
This value is approximately 71,000 uCi total release.
In order to reduce the total plant releases of iodine and particulate, the reactor building ventilation systems on both units were rebalanced such that the equipment cell exhaust was processed through one train of the Standby Gas Treatment System.
Rebalancing of these ventilation systems was performed during the week of October 25.
The iodine releases were significantly reduced.
Average release following rebalancing of the ventilation system was approximately 200 to 300 uCi per day.
These low values of release continued through the end of 1976.
River Temperatures As of December 1, 1976, the confidence limit of the measured river temperatures was reduced to 1.50F.
Thi s lower State limit, coupled with low river flows during the last part of November and early December, resulted in differential temperatures which exceeded the 6.SoF on numerous occasions starting December 3.
The load on both reactors was subsequently reduced in accordance with procedure.
It appears that the large variation in temperatures being experienced is primarily a result of flow variations caused by the operation of Muddy Run and conowingo and is not significantly affected by the actual power level of the Peach Bottom facility.
Because of the differential temperature problems, significant power reductions were taken during periods when high differential temperatures were p asent. ____ ____ - -
PBAPS l
l Reactor power was again increased in accordance with l
preconditioning requirements during periods when the differential temperatures were below 6.SoF.
I E.
Reactor Shutdowns, Power Reductions and Maintenance summaries Tables I-E-1 (a) through (c) and I-E-2 (a) through (e) j list the shutdowns and forced re8.ctions for Unit 2 and l
Unit 3 respectively.
The tables provide the cause of
{
the shutdown or reduction and list the component that was the cause.
For shutdowns, the duration was determined by the time that the main generator was off the line.
For forced reductions, the duration was determined by estimating the time the generator operated at less than 840 MW for greater than four hours.
If a Reportable Occurrence was associated with a shutdown, or f orced reduction, the identifying number of the Reportable Occurrence is included in the table.
Corrective action taken to reduce the probability of recurrence and a description of major safety-related corrective maintenance performed during each shutdown or l
forced reduction are included in Tables I-E-3 (a) through (k) and I-E-4 (a) through (g)
" Peach Bottom l
Significant Maintenance Summary." Tables I-E-5 (a) through (1) and I-E-6 (a) through (i) are a chronological listing of nuclear safety or plant reliability related instrument failures repaired during 1976. _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
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PBAPS i
II. MODIFICATIONS Modification 16-005 Motor Operated Circuit Change (Units 2 and 3)
This chansa af fects saf ety related valves previously equipped with 300% thermal overload devices.
It involves replacing 300% overload with 100% overload and adding circuity and a relay so that the 100% overload will trip the valve motor if the valve motor operation I
is initiated normally from the control room.
The relay will bypass the overload trip, if the valve motor is energized automatically, or if the spring-return-to-normal hand switch is manually held in the open or close than the motor in an, overload element will last longer position.
The 100%
actupl overload condition and, therefore, its faildre is not a safety problem.
The change is intended to reduce the large number of motor f ailures, has been designed by Philadelphia Electric Company Engineering Department, and does not involve an unreviewed safety question. '
i Modification 76-004 Steam Line-Bellows Repair (Unit 3)
This modification consists of installing a new bellows around the outer bellows and was required because the inner bellows failed, thus precluding bellows testing in the normal manner.
The modification will return the bellows to normal service, the design was by the origiani manufacturer Tube Turns, Philadelphia Electric Company Engineering Department had approved the modification, and it was determined than an unreviewed safety question is not involved even though the physical arrangement is slightly changed.
Modification 76-14 Removal of the Recire.
4" Bypass
_ Lines (" nit 2)
This change involves removal of highly stressed stainless steel pipe to preclude cracking in that pipe.
It is not an unreviewed safety question, and poses no operating problems for starting the recirculation pump.
PBAPS i
Modification 76-15 Venting of the RWCU Backwash Tank to the Standby Gas Treatment System (SGTS) (Unit 2)
)
This change adds piping to permit direct venting of the 1
l tank to the SGTS so as to preclude radiation relases from the vent stacks.
It is not an unreviewed safety question.
Modification 76-010 RWCU Return Path to Vessel (Common Plant)
This modification adds a valve in radwaste on a line discharging to the waste collector tank.
Closing this valve combined with other appropriate valving will permit RWCU ef fluent. return to the vessel via the RER system when the normal retain path through the feedwater lines is unavailable due to valve testing and maintenance.
An unreviewed safety question was not involved.
Modification 75-019 TIP Drive Interlock Addition (Units 2 and 3) 1 This change involves the addition of a, solid state circuit to the TIP system to back up the limit switch which stops TIP withdraw.
On several previous occasions, the limit switch has failed to stop TIP withdrawal.
This can result $n retracting the TIP chamber back into the cable reel housing, resulting in a very high radiation and, possibly, contamination conditipn.
The new circuit results in preventing.TIP cverrun during withdrawal by preventing withdrawal if TIPS exceed a civen withdraw position.
To provide a method of obtaining withdraw in the event of the failure of the new circuit, a DPDT pushbutton is used as a reset; one half to reset the logic, the other to interrupt the output.
Thus, even if the logic fails so as to prevent withdrawal, the operator may withdraw the TIP manually while holding in the reset pushbutten.
An unreviewed safety question was not involved. 1 4
PBAPS i
Modification 76-31 Radwaste Piping Change (Common Plant) l This change involves the addition of a 2 inch pipe and valves between the waste dmineralizer and the waste sludge tank.
This line permits processing waste j
demineralized resin to the centrifuge via the waste i
sludge tank and the condensate phase separators through l
large pipes instead of directly to the centrifuge through a one inch line.
This change should reduce time and water consumption, and preclude the line plugging.
An unreviewed safety question was not involved.
Modification 76-19 Torus Support Addition (Unit 2)
The addition of supports to the torus was necessary to upgrade the torus to account for pool swell effects, was engineered by the original designers, was reviewed by l
Bechtel and Philadelphia Electric Company Engineering Department, was a committment to the NRC as a result of 1
the Mark I containment studies, and was not an unreviewed safety question.
l I
Modification 76-22 RSCS Group Notch (Unit 2)
The RSCS was changed to incorporate " group notch" I
instead of a " Group c" between 50% rod density and 30%
l power.
This change will permit startup in either the
'A' or
'B' sequence yet prevent peak enthalpies from l
exceeding 280 cal /gm 6 ing the rod drop accident.
The 1
change was previously approved by the NRC.
Modification 76-26 RPV Stabilizers (Unit 2)
J Changes to be made during the repair of the bent reactor pressure vessel stabilizers:
1) increase clearance gap for 0.010 to 0.053 2) minimum clearance of 2" ovar nuts at ends 3) retension in accordance wi;h GE 730E853' 4) repair removed stablizer per M4,34 1
This was recommended by General Electric Company, approved by Philadelphia Electric Company Engineering Department, and was not an unreviewed safety question.
l 1 l
l
_ _A
PBAPS j
Modification 76-33 RWCU Check Valve Installation (Unit 2)
This change involves the installation of a new check valve in the RWCU return line between valves 62 and 68.
i This valve was installed because of the inability to repair the 62 check valve which had excessive through leakage.
The 62 valve has had the internals removed and the new valve essentially replaces it.
The valve meets the piping design, the installation was in accordance with the design, and it was determined that an 1
unreviewed safety question was not involved.
]
Modification 76-35 Radwaste Freeze Protection (Common Plant)
This change involves the addition of a one-inch crosstie with orifice, check valves, and block valve from the service water system to the radwaste discharge line to f
provide continuous flow ~and thus prevent line freezeup j
during winter.
The change was not an unreviewed safety
)
question and had been approved by the Engineering and i
Research Department.
)
j Modification 76-053 LPCI & Core Spray Logic (Units 2 and j
3)
J l
Wiring and calibration changes were made to the LPCI and i
Core Spray initiation logic to permit different pressure switch setpoints to be used to open the injection valves and to close the recirculation pumps valves.
The change I
to the Technical Specifications had previously been
)
approved by the NRC, as Amendment Nos. 28 and 27 I
respectively.
i Modification 76-48 Analog Switch Installation (Units 2
)
and 3) e 1 This change involves the replacement of various switch-type primary instruments with analog transmitter-electronics switch-type devices.
This had been designed by the Engineering and Research Department, and was similar to a change on the Reactor Pressure Switches, and involves a Technical Specification Change approved by the NRC as Amendment Nos. 30 and 29 respectively.
)
) 1
I PBAPS I
i Modification 76-58 RWCU Piping Modifications (Unit 2)
A line is installed bypassing the pumps and the regenerative heat exchanger on the reactor water outlet i
side and a valve is to be installed in the return line to the regenerative heat exchanger.
The purpose is to permit total isolatior, of the regenerative heat l
exchanger and thus stop leakage of radioactive reactor water.
The change had been designed by the Engineering Department, and is not an unreviewed safety question, because the system is still isolatable from the reactor.
Modification 76-65 Target Rock Ri}ief Valve Pilot Filters (Unit 2) i The reinstallation of these filters should reduce the i
potential for pilot valve leakage which, when large enough, causes valve operator and reactor blowdown.
General Electric Company had originally recommended j
removal of these filters to obtain the required valve l
response.
Subsequent modifications to valve vent and drain grooves and additional testing has resulted in acceptable valve performance even with a partially plugged filter.
General Electric Company has informed I
us that the vent and drain modification were made on our val ve s.
This modification should reduce blowdown accident probability, did not change Technical Specifications, and was not an unreviewed safety question because valve response would not be affected.
Modification 76-68 Rod Worth Minimizer (Unit 2)
This change involves a modification to the Rod Worth Minimizer which deletes the rod select block.
The new program for the RWM which is compatable with the Rod Notch change to the RSCS contains no select block.
The RWM still implements rod withdraw and insert blocks (which were also initiated whenever a Select Block was initiated) could be replaced by a second operator if necessary, and will still acceptably limit the rod drop accident consequences.
Because the RWM can implement no select blocks, it cannot clear then either, and the select block relays were electrically disarmed to prevent inadvertent latchup action.
1
' l l
l
PBAPS l
III PERSONNEL EXPOSURES & RADIOACTIVE RELEASES A.
Personnel Exposure By Job Punction i
A tabulation of station, utility, and other personnel receiving exposures greater than 100 mrem / year, and their associated mun rem exposure according to work and job function is presented for Units 2 and 3 in Table III-A.
i B.
Wholebody Exposures Annual wholebody exposures for the year are presented in Table III-B, in accordance with 10CFR20.407(b).
C.
Single In-plant Personnel Exposure Due to an Outage i
Which Resulted in 10% of the Annual Limit A review of the radiation work permit (RWP) logs for the j
Peach Bottom Units 2 6 3 was made to determine which RWP's had a potential of producing an exposure of 400-1 500 mrem in a single entry.
This determination was based on site HP personnel's evaluation of the radiation conditions and work performed under each RWP.
These
)
RWP's were then reviewed to determine if any individuals received a whole body exposure of greater than 10% of j
annual limits (12 Rem exposure with NRC Form 4, or 5 Rem 3
exposure without NRC form 4 being used as the annual i
limits).
No single exposures were found in access of 10% of these annual limits for the year 1976.
i D.
Single Gaseous Releases, Due to an Outage which Resulted f
in 10% of the Annual Limit A review of the gaseous, particulate and iodine (with
{
half-lives of 8 days) and liquid radioactive releases i
for the year 1976 has determined that no single release was made that was greater than or equal to 10% of the annual limit.
4 I j I
l l
1
1 4
i PBAPS' 2
E.-
Liould Radioactive Release Data See Table III-C.
F.
Isotopic-Analysis of Liquid Radioactive Releases See Table D.
Isotopic Analysis o_f, Gaseous Radioactive Effluents G.
f
~See Table III-E.
l
_y H.
Gaseious Radioactive Release Data l
l
'See Table III-F.
I I.
Solid ~ Radioactive Waste Shipments See Table III-G.
l l
1 i
l 1 l 1
I a
. TABLE !!!-A PEACH BOTTOM ATOMIC POWER. STATION UNITS 2 & 3 FOR CALENDAR YEAR 1976 STANDAR0 FORMAT FOR REPORTING NUMBER OF PERSONNEL & MAN-REM FOR WORK & JOB FUNCTION NUMBER OF ' PERSONNEL () 100 MREM)'
TOTAL MfM-REM CONTRACT WORKERS CONTRACT WORKERS
- WORK & JOB FUNCTION STATION UTILITY AND OTHERS STATION trTILITY AND OTHERS Reactor Operations &
Surveillance Maintenance Personnel 1
102 63 1.14 30.11
'23 68 Operating Personnel
' 37 18 12 17.18 4.39 1 75 Health Physics Per.
8 0
81 2 75
.00 47.00.
Supervisory Personvl 1
0 2
.43
.00
.24
' Engineering Persor,nel 17 18 12 12.01 4.61 4.20 Routine Maintenance Maintenance Personnel-2 466 290
.75 238.03 172 54 Operating Personnel 4
1 5
.88
.59 1.22 Health Physics Per.
0 0
5
.00
.00
.73 Supervisory Persennel 0
0 0.
.00
.00
.00 Engineering Personnel 2
5 11
.23 2.05 3 18 Inaervice Inspection 7 tntenance Personnel 0
22 31
.00 5 97 24.88 Operating Personnel 0
1 2
.00
.16 76 Heal th Physics Per.
0 0
2
.00
.00 32.
Supervi sory Personnel 0
0 1
.00
.00
.13 Engineering Personnel 0
0 1
.00
.00
.22 Special Maintenance Maintenance Personnel. 0 0
76
.00
.00 54.09 Operating Personnel 0
0 0
.00
.00
.00 Heal th Physics Per.
0 0
1
.00
.00 31 Supervi ery Personne1 0
0 0
.00
.00
.00 Engineering Personnel 0
0 13
.00
.00 6.67 Waste Proces sing Maintenance Personnel 1
10 7
.14 1 75 1.25 Operating Personnel 7
0 1
7 54
.00
.10 Health Physics Per.
0 0
6
.00
.00 1 34 Supervisory Personnel 0
0 0
.00
.00
.00 i
Engineering Personnel 0
0 0
.00
.00
.00 Refueling Maintenance Personnel U
28 32
.00 5 95 12.42 Operating Personnel 5
0 I
.65
.00
.18 Health Physics Per.
0 0
7
.00
.00 3 62 Supervi sory Personnel 0
0 0
.00
.00
.00 i
Igid (See Notes I & 2)
'l Maintenance Personnel 4
538 439 2 30 307.17 301.68 Operatino Personnel 43 21 22 29 36 6.41 4 38 Heal th Physics Per.
8 0
91 2 94
.00 57 35 Supervi sory Personnel 2
0 6
31
.00
.73 Engineering Personnel 17 22 35 12 78 7 52 20.75 CRAND TOTAL 74 581 593 47.69 321.10 384.81 Notes:
1.
T'he totals for " Numbers of Personnel 7100 MREM" represent the total number of individuals accounted for in the work & Jcb function categories of the report (where an individual may be accounted for more than once) plus individuals with total exposures >100 MREM who are not accounted for in the work & job -
function categ2 ries of the report because they did not accumulate >100 MREM in any one work - function.
2.
The totals for " Total Man Rem" represent the total dose for all individuals reported - not only those doses '/100 MREM in a specific work - function.
i TABLE III-B RECORDED AN!iUAL WHOLEBODY EXPOSURE FOR' CALENDAR YEAR 197'6 lj 1
I PEACH B0TTOM ATOMIC POWER STATION UNITS 2 & 3 LICENSE NOS.:
ANNUAL DOSE RANGES NUMBER OF INDIVIDUALS l
(REM).
IN EACH RANGE NO MEASURABLE EXPOSURE 1349 I
i MEASURABLE EXPOSURE LESS THAN.100 855
')
1
.100 -
.250 356 1
500 341
.230
.750 215 500 750 -
1.0 173 1.0 2.0 173 2.0 30 19 4.0 4
30 4.0 50 6.0 0
l 50 6.0 70 0
8.0 0
I 70 9.0 0
l 8.0 10.0 0
90 11.0 0
10.0 11.0 12.0 0
12.0
+
0 TOTAL NUMBER OF INDIVIDUALS REPORTED 3485
____-_aa---
9 7
8 8
56 0
2 0
0 0
5 2
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+ +
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40 4
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0 1
8 5
8
+
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1 3
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3 4
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1 7
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9 L
3 7
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3 6
1 3
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0 9
J 1
9 3
5 2
6 3
1 1
2 2
S T
I N
U M
O TTO B
9 0
0 0
9 H
4 5
9 4
3 8
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5 1
9 1
2 A E 4
7 1
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6 E
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m 3
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TABLE III-E I
Peach Bottom Units 2 & 3 - Isotopic Analysis of Gaseous Radioactive Effluents (in Curies) 1976 l
i ISOTOPE JULY A LE.
SEPT.
OC T.
NOV.
DEC.
Ci Total l.2E+01 3.hlE+01 1.14E+02 8.7E+00 8.45+00 1.77E+02
,Kr yp ton-85 M Xenon-133 4 3E+03 5.5lE+03 8.lE+03 9.0 E+0 3 1.0 9E+04 9.2E+03
- 4. 70 E+04 Xenon-135 1.4.E+02 5 32E+02 5.62E +02 1.64E+03 2.8C +0 2 3 5E+02 3.5lE+03 Kr-88 5 3E+00 l.lE+02 6.8E+00
- 1. 2 2E+0 2 Total 4.44E+03 6.06E+03 8.70E+03 1.0 9E +0 4 1.12E+0 4 9.56E+03 5.0 9E +04 l
Iodine 131 '
~
7 77E-02
- 9. 93 E-0 2 1 38E-01
.lE-02 1.85 E-0 2 f.64E-02 3 91E-01 lodine 133 4.54E-02 3.6 3 E-0 2 4.54E-02 1.llE-01 1.llE-01 1.3 9E-01 4.88E-01 Iodine 135
- 2. 78 E-0 2
- 2. 23 E-0 2
- 2. 78 E-0 2 4.10 E -02 4.10 E-0 2 5.13E-02 2.llE-01 Total 1.51E-01 1 58E-01 2.llE-01 1.93E-01 1.71E-01 2.07E-Ol 1.0 9E +00 Strontium-89 4.6 7_,-'
5 6.40E-05 1.0 3 E-04 1 35E-04 1.51 E-0 4 1 70 E-04 6.07 E -04 S t ront i um-90 1.72E-06
- 1. 76 E-06 2.66E-06 2.43E-06 9.82E-06 6.54E-06 2.49E-05 l
Cesium-134 4.42E-04 4.51E-03 3.5 9E-0 3 9 25E-04
, 2.10E-04 9.56E-05 9 77E-03 Casium-137
- 2. 36E-0 4
- 1. 4. E -0 4
- 1. 27 E-0 2 Lanthanum-140 f.83E-05 3.03E-05 2 312-05 5 37E-05 1.25E 04 Cobalt-58 1 33E -04 2.63 E-0 5 5 0?E-05 2.10 E -0 4 Cobalt-60 3,31E-04 2.5 9E-04 3 23E-04
- 7. 28 E -0 4
- 4. 54E -0 4 7.85E-04 2.88E-03 Zinc-65 1.25E-04 1.07E-03 6.0 l E -04 7.6 9E-04 4.3 5 E-0 4 7.44E-0 4 3 74E-03 Mangane se-54
- 5. l l E -05 2.12E-04 2.63 E -04 Chronium-Si 4.60E-04 3 75E-04 8.35E-04 Z i rconi um-95 9.56 E -0 5 6.27E-05 7.86E-05 1.0 4E-0 4 3.41E-04 Mol ybdenum-99 1.0 4E-0 4 3.17E-03
- 3. 20 E -0 4
- 1. 44E -04 2 57E-05 3.76E-03 Sodium-24 5.58E-05 3.47E-05 7.82E-05 3.45E-05 2.0 3E-04 Neptuntum-239 7.23E-05 2.30E-04 3 02E-04 As-76 2.86 E -04
2.19E-05 2.19E -05 Total 1.88E-03 1 50 E -02 1.06E-02 4.68E-03 1.65E-03 2.63E-03 3.64E -0 2
- Less than minimum detectable activity
~
~
~
4 0
1 1
26 0
0 5
1 2
L 0
0 0
0 00 0
0 06 0
0 A
+
+
+
+
+
+7
+
+
T E
E E
E EE E
E E -
E E
O 4
2 50 4
0 01 6
0 1
1 T
9 7
9 8
3 9
7 0
0 1
01 6
1 3
1 7
3 3
35 9
1 1
(
2 0
39 0
0 36 0
4 2_
1 1
0 0
0 0
00 0
0 07 0
0
+
+
4
+
+ -
+
+
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E E
EE E
E E8 E
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4 8
4 2
26 9
6 51 6
2 0
7 6
9 51 E
5
- 0. 2 5
4 1
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1 3
21 1
2 31 1
1 1
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31 3
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0 47 0
1
+
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+
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8 52 4
9
- 8. 1 8
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0
+
+
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+
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E E -
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4 0
4 04 0
2 08 4
8 C
2 3
2 29 5
- 8. 0 9
1 1
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5 3
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1 1
A ddd ooo
=
T 4
0 36 0
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0 0
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0 46 0
0 iih 3D
+
+
+
4
+
07
+
+
rrt T
E E
E E
EE E
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+ -
E E
eee OE P
0 4
8 0
37 4
8 E9 0
9 ppm NS E
9 8
3 8
71 8
2 4
0 51 AA S
hho 3
91 2
1 79 2
7 t ti E
1 2
1 2L
(
nnt F
E ooa SR mmr I
T 3
0 2
1 27 0
0 4
1 2
I IE 0
0 0
0 00 0
0 06 0
0 66 y I
NV
+
+
+
+
+
+7
+
+
b UI6 E
E E
E EE E
E E -
E E
rr E
T7 G
4 1
3 4
88 0
3 58 2
0 ood L
MC9 U
7 0
9 2
9 0
0 3
0 ff e 71 1
B OA1 A
n A
T0 9
3 9
3 38 8
5 1
1 3
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PBAPS IV IRRADIATED FUEL EXAMINATIONS During the Unit 2 refueling outage of March 27 to June 24, 1976, 277 fuel bundles were sipped using the out-of-core wet technique.
Of these 277 bundles, 231 were of the improved 7x7 type.
Of this group, 19 were determined to be leaking.
l "A detailed fuel inspection of 10 of the improved 7x7 leaker bundles was performed at the site and included overall bundle visual examination, bundle disassembly and eddy current and ultrasonic non-destructive test (NDT) of each fuel rod, and detailed visual examination of all perforated rods as well as other selected rods.
l The inspections performed resulted in the detection fo 15 perforated rods.
The locations of the perorated rods in the bundles, visible characteristics of the perforations (short tight cracks), and the NDT signal response all indicate that the primary mode of rod perforation was Pellet Clad Interaction.
Visible evidence of minor secondary hydriding was observed on most rods.
j "A similar detailed pool-side examination was performed on two of the four unimproved 7x7 leaker bundles.
The results of this inspection revealed a total of 3 leaker
]
rods.
Two of the leaker rods (both in one bundle) j sustained open long splits characteristic of the PCI j
mechanism.
These rods were in the higher duty locations i
of the bundle.
The third leaker rod (in the second bundle inspected) exhibited a typically hydrided l
appearance and was in a relatively low duty location of the bundle.
The apparent severity of the PCI cra'ks in c
the unimproved 7x7 fuel was significantly worsa than that observed in the improved 7x7 fuel."
I f
Samples of leaking and sound fuel rods have been retrieved and have been sent to the General Elect.ric hot l
cell facility for further analysis.
i i
I
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l I l
4
0
.PBAPS APPENDIX A CORRECTIONS TO SEMI-ANNUAL EFFLUENT RELEASES REPOR_T
.W3 1 JANUARY _'1, 1976 THROUGH JUNE 30, 1976 J
i
.i 1
-__--.-..___.---=.N--E
PBAPS Two tables included in the Peach Bottom Atomic Power Station
" Semi-Annual Effluent Releases Report No.
1, January 1, 1976 through June 30, 1976," contained errors resulting from erronious sample flow. rates.
This Appendix contains a revised Table A " Peach BDttom Units j
2 and 3 - Gaseous Radioactive Release Data - 1976" and a revised Table B " Peach Ecttom Units 2 and 3 - Isotopic Analysis of Gaseous Radioactive Effluent (in Curies) 1976".
These revised tables reflect corrected release rates resulting from a recalculation using maximum attainable sample flow rates.
The matter was a'*o the subject of a letter dated January 3, i
1977 submitted under License Numbers DPR-44 and DPR-56 to the Puel Facilities and Materials Safety Branch, Region I, NRC Office of Inspection and Enforcement in response to Inspection Numbers
)
50-277/76-39 and 50-278/76-29.
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