ML20238A524

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Semiannual Operating Rept 5 for Jul-Dec 1975
ML20238A524
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 12/31/1975
From: Ullrich W
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
References
EFF-75B, NUDOCS 8708310113
Download: ML20238A524 (54)


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PHILADELPHIA ELECTRIC COMPANY -

PHILADELPHIA PEACH BOTTOM ATOMIC PO1"ER STATION UNIT NO. 2 AND UNIT'NO. 3

(

i SEMI-ANNUAL OPERATING REPORT NO. 5 ,

July 1, 1975 through December 31,.1975 ]

.)

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i Submitted to l 1

The United States Muclear Pegulatory Commission- l l

Pursuant to  ;

1 Facility Operating License'No. DPP-44 & 56 l

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J I 8708310113 751231 l

PDR ADOCK 05000277  ;

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Philadelphia Electric Company ,

J Peach Bottom Atomic Power Station l

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4 SENI-ANNUAL OPERATING REPORT 4 I NO. 5 4 I

July 1, 1975 through December 31, 1975 i l

l Submitted To The United States Nuclear Regulatory Commission l

Pursuant To l Facility Operating License No. DPR-44 & 56 l

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Preparation Directed by:

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

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.t a PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET o 1

- PHILADELPHIA PA.19101

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(215:94i 4ooo ]

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February 27, 1976 .y

& E y( p, Mr. John G. Davis CN 87 N6 'I 8

Acting Director Office of Inspection and Enforcement g' United States Nuclear Regulatory Conrnission j g Washington, DC 20555 j

O b g

Dear Mr. Davi s:

Subject:

Semi-Annual Operating Report ~!

Peach Bottom Atomic Power Station .

Docket Nos. 50-277 and 50-278 3 Enclosed are forty (40) copies of the fifth i '

Semi-Annual Operating Report for Peach Bottom Atomic Power Station Units 2 and 3 covering the period of July 1,1975

  • through December 31, 1975 I

This report is being submitted in compliance with the Technical Specifications of Operating License DPR-44 and DPR-56.

Very truly yours,  ;

W. M. Iden .

Engineer-In-Charge ,

Nuclear Section  ;

Generation Divi sion - Nucl ear t

Enclosures 1997

TABLE CF CONTENTS Pace ]

I. OPERATIONS A. Summary 1 B. Unit 2 Operations 3 C. Unit 3 Operations 5 >

D. Reactor Shutdowns 8 5 E. Power Generation 11 l

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II. EQUIPMENT I

A. Equipment Performance 12 B. Changes to the Facility 12 C. Design Fatigue Usage Evaluation Report 14 III. MAINTENANCE

SUMMARY

A. Equipment Maintenance 15 '

l B. Instruments and Controls 26 IV. RADIOACTIVE. WASTE A. Liquid Pedioactive Release Cata 37 <

l B. Isotopic Analysis of Liquid Radioactive Releases 37 C. Gaseous Radioactive Release Data 37 i

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TABLE OF CONTENTS j Page D. Solid Radioactive Waste Shipments 37 V. PERSONNEL EXPOSURE 43 I

VI. ADMINISTRATIVE I l

l A. Technical Specification Compliance' 44 L

B. Organization Changes 44' C. Training 45 1

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l TABLES l I

i Page I-D-1 UNIT 2 REACTOR SHUTDOWNS 9 ,

I- D- 2 UNIT 3 REACTOR SHUTDOWNS 10 III-A-1 PEACH BOTTOM UNIT 2 SIGNIFICANT 15- 1 (a) thru (e) MAINTENANCE

SUMMARY

20 I III-A-2 PEACH BOTTOM UNIT 3 SIGNIFICANT 21- 4 (a) thru (e) MAINTENANCE

SUMMARY

25 III-B-1 PEACH BOTTOM UNIT 2 SIGNIFICANT 27- I (a) thru (e) MAINTENANCE

SUMMARY

- INSTRUMENTS 31 AND CONTROLS III-B-2 PEACH BOTTOM UNIT 3 SIGNIFICANT 32- 1 (a) thru (e) MAINTENANCE

SUMMARY

- INSTRUMENTS 36 l AND CONTROLS IV-A PEACH BOTTOM UNITS 2 ANC 3 - 38 LIQUID RADIOACTIVE RELEASE DATA IV-B PEACH BOTTOM UNITS 2 AND 3 - 39-ISOTOPIC ANALYSIS OF LIQUID 40 RADIOACTIVE RELEASES l ACTIVE RELEASES i IV-C PEACH BOTTOM UNITS 2 AND 3 41 GASECUS RADIOACTIVE RELEASE DATA IV-D PEACH BOTTOM UNITS 2 AND 3 42 SOLID RADIOACTIVE WASTE SHIPMENTS FIGURES ,

PAGE 1 UNIT 2 REACTOR POWER LEVEL 46 2 UNIT 3 REACTOR POWER LEVEL 47 iii -

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'I . . OPERATIONS A. Summary .

Histograms displaying the power history of both reactors from July 1, 1975 through December 31, 1975 are provided as Figures 1 and 2 on pages 46 and 47 respectively.

At the beginning of the reporting period Unit 2 was operating at full load and Unit 3 was operating at 50%

power, 50% core flow in accordance with NRC direction.

By 7/4, Unit 2 was in the procens of reducing core flow in 10% increments, with power spectral density data being taken on selected local power range monitors, in an attempt to correlate power and flow conditions to LPRM tube-channel box interaction.

As the phenomenon continued to be studied by the NRC and the industry, both units were operated derated' at a nominal 490 MWe after 7/6. Cn 7/25, new limits were issued by the NRC which permitted a small power increase. On 7/26 both units were at 540 MWe and 40% '

core flow. On 8/2, NRC permission was received.to  ;

increase Unit 2 core flow to 100% for a short time to accommodate acoustic monitoring of several selected LPRM instrument strings.

The Unit 2 testing program continued until late on 8/5 when plant electrical problems resulted in the tripping.

of both units. The units were returned to operation the following day. Unit 2 operated at a nominal 540 MWe througn 8/16, when it was shutdown for the installation of accelerometers on LPRM instrument tubes and returned to power the next day. Testing continued.until 8/22.

Unit 3 experienced three short shutdowns and returned to power operation and its previously interrupted testing program, on 8/10. By late on 8/16, Unit 3 testing was completed. After testing, both units returned to operation at a nominal 540 MWe. Unit 2 operated until l the end of October at nominal 540 MWe with the exception i

of a one day outage for a steam leak inside the drywell and two short power reductions.

l Unit 3 tripped on 9/2 as a result of instrument surveillance testing and was immediately returned to power until 9/17 when a shutdown was initiated for '

reactor recirculation pump seal maintenance. The unit was returned to power on 9/23 and operated at a nominal 500 MWe until 10/28 when dust accumulation on the

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generator field slip rings forced a generator shutdown.

The brush rigging fan assembly was cleaned and the unit  !

was returned to operation on 10/28.  !

On October 31, Unit 2 was removed from service to ,

accommodate plugging of the core plate bypass flow holes, All fuel bundle channels adjacent to instrument strings were inspected and rechannelled as necessary. There were indications of channel corner wear, but there were  ;

no through-wall holes or missing pieces. The outage provided time to perform significant maintenance on the i unit, including removal and testing of the generator j rotor. By late on 11/28, a reactor hydrostatic test had 1 been completed and by early on 11/30 the reactor was j critical. j Unit 3 operated during the entire month of November at a nominal 580 MWe.  !

Unit 2 was at 550 MWe by 12/2 and preconditioning fuel at 5 MWe/HR. Power was varied between 500 and 950 MWe )

from 12/3 to 12/11 due to conventional plant problems.

Late on 12/11, the unit was removed from service to repair a feedwater heater leak and was returned to service on 12/15. By early on 12/21, the unit had reached full load end remained at that level until a turbine trip and associated scram from low EHC oil pressure on 12/25. Tne unit was returned to service early on 12/27 and had reached a nominal 830 MWe on 12/29 when a scram occurred during MSIV surveillance testing. The unit was restarted and was at a nomin61 900 MWe as the reporting period. ended.

Unit 3 operated at a nominal 590 MWe until early on 12/25 when it was removed from service to repair minor steam leaks inside the drywell. The unit was restarted on 12/25 and was at 900 MWe on 12/30 as part of an NRC approved test program, when a scram occurred from low condenser vacuum. The unit remained out of service for the scheduled core plate bypass flow hole plugging outagc, which was in progress as .the reporting period ended.

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n. Unit 2 2D9Eations 4 July 1975 The month started with the unit at full load in the'"A" 1 rod sequence. By late on the 4th, the fuel j preconditioning soak had been completed and 10% core flow decreases were being started. University of Tennessee personnel were taking Power Spectral Density 3 (PSD) data froh selected local power range monitor j ficcion chambers as a function of core . flow. The 1 I

purpose of the data gathering was to determine at what core flow the physical interaction between fuel bundle .

channels and in-core tubes' ceased. By late on the 6th,  !

the testing had been completed and the unit had reached , l the operating conditions of 50% core flow and 50%

, thermal power.

l l The unit remained at a nominal 490 MWe until the 25th ]

when word was received from NRC that new limits were >

\

being imposed as a result of additional information concerr.ing the channel box - LPRM instrument tube 4 interaction phenomena. y I The limits were:

a) 40% core flow b) 85% MAPLHGR limit c) 3.35 MWt maximum per bundle d) recire M-G set units mechanically prevented from exceeding 50% core flow These limits were accommodated by the 26th and the unit remained at a nominal 540 MWe through the end of the month.

August 1975 The unit continued to operate at approximately 55% power and 40% core flow until late on the 5th when a scram occurred due to a loss of voltage on both Reactor Protective System buses. This voltage loss was due to an electrical transient resulting from a delayed transfer of auxiliary power following a Unit 3 scram on low condenser vacuum (See Unit 3 operations' report) .

The unit was returned to service the next day and continued to operate at 55% power until the 16th when it 1

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wasi .chutdown to permit installation of accelerometers on selected LPRM housing flanges in the sub-pile room. The unit was returncd to service the next day and flow and power were increased in steps to permit acquisition of LPRM vibr4 tion data. The data acquisition program brought the. unit up to full core 11ow (approximately 80%

L power) by the 21st. Data continued to be taken as core flow and power were decreased, with the test program reaching completion by midnight on the 22nd. The Unit remi;ined at 40% core flow and 55% power for the i remainder of the month.

September 1975 l The unit continued te operate at a 40% core flow and approximately 55% power until the 5th when a shutdown was initiated to repair a valve packing leak which had increased the drywell pump-out rate significantly. The unit was resynchronized on the 7th and returned to 55%

power and 40% core flow.

Operation' continued at this level until the 27th when a power reduction to about 430 MWe took place while exchanging from control rod sequence A to sequence B.

As power was being increased early on the 29th, r following the sequence exchange, a routine surveillance test revealed that outboard MSIV's CSD had closing times l } less than those required by the Technical Specifications. . Power was reduced to about 335 MWe to permit entry into the. steam t unnel' to add hydraulic i fluid to the MSIV operating pistons and to reset the valve stroking time. The palves were left with acceptable stroking times, and the plant was back to 55%

power and 40% core flow on the 30th.

October 1975' The unit operated in the 'B' control rod sequence at 40%

core flow and approximately 55% power from the beginning l of the month until late on the 31st, when the plant was l shutdown for the core plate bypass flow hole plugging l outage.  !

November 1975 Movement of fuel from thh reactor to the spent fuel pool began on the 6th. Fuel channel inspection and core I i plate bypass flow hole plugging proceeded to completion i on the 19th. By the 23rd, all fuel assemblies had been -l returned to the reactor and all required verifications  !

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were completed. The reactor vessel operational hydrostatic test was successfully completed on the 20th.

Other significant items accomplished during the outage were the modification of the circuitry required to eliminate the LPCI loop selection process and the removal of the main generator field for inspection and testing. The generator was synchronized late on the 30th.

December 1975 On the 1st the unit was at 360 MWe, moving up to the preconditioning envelope on "An sequence control rods.

By the 2nd the "A" rod pattern was established and fuel preconditioning via reactor recirculation flow was in progress. Between the 3rd and the 11th the unit power was varied between 500 and 950 MWe due to several mechanical problems in the conventional plant. Late on the 11th the unit was removed from service to repair tube leaks in one feedwater heater. Two additional feedwater heaters were inspected, with a few tubes being plugged in each.

By late on the 14th, the reactor was critical, and the generator was synchronized early the next day. By late on the 15th, the unit was at 500 MWe. The final rod pattern was set by early on the 17th at 600 MWe. Fuel preconditioning was started with the Unit reaching full load on the 21st. The unit operated at full load until late on the 25th when low pressure in the main turbine control oil system initiated a turbine' trip and associated scram.

The defective oil pumps were repaired and the reactor was critical by late on the 26th. The generator was synchronized by early on the 27th and had reached 830 MWe when a reactor scram occurred during MSIV surveillance testing early on the 29th. The unit was back on the bus within a few hours and was at a nominal 900 MWe as the reporting period came to a close.

C. Unit 3 operations J'uly 1975 The month started with the unit operating at the limits of 50% power and 50% core flow, in consideration of the channel box LPRM instrument tube interaction, in the "B" control rod sequence. Operation continued at this 3 evel

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until late on the 18th when a rodhswap from B to A sequence was initiated to accommodate fuel management req uirema t s . By early on the 19th, the rod swap was complebad and operation continued at 50% flow, 50% power until tne 25th when the new NRC limits were implemented,.

, exactly ac with Unit 2. By the 26th the new limits were achieved and the unit operated at a nominal 540 MWe through the end of the monM Aucuat ,1975 N3 Operation continued at 40% corb flow and 55% power until s the 2nd when NRC permission was obtained to go to full flow to accommodate acoustic monitgrjng or 9 selected LPRM strings (data obtained froin accelerometers on TIP d

drive cablcs) . .The power ascension ind testing .

continued until'tne 5th when 4 condenser lor vacuum /

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scram occurred from approximately 10?p MWe. The loss of I

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vecuum was the result of tripping of the Unit 3, s '

recombiner, which resulted from ad t electrical t ransient caused by tlhe overloading of a 4KV electrical bus which

, was beinq ;;owered solely by the emergeticy Diesel '

3 Generatore This electrical line-up was required by. the scheduled cutaie of the No. 2 start-up feed,7and was s

dcne according to procedure ,' ,

The unit was returned to operation on ths 6th,'but had to be shutdown almost immediately to tiepair a leaking feedwater check valve. s 3 3' a

Following repairs the unit was returnet to Operdtion on the 9th only to experience a scram due. to-inappropriate valving of a moisture separator level switch. The unit was restarted immediately and synchronized early on the 10tn. The turbine trippeC shortly thereafter due to the failure (closed) of a moisture separator ledel control dump valve. Following temporarysrepairs to 'the valve, the turbine generator was re-synchronized and the LPRM vibration data test program resumed. [Dy the 14th, the unit had reached full flow and 1065 v.Wo. Tne test program continued as flow and load were red 6ced in steps the following ddy. By the 17th, the unit uds itable at 400 pore flow end 55% power where it rdmained for the balance of the 6onth.

[ijptember 1975 ,

Unit operation at 55% power and 40% core flow was 4

' ititerrupted on thO 2nd by a scram and Group I isolatior.. I due to instrum ent surveillance testin7 The generator ,

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.1 was synchronized 'early the following day and was being -

brought back to a nominal 55% power when an increase in 1 reactor coolant conductivity, apparently due to Reactor Water Clean-Up (RWCU) -system resin breakthrough,- ]D required a temporary power. reduction. The reactor.

coolant chemistry was normalized by changing'the mode.of.

the RWCU system operation, and the unit was-back at 55%

power on the.Sth. Operation continued.at this power level until the 17th,_ when the unit was removed from service to replace the 3A ' recirculation pump seal. This L seal had indicated a stea'dy increase in .first stage

. leakage for_several days. The outage work was completed and the unit was returned to service,on the 23rdi Power was stabilized. at 55% and . remained there for the balance of~the month except.for a brief' reduction to 421 MWeLon the 28th to repair a circulating water _ pump traveling screen. This last incident occurred during the very high river flows due to Hurricane Eloise. 4 October 1975 4 The' unit operated in the "A"-control rod sequence at j approximately 55% power and 40% core flow from the beginning of the month until'the 8th when the core flow was increased to 65%. The increased flow, which resulted in an electrical output of 736 MWe, was ,

required to obtain LPRM vibration date from transducers I mounted on selected instrument nozzles. By late on the 9th the reactor was back at a nominal 55% power and 40%

core flow. Operations continued at.this level until the 18th when power was reduced to approximately 43% to I accomplish an "A" to "B" rod sequence exchange. By the 20th, unit output had stabilized at approximate]y 55%

power, where it remained until the 28th when the generator was removed from service. 'This outage was I required to clean the generator field collector fan j housing of a build-up of carbon dust, which had resulted ,

in a generator field grvund relay target. The generator 1 was resynchronized early the following day, and_ returned to 55% power where -it remained through the remainder of j l

the month. 1 November 1975

( The unit operated at a nominal 55% power and 40% core l flow throughout the entire month.

Plant electrical output was increased from 580 MWe to 1 592 MWe during the month by optimizing core power shape "

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while p.till staying within the temporary restrictions of- l 3.35 11W per bundic and 409, core flow.

l December 1975 j The unit continued to operate at a nominal 590 1We, 1 until late on the 24th when an increase was. observed in l the pump-out rate of the drywell floor drain sump, which necessitated a unit shutdown. Three valvo packing lenha  ;

inside the drywell were repaired and the reactor was critical by late on the 25th. The unit flow and power were subsequently raised to conduct an t'!RC approved testing program associated with the plugging of the core ,

plate bypass flow holes. The unit had reached 900'MWe I by the 30th, when a turbine trip and scram were initiated by low condenser vacuum.

Since the core plate plugging outage was scheduled for January 2, 1976, the unit was not returned to service and the outage was started two-days early.

D. Reactor Shutdowns

1. Reactor shutdowns for Unit 2 are listed in Table I- I D-1.
2. Reactor chutdowns for Unit 3 are listed in Table I-D-2.

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E. Power Generi1 tion July 1, 1975 through December 31, 1975-Unit 2 Unit 3 Gross Thermal Power, MWD 290,163 321,112 Gross Electrical Power, MW11 2,175,850 2,418,380 Net Electrical Power, MWil 2,070,655 2,309,418 Reactor Critical, firs. 3,574 4,154

, Generator Synchronized, 1]rs . 3,508 4,102 1 j 1

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II. EOUI PM Ef1T l

A. Equipment Performance

1. Fuel l Fuel performance for both units was good during the reporting period. Start-up operations were {

conducted in accordance with the General Electric Preconditioning Interim Operating Management l Recommendations (PCIOMR) . Control rod sequence l exchanges were made at 1000 to 1500 MWD /T average f core burn-up intervals to develop a more uniform { '

core exposure distribution.

The total control rod inventories at full power i have followed the predicted reactivity trends of j increasing core reactivity with increasing fuel i I

burnup during the initial part of the cycle and decreasing reactivity after 5500 MWD /T. Unit 2 and 3 average core burnups were approximately 8515 MWD /T and 5395 MWD /T respectively at the end of the reporting period.

B. Changes to the Facility Julv 11 1975 through December 31, 1975

1. RHR Heat Exchangers Stiffeners were installed in four tubes of the 2A RHR heat exchanger to reduce the probability of further tube failures. This change was recommended by the vendor. Because the modification enhances RHR heat exchanger reliability and has no significant effect upon heat transfer capability it has no adverse impact on plant safety. The remaining heat exchanger will be modified as conditions permit.
2. High Pressure Service Kater System Three-Stage orifices were installed in the High Pressure Service Water System in the effluent lines from the ASD RHR heat exchangers in both units.

These orifices provide the pressure drop previously supplied by a control valve. The control valve had been a very high maintenance item. Because there

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has been no change in system operation intent, and because system reliability has been enhanced, there is no adverse impact on plant safety.

3. Drywell ventilation system f Additional ventilation fans and ducts were added inside the Unit 2.drywell to correct localized high temperature problems. These additions were designed and constructed similar to the original ventilation system. Because these changes reduce local temperature and extend equipment lif e there is no adverse impact upon plant safety.
4. Fuel Pool Cask Support Fixture The fuel pool spent fuel shipping cask support l fixtures in both units were modified to accept a smaller cask. Because the changes were made in accordance with original specifications there is no ]

adverse impact on plant safety. 1 5, Fuel Grappl.e i I

on both units, the grapples on the fuel hoists were ]

modified to remove projections which could l inadvertently engage a fuel element and were  !

modified to add limit switches to sense fuel bail i position and correct grappling. The changes were i recommended and installed by the vendor. Because these changes increase the reliability of the fuel grapple there is no adverse effect on plant safety.

6. Third Reactor Level and Pressure Indicators A third instrument loop was installed in Unit 2 to supply indication only of reactor water level and vessel pressure. This change was recommended by the designer, makes both units identical in this respect, and is intended to assist the operator in determining the correct value of level and pressure during conditions of disagreement between the original two loops. There is no adverse effect upon plant safety.
7. Reactor Protection System Instrumentation The reactor pressure switches on Unit 2 were replaced with indicating transmitters and electronic bistable switches to preclude scrams 13 -

which had occurred due to panel vibration. This change vas approved by the NRC and appropriate license and Technical Specification changes were issued.

8. LPCI Logic On Unit 2, the electrical logic which initiates the LPCI system was changed to renovo loop celection ,

and to permit injection into both recirculation l loops at the same time. The change was approved by I the NRC and appropriate license and Technical l Specification changes were issued.

9. Core Plate Holes i On Unit 2, the holes in the lower reactor core l plate which were designed to supply cooling water l flow to incore instrument tul.es were plugged to l preclude tube vibration. This change was approved )

by the NRC and appropriate license and Technical I Specification changes were issued.

I C. Desian Fatigue Usage Evaluation Report

'There were no operational transients on either unit during the report period which were com.paralle to or

r. ore severe than the transients evaluated in the stress report code fatigue usage calculations.

1 1

s III. MAINTENANCE

SUMMARY

A. Equipment Maintenance d

1. Equipment maintenance for Unit 2 is shown in Table III-A-1 (a) through Table III-A-1 (e)
2. Equipment maintenance for LUnit 3 is shown in Table III-A-2 (a) through Table .III-A-2 .(e) l

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3. Instruments and controls maintenance for Unit 2 are shown in Table III-D-1 (a) through Table III-B-1 (e)
4. Instruments and controls maintenance for Unit 3 are shown in Table III-D-2 (a) through Table III-B-2 (e) 26 -

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. g E h n d C c i e r O N t p t e T E i . i a i R w h p . r f N R s h c y g i E U c t t a e l p

K C d t i o . l t A E e i w hh e n m

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N C C N es en e f a ,

B A E U p t o mn . o td i n t o i

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N T t r ai ee l r l phd ci r

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N I F

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t a t U ur i . e xs u e ee . er C

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i .

est ea t vt s ve iat ii O U i c t h h cr d , t ri t f T U A

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t c t t et t ep l a eec el

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wf S o io D c ew D s woe S t t a e F r eni em Di cD a

. H C

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=

l IV. RADIOACTIVE WASTE. '

A. _LJguid Radioactive Release Data See Table IV-A.

j B. Isotopi.c Analysis of Liquid Radioactive Releases See Table IV-B.

C. Gaseous Radioactive Release Data See Table IV-C.

l D. Solid Radioactive Waste Shir,ments )

See Table IV-D. 4

,1 1

l l

1

> l l

I l

l l l l

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2 2 5

7 r 9 e 00 1

b 1 9 8 8 66 1 1 m8 N N N N P P PP e7 6 8 2 2 7 56 0 9 9 c 8 7 0 9 A A A A 2 58 87 7 e S S S S .

D 1 2 4 5 N N N N 1 1 5 1 6 2 r

e 0 b 1 9 8 8 56 9 1 A m8 N N N N PP P P T e4 6 7 7 3 0 74 2 8 9 A v 0 6 1 5 A A A A 6 35 6 8 5 D o . S S S S N 1 3 2 7 N N N N 3 93 7 2 1 E

S A

E L ) )

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R e

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A 5 2 2 2 N N N N 5 73 72 1 O

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C

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P 3 3 3 3 t t 3

t t t f f f f f s e s e s e s e s e mt mt mt mt mt ees e ees e ees e ees e ees e e t t ait t t sit t t ait t t ait t t ait t aiaC s ai wC s aiwC s ai wC s aiwC s a d a d a d a d a d a d ff , w ff ,w f f ,

w ff ,w ff ,w gooy gooy gooy gooy gooy go n tf n tf n tf n tf n tf n i - ei o i reio i reio i r eio i reio pemv pemv pemv pemv pemv i

p pb uie pb uie pb uie pb uie pb uie ml t p pb

. i i ml t p i ml t p i ml t p i ml t p i

- h uoc y h u ocy h uoc y h uoc y h u oc y h S NvAT SNvAT S NVAT S NVAT S NvAT S R

E R B R E T M E B S E S M Y u T O E L 3 P U u T V E C 0

_ J A 3 C 1

\

V. PERSONNEL EXPOSURE' ]

The data tabulated below summarizes the recorded whole-body exposures for the site (Uni t 1 HTGR, DPR-12; Unit 2 BWR, DPR-44 and Unit 3 BWR, DRP- 5 6) during this report period.

I There were no personnel exposures sustained in any dose range I beyond those reported. The data furnished reflect the total number of individuals for whom personnel monitoring was j provided. )

l Individual values exactly equal to the values separating exposure ranges are reported in the higher range.

Exposure in ranges denoted by (*) are attributed to liTGR End- )

of-Life and Decommissioning activities.

SEMI-ANNUAL NUMBER OF INDIVIDUALS IN EACH RANGE _

DOSE RANGES HEALTH ]

MAINTENANCE OPERATING PHYSICS STAFF ENGINEERS TOTAL (R EM) l No Measurabic l Exposure - - - - -

1015 1

Measurable Exposure I

40.15 362 24 23 9 69 487 0.15 - 0.30 94 16 8 0 11 129 0 30 - 0.45 31 7 3 0 2 43 i

0.45 - 0.60 15 0 1 0 1 17 0 11 O.60 - 0 75 8 0 2 1 0.75 - 0.90 6 0 0 0 0 6 0 0 4 l 0.90 - 1.05 3 0 1 l

1.05 - 1.20 3 0 0 0 0 3 4 0 0 0 0 4 1.20 - 1 35 0 0 0 0 9 1 35 - 1 50 9 1.50 - 1.65 6 0 0 0 0 6 0 6 1.65 - 1.80 (*) 6 0 0 0 1.80 - 1 95 (*) 3 0 0 0 0 3 0 0 2 1.95 - 2.10 (*) 2 0 0 2.10 - 2.25 (*) 1 0 0 0 0 1 0

2.25 - 2.40 (*) 1 0 0 0 1 SUMMARIES 554 47 38 9 84 1747 l

VI. ADMINISTRATIVE A. Technical Specification Compliance i

Plant operations and surveillance testing were carried  ;

out during the reporting period in accordance with applicable Technical Specification requirements. System ,

deficiencies noted during the surveillance testing were 1 corrected as soon as practicable. Deviations from the required limiting conditions for operation and criteria established by the surveillance requirements were  !

reported to the NRC as Abnormal Occurrences in accordance with the Technical Specifications. .

l The following Technical Specification changes and license amendments were issued by the NRC: )

October 30 for both units The PS-55 modification to remove the " blind" pressure switches and install analog loops with electronic trip units.

November 28 Unit 2 ONLY j Amendment do. 15 to Unit 2 Operating License l l incorporating the LPCI modification and 10CFR50 )

Appendix K safety analysis with an accompanying order i which permits operation with plugged bypass flow holes.

December 30 Unit 3 ONLY  ;

I Amendment No. 12 to Unit 3 Operating License to '

authorize modifications to improve the functioning of the LPCI system in partial response to the 10CFR50 Appendix E application.

B. Organization Changes Name Previous Assignment Present Assignment J. G. Casey Asst. Control Operator Control Operator W. H. Truax Asst. Control Operator Control Operator W. A. Bradley Asst. Plant Cperator Plant OperatoJ.

D. R. Hooper Asst. Plant Operator Plant Operator G. A. Paton Asst. Plant Operator Plant Operator l

l

C. ' Training Three individuals are in training for the NRC Senior Operator License. Two Assistant Control Operators are in training to qualify as Control Operators. Three  ;

Assistant Plant Operators were-trained to qualify as Plant Operators. Six operators from Unit 1 are in training to qualify as Plant Operators. Audit tests were completed for the Senior Operator Licensee Trainees.

The Operator Requalificaticn Program continues. Annual l written tests .and oral tests have been completed or are in progress. On-shif t training is i n progress and includes the review of normal and emergency operating I procedures and Technical Specifications.

l Personnel safety meetings are conducted approximately each month. Additional training on site was completed

')'

in the areas of:

a. Radiological Safety and Plant Security
b. Respiratory Equipment
c. Fire Fighting
d. First Aid i

4 1

i l

I 1

1 l

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