ML20067B592

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Annual Operating Rept 2 for Jan-Dec 1977
ML20067B592
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 12/31/1977
From: Alden W, Ullrich W
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To: Grier B
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20067B523 List:
References
FOIA-90-452 NUDOCS 9102080297
Download: ML20067B592 (38)


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l PEACH BOTTOM ATOMIC POWER STATION UNITS-NOS. '2 and 3 1

ANNUAL ~ OPERATING" REPORT 1

NO. 2 Janu a ry 1, 19 77 throughi December ' 31,-'1977 Submitted to The United States Nuclear Regulatory Commission

' Pursuant-to l

Facility Operating Licenses Nos. DPR-44: 6: DPR-56 l

Preparation = Directed by:

[

W.

T. Ullrich,'Superint_endent-Peach Bottom Atomic Power ' Station

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TABLE OF CONTENTS PAGE 1

Intr od uc tion 2

Summary of operation 4

Unit 2 operation 15 Unit 3 operation Personnel Exposure by Job Punction 24 24 Wholebody Exposures Liquid. Radioactive Release Data 24 Isotopic Analysis of Liquid Radioactive Releases 24 Ga seous Railoa ctive Release Data 24 Isotopic Analysis of Gaseous l

Fadioactive Effluents 24 solid Radioactive Waste Shipments 24 Revisions to Previous Semi-Annual 25 Ef fluent Reports

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, P BA PS

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. TABLE OF CONTEN7S j'

TABLE PA'Q;

.:j STANDARD - FORMAT FOR-R EPORTING 26 l

i NUMBER OF PERSONNEL' 0 MAN-REM y

- FOR WORK; & JOB ' FUNCTION 2

RECORDED ANNUAL WHOLEBODY: FOR 27 CALENDAR YEAR-1977 3

LIQUID RADIOACTIVE: RELEASE ~1

-28

. DATA 4

o ISOTOPIC ANALYSIS OF LIQUID :

29 RADIOACTIVE RELEASTS 5

GASEOUS-RADIOACTIVE: RELEASE' 30 DATA i

6 ISOTOPIC ANALYSIS OF GASEOUS

' 31..

RADIOACTIVE-EFFLUENTS 7

. SOLID RADIOACTIVE WASTE

'32 SHI PMENT--

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PBA PS I NTRODUCT! ON Peach Bottom Atomic Power Station consists of two Boiling Water Nuclear Power Units with each having a

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

Philadelphia Electric Company is the f acility o pe r ator.

This report covers the period f rom January 1 through December 31, 1977 and contains the last yearly compendium of 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 pr epared.

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

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0 PBAPS SU MMM_ Y Peach botto-Unite 2 and 3 JiiOO MWe bht'r)

Ea ch Pea ch Bot tom unit experienced a refueling / maintenance outage during 1977 In addition to a 140 day re f ueling/raintenance outage, Unit 2 experienced eleven other outages with durations extending up to eight days.

Unit 3's refueling / maintenance outage was in progress at the beainning of the year and extended '101 days into 1977.

There were eleven additional outages in 1977 with the longest being nine days.

The 1977 Unit 2 short outages resulted from (A)

Turbine control valve and combined intermediate val ve problems (Three outages - tatsling 13 days)

(B)

Fecombiner condenser leak (one outage - two days)

(C)

Power load unbalance relay operation (one outage -

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

(D)

Neutron monitoring noise flux spike (Two outages -

totaling less than three days)

(E)

General Electric Company End of Cycle testing (two oatages - less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

(r)

Recirculation loop valve packing leak (one outage

- two days)

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

During the Unit 2 refueling / maintenance outage, var'.r os pajor medifications and nsintenance jobs were completed such ast (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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(r)

RHR system flow orifice installation (HPSW side)

(G)

Diesel generator overhaul (four diesels)

(H) condensate demineraliser valve replacewsnt (1) local leak rate testing of containment penetratf or.r.

(.1)

Puel handling and sipping The 1977 Unit 3 short outages resulted from:

1 (A)- condenser vacuum leak (One outage - two days)

(B)

Primary containment isolation syst.em relay fire.

(one outage - two days)

(C)

Electrical switching transients (Two outages -

totaling - two days)

(D)

EHC system oil leak (one outage - two days)

(E)

RCIC system inoperable (One outage - three days)

(r)

Drywell to torus leak (One outage - two days)

(G)

Steam leak in the drywell. (one outage - six-days)

(H) reedwater heater leaks, (One outage - nine days)

(1)

Reactor water chemistry problems. (Two outages -

tot aling six days)-

During the ref ueling/ maintenance outage for Unit 3 the f ollowing additional ma jor work was completedt (A)

LPRM replacement (D)

Inspection and maintenance of feedwater spargers (C) core spray nozzle spool piece replacement (D) control rod drive return nozzle crack repairs and special testing (E)

Local leak rate testing and containment ILRT -

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Unit 2 l

UNIT 2 - oPERAT. tons on January 1, 1977, a Unit 2 shutdown was initiated becausa 4

of dif ficulties 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 water contamination of the EHC fluid.

During the shutdown a Group I isolation and scram occurred.

These were initiated by reactor low pressure which was caused by the spurious opening of another turbine control valve.

Investigation of the initial problem indicated significant water in the El:C 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 f ollowing a November 197 6 outaco 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 operators were removed from the valves, were inspected and repaired.

Several of the diaphragms removed during chis inspection showed 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, af ter the reactor had achieved normal operating pres sure, the. 71E relief valve spontaneously opened causing a reactor blowdown.

The reactor scramed shortly af ter the blowdow1 on a safety system action caused-by low level.

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

These tests indicated that the D,

E and F pilot valves had excessive leakage.- These valves or their valve operators were replaced and the reactor was taken critical on January S.

The relief valves were again tested at approximately 150 peig during the startup.

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Turbine operation war delayed when an Eyc oil sample again indicate $ wat er cont amination.

An additional leak war identified in the heat exchanger.

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

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

The pH of the reactor water was within limits.

Within sixteen hours following the transfer of water to ths Radwarte system the conductivity was within lindts.

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

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

Three main condenser waterboxes were inspected and a leak was identified and repaired in the C-2 water box.

A drywell entry was r.ade and a packing leak on the vessel head to main steam line vent valve was repaired.

The reactor was critical at 2: 30 p.m. on February 6, and the generator synchronized at 8:45 p.m.

Operation during this period was with a reactor water conductivity of approximately 3.5 umhos.

The pH values were within limits.

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

Ly February 8, the reactor conductivity had increased to B.5 umhos.

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

A large tube leak was subsequently identified and corrected in the A-2 water box.

No leakage could be found in the A-1 water tcx.

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

Reactor conductivity subsequently returned to less than 5 umbos..

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At 1:15 p.m. on February 9, the turtine generator tripr>ed a+. a loa 3 of approximately 600 MWc.

Plant instrument at f or.

indicated that the turbine trip was caused by the power load unt41ance relay associated with the turbine generator protection system.

The turbine trip caused a reactor scram as well as-a Grou9 I, II and III isolation.

An l

investigation of the power load unbalance trip did not identif y any equi eent or system malfunction.

Following the l

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

Generator circuit breaker No. 225 was successfully closed:

i however when Generator circuit breaker No. 215 was cloced both circuit breaker Nos. 225 and 215 tripped.

No cause for this tripping was determined.

These breakers were successf ully closed a f ter the investigation.

The reactor was made critical at 10:50 p.m. on February 9, 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 c/.V stuck in the 85%

open position.

The Unit was removed from service at 5:03 pm

~

F on March 3 to identify the cause of the problem and make i

r epair s.

The number 5 CIV Servo was replaced and the valve tested satisf actorily.

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

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During the ris e 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 I

l heater as a result of tube-f ailures.

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

With tt? 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 i

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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 perfor~ed on the 3A f eedwater heater.

Investigation of th -

nurber 5 c2v f ailure, identified a badly scored hydraulic actuator cylinder.

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

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

Full load was achieved on 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 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 f'oor drsin sumps pump rate was identified as a packing leak o,i the

'A' recirculation pump suction valve.

Additionally, the Reactor Water cleanup System isolation valve it.3ide the drywell had f ailed in the closed position.

The leaks identified were corrected and the isolation valve motor r e pl ac ed.

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-r provided proper range from the various core parameters such that no Technical Specification violations would occur as a result of the testing.

The End-of-cycle pressure perturbation and tarbine trip testing was performed on April 9.

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

A restart was immediately undertaken and the Unit was resynchronized at 11: 40 p.m.

Power was increased L

toward the first stability test condition of 625 power and 50% flow.

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Unit 2 Feactor Power operation centinued through April 14,- at increasino power levels.

On this date, load was reduced to permit End-of-cycle etability testing.

This testing continued through April 16.

No operating conditions were exterienced which resulted in unstable operation.

On April-16, reactor power was f urther reduced in order to permit a turbine trip test f rom 15%. power.

This test was performed i

at 1:15 p. m.

The turbine generator was resynchronized at 1

1: 25 p.m.

FJ11 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 a 601 - 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 112 34 a.m.

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

The End-of-cycle turbine trip test from 701 power was conducte d on April 27.

Unit 2 load was reduced at l

approximately 5:00 a.m.

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

The i

turbine trip test was perf ormed at 11: 05 p.m.

Following the turbine trip, a special shutdown procedure was used to 1

maintain reactor pressure as long as possible.

This

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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 shutdovn.

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

By April 2 9, the reactor l

was depressurized and chutdcwn cooling established.

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

The reload for this refueling consisted of 172 bundles of BX8 fuel.

During the out age, 421 tendles were sipped with twelve bundles being identified as leakers and removed f rom further service.

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

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and the dryer and serarator were removed in preparation f or i

in-ves rel w'r t.

Following hydrolasing and draining of the reactor vessel to i

approximately twelve inches below the control rod drive i

return line norrie, a PT inspection was performed on the nottle.

This insrection identified multiple cracks on th*

blend radius and in the tore section of the nozzle.

1 Following this exar.ination, grind-out of these cracks was l

initiated.

This work was perf ormed f rom May 10, through May 17 Dn May 17,1977_ additional cracks were identified below l

the return nozzin in the vessel cladding.

These cracks were removed by grinding in accordance with recommendations of the nuclear steam supply system vender.

The maximum depth of any one crack was 0.9 inches, measured from the surf ace of the cladding.

Daring one of the PT examinations, additional cracks below the CRD nozzle outside the eight 1

inch diameter inspection circle were identified.

These cracks (approximately eight) were from one inch to seven inches long and generally were in a horizontal direction.

The cracks extended down below the bottom edge of the CRD j

nozzle approximately eight inches.

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 '38 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 metalurgist.

Agreeme nt was reached that the suspected weld on tho'B' loop would be renoved and investigated and proper repairs-made.

Additional consideration was given to performing the same tyre of work on the ' A' loop.

Following the completion of CRD nozzle grinding, the reactor vessel and rea ctor - head cavity were flooded.

CRD replacement operations were also started on this date.

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

The control rod drive j

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 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'specisi 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 RHP 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 mair.tenance work was begun on June 2.

With the water level reduced to approximately one f oot below the control rod drive return line nozzle, work proceeded on both the core spray piping repair as well as i

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 f or 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 l

that the cracks were caused by high cyclic thermal f atigue.

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i PBAPS Unit 2 Core spray vor k continued through July 9, at which time both spool pieces were w-1ded into the line such that ISI wor k in the vesrel could beain.

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

On July 14, the core spray system was tested successf ully.

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

Tids work was done using a special tool which gripped the source holders f rom the tottom.

Work was completed on July 20, except for the removal of the debris lef t above the f uel support pieces f rom the broken source holders.

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

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

Following clarification of the water, f uel movement associated with the emptying of the cells surrounding the f our damaged source holder locations was initiated.

Removal of fuel support castings, control rod drive blades, and debris from these locations was in 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 re-oved.

The performance of this work also 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.

Tnis identified two fuel cells which were not properly seated.

Puel f rom these cells was removed, the f uel 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 f rom September 8, until criticality was achieved on September 12, at 7: 03 p.m.

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

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I PBAPS Unit 2 Se pt embe r 14 Load was increased to approximately 100 MWe and held at this level during LPFN calibration.

Balancing data taken daring initial turbine generator operation indicated a need f or additional weights in the number 2 coupli ng.

The Unit was theref ore renoved -from service at 4: 15 a.m. on September 16.

The length of time the Unit was out of service was extended until 5:34 p.m.

for repair of an-EHC oil leak in the relay trip valve in the turbine f ront standard.

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

It was determined that the 'r' condensate demineralizer had several defect 2ve 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.

Ioad was again increased.

Electrical power of 868 MWe was achieved on September 24 U.d was then limited by the availability of condent te demineralizers.

Jnit 2 operated at atout 921 power (1000 MWe) through october 28 when load was reduced to 400 MWe for_ a rod sequence exchange.

The Unit was then placed on a preconditioning ramp until 1:25 a.m. on tevember 3, when a reactor scram occurred.

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

i f eedwater flow.

The flow of cold water increased flux to l

the trip point of the APRMs and the scram occurred. _The Unit was at about 511 power (515 MWe) 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 p' ump tripped-on

'B' phase dif ferential relay operation.

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r FEAPS Unit 2 increases were not affected at the time of the trip.

Due to the lor s of *,h e condensat e pump, Unit 2 load wat lini+ ed to aboat 900 MWe.

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

on November 10, and preconditionine was resumed.

Unit 2 reached a raximum load of 96% pcVer (1035 MWe) on Novemter 12.

At 7:00 p.m. on November 17, 1 cad was decreased to 33% (300 MWe) to allow access to the MSIV room to repair a RCIC motor operated valve which had f ailed ar. opersbility test.

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

The motor and torque cwitch were replaced.

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

on November 16.

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

Load increases were stopped due to condensate demineralizer dif ficulties.

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

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

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

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

A normal plant shutdown was begun.

The turbine g<mtrator was of f the bus at 2:05 p.m.

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

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

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

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

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

The unit reached 98.5% power early on December 11.

Unit 2 operated at about 9 8% powe r (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 Decemt* 3 l

13 -

a PBA PS Unit 2 14 The generator was synchronized at 8:50 a.m. on December 1u.

Unit 2 continued ~ power increater, reaching 965 power December 21 when it was necessary to drop load (1034 MWe) on 100 MWe due to a vacuum loss f o31owing a recombiner mechanical compressor oil-change.

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

Unit 2 continu ed operation at about 971 power (1055 MWe) until 7: 30 p.m. on December 27, when a 100 MWe load-drop _ was taken to regenerate a condensate demineralizer.

Load was-increased at 5 MWe per hour f ollowing the regeneration.-

On Def. ember 28 at 10:03 p.m., Unit 2 load was decreased. to 900 MWe to per form 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: 40 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' hea ter string.- The ' B ' heater string was isolated and operation continued at reduced load through the' end of the year.

l I

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PBAPS Unit 3 Unit } Operati onr The first ref ueling/ maintenance outage for Unit 3 began ori December 24, 1976.

Following MSIV testing, work was begur.

on December 30, 1976 in preparation for entering the reactor vessel f or ref ueling and maintenance.

Fuel handling began on January 3, 1977 in preparation for a reload consisting of 188 bundles of 8X8 fuel.

LPRM replacement and fuel sipping were performed simultaneously.

Of the 172 f uel eleme nts that were sipped, three elements were identified as leakers.

Fuel sipping was completed on January 16.

Following the verification of LPRM latching, the reactor vessel level was lowered to inspect the feedwater sparger nozzles.

A PT examination of selected areas of the

'D' and

'F' nozzles identified slight indications which were renove5 by a flapper wheel.

A repeat of the UT examination showed 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' spargers were removed from the nozzles on Januar y 19, 1977.

An extensive PT examination of the

'D' an d ' F ' bl en d r a di us a nd bore ar ea s howed minor -

indications which were removed by light grinding.

None of these indications penetrated the cladding.

The condition of the spargers was excellent.

Both spargers were reinstalled.

UT examination of the nozzles still showed the reflectors.

Since these indications could not te located by PT examination, a decision was. made to return the Unit to l

service when this refueling outage was complete and re-examine these two spargers at the next refueling outage.

)

l 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 ef f ected 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 Farmit f uel handling operations.

The refueling operation tcok place between January 23 and February 3.

1 l

PEAPS Unit 3 locations were verified on February 5, ar,d The fuel element the reactor head cavity level was lowered to a level b(low the core spray nozzle to permit repair of the weld crack core spr ay loop.

Radiography en the identified in the ' A' core spray piping verified UT data which indicattd 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 pire stools removed.

Additional circumf erential and axial cracks were identified in both the ' A' and

'B' core spray elbovs in the heat ef fected zone of the veld to the stool piece.

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

Fadiography and repair welding continued through March 18.

A PT inspection on the Unit 3 control rod drive return line nozzle was performed on March 6, 1977.

This inspection identified a number of cracks.

These cracks were f airly unif ormly distributed around the circumf erence of the Grind-out of the cracks was started on March 9, and nozzle.

The deepest crack was was completed on March 17.

approximately 7/B inches deep including the thickaess of the cl a d di ng.

on March 10, f ollowing the completion of in-vessel work and 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 l

the control rod drive system were performed to verif y that acceptable drive perfermance could be attained with the control rod drive returr. line isolated.

The results of this testing were satisfactory.

On March 29, preparation for a vessel hydrostatic test was l

Additionally, twelve selected control rods were begun.

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

No significant di f f erence in control t od drive per f ormance was noted during these tests. 1

1 I

PEAPS i

Unit 3 Startup activi tier continued and - on April it, the reactor wa s ma de cri ti cal.

Several dif ficulties associated wit h recorbiner operations, high radwar.te inputs and high 4

condenser air in-leakage were identified, on April 12, because of continued dif ficulties 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 very loose.

To safely tighten the packing fol'1ower, 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 cloeure of the MSIVs.

During this period a blockage in the recombiner line between the mechanical compressors e.nd the holdup volume was determined to be in a dif fuser inlet pipe to_ the holdup volume.

The dif f user was found essentially rusted closed.

Following repair of the air ejector valves. and renoval of the dif fuser, the MSIVs were reopened, operating pressure established and the turbine generator resynchronized at 1:25 p.m. on April 15,<1977.

Power was t

increased to 2 5%.

on April 18, with the Unit at 255, a small fire occurred in f

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.n.

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

The damaged relay s and wiring were removed and replaced with new components.

Propagation of the fire was traced to the relay _

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

i 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 L

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

Following condenser leak re pa ir s, the power level of' Unit 3 continued to be 17 -

L r

s 4

PBAPS Unit 3 increased.

A maximur load of 996 MWe was achieved on April 29, wit h the existing rod pattern.

Since the recirculatf or, flow was at a maximur, power WaE reduced and additional rois we re wi thdrawn.

This operation resulted in achieving 96% of f ell load (1060 MWe) on May 4 On May 2 2,197 7 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.

Follosing these repairs, the Unit was returned to f ull load.

Operation of Unit 3 continued at approximately 98% power f rom May 25 through 27.

On May 27, the

'B' reactor f eedpump ran bsck to minimum speed.

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

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

Later on the same day, power was reduced to approximately 342 KWe 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 O n J un e 7, at approximately 2: 40 a.m. the Number 2 startup feed (220KV line 220-0 8) tripped due to dif ficulties in the Graceton Substation.

This resulted in an automatic transfer o f t he 4 KV ba s s es,

rollowing 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 trans f er.

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

The E3 diesel had been removed f rom service-on June 6, to start its annual maintenance outage.

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

The high pressure service water had been used on the previous day to supply cooling to the Unit 3

'D' RHR heat exchanger, thereby permitting mud removal operations on the Unit 3 intake s t ruct ur e.

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

The reactor was r.ade

' f l

PRAPS Unit 3 critical and t he 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, ar.d the turbine generator synchronized.

Full load was achieved on J un e 2 0.

Unit 3 continued f ull power operation from June 20, t hrough July 1 when leakage of resin through two condensate derdneralizers 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 Nitr ogen system f or 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.

Instrunent 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 dif ficulty was experienced with the RCIC inner isolation valve.

This 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 f rom the Nuclear Regulatory Commission on July 19, to permit 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 operation provided the HPCI was tested daily.

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

A shutdown was !

PBAPS i

Unit 3 initiate d inmediately.

The turbine generator was removed f r o.m s e r vi ce a t 1:32 a.m. on July 22.

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 inspection was made.- No obvious cause could be identified.

A zero dif ferential pressure test for vacuum breaker operability did indicate some friction in the mechanism.

The torus to drywell vacuum breakers were then cleaned and lubric ate d.

A torus to drywell leak test was performed and f ound 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 PCIC System declared i noperable.

Surveillance testing of the HPCI was successful.

Reactor power was reduced to 314 MWe-on August 12, to repair this valve.

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

I During surveillance testing of the HPCI on September 1, the steam supply valve f ailed to lopen. ~ The HPCI was declared inoperable and the required surveillance-testing performed.

During the performance of the ADS Logic System Functional Surveillance T est, (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.

, i i

[

PBAPS Unit 3 operation of Unit 3, at essentially full load, continued through Se;4 ember 24 On September lo, the HP01 was again declared inoperable because of the f ailure of the in1t t steam supply valve to open.. Repairs were mace and th( HPCI i

r eturned to service on September 12.

on Septemter 2 5, - Unit -3 was removed f rom service a t 4: 20

a. N to accommodate a maintenance outage.

The prirary activities during this outage were associated with correction of leaks in the heat cycle, correction of several steam leaks in the drywell, repair of RPIS instrumentation, and tack walding 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 start 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 was successfully completed.

The turbine generator was synchronized on October 1.

Approximately 50% power was attained on October 3.

At that tina, 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 aor resin into -the reactor vessel f rom the RWCU system following its return to a vessel to vessel node 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 iodine release.

The rate of release vas approximately 80,000 uCi per day which is 2341 of Technical 1 Specification limits.

By the-morning-of october 5, the rate was about.1200 uCi per day which is less than 41 of the Technical SFecification limit (See bER 77-04 9/1T-0 f or Unit' 3).

Iodine levels continued This problem was caused by the venting of the RWCU to drop. -

i _. _ _ _ _ _ _, _ _ __ _ _ _.. _ __ _.-..._ _._ -

PBAPS Unit 3 system heat exchangers.

On Oct oter 4, a load reduction was required due t o a trip of the recortiner on indication of high hydrogen concentration and a slight loss of vacuum which f ollowed.

The problem was traced to a closed valve on steam line return f rom the recombiner preheater.

The avalve was opened and normal operation continued.

Unit 3 reached f ull power operation on October 10.

On Hovember 7, a feedwater heater leak was suspected based on a disparity between dif f erent flow indications.

An i nve st ig ation f ollowed, which indicated a leak in the

'A' heater string.

Load was reduced to about 800 MWe and the

' A' heater string was removed f rom service.

Load was then increased to about 980 MWe.

Unit 3 continued operation at about 93% power (980 MWe) j until Hovember 26, when a controlled shutdown was initiated to repair heater leaks.

Unit 3 was removed from service at 10:16 a. m.

Unit 3 remained shutdown until December 5 to allow repairs.

During the Unit 3 startup on December 5, a reactor scram f rom about 5% tower occurred.

The scram was caused by low water level when two reactor f eedpumps f ailed to respond to control signals.

The reactor was restarted at atout 6:25 p.m. on December 5.

The generator was on the line at-9:07' a.m. on December 6, but was removed from service at 3:25 p.m. because of a reactor water chemistry problem caused by resin f rom a condensate /demineralizer.

This resin had leaked f rom loose elements and was carried into the reactor vessel.

At about the seme time, the HPCI was declared inoperable due to a foreign material (cap screws) being f ound downstream of the HPCI turbine exhaust.

After consultation with the turbine manuf acturer, the bolts were determined to te 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 mechanical compressors.

A shutdown was begun on December i

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".ho air eiector and recorbiner woro placed in service and load increases bonun.

The 1 r.1 t :t t ni net' 99' r.ouar-en rocerber 2' an't enerati-* a' that 1cve' conti:.urf throunh the end of the v a.* r.

23 -

.4 PERSOMNEL EXPOSUPES 6 F.ADIOACTIVE FELEASES A.

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

-r i

B.

Whol etody Ex posur es 1

Annual wholebody exposures for the year arc presented in Table 2, in accordance with 10CFR20.4 07 (b).

C.

- Liquid Ra dioactive-Release Data See Table 3 D.

Isotcpic Analysis of Liquid Padioactlye.3eleass; See Table 4

'E.

Gaseous. Radioactive Release Data See Table 5 l

F.

Isotopic Analysis of Gaseous Radioactive Ef fluents 1

See Table 6 G.

Solid Radioactive Waste Shipments See Table 7 f

l l-L l

l l-24 -

F EvlsioMs E PkEV1ous SEMS-AmiUAL trn,Unil REPoPTS Tables III-D and III-T f rom the July th ough December, 1976 semi-Annual Effluent Report are attached as pages 33 and 34 r e spec ti vely.

The Decemter entry f or Cesium-137 has been corrected on Table III-D.

The Mixed Noble Gas value f or July has been correct ed en Table 111-T.

Thble B f rom the January throuch June, 1977 Semi-Ar.nual Effluent kepart is attached as page 35.

The " Total" value of the Noble gas totals (Krypton and Xenon) was corrteted on this table.

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IN L ACH RANGE NO M ASUR ABLE E XPOSLAE 1072 ME AStAABLE E XPOSURE LESS THAN.100 8 97

.250 403 100

.2 53

.500 359

.500

.750 245

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19);

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lodine.1)I 1.64t 02 4.5 9(.0) 1.60t.02 9.6) t.0 2 3 42t.c) 8.04t 4)

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6. 90t.0 2
6. pot -0 2 8.60t -0 2 3 40t-01

'. W C -0 2

1. Wi-0 2
2. 40 t-01 8.901 01 2

total 5.60t 02

4. 40t.0 2 6.50t-02

~

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Cest e l)4 1.6 9t.c 4 1.67t.04 2.67t.04 1.26E4) 3 69t 05 7.lkt 05

1. 99t.0 )

Ce s i c.137 2.77t.c4 3 91t.04 3 26t.04

1. 4 9t -0) 3 5)[.05 1.0 9t -0 4 2.6)t.0) 9 8)t-05
9. 6)t.05 L an t h a n e-140 Cooa1t.56 C ob a l t.6*,

3 12 t -04 4.06t.04 7 06t 04 3 5)t 04

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1. lit.0 2 Les s than minime estect able (1) titinoted dat e ottelend f r cs of f-ges data (et $W) as suming a 2 day hold-up, laryling problems dsring this period prevented the ObtelnIng Of e reg resentative s ertple.

(2) inc l ude s sone e s t led te d d a t a pe r (1)

~31-

i e

J a

?

.g

.:, }

T ARI.t 7

.I l

PE ACH BOTT ON UNITS 2 & 3'. SOL ID R A010AC T fvt stASTE SMIPq ast apy

.q JULY AUG' SEPT OCT

  • s0V O((

r rat.

l sewdper of shisements

' t7 :

29 24 26 22 f52

. 7 vote of waste (f t)J 9 15E *03 4.21.E+03

1. 40 f *04 6.86E*93 f.9?E*03 4.2e,[-93_

..gt _-_,

Activity, curies 4.33E-Os

-f.00E-el' 9.38.E-UI 5.J 8 E -0 2 1.651 -s e 6.$ t t *

a. m. -

wipping dates (s of shipwnt s) 8 (8) 4 il) 8 (8L.

J (1) 8 i8; 3 (81 5 (t) 2 (t)

'2 (t).

4 (8) 2 (2) 2 (s)

.[

6 (2) -

4 (I) 7 (8)'

5 (3) 3 (2)..

' 6 (23 i

7 (s) 8 (i) 9.(2) 6 (t).

4 (2)

-7 (e)

. shipped by atttaan wucie c-(2)

. ::(t)

}{s); '

- sott)'

a (3)

' 9 (2).

on.p..ttsan A s s...t.

a (7) 9 (s).

' arts) 7 (2)-

p (s),

p.(i)-

}

and Deve lo,Wat Corporet t o*i.'

12(1)-

12(l) 18.(3) 11(3)-

. 9 (1) 12(l)

., I r

f in trucks.to the Chem. Nuclear 13(1) 16(l) 15 (!)

82(4)

t1(2) 13 (f) 3 Corporetton, earn-ell, south.

Ih(3).

17 (5) -

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IL(if.

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18 (2).

' 22 (l) 19(l)

IB(1)

' 18 (t) 16 (8) 19(1)'

24(I)

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20 (!)

. 22(3) 39(e);

20 (2)

25(1) '

23 (2).

74(1)'

23 (I) :

- 29 (I) 21(1).

30(I) ~

22(2)

.25(t) 25 (2) c 2t(s) i

. 22 (2) 31 (1).

23(2) 26 (2) l

. 29 ( t ),

23(t) i 25 (2):

24(2) 27(3) '

' 33 (2) 21t1)-

- i 26 (t)

' 28 (t)

! 26(2)

' 26 ( 3 ) -~

27 (t) -

??(3) ya t t '

27(3)

L 28(2):

28 (l).

'. 3 8 ( s )

29(2).

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TABLE III-D LIQUID RADICACTIVE Retf ASE3 ( ra c=-**s) 19/6 PEACH BO' TOM UNITS 2 & 3 - ISOTOPIC ANALYSIS OF 15_0_ T C ',' E JULY AU_G.

SEPT.

OCT_

[ NOV_

0 nrc.

ri

Tor, Strontium-89 1.39E-94 1.061-D6 1 J6 f. -05
2. 5 7E 45 3_18E-Os 1 58Cp5
2. 3 7r-06 Setontium-90 9.73E 06 5.16E 06
5. 2 4 E -05 4.15E 06

' 5.91E 06 S. 3 7E- 06 3.56E-03 Cesium-134

3. 4 4 E-02 5. 8 2 E -04 1.97E 03 9.9E- 04 7.3E 03 4. 5 2 E 02 Cesium-137 5.IE-02
9. 54 9 E 04 -
3. 2 7E- 04 2.7E- 03 1.8E 03 9.8E- 03 6.665 03 5. 2 E - 04 7.9E 03 3.SE 03
1. 51E 0~

Iodine-131 2.33E-03

8. 8 7E- 04 l

J i

Cobalt-58 i

Cobalt-60 6.61E-04 8.6E-05

6. 4 5E 04 2. 2 E 04 1.61E C Zine-65 3.17E-03 4. 5 E-04
8. 5 E 04
4. 5 E 04 4.92E 0 w Manganese-54 W
9. I E 04 9. I E - 04 Chromium-51
3. 0 E 04
3. 0 E 04 f

Zirconium-95 4. 0 E 04 :

4.0E 04 Molybdenum-90 L 5. 7E 04 4 E - 04

1. 7 E -04

+

Lanthanum-140

8. 4 E 04 4,.78E 0 Arsenic-76 1.4 E- 03 2.04E-03 1

Sodium-24

1. 05 E- 01
5. 3 E -03
1. 2 E - 01
6. 5E - 03 1.29E 0 5.56E C Neptunium-239 2.99E-03 c
4. 7E-04
2. I E- 03 I. 2 E Oi Iodine-132
1. 2 E- 04 e

3.22E44

1. 3 E. 04 1.18E C Iodine-133 6. 6 E - 04
6. 3E- 05 i

i

1. 9 E - Of 1.9E-04 Iodine-135
2. 8118~-07 J. 7 C r.-o Total (Curies) 2.02E 01 5.850-03 6.041. 04 1.16 E -02
2. 7 3r o2

.,, + s..

t e t...- detectable activity.

proised ina nee _ entry for Cen t erm-l 'l7 correct ed.

i-MM-

' i

~.

TABLE III-F P

PEACH BOTT0H LNITS 2 ANO 3 GASEOUS RA010 ACTIVE RELEASE DATA 1976 OCT.

NOV.

DEC.

TOTAL

{

JULY AUr..

SEPT.

f Mixed Noble Cases Ci 9.81E4)4 9.74E+03 1.90E*06 1.26E*0f*

7.80E403

1. 'h r > < f*

6.94E*04-

% of Tech. Spec. Limit (1) 1.5IE+00 3 01E*00

2. A'4E + 00 1 3hE*00 7.68E-Of
?.78F-01 1.72E*00 Iodine 131 Ci 7.77E-02 9.9 3E-02 1 38E-01 4.10E 1.85E "
1. '.'* E ' " 2 3 91E-01

% of Tech. Spec. Limi t (2) 1.67E401 3 24E+01 3.80E+01 1.24E+01 3 84E40')

_J.;'2E

'1.81E=01 Particulates>8 Day Hair Life 1.68E-03 1.18E-02 9 73E-01 4.20E-03 1.52E-03 7.c2E

Alpha Ci (6) 46.92E-08 (8.98E-07' (1.17E-Ot, 52 94E-06~

(2.20 E-M

<l.16E-09' (5 30E-06 w -

8 % of Tech. Spec. Limi t (2) 4.79E-01 5 00E+00' 2.84E+0C 1.50E*00 L.61E-01 1.SoE'0?

1.0LE=00-l.

1.20E+00 1.03E+00 1.28E400-

'2.12E+00-1.91E*00' 2.16E *M 4 9 70E+00' Trititen C1 (3) 7 5E+04-1.80E+04 5.8E*0h 3 95E*03 1.00E*05 Max. Noble' Gas Release Rate pCi/sec 1.00E405 3 75E+0f4 7-11-76 8-18-76 9-19-76 10-8-76 11-15-76 12 18-76 7-11-76 Date:

' Tech. Spec. Limit for Maximum i

A le Gas Release (1) 3 06E*01 1 32E+01 2.40E+01 ~

'5.94E*00-1.81E+0T 1 56E*00 3 06E*01

~.62E*01

'3 00E402.

1 Maximum % of Tech. Spec. Limit'(1)

'3.06E401 3 00E402 7 09E401 3 18E+01-2 9E*01-(1) Basi s Tech. Spec..3 8.C.1.

'(4) Average for 6 month period.

(2) Basi st - Tech.. Spec. 3.8.C.2

( 5) Maximum for' 6 mont h period.

(3) Quarterly

'(6) Determined by' ratio method.

3 i

Revised 2/78 - Mixed Noble Gases value for~ July corrected?

i 4

.j

0 0

4 2

4 2

1 1

1 4 5 3

3 4

5 3 3 3

3 4 4 4 5 5 2

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0 0

0 N

0 0

0 0

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0 0

0 0

0 0

0 0

0 0

0 0

0 d

+

+

+

+

L E

E E

E E

E E

E E

E E E

E E

E E E

E E

E E

E E E E

e 2 4 0

7 4 8 0

0 5

5 9

4 3

f 5

t 0

7 9

2 2

3 3

t A

6 0

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PHILADELPHIA ELECTRIC COMPANY PHILADELPHIA,

4 PEACH BUTTOM ATOMIC-POWER STATION 1

UNIT NOS. 2 ND 3 DOC KET. NO S. 50-277 6 50-278 1-4

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4 SEMI-ANNLAL EFFLUERT RELEASES REPORTJ-NO. 3 ' "

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JANUARY I, 1977 'THROUGH JUNE' 30,' 1977

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SUBMITiED. TO THE UNITED STATES, NUCLEAR -REGULATORY COMMISSION 3

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PURSUAMT TO y

FACILITY'_0PERATISG dCENSE NO.- DPR-% GT 56.

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PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION Unit Nos. 2 and 3 Docket Nos. 50-277 & $0-278 '

SEMI-ANNUAL EFF LUENT RELEASE S REPORT NO. -3 JANUARY.1,.1977 THROUGH JUNE 30,-1977 Submitted to The United States Nuclear-Regulatory-Corrmission Pursuant to Facility Operating License No. OPR-M & $6

)

i 1

Preparation Directed-By:.

W. T. Ullrich, Superintendent Peach-Bottom Atomic Power Station

[.

t

6 I

TABLE OF CONTENTS Pace No.

I.

Introduction i.

II.

Tables A - Gaseous Radioactive Release 1

8 - Isotopic Analysis of Gaseous Radioactive Releases 2

C - Liquid Radioactive Release Data 3

0 - Isotopic Analysis of Liquid Radioactive Re leases 4

l E - Solid Radioactive Waste Shipment

5 i

4 I. INTRODUCTION In accordance with. the Unique Re-porting Requirements of Technical Specification 6 9 3.,- this report summari zes the Ef fluen t Release Data.

for Peach Bottom Atomic Power' Station Units 2 & 3.

This data covers the period January.1, 1977.through June 30, 1977.

The not'ations E+ and E-are used to denote positive and negative exponents to the base 10.

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2-

Table C PEACH BOTTOM UNITS 2 & 3 - LIQUID RACIDACTIVE RELEASE DATA 1977 JAN.

FEB.

HARCH APR IL MAY JUNE TOTAL Gross Activity (P.y)

Total Curles Except Tritiwa 3.22E-01 2.22E-Ol 2.56E-02 2.70E 42 3.24E-08 5.89E-01 1.58E+00 Ave. gil/mi Gross AcElvity (except Trittura) at Point of Release 1.10 E-08 1.20E-08 1.66E-10 1.78E-08 3 72E-08 la.24E-08 1.40E-CP Total Cur tes of Tritium 1.16E +01 8.85E+00 5.59E +00 9.92E+00 7 74E+00 4.33E+00 4.ME+08 Average pct /ml Tritfue at Point of Release 3.93E-07 4.63 E-07 4.40E-07 4.20E-07 8.90E.07 3.12E-07 4.45E-07 Total Curtes Alphe 3.69E-05

2. ISE-04 8.48E.07 6.40E-07 5.36E-07 1.66E-06 2.36E-04 Average pCi/ml Alpha at e

Point of Release 5 73E-13 1.83E-II 6.68E-P 2.7tE-14 5.93E-14 1.19E-13 2.38E-12 g

Total Curles of Olssolved Noble Gases 3.40E +00 1 74E-03 1.15E+00 1.44E-01 2.83E-Ol

3. 28E-01 3.48E+00 Average pEl/ml of Noble Gases At Point of Release 4.75E-08
9. l lE-09 9.06E-08
6. 80E-09 3 25E-08 2 36E-08 3 22E-08 riaximu1n pct /mi Released except Trittura - At Point cf Release 1.08E-07 2.45E.07 1.09E-08 1.7BE-08 2.35E-07 3.08E-07 3.08E-07 (3)

Total Volume Gallons:

1.40E+06.

I.17E*06 7.13E+05 1.02E+06 1.29E+06 9.47E+05 6.54E+06 of Wastes Liters:

5.24E+06 4.42E+06 2.70E+06 3.86E*C6 4.87E+06 5 58E+06 2.47E+07 i

Total Volune Gallons:

7.84E+09 5.64E +09 3 35E+09 6.23E+09 2.30E+09 3 67E+09 2.90E+10 of Ollution Literns 2.95E+ IO l.91E+10 1.27E*lo 2.36E+10 8.70E+09 1 39E+10 1.08E+11 (1) % of Tech. Spec. Curie Limit 4.83E+00 3 33r*00 4.47E.OI 4.05E-01 4.86E+00 8.84E+00 3 78E+00 (2)

(1) Basis - Tech. Spec. 3.8.8.2 (3) Hamlesas for 6 enonth period.

(2) Average for 6 month period l

T bl+

PEACH BOTTON ise!TS 2 & 3 - !$0 TOPIC ANALY515 0F LIQUID RADICACTIVE RELEA1ES (in Curies) 19J7 ISOTOPE JAN.

FES.

MMt.

APR.

I MAY JUNE C1 TOTAL i

Strontits= - 89 5.47E 04 2.13E-03 3.5f.E-05 2.28E-05

7. f.0 E-c 4 2.66E-03 6.84E-0)

Strontium - 90 2.97E-05 5 70E-05 7.67E-06 1.12E-05

1. 5 9E-04 8.27E-04 1.0 9E-0 3 Casitan - 134 6.20 E-0 2 5.83E-02 8.04E-03 8.37E-03 3 56E-01 2.40E-Ol 7.26E-01 Cestum - 137 8.70E-02 6.34E-02 1.16E-02 1.18E-02 4.58E-OI 3.12E-01 9.44E-01 fodin - 131 5.00E-01 1.31 E-02 3.3 5E-04 1.83E+00 2.75E-02 2.37E+00 Cobalt - 58 1.00 E-0 2 1.92E-04 1.02E-02 Cobalt - 60
2. 70E-02 3.89E-03 8.63E-04 2.28E-03 3.05E-02 2.53E-02 8.79E-02 Zinc - 65 9.30E-02 4.59E-02 9.14E-03 4.91E-03 3.96E-CI 3.52E-01 9.0lE-01 a

Manganese - 54 3.90E-03 4.30E-0) 1.51E-03

$.59E-03 3.93E-02 Chromium - 51 6.70E-02 6.09E-03 6.28E-04 1.26E-02 8.63E-02 Zircontum - 95 1.60E-02 6.22E-03 3.58.E-03 2 76E-04 2.80E-0 2 Holybdenum - 9')

1.70E-03 8.48E-03 1.75E-03 4.93E-03 Lanthanurs - 140 8.20E-04 3.le7E-04 1.61E-03 2.78E-03 Arsenic - 76 2.20E-CI 3.07E-02 3.49E-02 4.9f E-03 3.49E-0 2 3 25E-01 Sodlura - 24 6.20E-03 4.88E-03 8 77E-04 1.63E-02 1.89E-02 4.65E-02 Neptonium - 239 1.00E-43 1.0CE-03 Iodine - 132 l.18E

  • I.lBE-04 Iodine - 133 4.90E-03
1. ISE-03 8.91E-03 6.13E-03
2. I I E-02 Iodine - 135 3.IIE-03 1.13E-03 4.24E-03 Silver - IlOM 1.40E-02.

2.38E-03 4.61E-03 1.43E42 1.14E-02 8.02E-02 1.27E-01 Ceritsu - 144 3.20E-03

2..

3.20E-03 Stront f ura - 92 2.62E-04 2.62E-04 Total 3.12E+00 2.33E-05 7.57E-02 4.46E-02 3.ls.E*00 1.10E*00

,5 72E+00 toss than mininuma detectable activity i

Table E l E ACH BOTTOM UNITS 2 6 3 - SOLID RADIDACTIVE WASTE SHIPttENT 19/7 JAN Fte tout CH AntIL MAY JmE TOTAL Namber of shipaments 17 20 27 26 28 33 18 9 Voksne of waste (f t )

4.74E+03 6.03E+03 6.84E+03 7.64E +03 6.99E+03 1.02E*C4 4.24E404 3

Activi ty, Cu-te s 4.86E+01 4.90E+0I 1.06E+c2 7.25E+01 2.42E+02 1.06E+02 6.24E +02 shipping Dates /# of shipnents 4/2 t/l I/I 1/2 1/8 1/3 5/t 2/2 2/1 4/2 3/I 2/I 6/1 3/

3/I 5/l 1/2 3/

9/1 7/I 6/I 6/2 6/3 6/l 10/I 9/2 8/I 8/2 WI 8/3 82/2 II/t WI 10/1 10/1 VI I4/2 15/2 11/I 11/i al/t to/

18/l 17/l If/I 12/2 12/3 13/2 23/I 18/l 15/2 17/1 13/1 Ildt m

25/3 IW1 86/s IWI 16/1 35/2 i

26/l 20/

17/2 20/1 17/1 I6/2 27/l 22/3 18/l 21/2 18/l 17/l 30/

23/l 20/l 22/I 20/2 20/2 3s/I 2tdl 21/2 24/l 2/

21/1 25/

23/2 26/2 2t/l 22/2 27/2 25/2 27/2 25/2 23/2 27/1 2V2 26/2 21/ 2 27!I I!

All Solid Radioactive Weste 3I!I I

Shipped by Hittman Nuclear and 30/2 Development Ecrp. in Truds to ths Chem. Nuclear Corp.

Berrn ell, South Carollia.

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