Text

Annual Operating Rept 2 for Jan-Dec 1977
	ML20238A437
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1977
From:	Ullrich W; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
References
	EFF-77, NUDOCS 8708310077
	Download: ML20238A437 (42)
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PDR ADOCK 05000277 j

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PEACH BOTTOM ATOMIC POWER STATION UNITS NOS. 2 and 3 ANNUAL OPERATING REPORT NO. 2 Janu a ry 1,

1977 through December 31, 1977 Su bmit ted to The United States Nuclear Regulatory Commission.

Pursuant to Facility Operating Licenses Nos. DPR-44 6 DPR-56 Preparation Directed. by:

W.

T.-Ullrich, Superintendent Peach Bottom Atomic Power Station

PBAPS TABLE OF CONTENTS PAGE Introduction 1

Summary of operation 2

4 Unit 2 Operation Unit 3 operation 15' Personnel Exposure by Job Function 24 Wholebody Exposures 24 Liquid Radioactive Release Data 24 Isotopic Analysis of Liquid Radioactive Releases

.24 Gaseous Radioactive Release Data 24 Isotopic Analysis of Gaseous Radioactive Ef fluents 24 Solid Radioactive Waste Shipments 24 Revisions to Previous Semi-Ar.nual 25 Ef fluent Reports 1

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PBAPS l

1 TABLE OF CONTENTS PA'I4 TABLE 1

1 STANDARD FORMAT FOR REPORTING 26 NUMBER OF PERSONNEL & MAN-REM FOR WORK & JOB FUNCTION 2

RECORDED ANNUAL WHOLEBODY FOR 27 CALENDAR YEAR 1977 3

LIQUID RADIOACTIVE RELEASE 28 DATA i

4 ISOTOPIC ANALYSIS OF LIQUID 29 RADIOACTIVE RELEASES 5

GASEOUS RADIOACTIVE RELEASE 30 DATA 6

ISOTOPIC ANALYSIS OF GASEOUS 31 RADIOACTIVE EFFLUENTS i

7 SOLID RADIOACTIVE WASTE 32 SHI PMENT i

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PBAPS l

. INTRODUCTION l

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.f acility is owned by, and licensed to, Philadelphia l

Electric Company, Public Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electric Company.

Philadelphia Electric Company -is the facility operator..

This report covers the period from January 1 through December 31, 1977 and contains the last yearly compendium of l

Peach Bottom Unit 2 and 3 operations.. Starting with January, 1978, monthly narrative descriptions will be prepared as part of the NRC Monthly reporting obligations and no yearly summaries will be prepared.

This change to the annual and monthly reporting requirements became ef fective with the issuance of Amendment Nos. 37 and 37 to License Nos. DPR-44 and DPR-56 on December 13, 1977.

1

PBAPS SUMMAPY geach~ Botton Units 2 and ' 3 LJ1100 MKe BWP's)

Each Peach Bottom unit experienced a ref ueling/ maintenance outage during 1977.. In addition to.a1140 day refueling / maintenance outage, Unit 2 experienced eleven other outages with durations extending up to.eight days.

Unit 13's refueling / maintenance outage was in progress' at the i

beginning of the year 1and extended 101 days into 1977.

I There were eleven additional outages in 1977 with the

.l longest being nine days.

The 1977 Unit 2 short outages resulted from:

(A)

Turbine control valve and combined intermediate valve problems (Three outages - totaling 13 days) ~

l (B)

Recombiner condenser leak (One outage '- two days)

(C)

Power load unbalance relay operation (One outage. -

less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

(D)

Neutron monitoring noise flux spike (Two outages -

totaling less than three days)

(E)

General Electric Company End of Cycle testina (Two outages - less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

(F)

Recirculation loop valve packing leak (One' outage

- two days)

(G) surveillance testing (one outage - less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

During the Unit 2 refueling / maintenance outage, various major modifications and maintenance jobs were completed such j

i as:

l (A)

Core spray system spool piece replacement' (B) control rod drive return nozzle crack removal (C)

Removal of source holders (D)

Control rod drive replacement. (20 drives)

(E)

LPRM replacement.

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PBAPS' 1

(F)

RHR system flow orifice installation (HPSW side)

(G)

Diesel generator overhaul (four diesels)

(H)

Condenrate demineralized valve replacement (I)

Local Icak rate testing. of containment penetration' i

(J)

Fuel handling and sipping i

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l The 1977 Unit 3 short outages resulted from:

f (A)

Condenser vacuum leak (One outage - two days) j (B)

Primary containment isolation system relay fire f

t (One out age - two days)

(C)

Electric ~.1 switching transients (Two outages -

totali'49 two days) j (D)

EBC eyhtem oil leak (One outage - two days)

(E)

RCIC system-inoperable (One outage - three days)

(F)

Drywell to torus leak (One outage

.two days)

(G)

Steam leak in the drywell (One outage - six days)

(H)

Feedwater heater leaks (One outage - nine days)

(I)

Reactor water chemistry problems (Two outages -

totaling six days)

During the ref ueling/ maintenance outage for Unit 3 the f ollowing additional major work was completed:

1 (A)

LPRM replacement (B)

Inspection and maintenance of feedwater spargers (C) core spray nozzle spool piece replacement (D)

Control rod drive return nozzle crack repairs and l

special testing (E)

Local leak rate testing and containment ILRT l

c 3-l l

I

PBAPS Unit 2 UNIT 2 - OPER ATIONS on January 1, 1977, a Unit 2 shutdown was initiated becausa of difficulties associated with the Turbine Electro-Hydraulic Control '(EHC) System.

The number 3. turbine control valve had f ailed open.-

This f ailure was caused by l

water contamination of the EHC fluid.

During the shutdown a l

Group I isolation and. scram occurred.- These'uere initiated by reactor low pressure which was caused by the spurious opening of another' turbine control valve.

Investigation of l

the initial problem indicatedisignificant water in the EHC -

fluid which was traced to an EHC, cooler. leak.

Two leaking tubes in the EHC heat exchanger were plugged.

On January 3, information was received from Target Rock Corporation, which indicated that the air operator diaphragms removed from four valves which were rebuilt following a November 1976 outage showed signs of-dete ri or ation.

Two of these valves had been in service. for.

over two years, while the other two had been installed during a May 1976 outage.

An inspection of the diaphragms on the Unit 2 relief valves was begun.

The air operaters were removed f rom the valves, were inspected and repaired.

Several of the diaphragms removed during this inspection shcwed some sign of deterioration of the diaphragm material.

The valve operators were reinstalled and leak tested.

The reactor was taken critical on January 5,1977 and all eleven relief valves were test operated at approximately 150 psig during the reactor startup.

At approximately 5:30 p.m. on the same day, after the reactor had achieved normal operating pressure, the 71E relief valve spontaneously opened causing a reactor blowdown.

The reactor scramed shortly after the blowdown on a safety system action caused by low level.

Following reactor cooldown, a11' relief valve pilot valves were leak tested in place.

These tests indicated that the-D, E and K pilot valves had excestive leakage.

These' valves or their valve operators were replaced and the reactor was taken critical on January 8.

The relief valves.were again tested at approximately 150 psig during the startup. l 1

i

i PBAPS Unit 2 l

Turbine operation was delayed when an EHC oil sample again indicated wat er cont amination.

An additional leak war identified in the heat exchanger.

This heat exchanger wac replaced with a heat exchanger from Unit 3 which was being refueled at that time.

The EHC oil reservoir was drained, cleaned and refilled with new EHC fluid.

The generator was synchronized on January 10, 1977.

Full power was achieved on January 16.

On February 1, increased conductivity in the condenser and reactor were noted.

Af ter some investigation, the source of high conductivity was traced to a leak in the recombiner condenser.

Temporary corrective action was taken by valving the recombiner system drains to the Radwaste System.

The i

maximum conductivity in the reactor during this transient was approximately 3.7 umhos above the specified limit of 5 umhos.

The pH of the reactor water was within limits.

Within sixteen hours following the transfer of. water to ths Radwaste System the conductivity was within limits.

Late on February 4, the plant was shutdown to permit repair of the recombiner condenser.

During the outage two leaking U tubes in the recombiner condenser were plugged.

Three main condenser waterboxes were inspected and a Jeak was l

identified and repaired in the C-2 water box.

A drywell

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entry was made and a packing leak on the vessel head to main l

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steam line vent valve was repaired.

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The reactor was critical at 2: 30 p.m. on February 6, and the

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generator synchronized at 8:45 p.m.

Operation during this period was with a reactor water conductivity of l

approximately 3.5 umhos.

The pH values were within limits.

Ef forts were made to locate the source of in leakage which had created the conductivity above that usually experienced.

l By February 8, the reactor conductivity had increased to 8.5 umhos.

Load was reduced to 300 MWe to permit removal from service of the A-1 and A-2 condensers simultaneously.

A l

large tube leak was subsequently identified and corrected in the A-2 water box, tio leakage could be found in the A-1 water box.

Following repair of the A-2 water box, load was increased in accordance with preconditioning requirements.

Reactor conductivity subsequently returned to less than 5 umhos.

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PBAPS Unit 2 i

At 1:15 p.m. on February 9, the turbine generator tripped at a loa 3 of approxi

'+.ely 600 MWc.

Plant instrume nt at ior, indicated that the tut oine trip was caused by the power load unrelance relay associated with the turbine generator protection system.

The turbine trip caused a reactor scrum as well as a Group I, II and III isolation.

An investigation of the power load unbalance trip did not Following the j

identif y any equipment or system malfunction.

turbine generator trip, the generator disconnect was opened and an attempt made to reclose the generator breakers.

i Generator circuit breaker No. 225 was successfully closed; however when Generator circuit breaker No. 215 was closed both circuit breaker Nos. 225 and 215 tripped.

No cause for this tripping was determined.

These breakers were successfully closed after the investigation.

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The reactor was made critical at 10:50 p.m. on February 9, I

1 and the generator was synchronized at 3: 50 a.m. on February 10.

Full load was achieved on February 16.

During this period the reactor conductivity varied from 1.4 umhos to 2.2 umhos.

Several small leaks in the condenser were causing this higher than usual conductivity.

The pH values remained within Technical Specification requirements.

on March 2, during routine testing of the turbine combined intermediate valves (CIV), the number 5 CIV stuck in the 85%

open position.

The Unit was removed from service at 5:03 pm on March 3 to identify the cause of the problem and make repairs.

The number 5 CIV servo was replaced and the valve testes satisf actorily.

The reactor was critical at 5:10 a.m. on March 4, and the generator was synchronized at 9:25 a. m.

During the rise to power, a discrepancy in the heat cycle flow was identified.

This resulted in the determination that significant bypassing was occurring in the 3A feedwater heater as a result of tube f ailures.

A power reduction was taken and the ' A' feedwater heater string was isolated on March 8.

With the heater string isolated, power increase continued to a maximum of 942 MWe.

Operation at this power level continued from March 10, through March 13.

On March 13 the number 5 CIV on the main turbine again stuck in a PBAPS Unit 2 I

partially closed position.

Shutdown was initiated with the Unit. being removed f rom service at 10: 35 p. m.

During this outage, corrective and preventative tube plugging was performed on' the 3A f eedwater heater.

Investigation of th^

number 5 CIV failure, identified a badly scored hydraulic actuator cylinder.

Since Unit 3 was-being refueled, a similar assembly was transferred from the Unit 3 turbine to expedite restart of Unit 2.

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On March 16, the reactor was made critical and the turbinbe was synchronized at 2: 40 p.m.-

Full load was achieved on l

March 24.

On April 3,1977 the drywell floor drain sump pumpout volume showed an increased value.

By April 4, the unidentified leakage had reached 4 gpm.

During checkout of the LPRM inputs to the temporary data collection computers associated with special end of cycle testing, a false APRM flux spike developed inadvertently which resulted in a l

reactor' scram at approximately 10: 40 p.m. on April 4 The plant was depressurized and a drywell entry was made on April 5.

The primary source of the increasing drywell floor I

drain sumps pump rate was identified as a packing leak on the

'A' recirculation pump suction valve.

Additionally, the Reactor Water Cleanup System isolation valve inside the drywell had f ailed in the closed position.

The leaks identified were corrected and the isolation valve motor replaced.

The reactor was critical at 10:00 p.m. on April 6.

The turbine generator was synchronized at 2: 42 p.m.- on April 7.

Power was increased toward the desired End-Of-Cycle turbine trip test condition (50% power,100% flow).

This condition provided proper range from the various core parameters such that no Technical Specification violations would occur as a l

result of the testing.

The End-Of-Cycle pressure perturbation and turbine trip testing was performed on April 9.

The turbine was tripped at 5:43 a.m.

A restart was immediately undertaken and the i

Unit was resynchronized at 11: 40 p.m.

Power was. increased toward the first stability test condition of 62% power and 50% flow. _ - _ _ _ _ _ _ _ -

PBAPS Unit 2 Reactor power operation continued through April 14, at increasing power levels.

On thi s date, load was reduced to permit End-of-Cycle stability testing.

This testing continued through April 16.

No operating conditions were experienced which resulted in unstable operation.

On April 16, reactor power was further reduced in order to permit a turbine trip test from 15% power.

This test was performed at 1:15 p.m.

The turbine generator was resynchronized at 1:25 p.m.

Fall load was achieved on April 21.

On April 22,. load was again reduced to establish reactor conditions to support an End-of-Cycle turbine trip test f rom o 60% power, high flow condition.

The trip test was preceded by perturbation tests.

The turbine trip test was performed at 1:15 a.m. on April 24.

The reactor was critical at 11:34 a.m.

The generator was synchronized at 4: 26 p.m.

The End-Of-Cycle turbine trip test from 70% power was l

conducted on April 27.

Unit 2 load was reduced at l

approximately f:00 a.m.

for pressure perturbation testing which was perf ormed prior to the turbine trip test.

The turbine trip test was performed at 11: 05 p.m.

Following the turbine trip, a special shutdown procedure was used to maintain reactor pressure as long as possibic.

This procedure permitted the testing of various relief valves to correlate torus reaction to relief valve operation with other tests performed on Mark 1 containments.

Maintaining the reactor pressure also tended to reduce the increase in iodine concentration in the reactor water following the shutdown.

With the completion of relief valve testing, reactor pressure was permitted to decay.

By April 29, the reactor was depressurized and shutdown cooling established.

This was the beginning of the Spring 1977 Unit 2 refueling / maintenance outage.

The reload for this refueling consisted of 172 bundles of 8X8 fuel.

During the outage, 421 bundles were sipped with twelve bundles being identified as leakers and removed from further service.

Between April 30 and May 10, the drywell head, the reactor vessel head, 8-

1 PBAPS-Unit.2 and' the dryer and separator were removed in preparation for.

in-ves sel wort..

Following hydrolysing and draining of the reactor vessel to approximately twelve inches below the control. rod drive return line nozzle, a PT inspection was performed on the' nozzle..This inspection identified multiple cracks on tha blend radius and in the. bore section of the nozzle.

Following this examination,. grind-out of these cracks was

~

initiated.. This work was performed from May 10, through May

17.. On May 17, 1977 additional cracks were identified below the return nozzle in.the vessel cladding.- These cracks wers L

removed by. grinding in accordance with recommendations 'of.

the nuclear steam supply system vendor.

.The. maximum depth of any. one crack was 0.9 inches, measured from the surface

)

of the. cladding.-

During one of the PT examinations,-

additional cracks below the CRD nozzle outside the eight' inch diameter. inspection circle were identified.

These 1

1 cracks (approximately eight) were from one inch to sevr:n inches long and generally were in a horizontal _ direction.

The cracks extended down below the. bottom edge of the CRD nozzle approximately eight inches.

l During this period, In-Service Inspection (ISI) of the core spray piping was' conducted.

On May 17, the inservice inspection agent stated that a crack-like indication was present on the

'B' loop spool piece to elbow weld.

Additional radiographs were taken which tended to confirm the UT analysis.

A meeting was held on this topic on May 20 with the ISI agent, the nuclear steam system supplier and the Philadelphia Electric Company. metallurgist.

Agreement-was reached that the suspected weld on the'B' loop would be removed and investigated and proper repairs made..

Additional consideration was given to. performing the same -

type of. work on the ' A' loop.

Following the completion of CRD nozzle grinding, the reactor.

vessel and reactor head cavity were flooded.

CRD Fuc3 replacement operations were also. started on this ~ date.-

handling operations were started on May. 22.

The inspection of the control rod drive units removed from the reactor identified five collets with circumferential cracks in the

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PBAPS Unit 2 i

crea of the lower shoulder.

These cracks varied in length with a maximum of one inch.

The control rod drive replacement operations were completed on May 27.

Control rod drive units were replaced on twenty rods, on May 26, during removal of source holders, the fourth source holder to be removed came apart.

The upper section i

was removed and the lower section remained in the reactor.

A TV inspection of the remaining portion, as well as the other three source holders, indicated that significant cracking of the stainless steel sleeves existed.

A special tool which permitted removal of the remaining source holders intact was requested of the nuclear steam system supplier.

Procedures for recovery of portions of the broken source holder were written.

LPRM changeout was started on May 28.

Work was interrupted because of a cloudy condition in the reactor water on May 31.

The deterioration of water quality is believed to have been caused by an RHR heat exchanger leak.

Because of the desire to proceed with the core spray piping work and the time required to re-establish water clarity, reactor level was reduced and core spray maintenance work was begun on June 2.

With the water level reduced to approximately one foot below the control rod drive return line nozzle, work proceeded on both the core spray piping repair as well as the removal of cracks below the control rod drive return line nozzle.

By June 9, 1977 both core spray spool pieces had been removed and buttering of the inside of the elbow and reducer was in progress.

During welding, some dif ficulty with weld porosity was encountered.

Cracks in the vessel wall below the control rod drive nozzle had also been removed except for an area marked for boat sample removal.

This boat sample was removed on June 9.

Maximum depth of the cracks identified was approximately 0.4 inches.

The nuclear steam system supplier has indicated that the cracks were caused by high cyclic thermal fatigue.

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PBAPS l

Unit 2 1

i Core spray work continued through July 9, at which time both spool pieces were welded into the line such that ISI work in-the vessel could becin.

Final radiographs on the core spray spool pieces were taken on July 12.

On July 14, the core spray system was tested successfully.

On July 17, operations associated with removal of the remaining source holders was started.

This work was done using a special tool which gripped the source holders f rom the bottom.

Work was completed on July 20, except for the removal of the debris left above the fuel support pieces from the_ broken source holders.

Removal of the debris from cbove the fuel support castings was completed by July 22.

l Reactor vessel water level was increased to the refueling floor elevation on July 22.

Following c'..arification of the water, fuel movement associated with the emptying of the cells surrounding the four damaged source holder locations was initiated.

Removal of fuel support castings, control rod drive blades, and debris from these locations was in i

progress from July 25, through August 5.

The cells surrounding the f ailed source holders were emptied and debris on the core support plate and in the control rod guide tube removed.

The performance of this work also 4

resulted in the requirement to replace CRD 42-35 since a part of the debris became lodged in the upper spud of the drive.

With the completion of fuel handling, core verification was initiated.

This identified two fuel cells which were not properly seated.

Fuel from these cells was removed, the fuel support piece properly seated and the fuel returned to its original position.

This portion of the core was reverified on August 24 On August 24, recovery f rom in-vessel work was begun.

A reactor hydrostatic test was started on September 4, and continued through September 6.

Startup activities on Unit 2 continued from September 8, l

until criticality was achieved on September 12, at 7:03 p.m.

l The turbine generator was synchronized at 11:50 p.m. on l l

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i PBAPS

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Unit 2 September 14.

Load was increased to approximately 180 MWe i

and held at this level during LPRM calibration.

Balancing l

data taken during initial turbine generator operation indicated a need for additional weignts in the number 2 coupling.

The Unit was theref ore removed from service at j

4: 15 a.m. on September 16.

The length of time the Unit was i

out of service was extended until 5:34 p.m.

for repair of an l

EHC oil leak in the relay trip valve in the turbine front standard.

With the Unit operating at 448 MWe on September 18, a high activity alarm on the main steam lines occurred.

Reactor l

power was reduced.

The high activity was accompanied by an increase in reactor water conductivity.

Investigation indicated that this disturbance of primary coolant purity was coincident with placing the F' condensate demineralized in service.

It was determined that the

'F' condensate demineralized had several defective gaskets permitting resin to enter the reactor.

Load was reduced to approximately 303 MWe and held at this level until primary coolant purity was again within Technical Specification limits.

Load was again increased.

Electrical power of 868 MWe was achieved on September 24.

Load was then limited by the availability of condensate demineralizers.

Unit 2 operated at about 925 power (1000 MWe) through October 28 when load was reduced to 400 MWe for a rod seguence exchange.

The Unit was then placed on a preconditioning ramp until 1:25 a.m. on November 3, when a l

reactor scram occurred.

The scram was caused by a loss of reactor water level signal and subsequent increase in feedwater flow.

The flow of cold water increased flux to the trip point of the APRMs and the scram occurred.

The Unit was at about 51% power (515 HWe) at the time of the trip.

The source of the problem was subsequently traced to a number of loose connections in a control room panel.

A restart commenced at 8:09 a.m. and the Unit was on the bus at 5:41 p.m.

At 2:10 p.m. on November 9, the 2B condensate pump tripped on

'B' phase differential relay operation.

Plant load,

PBAPS Unit 2 increases were not af f ected at the time of the trip.

Due to the loss of the condensate pump, Unit 2 load was limited to about 900 MWe.

The 2B condensate pump was returned to service at 7: 00 p.m.

on November 10, and preconditioning was resumed.

Unit 2 reached a maximum load of 96% power (1035 MWe) on November 12.

At 7:00 p.m. on November 17, load was decreased to 33% (300 MWe) to allow access to the MSIV room to repair a RCIC motor operated valve which had f ailed an operability test.

The problem was traced to moisture in the valve control housing causing damage to the motor and torque switch.

The motor and torque switch were replaced.

The valve was repacked with modified packing and returned to service by 8:30 p.m.

on November 18.

Unit 2 reached a maximum load of about 93% power (1010 MWe) on November 22.

Load increases were stopped due to condensate demineralized dif ficulties.

At 11:30 a.m. on November 30, a drywell high pressure alarm annunciated.

The RWCU system was manually isolated and other suspect valves in the drywell were backseated to stop the leak.

A load reduction was begun and venting of the drywell to SB3Ts initiated.

The drywell sump pumpouts increased to a rate of approximately 8 gpm.

A normal plant shutdown was begun.

The turbine generator was off the bus at 2:05 p.m.

The cause of the leak was a blown packir.g on recirculation loop valve 65B.

Repairs were made and ' cod withdrawal was begun at 6: 22 p.m. on December 1.

A 6rywell inspection took place at 350 psig reactor pressure and verified no leaks.

The Unit was back on the bus at 8:53 a.m. on December 2.

Power level was limited until December 10, due to condensate demineralized availability.

The unit reached 98.5% power early on December 11.

Unit 2 operated at about 98% power (1068 MWe) until 3:14 p.m. on December 13, when a scram occurred during instrument valving associated with surveillance test.

A restart was begun and the reactor was critical at 3:10 a.m. on December - - - _ _ _ _ _ _ _

PBAPS Unit 2 14 The generator was synchronized at 8:50 a.m. on December 14.

Unit 2 continued power increaseE, reaching 96% power (1034 MWe) on December 21 when it was necessary to drop load 100 MWe due to a vacuum loss f ollowing a recombiner mechanical compressor oil change.

The unit was returned to 96% power early on December 23.

Unit 2 continued operation at about 97% power (1055 MWe) until 7:30 p.m. on December 27, when a 100 MWe load drop was taken to regenerate a condensate demineralized.

Load was increased at 5 MWe per hour following the regeneration.

On December 28 at 10:03 p.m., Unit 2 load was decreased to 900 MWe to perform a feedwater heater leak test.

The results of this test were inconclusive and the load was increased at 5 MWe per hour.

On December 29 at 8:4 0 p.m.,

Unit 2 load was again decreased to about 840 MWe and a series of tests revealed a significant feedwater heater leak in the

'B' heater string.

The

'B' heater string was isolated and operation continued at reduced load through the end of the year.

14 -

PBAPS Unit 3 Unit 3 operati one The first refueling / maintenance outage for Unit 3 began on December 24, 1976.

Following MSIV testing, work was begur.

on December 30, 1976 in preparation for entering the reactor vessel for refueling and maintenance.

Fuel handling began i

on January 3, 1977 in preparation for a reload consisting of l

188 bundles of 8X8 fuel.

LPRM replacement and fuel sipping

{

were performed simultaneously.

Of the 172 fuel elements i

that were sipped, three elements were identified as leakers.

l Fuel sipping was completed on January 16.

l Following the verification of LPRM latching, the reactor I

vessel level was lowered to inspect the feedwater sparger i

nozzles.

A PT examination of selected areas of the

'D' and

'F' nozzles identified slight indications which were removes j

by a flapper wheel.

A repeat of the UT examination showed l

that the UT reflectors had not been completely removed.

The f eedwater sparger work platform was installed on January 16, 1977 and the ' D' and ' F' cpargers were removed from the nozzles on Jan uary 19, 1977.

An extensive PT examination of the

'D' and ' F' blend radius and bore area showed minor indications which were removed by light grinding.

None of these indications penetrated the cladding.

The condition of l

the spargers was excellent.

Both spargers were reinstalled.

UT examination of the nozzles still showed the reflectors.

Since these indications could not be located by PT examination, a decision was made to return the Unit to service when this refueling outage was complete and re-examine these two spargers at the next refueling outage.

The NRC concurred with this decision.

During the f eedwater sparger work, additional UT examination of core spray piping indicated possible cracks in the heat effected zone of the core spray line at a reducer to pipe spool weld.

This UT data was verified by radiography on January 22.

Following completion of the radiography, the reactor vessel and head cavity were again flooded to perndt fuel handling operations.

The refueling operation took p3 ace between January 23 and February 3.

. l

PBAPS Unit 3 locations were verified on. February 5, and The. fuel. element the reactor' head cavity level was -lowered to a level bclow the core spray nozzle to permit repair of the weld crack identified in the

'A' core spray loop.

Radiography on the core spray piping verified UT data which indicated a

'B' crack in a similar location on the

'B' line.

The primary activities associated with the Unit 3 outage f rom. February 10, to March 7, involved the repair of the core spray piping.

Both core spray lines were cut and the l

l pipe. spools removed.

Additional circumf erential and axial i

l cracks were identified in both the ' A'. and

'B' core spray elbows in the heat effected zone of the weld to the spool piece.

.The core spray piping repairs were essentially completed by March 7.

Radiography and repair welding continued through March 18.

A PT inspection on the Unit 3 control rod drive return line nozzle was performed on March 8, 1977.

This inspection identified a number of cracks.

These cracks were f airly uniformly distributed around the circumference of the Grind-out of the cracks was started on March 9, and i

nozzle.

was completed on March 17.

The deepest crack was approximately 7/8 inches deep including the thickness of the cladding.

f ollowing the completion of in-vessel work and On March 18, l

radiography on the core spray piping, the reactor vessel' level was increased and control rod drive replacement started.

During this period, preliminary special tests on the control rod drive system were performed to verify that acceptable drive performance could be attained with the control rod drive return line isolated.

The results of this testing were satisfactory, on March 29, preparation for a vessel hydrostatic test was Additionally, twelve selected control rods were begun.

l stroked both with the control rod drive return line in service and isolated.

No significant difference in control rod drive performance was noted during these tests.

1,

l PBAPS Unit 3 l

Startup activities continued and on April 11, the reactor I

wa s made criti cal.

Several di f ficulties associated with recombiner operations, high radwaste. inputs and high condenser air in-leakage were identified.

On April 12, because of continued difficulties in-maintaining condenser vacuum, reactor pressure was decreased to less than 600 psig.

Several condenser vacuum leaks were identified and corrected.

Following these repairs, reactor pressure was increased to operating pressure and the turbine generator synchronized.

At 9: 30 a.m. on April 13, the packing on one of the turbine stop valves was found to be i

very loose.

To safely tighten the packing follower, the turbine generator was removed from service and. the stop valves closed.

Additionally, the repair of several valve packing leaks in the air ejector system required closure of the MSIVs.

During this period a blockage in the recombiner line between the mechanical compressors and the holdup volume was determined to be in a diffuser inlet pipe to the holdup volume.

The diffuser was found essentially rusted closed.

Following repair of the air ejector valves and removal of the diffuser, the MSIVs were reopened, operating pressure established and the turbine generator i

resynchronized at 1:25 p.m. on April 15, 1977.

Power was increaseG to 2 5%.

On April 18, with the Unit at 25%, a small fire occurred in a relay cabinet associated with isolation circuitry.

A reactor shutdown was initiated because the full extent of the damage was not known.

The Unit was tripped at 2:11 p.m.

and all rods f ully inserted by 4:30 p.m.

The damaged relays and wiring were removed and replaced with new components.

Propagation of the fire was traced to the relay l

manuf acturers use of a flamable relay contact-arm retainer.

The replacement of such contact-arm retainers was initiated.

Surveillance tests were performed on the isolation circuitry and preparation was made for a restart of the reactor.

The reactor was critical and the generator synchronized on April 20.

Power had reached 726 MWe by April 26.

Power increase was then temporarily halted to identify and repair a condenser leak in the B2 waterbox.

Following condenser leak repairs, the power level of Unit 3 continued to be,

i

~

PBAPS I

Unit 3 increased.

A maximum load of 996 MWe was achieved on April 29, with the existing rod pattern.

Since the recirculation.

flow was at a maximum, power was reduced and additional roir were withdrawn.

This operation resulted in achieving 98% of f ull load (1060 MWe) on May 4.

On May 22,1977 a leak in the B1 condenser necessitated a load reduction to approximately 750 MWe to allow entry into the condenser waterbox and plugging of the failed tubes.

Following these repairs, the Unit was returned to full load.

Operation of Unit 3 oontinued at approximately 985 power f rom May 25 through 27.

On May 27, the

'B' reactor feedpump ran back to minimum speed.

To maintain reactor vessel level the operator immediately dropped approximately 270 MWe.

Repairs were made and load was again increased in accordance with preconditioning requirements.

Later on the same day, power was reduced to approximately 342 MWe to accommodate a rod sequence exchange.

This sequence exchange was completed and the power level was increased with f ull load achieved on June 4 On June 7, at approximately 2: 40 a.m. the Number 2 startup feed (220KV line 220-08) tripped due to difficulties in the Graceton Substation.

This resulted in an automatic transfer of the 4KV busses.

Following completion of this transfer, the reactor scrammed on high neutron flux, because of a speed increase on the recirculation pumps caused by an instrument upset associated with the loss of power during the transfer.

Rod withdrawal and return to power were delayed until the E3 diesel could be returned to service and the high pressure service water system could be normalized.

The E3 diesel had been removed f rom service on June 6, to start its annual j

maintenance outage.

The diesel was reassembled and tested following the scram, prior to rod withdrawal.

The high pressure service water had been used on the previous day to

{

supply cooling to the Unit 3

'D' RER heat exchanger, thereby i

permitting mud removal operations on the Unit 3 intake structur e.

The mud removal operation was halted and the system normalized prior to startup.

The reactor was made I !

l

PBAPS Unit 3 R

critical and the Unit synchronized on June 8, 1977.

Full power was' achieved on June 13, 1977.

On June 14, at 12: 45 a.m., the turbine generator tripped and the reactor scrammed.

Turbine trip was initiated by a f alse power-flow unbalance due to the simultaneous closure of number 5 Combined Intermediate valve (CIV). and number 2 CIV.

The number 2 CIV had been closed as part of the routine turbine testing program.

The number 5 CIV closed because of an EHC fitting leak on the valve control piping.

The oil leak was corrected and similar fittings in other valves were replaced.

The reactor was taken critical on June 15, and the turbine generator synchronized.

Full load was achieved on June 20.

l Unit 3 continued full power operation from June 20, through July 1 when leakage of resin through two -condensate demineralizers resulted in an increase in reactor coolant conductivity and a decrease in pH values.

This required a plant load reduction until reactor coolant chemistry could be returned to within Technical Specification values.

On July 5, at 8:30 p.m., the Number 2 startup feed (220-08) line tripped.

This resulted in isolation of the Instrument Nitrogen System for containment.

Loss of air to the main steam isolation valves eventually permitted two of the valves to drif t closed.

This caused an increase in reactor pressure such that a reactor scram was caused by high flux.

Instrument nitrogen was restored and the reactor was made critical on July 6, and the turbine generator synchronized.

Full load was achieved on July 11, at which time difficulty This was experienced with the RCIC inner isolation valve.

led to declaring the RCIC inoperable on July 13.

The Unit continued full power operation through July 21.

A temporary Technical specification Change was requested from the Nuclear Regulatory Commission on July 19, to permit l

continued operation of the unit during a peak power. demand period on the East Coast.

This Technical specification Change was approved on July 20, and permitted continued I

operation provided the HPCI was tested daily.

On July 21, the HPCI f ailed the surveillance test.

A shutdown was I l

s

.?

I PBAPS' Unit 3-

' initiated immediately.

The turbine generator was-removed from service at 1:32 a.m. on July.22, h

During this shutdown,'the HPCI turbine control valve. shafts were replaced and adjustments were made to the HPCI turbine linkages.

A checkout of the HPCI turbine and testing during the' subsequent startup indicated.that adjustments made to the turbine linkage were successful.

Additionally, the RCIC.

isolation valves were repaired and proven. operable.

. Maintenance was completed and the reactor made critical and the Unit was synchronized on July 24.

By July 28, power had been increased. to 790 MWe.

Following the startup on July 24, high nitrogen makeup.

requirements to the drywell prompted an. investigation.

A shutdown was initiated on July 28.

A torus entry and.

J inspection was made.

No obvious cause could be identified.

L A zero differential pressure test for vacuum breaker l

operability did indicate some friction in the mechanism.

The torus to drywell vacuum. breakers were then-cleaned and lubricate d.

A torus to drywell leak test was performed and found to be satisf actory.

The~ reactor'was returned.to service and the turbine generator synchronized on July 31.

Full power operation was achieved on August 8.

On August 9, surveillance testing of the RCIC System identified an inoperable outer isolation valve.

The inner isolation valve was closed and the RCIC System declared inoperable. - Surveillance testing of the HPCI was i

successful.

Reactor power was reduced to 314 MWe on August l

12, to repair this valve.

Following repair and testing, power level was again increased with full load being achieved on August 17.

l During surveillance testing of the HPCI on September 1, the l

steam supply valve failed to-open.

The HPCI was declared L

inoperable and the required surveillance testing performed.

During the performance of the ADS Logic System Functional Surveillance Test, (required by HPCI being inoperable)

- setpoint drif t of the timers on this system was noted.

A power reduction was initiated until the timers could be properly adjusted and the surveillance test repeated. l

PBAPS Unit 3 i

l i

l operation of Unit 3, at essentially full load, continued through September 24 On September 10, the HPCI was again declared inoperable because of the f ailure of the inlet steam supply. valve to open.

Repairs were made and the HPCI returned to service on September 12.

On September 25, Unit 3 was removed from-service at 4:20 a. nu to accommodate a maintenance outage.

The primary activities during this outage were associated with correction of leaks in the heat cycle, correction of several l

steam leaks in the drywell, repair of RPIS instrumentation, and tack welding of snubbers in the.drywell.

During the outage, surveillance testing identified two MSIV problems.

One valve had a bad limit switch and another failed to reopen after test closing.

Both problems were repaired and reactor startup was begun with criticality achieved on September 30.

Startup operations included surveillance testing of the HPCI System at approximately 150 psig.

During the quick s'. art test of the HPCI, the turbine failed to produce the required flow. _This was caused by f ailure of the automatic control module in the flow controller.

Following replacement of this module, the test 3

was successfully completed.

The turbine generator was j

l synchronized on october 1.

Approximately 50% power was attained on october 3.

At that tire, load was reduced due to a main steam line activity increase caused by a primary coolant chemistry upset.

This was caused by injection of air or resin into the reactor vessel f rom the RWCU system following its return to a vessel to vessel mode of operation.

Following the return of primary coolant conductivity to normal on october 5, the power increases were continued.

On October 3 and 4, Unit 3 experienced an fodine release.

The rate of release was approximately 80,000 uCi per day which is 234% of Technical Specification limits.

By the morning of October 5, the rate was about 1200 uci~ per day which is less than 4% of the Technical Specification limit (See LER 77-04 9/1T-0 for Unit 3).

Iodine levels continued to drop.

This problem was caused by the venting of the RWCU

H 11 i

j y'

PBAPS 1,

i

[

'T Unit 3 T

ib(

.On October 4, a lo' ad. reduction was system heat exchangers.

required due to a trip of the recombiner on indication'of

-high hydronen concentration and' a slight loss of vacuum which followed.

The problem was traced to a closed valve on The.

a.' steam line return from the recc;nbiner preheater.

UnitJ3 valve: wasfopened;and normal operation continued. -

reached.f ull power ~ operation on October 10.

1 ni on November 7, a feedwater heater leak was suspected, based i

i on a disparity between dif ferent-- flow indications.

An investigation followed, which indicated a leak in the ' A'.

y" heater string.

Load was reduced to about.800 MWe and the

'A' heater string was removed from service.

Load.was then increased to about 980 -MWe.

Unit 3 continued operation at about 935 power '(980 MWe) until November' 26, when a controlled shutdown was initiated to repair heater leaks.

Unit 3 was. removed from service. et Unit 3 remained shutdown until1 December 5 to 10:16 a.m.

l allow repairs.

l 0

l 1

1 During the Unit 3 startup on December 5, a reactor scram l

f rom about 5% power occurred.

The scram was caused - by low '

l water level when two reactor feedpumps f ailed to respond to control signals.

The reactor 'was restarted at about 6:25

{",

The. generator was on the line at 9:07 p.m. on December 5.

i a.m. On December 6, but was removed from' service at 3:25 l

p.m. because of a reactor water chemistry problem caused by.

i resin f rom a condensate /deFdneraliEer.

This resin h8d l

1eaked from loose elements and was carried into the reactor -

vessel, At about the same time, the SPCI was declared inoperable due to a foreign material (cap screws) being.

f ound downstream of the HPCI turbine exhaust.

After

!i consultation with the turbine manuf acturer, the bolts were-l determined to be from support brackets. for flow reversing chambers in the turbine.

Unit 3 remained shutdown f or HPCI turbine repairs until early on December 11.. Unit 3 generator returned to service on December 12.

However, the Unit was limited to 60 -MWe due to cracked low pressure crossheads.on both recombiner l

mechanical compressors.

A shutdown was begun on December.

a s

5 1

1 l

P'itl'"

]

4 I

l t~ni t 3 l

l l

1 j

3 3 du i

tv r voi r '

c' u r i ': ' r '.'.

r n"-

r*

deprer.. 'r;;.

but * ' -

r. a r+ c r rortirnd criti m'.

7 T * ' 'T uer.iritralirer wrs r'*rc<' i ri snrvi rr-t r' r } "" un t h.

water.

':'hc c;)cris t rv irnrov"'! o ui cb.1v a n e' n nn.mr i-cren.'

was initiater!.

The orncrat or was returner? tn servic" n-De cerbe r i t,.

I, oar was sti]3 liritcd tr about 5; 0

"n dur t e-

.the recorbiner corpresr.cr nrohlern.

One cer"rcFrnr vne 1

returMu' te service at 11: On n.r.

on Dece-ber 1/ br+ n satis fi.ctory vacuor could not be estab}ished for newer Je"n' increases un til the second cornrcssor was ret"r-r0 to service at 11:30 p.rn. on December.16.

.The air eiector and j

recorbiner were placed in service and load increases benun.

l The init :tttained 995 pounr en recerber 25 and annrati'r at l

that Invel contipuce throuch the end of the v'er.

l 1

l l

i PERSONNEL EXPOSURES & RADIOACTIVE RELEASES l.

A.

Per sonnel Exposure by Job Function A tabulation of station, utility, and other personnel receiving exposures greater than 100 mrem / year, and their associated man rem exposure according to work and job f unction is presented f or Units 2 and 3 in Table 1.

1 l

B.

Whol ebody Ex posur es Annual wholebody exposures for the. year are presented in Table 2, in accordance with 10CFR20.4 07 (b).

1 C.

Liquid Ra dioactive Release Data See Table 3 D.

Isotopic Analysis of Liquid Radioactive Releases

)

See Table 4

E.

Gaseous Radioactive Release Data See Table 5 F.

Isotopic Analysis of Gaseous Radioa ctive Ef fluents See Table 6 G.

Solid radioactive Waste Shipments See Tabl9 7 1

l 24 -

..I

r7-i

' e'

,L REVISIONS LU F NEVI 6DS sE.MI-REIM EFFLUENT REPOR% '

Tables III-D and III-F f rom the July.,through - December, 1976.

m.-

i, i

semi-Annual Ef fluent Report are attached' as pages 3?/ and 34 '

respectively.

The Decu.ber entry for Cetity,m 137 tas been corrected on Table' II7~D

'Ihe' Mixed Noble Ga:s value for

^

1 Jply has been corrected;on Table III-F.

Table. B f rom the January t hrouch June, 1977 Semi-Annua 3 Effluent Report is attached as page_ 35.

The " Tota l" value !

of the Noble gas totals (Krypton and Xenon) was corrected on this table.

1 4

1

. l n

I l

. 1 l

l l

1 l

j l

1 l

i t

l

' I l

a. i l

i 1

4i a

u

T AB LE 1 PE ACH BOTTOM ATOMIC POTtR STATION UNITS 2 f, 3 FOR CALENDAR YEAR 1977 ST ANDARC FORMAT FOR REf 0R11N: NUMLER OF PER$0N'i!L AhL Mh+RER f 0f. W3ff A!d J01 Fuh',110N h;' ICE Or IEk',0n!L t > 1H KCW 101AL Ms.r.Ee (cht t F T W:t.US (Oc ;:. 1 w '... [ t.

W)k

JOE FL'hpl A 51 AllP'.

Liltllt Ahi Cidii 51 Aii h OllLlit f.h'. y h y ;

REA:10R 0; IRA 110'.: i i t h i h J'. i NAIN1EhAN;E F Ek 'Ji'.[,.

I t.5 il M.d

..J 43 27 14 M.75 ft.N

. i. ?

OPERAtlNOFERL0%.L HE A 1H FH151(t FER50uh 1;

3 (f

9.4i 1.t' 55.E SUFERVliOR; FER50m!L

'l 2

2

.M

.2 ::

ENOINEEF.!h', PERT 0NNEL If 19 22 5.0L 9.ft

!!.73 R0011NE MlNlEhn';E MINTENAh;E FER$ChNEL L

Sf1 7f3

3. d 414.5C 745.I,6 0FERAllNO FER',0hF.L 7

9 9

1.d4 1.56 i,Cl HEALTH PHYSICS PEH50k T.

5 0

96 1.di

.fJ ift.21 SUPERVISORS PERSONNEL f

f 1

.fd

.fi

.17 EN0lNEERik0 FERSONNEL 2

14 0

.29 4.C3 2.34 INtERVICE INiFEC110N MihTEhAN;E PEkt0inEL e

40 02

.61 45.7!

07.!!

ORERAi!N0TErichtd.L f

2 3

.4 6.5f 2.?f HEALTH FkvE!CS FEiit0 EEL f

f 2

.d.1

.fc

. 3t, EUFEFVliOR1 FERi0a'dL f

f 3

.H

.H 3.f5 EN0thEERlh0 FER50'idt i

i

.H

.21

.25 SFEtlAL MINTENANlE MINTENAh0E PERiON!EL f

1 tif

.H

.21 232 E7 0FERA11NO FERtCh!/.L f

f f

.H

.ft HEALTH FH151CS FERSONNEL f

f 5

.H

.d?

1.01

$UPERVis0R1 FER$0N8EL f

f f

.O

.22 EN0lNEERINO FERiONMEL f

f 7

.de

.N C 77 WRt1E FR0CEttlN; MINTENANCE PERSONNEL 6

4

.ff

.C5

.83 0FERAllNC PERSONNEL 9

e f

6.65

.H

.ft HEALTH PHitlt! PERSONNEL f

f 16

.N

.N 9.ft SURERVic,0R1 FERSONNEL f

f f

.H

.ff EN0INEERINOf'ER50NNEL f

f f

.H

.63

.ft REFUELIN0 MINTENAN;E PERSONNEL f

2 9

.H 1.87 1.21 1

0FERATINC FERSONV.L 4

e f

.66

.H

.N HEALTH PHISICS PERSONNEL 14

.4

.dl 7.H

$UFERVISOR1 FERt0NNEL e

i e

. f.1

.f!

.91

]

ENGINEERING PERT 0NhEL 0

f f

.H

)

8 l

foIAL (See Notes Next Page)

MINTENANCE PERSONNEL 7

574 953 3.99 526,22 Si,.f C OPERATING PERSONNEL 47 35 34 37.04 20.54 15.7%

)

HEALTH PHYSICS PERSONNEL tt 3

169 11.3?

1.19 106.15

)

SUPERVISORS PERSONNEL 4

3 7

.63

.30 3.06 ENCINEERlNC PER$0NNEL if 38 33 6.47 15.26 29.40 l

GRANDTOTAL 86 645 1196 59.48 571.59 1293.36 l

bl. _ _ _ _ _ _ _

i TABLE 2 RECORDE D ANNUAL WHOLE B00Y EXPOSURE FOR CALE ND AR YE AR 1977 l

PE AC H BOTTOM. ATOMIC POWE R ST ATION.UNITL 2 6 3 J

LICENSE N05.: OPR.44 L.0PR-56 ANNUAL DOSE. RANGES NUMBER OF INDIV100ALS

-(R E M)

IN E ACH RANGE i

NO ME ASUR ABLE E XPOSURE 1072 j

ME ASURABLE E XPOSURE LESS THAN.100 8 97

.250 403

.1. 0,0

.250

.500 359

.750 245 j

(

.500 1.0 171 J

l

.750 l

1.0 2.0 498 3.0 185 2.0 l

4.0 47 30 50 13 4.0 i

6.0 7

5.0 7.0 2

)

6.0 6.0 0

7.0 90 0

6.0 1

90

- 10.0 0

]

10.0

- 11.0 0

)

11.0

- 12.0 0

12.0

+

0 i

TOTAL NUMBER OF INDIVIDUALS REPORTED 3899 l

t I

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l l 1C___.__._

T ABLE 6 Y

PE ACH BCTETOM tDCis 2 6 3 IS0 TOPIC ANALY115 Of GASE OUS k A0!GACT!VE EFF LtKNT5 (in Curies) 1977 i

15010FE J UL Y A A.,

SE Pl.

OC T.

N Qw,

OEC.

C i 7 07 4

[

8. iK> E + 0 i 9.50 E -01 3 00E+00

9. 20l +; l

3 p ton.6 %

f e

Aenon 13) 1 50E +C 3 0) 6.el 'J U) 4.10E 02 4.20E*03 3 40E+03 8.%E+02 1.1'A ; R.

xenon 135 1.10E 02 0) 3.)DE *01 (2) 6.60E+00 1.60E 03 1.70E +02 1.6CE*02 2.10 ! +' ) ( 2, 3 00E +0!

f 3 00E+06 krypton-86 l

Total 1.60E +03 (l) 8.70E+02 4.20E

0 2 5.80E+03 3 60E +03 1.10E+03 1 30E +04 (2 )

lodine-131 1.64E-02 4.5 9E 03 1.60E -0 2 9.83E-02 3 42E 03 8.04E-03 1.47E-01 lodine 133 4.10 E-0 3 4.10E-03 5 20E-c3 1.20E-Ol 1.20E -01 1 50E-01 4.00E-01 l

f Iodine 135 3.5e t -0 2 3 50E-02 4.40E-02 6.90E -0 2 6.90E-0 2 8.60E-02 3 40E-Ol Total 5.60 E-0 2 4.40E-02 6.50E-02

2. 90E-0 2 1 90 E-02 2.40E-01 8.90E-01 i

St r on t i e-8 9 2.70E 45 6.10E-05

1. 20E -0 4 1.60E-04 1.70E-04 3 00E-04

8. 40E -04 17.80E-06 16.00 E -06 16 30E-06 g,3 9C E -05 S t r on t i um 50

16. 90 E -06 5.5 30E-06 16.10E-06 md34 1.6 9E -0 4 l.87E-OL 2.67E-04 1.26E-03 3 69E-05 7.14E-05 1.99E 03 Ce s t a Cesium-137 2.77E-04 3 91E-04 3.26E-04 1,49E -0 3 3 53E-05 1.0 9E -04 2.63E 0) 9.83E-05 Lenthena -140 9.8 3 t -0 5 m

CodeIL-58 todelt-60 3.12 E -0 4 4.06E-04 7 06E-04 3 53E-04 3 93E-04 4.0lE-04 2 57E-03 Zinc-65 4.65 E-04 5 07E-04 7 12E -04 2.00 E-0 4 1 72E-04 2.61E-04

2. 3 'E -0 3 m ngenese-54 4.54E-05 4.54E-05 l

Strontium-91 1.04E 04 1.04E-04 Zirconium 95 l.55E-04 1.55 E-04 2.85 E-04 2.85E-04 Molybdenum-99 Sodic-24 4.06E-04 4.06E-04 Cesium-138 6.05 E -05 6.05E-05 8.75E-05 8.75 E -0 5 8erf e-140 3 28E-05 3 28E-05 S i l ve r-110m Rubidium-88 2 50E-06 2.50E-06 TOTAL 1.26E-03 1.56E-03 2.17 E -0 3 4.47E-03 1.06E-03 1.15E-03 1.17E-02 Less then minimum detectable (1) Estimated dets obtained f rom off-ges dote (et $JAE) es suming a 2 say noia-up. Sempting problems during this period prevented the obtaining of a representative senple.

(2) includes some estimated data per (1) i l l

l

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FORWARDING CORRECTED: COPY:-OF TABLE 7 OF SUBJECT FACILITY"S3sNNUAC CPERr. TING OFERATING REPT NO 2 FOR01977:

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g PHILADELPHIA ELECTRIC COMPANY i.

2301 MARKET STREET P.O. BOX 8699 O IW k[CffyES l

PHILADELPHIA A. PA.19101

'MR f yjg,g k taisi c4t 4oco M

/

y

/% ~

2 1

Mr. Boyce H. Grier Director, Region 1 Office of Inspection & Enforcement U.S. Nuclear Regulatory Comission o31 Park Ave.

Aing of Prusaia, PA 1.H.0 6 "UBJECT:

Correction to Table 7 of Annual Operating Repor:

No. 2 January I through Decemcer 31, 1977 Peach Bottom Atomic Power Station.

Units 2 and 3 Docket Nos.:

50-277 and 50-278 Oear Mr. Stier:

Attached are two corrected copies of Table 7 of the Peach Sottom Annual Cperating Repcet No. 2 for 1977 Please insert enem into the ecpies of the Reocrt which were transmitted to you previously.

Corrections have been made to the " Activity" values shown for the Solid Radioactive W.iste Shipments. A conversion error had been made in converting picoeuries to curies f or tnis listing.

Very truly yours, W. M. Alden En gi nee r-In-Charge Nuclear Section Generation Division GJTabgb Att achment cc:

Mr. Ernst Volgenau

&. Richard A. Hartfield Director Acting Director Office of Inspection & Enforcement Of fice of Management Information J.S. Nuclear Regulatory Comission and Program Control Washington, oc 20535 (40 copies)

U.S. Nuclear Regulatory Comission 'i Washington, DC 20555 (2 copies)

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