ML20037A889

Annual Rept of Station Operation,1969

ML20037A889
Person / Time
Site:	Dresden
Issue date:	01/21/1970
From:	COMMONWEALTH EDISON CO.
To:
References
NUDOCS 8008070706
Download: ML20037A889 (61)

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l DRESDF.!! 1,"JCLEAR PC'.1ER STATIC:i A!C:UAL REFORT OF STATION OPERATICli FOR Tale ' LEAR 1969 January 21, 1970

TABLE OF CONTENTS PACC I.

Introduction 1

II.

Summary of Operations 1

A.

Scope of Operations 1

B.

Shutdewns 1

C.

Load Restrictions 5

III.

Discussion 5

A.

Operating Experience 5

1.

Generation 5

2.

Scrats 5

3.

Incicents 9

4.

Control Rod Drives 10 a.

Coatrol Rod Drive Operation 10 b.

Control Rod Drive Tests 13 c.

Control Rod Drive Inspection 13 5.

Control Rod Blades 21 a.

Blade Following Checks 21 b.

Control Blade Inspection 21 6.

Changes in Facility Design 22 a.

v=tari,1 Ta=r t.non in s tal l a t i on 22 b.

Rad-iiaste Discharge Line !!odificatica 22 c.

Unit 1 Diesel Generator !iedification 22 d.

Fuel Stcrage ?cci Modifications 23 e.

"A" Demineraliced '.iater Tank Tie Line 23 f.

ev Laundry Installation 23 g.

Auxiliary Steam Tie f ines to Unit 2 23 h.

River unter Sacple Station 24 1.

A.D.S. Equipment Installation 24 7.

Persennel Radiation Exposure 24 8.

Liquid Poisen System 24 9.

Radioactive '..'aste Disposal 25 10.

Fuel Assembly Cleaning and Testing 29 a.

Fuel Cleaning 29 b.

Flow Testing 29 c.

Bow Checking 29 d.

Orifica Changes and 1:odifications 29 11.

New Fuel and Plutonium Rod Installation 32 12.

Inspe :tions 33 a.

Irradiated Fuel Inspection 33 b.

Metal Surveillance 33 c.

Fuel Transfer Carrier Inspection 35 d.

Reactor Vessel Inspection 35 e.

Primary System Reid Inspection 35 1.

Reactor Flange Incpection 35 2.

Reactor Flange Stud Bolts 36 3,

Reactor Thimble.l eld Testing 36

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TA13LE OF CONTENTS (Con't.)

PAGE 4.

Primary Piping 36 13.

Emergency Condenser 40 a.

Sand Blasting and Repainting 40 b.

Hanhead Installation and Tests 40 14.

L.P. and I.P. Turbine overhaul 40 15.

Main Condenser Retubing and Feedwater Hester Replacement 43 a.

Condenser Retubing 43 b.

Replacement of "D" Secondary Feedwater Heater 43 16.

Tests 44 a.

Sphere Integrity Test Program 44 b.

Primary Steam Drum Safety Valves 45 c.

Temperature Coefficient of Reactivity Check 45 d.

Fuel Sipping 45 e.

Mini =um Critical Tests 45 f.

Shu.down Margin Checks 49 3.

License DPR-2 50 Correspondence References - 1969 52 w

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'1,

DRdSDEN NUCL?> R FC P 9TATICN M:NUAL REPORT I INTRCDUCTIO:!

This eighth annual report is submitted in cenpliance with paragraph 3.C(2) of the Utilization Facility License DF3-2, as n= ended, and covers operation of Dresden Nuclear Peuer Station during the year 1969.

II S'ZARY OF OPERATIONS A.

Scene of Operations Dresden Station operated during 1909 with a total cf 11 shutdowns.

The last of these uas the sixth partial refueling, inspection, and maintenance outage, which extended f rca Septcmber 6, 1969, until January 1, 1970. This cutage vas quite extensive and included items such as:

Complete retubing of the nain condenser; overhaul of the inter =ediate and low pressure turbine; replaccmant of "D" secendary feeduater heat 2r tube bundle; refueling of 9e fuel assemblies and cleaning of all fuel returning to the core for Cycle 7 operaticn; overhaul of 28 control rod drive s; inspection and testing of major primary systen welds; and general raintenance and incpections made available by the shutdown.

During the year, additions to and changes in facility design were made by:

. Complete retubing of the main condenser; replacement cf "D" secondary feeduater heater tube bundle; installation of a emterial test loop in "B" secondary stec_t geaerator compartment; codificction of the Unit al diesel generator; redificr tion of spe n fuel storage pool rerainers and rack reintorcement; installation of ' a" dan.ireraitzed wa ter tank tie-line to Unit #?: installation of Automatic Dispatch Sys tem (l.D. S.) equipment; cerpleticn of the Dresden St& tion Illinois River Sampling Unit; laundry discharge tie to Unit #2; new laundry addition; addition of a hypochlorite system; and a heating steam tie line to Unit #2.

One shipment consisting of 19 spent fuel assemblies cas shipped to t're Chemical Processing Plant of Nuclear Fuel Services, Inc., at Uesc Valley, Nau York during the period. Two shipments, consisting of 32 assemblies and three special containers containing single leaking elements were shipped to the U. S. AEC, Duront Dene= curs, in' Dun 3nrton, South Carolina.

B.

Shutdowns The plant was shutdoun 11 times during 1969 as shown in Table 1 and Figure 1.

Seven of these were forced outages, fcur of which were due to turbine condenser tube leak repairs.

There were four schedaled outages: ona for operator training; anc for condenser tube inspecticn and fee ' water heater repair; one for diesel genera tor modificatict, control red drive scram and f riction tests, and operator licensa exams; and one for the enjor refueling, inspecticn, and maintenance outage.

One of the four scheduled outages uas tenporarily extended during the year.

It occurred en December 23, fer a prirary pipir.g leak repair and turbine thrust bearing instrumentatico inve s t iga t ion.

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TABLE 1 DPERATING PEllyO3MANCE 1969 NO. OF OFW SYSTEM ON SYSTEM OUTAGE DATE TIME DATE TIME OUTAG8 IlotIRS OUTAGE CAUSE 99 1/23/69 2245 2/ 3/69 0304 24411tra. 19 Min.

Scheduled - Turbine - (228 lies. 31' Min.) Con-denser Tube Inspection and Feedwater tienter Repair' Available, Not Operating - (15 lies. '48 Min.)

Operator Training 100 2/15/69' 0113 2/16/69 1552 38 lira. 39 Min.

Forced Turbine - (38 IIrs. 39 Min.) Condenser Tube Leak Repair 101-2/18/69 1153 2/19/69 0740 19 liri. 47 Min.

Forced Reactor - (19 Hrs. 47 !!in.) Primary Systern Drain Line Valve Repair 102 4/ 9/69 2330 4/13/69 2359 96 I!ri. 29 llin.

Scheduled Reactor - (96 lirs. 29 Min.) Diesel u

Generator Modification, Scram and Friction Tests, and Operator License Exams 103 4/14/69 0320 4/18/69 0544

'98 IIrs. 24 Min.

Forced Turbine - (98 lirs. 24 Min.) Secondcry St eam Chec t Drain llender and Valve MO 118 Repair 104 5/10/69 0055 5/11/69 0728

. 30 lies. 33 Min.

Available, Not Operating - (30 lirs. 33 Min.)

Operator Training 105 7/27/69 1940 7/28/69 '0408 8 lirs. 28 Min.

Forced Plant - (8 Hrs. 28 Min.) Reactor Scram Due' To Voltage Transient While Switching 106 3/ 2/69 1322 8/ 3/69 2334 34 IIrs. 12 Min.

Forced Turbine - (34 lirs.12 Min.) Condenser Tube Leak Repair f

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Table 1 (Cont.)

NO. OF OFF SYSTEM ON SYSTEM OUTAGE DATE TIME DATE TIME OUTAGE HOURS OUTAGE CAUSE 107 8/ 4/69 1210 8/ 5/69 1323 25 lirs.13 Min.

Forced Turbine - (25 lirs. 13 Min.) Con-denser Tube Leak Repair 108 8/14/69 2306 8/16/69 1105 35 !!rs. 59 Min.

Forced Turbine - (35 IIrs. 59 Min.) Con-denser Tube Leak Repair, 2,803 iles. 38 Min.

Scheduled Plant - (2,606 Hrs. 38 Min.)

109 9/ 6/69 0022 Sixth Partial Refueling, Major Control Rod Drive Overhaul, I.P. and. L.P.

Turbine Overhaul, Ec.crgency Condenser Painting and Fuel Cleaning Forced Reactor - (157 Iles. 00 Min.)

Primary Piping Leak and Primary Drum Level Instrumentation Forced Turbine -- (45 lies. 00 Min.)

Turbine Thrust Bearing Instrumentation TOTAL OUTAGE TIME 3,440 HRS. 41 MIN.-

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5 C.

Load Restrictions The load restrictions it.iposed during the year are listed in Table 2.

Restrictions were due to feedwater heater outages, rod worth tests, secondary steam generator outages, and fuel burnup limitations.

III DIS CUS SION A.

Onerating Excerience 1.

Generation The total reactor operating (critical) time during the year was 5,468 hours0.00542 days <br />0.13 hours <br />7.738095e-4 weeks <br />1.78074e-4 months <br /> and the total pewer for the period was 120.493.2 1MD. The gross electrical generation during the year was t

873,284,92 MW"de; net generation was 825,452.2 InTde. As of December 31, 1969, the total gross generation since cenmence-nent of pouer operatica on April 15, 1960, was 9,294,4C5.19 M',Gie.

2.

Scrnns a.

At 4:32 p.m.,

on February 2, a reactor scram was initiated by a fast period on channels 9 and 11 while uithdrc9ing control rod E-4 from position 2 to 3.

All instrumentatien indicated that a previous withdraw signal given to control red E-4 had failed to nova the red, although the rod ims cetually unictched and was drifting sioal' to nosition 3.

A seccad eithdraa sicnal n.oved the rod rapidly to position 3 and the reactor scramned on fest pericd.

3 The,reccter nas ct.320 ? and heating uith 34 rods cnd four notchee uttacrawn at the ti:.e or the s crcn.

b.

At 1: 13 a.m., on February 19, a reactor scram una initiated by inadvertent eperation of No. 2 vacuum trip.

The reactor was at 45 M7t with 1-1/4 bypass valves open and 60 control rods withdrawn.

Praparations were being rade to roll the turbine with steam when "B" cican-up purp tripped on high water temperature to the demineralizer, 'Ihe hi2h water temperature was dae to a large amount of blowdet:n.

When the pump tripped, bloudcun was stopped - and the water levcl increasad, tripping the turbinc. Uhen icyc1 was restored, the operator recched for to. 1 vacuum trip to reset the turbine, and tripped no. 2 vacuum trip by mistake, scremming the reactor.

c.

At 10:59 p.m.,

on April 17, while returning Unit 01 to operation follouing an cutage for centrol rod scram and friction tests, the recctor scrc=r.ed.

This scram was initiated by the No.1 vacuum trip, uben reactor pressure exceed 4d 200 psig at a =ain condencer vacuum of 21 inches Hg.

6 TABLE 2 LCAD RESTRICTIC"S Fon 1969 Reductica f ccc 1:axia: uni Date Ca pa b ility of 210 G Condition January 1 - January 8 60 "B" secondary steam generater tube leak repair and feedater heater tube leak repair.

January 9 - January 16 18 "B" and "C" p r it.a ry and "A" and "B" secondary fee? water heater tube lect: rcpair.

January 17 - January 23 45 "C" secondary s ten:a genera ter tube leak repair.

February 3 - February 6 18 "D" and S" primary feedwater heater tube Iqah rqairs.

February 7 - February 8 50 "C" secondary steam generator tube leak repair Fah aary 9 - February 13 18 "D" and

'd" primary f eedra ter hea ter tube leck ;epair.

February 14 - April 26 16 Fuel censervation April 27 - ::ay 2 30 Incore calibra+1ct.

1:a; 2 - :-:ay 11 16 Fuel Conservation

n;: 12 - I:cy 28 40 Fual Ccnservstien Iby 29 - June 9 30 Fuel Conservation June 10 - June 23 20 Fuel ccusarva tion June 24 - Septenber 6 50 Fuel Conservation j

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7 At the time of the scram, the primary system was being heated at 30 Mit and 50 F/Hr. Thirty-five rods and five notches were uithdrawn.

d.

At 1:04 a.m.,

on April 18, while returning Unit #1 to operation following a low vacuum scram on April 17, the reactor scrame.ed again on low vacuum.

Prior to the scram, the reactor was critical on thirty-six rods and four notches, and heating.

A" air ejector was in

service, "3" air ejector and the mechanical vacuum pump were out of service. Uhen the reaetor pressure exceeded 200 psig, at a main condenser vacuum of 23.2 inches Hg., the reactor scrarned on low vacuum.

The investigation revealed that the low vacuum scram switches do not recct until main condenser vacuum is between tuenty-four and twenty-five inches Hg. and increasing.

e.

At 7:40 p.m.,

on July 27, the reactor scrarned due to low voltage on busses 11 and 12 during a 4KV switching voltage tra ns ien t.

Lead at the time of the serem uas 150 Etc.

Prior to the scram, the excitars had been changed over in accordance uith the monthly schedule.

LA exciter was paralleled with 1D excitc; and 13 removed frca service. At this tima bus 11 was fed f rca transformer 12 and bus 12 uas fed from transformer 11.

Accordirg to procedure, the exciter in service is to be supplied by transformer 11.

The folicwing switching uos done:

1)

The auto transfer owitch was placed in the off position end transformer 12 to bus 12 breaker was closed.

2)

An atterpt to close transformer 11 to bus 11 failed because the synchroscope uas not turned on.

The scope was turned on and the breaker closed again. At this time the feed fren transformer 12 to bus 12 tripped due to an over current ralay. Due to the protective circuitry, this caused transformer 11 to bus 12 breaker to open, resulting in a dead bus 12.

The over current uas due to circulating currents uhen the tuo auxiliary power systems uere tied together. This resulted in tripping the "C" and "D" reactor recirculating pumps and caused the secondary load limit to trip dropping the load to 30 :T.le.

Also "5" safety system IS set tripped giving a half scram.

The "B" system was switched to the alternate power source and the safety system was reset.

,1 The over current relay was reset eight minutes later and transforner 12 to bus 12 breaker was closed and the reactor

)

8 scrnmecd. This was due to closing in on a dead bus.

The startup load on bus 12 caused the voltage to cag resulting in low voltage on bus 11 and 12.

"B" secendary steam generator had been taken out of service previously to reduce the core A P.

The tripping of Eus 12 resulted in tripping "C" and "D" recirculation pumps. Th is resulted in operating for eight minutes with onc recirculatica pu p at 80 INe.

The locest !!CilFR during this period uas approximately 3.5.

In conclusion, when both transformers are connected to the same bus, the procedure is to open ene of the breakers bume dia tely. This was not done and in ef fect both husses were supplied by two sources for a period of tire, f.

A t 9: 40 p.m., on July 27, during rod withdrawal for criticality, the reactor scrat=ed.

The micro micro anneter switch vas in the non-coincident position and a spurious scran occurred on channel 3.

The safety systen was reset and rod withdrawal was restarted, g.

At 1: 47 p.m., on August 2, the reacror scrammed due to a spuricus signal from the cut-of-core pcuer range channel No. 3.

The unit uas being shutdoun te repair tube leaks in the condenser and.Jas off systen. Control rods were being inserted to corpl;te a normal shutdown, when a spurious signal f ter cuuci-core po.rer range channel I;o. 3 scracned tha reactor (in the non-coincident mo/c).

h.

At o:25 p.n.,

on August 3, the reactor scra==ad due to low steam drum water level.

The reactor was at 1,000 psig and the turbine uns being brought up to speed. During the startup, difficulties were experienced in controlling drum level due to leahage through a bypass valve. The final feedvater valve (M0-9) was opened and closed to control this leakage. Daring one opening cparatien of 1:0-9 the valve motor tripped on thertal overload preventing the valse from opening.

The reactor subsec,uently scram =ed on Icv drun water level.

1.

At 7:15 p.m.,

on August 3, following the earlier scram on this date, the reactor scratred again.

Control rods had been uithdrawn and reactor heat-up uas contir.uin;.

An attempt was made to open the turbine stop valves by resetting No. 2 vacuum trip. The suitch was accidently turned to the

" trip" position instead of the " reset" positica and the reactor scratmed.

j. At 3:13, a.m., on Dececher 29, the reactor scr9tmed due to a spurious signal from the channel Mc. 2 out-of-core ins trunentation.

The reactor was suberitical and rode were being withdrawn to bring the reactor critical af ter the sixth parti;l refuel in g,

ma in: e nance, and inspection enage.

The micro-nicro an:.et r switch was in the non-coincident position, with 13 ccatrol rods withd rawn.

9 Uithdraual of control red H-4 to notch 5 uns in progress, when a spuricas signal on channel 2 cat-of-care initiated a reactor scram. The safety system uas subsequently reset and rod withdrawal uns continued.

3.

Incidents a.

Unit #1 uas shutdown with all fuel removed to the fuel building, when in the course of maintenance,the failure of tuo metor operated valves uns discovered.

In both cases, the valve failure was due to burned out control transformers.

In the case of MO-109, the emergency condenser condensate ;alve, the transforrer burned out because an indicating lamp sccket at the local centrol s tatien was shorted.

In the second inatance, MO-123, the transformer burned cut because the lamp uas twisted off ita base and the wires were shorted across. This is the "A" unloading heat exchanger upstream discharge valse.

Tc prevent a reoccurrence of this probicm, the indicating lamps were removed frca the local control stations that are norr. ally locked during plant operatien. When maintenance is being dene on thase valves, the lamps will be temporarily replaced and the valve tested af ter the lamps are rencved. Where the local control atations are accessible during operation, they will be regularly inspe:ted during operation.

The station Electrical Engineering Department has Scon requested to study the moi.cr cparated valvo circuits nnd recemmend : perranent solution to prevent tuture transtormer railures.

b.

On October 23, 1969, daring tha normal calibration of safety systen sensors done daring ev<.ry refueling outage, one of tbc fcur rcactor low water level IMgnetrol suitches, LSL-2, fail:W to operate properly. The other three level switches did operate preperly.

At the time the failed level switch was discovered, the unit uns shutdo'm for refueling and the reactor had been drained in preparation for control red drive removal work.

An inspecticn of LSL-2, the affected sensor, and LSL-3, a companien ubich hed cperated normally, showed the following:

1) Actuating magnets on bcth switches were in the operated position indicating that the ficat of LSL-2 had operated properly.

2) The mercury vial on LSL-2 had not tipped sufficiently to open the Safety Systen contact.

3)

High ambient temperatures affected some of the viring at the Magnetrol switch, but not to the point of incipient failure.

The conclusion was that the ncchanical lin'< age of the Ibgnetrol had deteriorated to a point where it failed to move the mercury vial the required distance. The defective !bgnetrol was replaced with a neu switch during the outage.

10 4.

Control Rod Drives a.

Control Rod Drive 0,cration 1)

On January 5, with the reactor at normal operating pressure and tenperature, contiel rod drive H-7 failed to withdraw from the fully inserted position. Attempts at flushing the drive in the insert and withdraw positions, varying the hydraulic pressures to the drive, cleaning and adjustments of the drive's Asco valves, and finally scramming of the drive in the fully inserted position were all to ne avail.

On January 24, with i;2 operation still inpaired, drive H-7 was lef t electrically disarmed and valved cat of service.

The drive rcrained out of service for the rest of Cycle 6 operation and uas removed from the reactor for overhaul during the Sixth Fartial Refueling Outage. As of the cine of this report, the disassembly of the drive has net been performed.

It is believed that an accumulation of foreign natcrici has built up in.the shuttle pisten - collet assembly area in the drive and is preventing normal shuttle pisten movenant.

2)

During reacter heating on January 24, control rod drives C-2, C-3, and 3-2 drif ted frem the fully uithdrawn position in the core to hotch 3(C-2), 7(C-3) and 11(E-2) while chif ting the cooling water flow frem the emergency feedwater pump to che prinary feedvater pump.

Investigatica revealed the drifting was caused by a merectary in:recse in the control red drive header pressure during simultaneous pump cpcration.

The energency feadwater pump, operating uith a icwer than nor:21 dicch7:ge prn::ur2, ra: t cing ::::"ed f rrn e:"ic a 9-d the prinary facd punp added in crder to clear 1cu pres;ure accumulator alarus. 'Jhen the switch in pumps uns rade, the sirultan:cus purp cperation ca2 ced a sudden surge in cooling water to the drive system. Due tc the low discharge pressurn from the e=crgency feedvater punp, the Pr + 15 contrcl red drive pressure regulator uns open nore then usual trying te sense the required 15 psi differential. Zecause of the physical lccaticn of the regulator sencing line downstreer of the pressure centrollar, the sudden surge of pressure, being greater than the 50 psi reactor pressure, wcs not throtticd irrediately and rushed threngh the cooling uater header to th2 three well scaled drives. The cmcrgency feedrater panp was rc=cved fron service end overhauled en January 25.

No further problems were exparienced.

3)

Frier to criticality on Fchruary 2, drive C-1 (accumulater 13) failed to uithdraw f rca position 0 at no mal er increased hydraulic prassures. The tuo other drives on accanulator 13 vere also tried and th2y too failed to cove frem positicn G.

Checks of the hydraulic piping on accumulator 13 ravcaled the normally open srpply isolation valve on' the uithdraw header closed. The valve was returned to the correct position and normal cperation continued. It is believed that tbc vntec was accidentally clesad during a previous outare (January 23 to February 2) vbea another accuuulator in close prcri;.ity to accumulator 13 was takca out of service for repair.

11

4) During functional scra= tests cn April 13, contr:1 rod drivas 11-5 and C-6 on accanulater 27 fatitd to scram ncn c ceran signal was initiated frem the drive test facili ).

The drives ware then su cassfully scranmed frc= the coatrol roen; heuover, the resulting scram tirc was abnornally long ct 2.9 seconds.

Subsequent investigation revected the screa cir solenoid valves on accumulator 27 piping to be malfunctier.ing, and dhen overhauled, revealed the "A" systc= valve to have a weak return spring and a damaged core assembly tip.

The cemhination cf the tuo problems caused the scram valve air solenoid to r.alfunction and thus prevented the inlet and cutlet scram valves from functienty; properly. The ability te scram the drives frem the control rece and not f rom th e t as t facility is due to the presence of a bcchup air solenoid valve dhich is actuated when an GO rod scram is initiated frc.? the centrcl room but not when an individual accumulator is scrarred as frem the test facility. Uhen a scram signal was initiated from the control room, the bachup air solencid valve opara.ed properly, cllowing II-6 and C-6 to scram cvan though the normal scrc=

nir solenoid valves had f ailed. The lenger than norral scram tine anperienced chan the drives' were scranned fror the control roce can be attributed uo the time delay in bleeding the air off the backup solenoid valve rather then the normal scram solenoid air valve.

Uau inner core assemblies ucre subrequently ordered and installed in all such valves in th; system.

Tha vcak retutn spring was cleo replaced in the defcetive valve end a test us cenducted and resulted in N r.a1 oprat ion of both h-6 anc C-6 5)

Cn April 1C, during plant startup, control red drive F-9 cas selected for uithdrawal frca position 0, but the drive did u*ot nova.

Subscquently, followin3 alterno te rod eeverents, F-9 dr-fted out to pcsition 5.

Investi::ntica of the centrel rod hydraulic systen revealed that tbc mcnual isciction valve en the insert header was closed. !:aintenance had been perferned on F-9 cccumulater during the outage, and the cut-of-service card had not been cleared prior to the startup.

Satisfactery operation of F-9 again continued when the card uas cicated and the val ~ve opened. Drifting of the drive was caused by a bydrculi:

lock of pressure trapped in the drive due to the closed isolction valve, which was bled f rcm the drive slowly arnand the pis ton seals ins tead of the nornal insert e: haust line. Until the pressure lock was relieved from the drive internal shuttle piston, the drive drifted out to positien 5 under its oun vaight.

6) On July 28, chile heating and prcssuricing the system en plant startup, control rod drive B-3 was withdrawn frem position 0 to positien 5 with the drive subsequently drif ting to positten 12.

As this "as the normal cpera ting positica of tPc drive, it uns lef t a. this location during the re ninder of the startup period.

Onc2 at pruer, the driva was flu.ched and encrcised, but all a t tempts at latching it at any position but position 12 care to no avail. Further attempts centinual until August 2,

.'h a n zhile shutting the unit doun, a spurious signal f rom poeer cbcnnel 03 (in the non-coincident node) initiated a r: actor scram.

B-5 uas scrammed in from position 12 and latched

12 l

normally at position O.

No attempt at moving the rod was made again until startup on August 5, when while withdrawing the drive in one notch increments, satisfactory operation was again resumed. Apparently, the crud accumulation thought to be holding the drive collet fingers in the open position was disledged by the reactor scram. To insure its continuad satisfactory operation, the drive was exercised daily during poter operation for the rest of Cycle 6 with no occurrence of further abnormalities.

7) 1 Mile increasing power on July 28, control rod drive F-6 failed to withdraw from position 2 on two attempts. When

~

successfully withdrawn to position 3, the driva drif ted oct under its own weight to position 5.

Subsequent flushing and exer-cising of the drive resulted in returning the drive to normal cper-ation. The movement and flushing of the drive had evidantly removcc the believed foreign material obstruction fran chuttle piston-collet assenhly area in the drive which was causing sluggish movcment of the piston.

To insure its removal, continued flushing and exercising of F-6 was performed during subsequent plant outages.

8)

During control rod drive exercising on August 12, drive E-7 could not be withdrawn to position 12, but all attempts at inserting or withdrawing tha drive between positions 0 and 11 vere performed with no difficulty.

Investigation revealed a calfunctioned reed suitch at notch 12 on the drive probe. which was confirmed by visual indicarich of incore instruuantation while n.cving the drive and subsacuent ins trument mechanic " coherence" checks. Normai eperatien was resumed af ter the drive probe uas replaced on August 15.

9)

On Dece=ber 29, while uithdrawing rods to bring the core critical, the reactor scrarmed due to a spurious signal on the out-of-core instrunenta tion (channel 2). After resetting the safety system and completing initial preparations for rod withdraual, control rod J-7 was found at positicn 10.

The drive is on accumulator 12 and at the time of the scran was at position 12.

While withdrowing tods for criticality, the lot accumulator pressure alarm uns conoistently alarming on accumulator 12 and an operator was assigned to the accuniulatet area to maintain the required charging pressure.

Prior to the serem, accumulator 12 was being charged and uhen the scram was initiated, it is believed that the accunulater ras not completely pressurized, thus resulting in the failure of the drive to scram. The inward novement of the drive to notch 10 apparently resulted from the accumalated prassure in the accumulator from the charging. The resultact beeder pressure was enough to move the drive inward, het was not enoug'.. to scram it to position O.

Opecation uith tuo accumulators.

out of service simulataneously during rcactor op 3 ration is lir:ited '

by staticn operating procedures. Bypass of acettmulator 12 therefore, did not viola te tl.ese procedures or t he 1: cme d ia te shutdoun capability of the reacter.

?

13 b.

_ Control nod Drive Tests On April 12, 1969, all control rod drives, except SN1229, core position H-7, were scram and friction tested and timed for normal insertion and withdrawal.

Drive 1229 was disarmed on January 7,1969, af ter its failure to withdraw from position 0, and was not tested during the outage.

As a routine check of drives to be overhauled during the fall refueling outage, all drives except for H-7 were again tested on September 6, 1969.

Based on this and previous tests, 28 drives were selected for inspection and overhaul.

Following the everhaul and replacement of the 28 drives dering October, initial scrom and timing tests were performcd on all 80 drives on November 17-22, to insure the operability of each drive prior to loading fuel. The tests were conducted using "durriy" fuel assemblies in the fuel cells. The tes ts care followed by normal timing, latching and scram and friction tests after fuel loading on December 6.

Data obtained frca these tests were satisfactory on all drives.

c.

Control Red Drive Inspection Tha Technical Specifications to Dresden's License DpR-2, as amended by change Uc.'12, dated November 23, 1966, states that during major outages, "Not less than tuo control rod drive mechanisms shall be removed, disassenbled, and thoroughly inspected at intervals not to eneced 24 montl.a."

In addition to licence require:ac n t s, drives are removed for inspectica on the ba;is of drive test results and malfunctions experienced during o;c ra tion.

Prior to shutdown for the sixth partial refueling outcge on September 6, 1969, twenty-eight control rod drives uere selected for removal and overhaul.

Four spare drives (1295, 1290, 1266, and 1270), were inspected prior to the outage and were satisfactorily tested in the control rod drive test facility following overhaul. The four drives were used to replace four of the twenty-eight drives removed for overhaul (1229,1291,1253, and 1235). The drives that were replaced will be everhauled at a future date.

Af ter the renoval of all fuel elements and all centrol red blades, cach of the twenty-eight control rod drives was removed wi'th its index tube fully withdrawn. One exception was drive 1229, which was stuck at Position (0). The drives ucre then trans-ported to the drive shop area at sphere elevation 565' for

~

disassembly and inspection.

.i

16 1)

Visual Insnection Af ter drive disassembly, all parts were visually inspected as closely as radiation levels would permit. A 1/4" plexiglass face shield was used to protect against beta emission while viewing the parts. The roller mount assembly was also submerged in one foot of water during inspection.

All movable parts of the roller nount assembly were tested to verify freedem of operation.

Each canned magnet was heated in a beaker of water to a temperature of 200 F to insure the integrity of the nagnet hous ing.

2) Flourescent Dye Penetrant Inspection A flourescent dye penetrant (Zyglo) inspection was conducted on major components of the twenty-four drives. A penetrant (ZL-2), developer (ZP-5), and an ultraviolet light were used for the inspection.

The inspection procedure consisted of a thorough cleaning and drying of the parts to be examined, application of the dye penetrant, removal of excess penetrant by wiping the surface, and finally, application of the developer.

The parts were then inspected for defects using an ultraviolet lighe ec"rea.

Tha dra 'e t al.le":d a penetratfen th:2 of 15 minutes beforci hand wiping with a clean cloth and applicatien of the developer.

If a surface defect is present, the dye penetrant is absorbed by the defect. 'Jiping the part removes the execes dyc frem the surface of the part but not from any cracks which r.ay be present. The opplication of the developer draus out the dye penetrant and accentuates the defective area.

Viewing with an ultraviolet light source pin-points caterial defects.

3)

Results of Insnection a)

Nitrided guide roller pins were used to replace all pins in the roller ecunt assembly.

Practically all of the pins r:=oved exhibited some waar.

None of the 112 pins removed had failed. No attempt was made to accurately measure the amount of pin wear experienced.

b) All rollers uere checked with a go-no-go gage. This gage has an outside diameter of.260".

.The maximum tolerance for the hole of a new roller is.259", thus only one thousandth of an inch was acceptable for roller wear.

None of the rollers inspected exhibited an excessive amount of wear.

15 c)

Dye checking revealed cracking or flaking chrene on three guide plugs, one collet, and one unlocking spring.

Aside from the chro=c plating ancmalics, no cracking was found in drive cenponents.

d) All seals were replaced on every drive overhauled.

In the absecce of gross seal ucar or breakage, the cendition of seals causing long insert and short withdrau times is described as " normal wear" in the re po r t.

e) Figure 2 summarizes the locations of the drives removed for overhaul while Table 3 sucmarizes the inspection results.

e e

pr-

16 FIGURE 2 Loce* ton of Control Rod Drives Overhauled During 1969 Refueling Outage i

10 1

9 8

i i

I 1

1, I

l 6

5 l

l 3

2 1

A B

C D

E F

G H

J K

m........

i Table 3 1969 Maior Control Rod Drive Overhaul and Ins mction Summary Core SN Drive SN Drive Position Removed Installed Abnormal Synproms of Drive Removed Inspection Results 11 - 7 1229 1276 Drive St ick at Position Zero Inspection will be performed at a later date B-5 1285 1228 a.) Driva Failed to Latch No obvious explana-b.) Drif ted out tion for drive symptoms F-6 122:

1280 a.) Driva Failed to Withdraw Broke drive & stop b.) Drif ted Out niston seals; guide C

sleeve on outer filter was cracked A-4 1234 1308 a.) Short Buffer Time Broken stop piston b.) I,ong Insert Time seals; scratches on c.) Short Withdraw Time shuttle piston ID J-7 1235 1296 a.) Long Insert Time Inspection will be b.) Short Withdraw Time performed at a later date i

E-2 1269 1241 a.) Long Insert Time Worn drive and stop b.) Short Withdraw Time pistons; cap off inner filter J-5 1253 1275 a.) Long Insert Time Inspection will be performed b.) Short Withdraw Time at a later date

Tab t e 3 (Cot'L.)

Core SN Drive SN Drive Position Removed Installed Abnortaal S naptoms of Drive Removed Inspection Results l

E-10 1291 1234

a. ) 1.on ; Insert Tirac Inspectioa will be l

b.) Short Withdraw Time performed at. a c. ) Sho c t Buffer Time later date l

D-9 1261 1261 Loag Insert Time Small scratches on l

chuttic piston I.D.

1 i

B-6 1275 1269 Long Insert Time Guide sleeve on outer filter cracked.

1/16" ball found be-tween finger and collet K-6 1247 1290 Long Insert Time Broken screen on inner filter, broken drive 3

i l

piston bushings and seals

!!-3 1283 1227 Long Insert Time Broken seal on drive piston l

l J-8 1228 1287 Long Insert Time Worn outer seals and bushings worn.

Shut tle piston I.D.

scratched G-6 1314 1225 Long insert Time Drive piston bushings,

and seals worn; stop piston seal broken; guide sleeve on outer filter cracked.

Tabic 3 (Con t.)

Core SN Drive SN Drive position Rrmoved Installed Abnormal Synproms of Drive Removed Inspection Results J-4 1289 1247 Long Insert Time No obvious explanation for drive symptoma C-4 1232 1282 a.) Long Insert Time Chrome flaking on guide b.) Short Buffer Time plug and broken bushing and seal on drive piston G-4 1280 1283 Long Insert Time Broken drive piston seal and guide sleeve on outer filter cracked C-5 1258 1240 Long Insert Time Broken stop piston seals H-4 1296 1289 Long Insert Time Guide sleeve on outer G

filter cracked E-3 1281 1295 a.) Long Insert Time Plaking chrome on guide b.) Short Buffer Time,

plug outside diameter.

Bushing and seal chipped on drive piston.

B-7 1308 1281 Long Insert Time No obvious explanation for drive symptoms H-8. ;

1287 1314 Long Intert Time llorn seals and bushings on drive piston and one broken stop piston seal C-7 1255 1266 a.) Lonf. Insert Time Worn drive piston. bushings b.) Short Buffer Time and broken stop piston seal.

~

S Tabic 3 (Con t.)

Core SN Drive SN Drive Position Removed Installed Abnormal Synptoms of Drive Removed Inspection Results J-6 1276 1210 Long Insert Time cap off inner filter D-7 1240 1258 a.) Long Insert Time Broken seal on stop piston E-7 1282 1277 a.) Long Insert Time Bent fingers on b.) Short Euffer Time collet assembly and broken bushing on drive piston-D-3 1241 1232 Long Insert Time Broken bushing on drive piston F-2 1277 1285 Short Withdraw Time Flaking chrome on o

guide plug outside diameter, scratches on shuttle piston I.D., broken seal on drive piston, metal shavings found on magnet

21 l

5.

Centrol Red E L1 des a.

Blade Follo ini Checka During perieds of operation, control rods have been verified for blade following on a '..eekly basis. During each startup, control rod patterns for criticality have been predicted and all blade folloaing verified.

b.

Control Blade Inspectien Control blades ware inspected in the fuel baiE ing stora3a pool during the refuelinc cu. age en !;nverber 7, 19 M. Six blades were checked wid go-no-go gegas for di: ansienal va ria tien s. All blades inspected wara found to be in acceptable condition. Tna 1, lades axanined and their locaticns in the core are e:,hibited belou:

A O C O C F G H 3 M

'0-l I

{

l

.i I

e i

_;_. ' _. ; ____: _.___ ____; _L e

l t

l

____'___________J,_._

I j

7 e

__.___ 7 6.

i_

t i

o

<________i_

{

4 i

{

I j

l 1

3

'.__,___.___t_

'_ _E:B 'B-lE_ 4 3 ' 13 '

2

}

f t

S-in-131 174 i

COMTRCL ELADES EYJ.!!II ED BY Go-!!o-Go Gage

,.i D

22 6.

Changes in Facili:v Desian a.

Material Test Leon 7.nstallation The material test loop, which General Electric began installing in 1968 in "B" recirculation loop, was ccmpleted except for wiring to the control room and completion of the installation of the dynamic test vessel. The system will not be in full operation until 1970.

b.

Rad-Waste Discharge Line Modification The Dresden Unit I rad-waste system previously discharged to the Unit 1 circulating water discharge canal via a 6-inch carbon steel underground line. When maintenance or replace-ment of portions of the Unit I circulating water system are undertaken, the circulating water flow is not available for diluting the rad-vaste discharge.

Hold-up of the rad-waste water during circulating water outages is not an acceptable ~

alternative.

Therefore, it was decided to provide an alternate discharge path to the Dresden Unit 2 and 3 circulating water discharge canal.

In addition to prcviding an alternate discharge path, the addition of this line avoids the accessity of using the present underground lire which has experienced crosion problems.

Either line may be used, depending upon Unit 2/3 operation.

The chang conci:::d Of adding : 3-1 ch c arbar s

• aa! ual. dad line which is attached to the e-inch carbon steel linc in the Unit I rad-waste handling building dcunstream of the flow metering equipmant.

The line is routed frca the rad-waste building through the rad-waste piping tunnel to the Unit 1 turbine building pipeway. The pipe crosses the Uait 1 turbine building pipeway and enters the Unit 2 area and connects into the Unit 2 circulating water discharge pipe, c.

Unit 1 Diesel Generator Modification With the construction of Dresden Unit 2 turbine building, the Unit 1 diesel generator is no lenger adjacent to an exterior wall.

Prior to construction of Unit 2, air intake to the radiator was througn the south wall of Unit 1 west auxiliary bay and discharge was through the west wall. Air intake was originally changed to the east-west corridor of Unit 1 and 2, with discharge into the Unit 2/3 makeup demineralizer area.

This arrangement presented a safety hazard with the hot exhaust air blowing into the domineralizer uorkit g area.

The diesel radiator was then subsequently moved up aine feet and south twenty feet to a new location over the cast-west corridor with the intake from tha south and discharge to the east.outside the building.

23 d.

Soent Fuel Storane pool Retninar !fodifications and Rack Re-enfereeren.

The portable' retainers that were originally used in the sterage pool were replaced by permanently installed, reversible retainers.

f These new retainers need only to be flipped over the bail of the fuel assembly with at actuator pole. The orignal retainers had a history of wedging between the fuel assembly and tha storage rack, which entailed considerable time for disengaging.

The new retainets eliminate this difficulty.

The fuel racks were also reinforced to prevent any possibility of tipping over.

e.

"A" Demineralized Water Tank Tie Line In November, 1969, a tie line was installed between the existing Unit 1 "A" demineralized water storage tank and Unit 2 and 3 "A" and "B" condensate water storage tanks.

The twenty-four inch tie line was installed to conform with the principle of generating station flexibility of operation.

To facilitate this tie, the Unit 1 storage tank had to be dra ined.

This operation eculd not be perferned until Unit I had shutdorn for its refueling outage, at which time the tuenty-four inch tie line and a two inch line from Unit 1 cakeup damineralizer were ins ta lled.

f.

New Laundry Instella tion During 1969, a neu laundry facility was constructed for the decontamination of protective clothin:; and protective breathing apparatus. The new facility was nacessitated by tha ad61 cions of Units 2 and 3.

The enclosure fer the new laundry was erected above the drain tank vault, in the crea betueen the Unit 1 turbine trackuny and sphers. All access is afforded through the north corrider of the fuel building ;c;4 age,.

The installation was completed and placed inte operation during the first ueek of Novenber, g.

Auxiliarv Steam Tie Lines to Unit 2 Installation of 50 psi and 200 psi steam tie lines between the Unit I and Unit 2/3 boiler house uas completed during the year. The primary purposes of these lines is to enable all the summer steam loads te be supp1Lud by either Unit 1 or Unit 2/3 boiler. The basic steam loads that cre requiced are

1) steam to the Unit 2/3 radwaste concentrator, 2) steca to the station laundry and 3) steam for initial startup tests for Unit 2 cut' 3.

A tie betueen the two condensate return lines pas also installed.

l

2 '+

l I

h.

River Water sample station The Dresden Station Illincis Rivec sanpling unic,vhich is located at the extreme northwest corner of the Dresden site, was placed in service l' y 10, 1969.

a This sampling unit ecmposites intermittent sacple, frem the Illinois River downstream from the Dres.en Staticn cischarge canals. Two one-liter samples are withdrawn and cre subqequently analyzed for gross radioactivity and certain specific radio-nuclides as required by the environmental coaitoring specification.

After collection of these samples, the composite bottle is drained and the system is visually inspected to assure continued proper operation,

i. Autoratic Disratch Svstem (A. D.

S.) Ecu inmen t Ins:allation The installatien of equipment necessary to previde automatic lead control of Unit 1 by the Autcmatic Dispatch System was completed during

• the refuelinn outage. Af ter completion of initial calibra tions and testing, the output circuit from _he A.D.S.

console te Unit 1 governor synchronizing notor rac openad and tagged out of service to preclude any possibility of A.D.S.

control until approval of its use is obtained.

The f ollowing equipment and all necessary ir.terconnacting cabling was installed and tes:ed during the cutage.

1) A station and rnchine control console were loca ted in Unit 1 control room.

2)

The governor control switch on consolh C-1 vas replaced with one compatible with A.D. S.

3) A governor selsyn unit was acunted on the turbine frcnt standard with a belt drive frem the govcenor synchronizing motor.

4) A valve position pickoff unit vas installed and belt connected to the secondary servo motor positien indicator on the turbine frcnt standard.

7.

,F_e_rsonnel Rndiation Excosure Personnel exposures to radiation during 1969 vere within limits specified in 10 CFR Part 20.

8.

Licuid Poison System The liquid poison system was cperative at all times during the year.

The boren poison uns sampled on Jcnuary 28, April 11'and 1:cvember 13.

There were no conditions which wculd indicate a los, of borca from the solution tenh.

Leron con:entra tions in the reactor water remained low throughout the year.

25 9.

nadicactive Uaste Disposal Reicase of radioactive liquid waste was accomplished in batch quantities at controll2d release flow rates according to establish-ed procedures. The centribution to the activity of dilution water was always maintained within the limito specified in the applicable federal regulations. The average contribution tc the unidentified activity in the water utilized for radioactive liquid waste dilution during the year was calculated to be.225 x 10-7 uC/ml (22.5 uuC/L) ccmpared to an average limit of 1.00 X 10-7 ue/ml (100 uuC/L) for unidentified mixtures containing no Ra 226, Ra228, or Il29 as specified in 10 CFR Part 20.

solid radioactive wastes were stored on-site pursuant to License DPR-2.

Table 4 shcus the radioactive content, shipment destinations, and dates of radioactive waste shipments made during the year. Three radicactive waste shipments were made in 1969.

The concentration of noble fission genes in the stack discherge to accosphere was maintained well within license Ibnits of 700,000 microcurics per second. The annual average activity release rate free the stack was appro::imately 25,593 uCi/sec.

One ::ucicar Fuel Servicec, Inc. - Stanray rail cask shipment, totaling 19 sp2nt fuel assemblics, uns nada during the year to the Chemical Processing Flant of Puclear Fuel Services, Inc., in West Valley, I!cw Yori:. Two shipments totalir.g 32 asserblies and thr:2 sr ceial cortainers, concicting of single " leaker" c' ements, were shipped to the U.S. AEO, DuPont veHemours, Inc. In uunnarton, touth Caroli Table 5 is a breakdcun of theca shipments in addition to all shipments made since initiation of fuel shipping in June, 1965.

m

~.

26 TABLE 4 RAD'. ASTE SHITIENTS Destination Activity (Curies)

Volume (n.J) 7/;

N.

E. Co. Inc. Sheffield, Illinois 280 689 7/:

39 N. E.

Co. Inc. Sheffield, Illinois 1.196 2133 8/22/69 U. E. Co.

c. Shef ficid, Illinois 0.367 550 e

27 Table 5, SI'ENT FUEL SilIPMENT SLU24ARY Mumber of Assemblics or Containers Total Shipment Number Date Batch To Rail Truck Shinned 1

2 3

4 5

Ra il Truck, Date 1

6/11/65 24 0

0 0

0 24 24 2

6/30/65 2 '+

0 0

0 0

24 48 3

7/.16/65 24 0

0 0

0 24 72 4

8/ 3/65 24 0

0 0

0 24 96 5

8/16/65 24 0

0 0

0 24 130 6

9/ 2/65 24 0

0 0

0 24 154 7

9/22/65 24 0

0 0

0 24 168 1

8/ 1/66 0

0 4

0 0

4 172 8

8/ 5/66 16 0

6 0

0 22 194 2

8/15/66 0

0 4

0 0

4 198 3

8/24/66 0

0 4

0 0

4 202 4

8/28/66 0

0 4

0 0

4 206 9

8/31/66 0

0 12 8

0 20 226 5

9/ 5/66 0

0 4

0 0

4 230 6

9/12/66 0

0 4

0 0

4 234

.,7 9/14/66 0

0 4

0 0

4 238 10-9/16/66 0

0 0

20 0

2G 23G 8

9/19/66 0

0 4

0 0

4 262 9

9/21/66 0

0 4

0 0

4 266 10 9/25/66

-0 0

4 0

0 4

270 11 9/26/66 0

0 4

0 0

'4 274 12 9/28/66 0

0 4

0 0

4 278 13 10/ 2/66 0

0 0

0 4

282 14 10/ 3/66 0

0 4

0 0

4 286 11 10/ 7/66 0

0 0

24 0

24 310 15 10/11/66 0

0 4

0 0

4 314 16 10/12/66 0

0 4

0 0

4 318

-17 10/20/66 0

0 4

0 0

4 322 18 10/23/66 0

0 4

0 0

4 326 19 10/26/66 0

0 4

0 0

4 330 12 10/28/66 0

0 3

16 0

19 349 20 10/30/66 0

0 4

0 0

4 353 21 11/ 1/66 0

0 4

0 0

4 357 22 11/ 6/65 0

0 4

0 0

4 361 23 11/ 8/66 0

0 4

0 0

4 365 24 11/10/66 0

0 4

0 0

4 369 25 11/13/66 0

0 4

0 0

4 373 13 11/14/66 0

0 0

23 0

'323 396 26 11/15/66 0

0 4

0 0

4 400 27 11/17/66 0

0 4

0 0

4 404 28 11/20/66 0

0 4

0 0.

4 408 29 11/27/66 0

0 4

0 0

4 412 30 11/29/66 0

0 4

0 0

4 416

~,-w,-

y--.

+,w r

-_e

--r--

--,h

+ - -

w-

28 TABLE 5 (Cont.)

SPENT FUEL SHIPMENT sum 1ARY Number of Assemblies or Containers Total Shipment Number Date Batch To Rail Truck Shipped 2. 3 4

5 6

7 Rail Truck Date 14 12/ 2/66 4

19 0.

0 0

23 439 31 12/ 6/66 4

0 0

0 0

4 443 32 12/11/66 4

0 0

0 0

4 447 33 12/15/66 4

0 0

0 0

4 451 34 12/18/66 4

0 0

0 0

4 455 35 12/20/66 4

0 0

0 0

4 459 36 12/27/66 4

0 0

0 0

4 463 37 1/ 3/67 4

0 0

0 0

4 467 38 1/ 5/67 4

0

'O O

O 4

471 39 1/ 8/67 4

0 0

0 0

4 475 40 1/10/67 4

0 0

0 0

4 479 41 1/15/67 4

0 0

0 0

4 483 42 1/22/67 1

0 0

0 0

1 484 15 7/15/68 0

0 0

24 0

24 508 16 8/ 9/68 0

0 0

24 0

24' 532 17 8/29/6S- 0 0

0 21 0

21 553 18 9/19/68

)

0 u

21 0

21 574 19 10/ 9/6t; 9

o o

24 G

24 590 20 10/10/68 0

0 0

24 0

24 622 21 11/20/68 0

0 0

24 0

24 646 22 12/11/68 0

0 0

22 0

22 668 23 1/ 3/69 0

0 0

19 0

19 687 24 1/29/59 0

0 24 0

0-24 711 25.

2/26/69 8 0 0

3 0

0 11 722

,o

29 I

10.

Fuel Assembly Cleanina and Testing k

a.

Fuel Cleaning Fuel assembly inlet orifice cleaning was initiated on October 1, 1969, on all assemblies scheduled for return to the reactor for Cycle 7.

All orifices were manually cleaned and the effective-ness of the cleaning determined by flow testing.

b.

Flow Testing Data obtained by flow testing of onc new Type V assembly with an "E" orifice in 1967 and sin Type VI assemblies with UNC Type 1 and 2 orifices in 1968 were used as criteria for cleaning effectiveness. A comparisen of this data with data obtained from cleaned assemblies during the present outage is shown in Figures 3 and 4 All seven hole "D" orifices were also replaced with new single hole "F" orifices in order to remove the worst crud collectors in the core. Data obtained from both orifice types are also shown in Figure 4.

Several UNC Type VII fuel assemblics ware flow tested clean to provide " clean condition" data for subsequent fuel cleaning.

c.

Bow Checking All fuel assembly channels on fuel scheduled for Cycle 7 operation vere bow checked in the bow check facility. There were diu chauumi. fcend :: he hencd and th::: fere diceerded and replaced with spare channels, d.

Orifice Changes and Modifications One hundred thirty-nine orifice changes were made in accordance with United Nuclear's refueling plan. These changes were necessary to f acilitate moving of the higher expoced fuel to the periphery and the lower exposed fuc1 to the center of the Sixty-six orifices were changed. from Type "E" to Type "F",

core.

forty-six vere changed from Type "F"

to Type "E", thirteen were changed from Type 1 to Type 2, and fourteen were changed from Type 2 to Type 1.

The results of the flow testing and orifice cleaning showed a somewhat unexpceted degree of crud buildup in orifices on UNC Type VI fuel assemblies.

These orifices are tripod type design, as opposed to the single bail type employed by Cencral Electric Company. United Nuclear performed a series of calculations regarding the previous and upconing cycles to determine the effect of the unexpected crud accumulation and made the following conclusions.

6

1) Cycle 6 was operated within license limits even though there was more crud accumulation in Type VI fuel orifices than anticipated.

'N 30 Figure 3 Fuel Acuenhly Pressure Drop vs Flow C2 Type "E" Crifice

!!aninun-Before C1 caning Averac -Af ter Cleaning 180 160 140

11nittun-Beferc Clconing 120

/,'

100

//

'll E

. l/

/ g /j/

UO o

C b

.//

o E

60 40 20 0

1 2

3 4

5 6

7 8

9 10 11 12 Assechly Pressurc Drep-r31 180 Fuel Assembly Pressure Drop vs Flow 160 UI:C Type "2" Orifice Avera3c-/.f ter clenning t

140 120 t

100 1:a:&.:un-Bu f orn C1 ecais::

E 80 1

.inican-Before Cleviing 6

[

60

,,v 40 20 O

1 2 3

4 5

6 7

8 9

10 11 12 A.ssentely Iresvire Drep-rSI r-

-e a,e=vw-

+,,

a w-,, - -

m w

-+---w

~-~ 9

T 31 FICU:C 4 Fucl Asscchiy Pressure Drep vs Flcu G2 Types "D' and "F" Orifices 180 160 140 thw CE Type 120 npo orific3 E:

0 100 Maxirum Ecfore Cleaning O' Typa w

80 up" orifice 60 x

M* I"

• d ? ' *

• 40 Cleanin;; G2 Type "D" Orifice N

20 0

1 2

0 4

5 5

7 c

?

?.0 1.1 I?

A s s embl-f Tressure Drep-FSI Fael Assenbly Pressure Drop vs Flev 180 U 'C Type "1" Crifice Average-Af ter Cleaning 160 140 120 FG 100 L

6 E

80 I'axi:nna-3ciore Cleaning

inimum-Defore Cleanir:

60

..)

40 20 0

1 2

3 4

5 6

7 8

9 10 11 12

32 2)

Cycle 7 operation may result in IfCHFR's unce.nfortably close to the license limit towards r.he end of the cycle unless the tripod nose pieces on the orifices of the new Type VII fuel are modified.

3)

Cycle 7 may be operated well within the h'CHFR license limit with the Type VII fuel orifices modified and with Type VI fuel retaining its original tripod type orifice nose piece.

In light of these conclusions, it was decided to codify all Type VII orifices to the single bail type.

The modificatica process included cutting off the tripod bail and welding on a 304 stainless steel bail prepared by United Nuclear.

The nedified bail is similar to the type presently being employed in approxi-mately three-fifths of the core.

11.

Ucu_ Fuel and plutonium Rod Insta11atten Ninty-s i:: (96) Type VII fuel assemblies uele purchased frem United Nuclear Corpor5 tion of Elmsford, New York. This fuel is similar in nuclear and thernal hydraulic characteristics to the fuel uhich U. N. C. supplied for Cycle 6.

The Cycle 7 core reload consists of 85 Type VII fuel assemblies and 11 Type VII - pu fuel assen.blies. The 85 Type VII fuel assemblics censist of 50 "norral" uraniun asse.-blies (Type VII) and fivo ins tomented and source bearinc assen.blies (Type VII-I).

The Type VII and VII-I assemblies are essentially identical to the Cycle 6 Type VI and VI-I fuel.

The 11 Type VII - Tu plutentum fuel asserblics consist of three

" no rt:a l ' plutonium assenblics (Type VII - Iu), six (6) " low gadolina -Flu ccuiut." assemblies (Type VII - Fu-L), and tuo

" ins trumentad" plutonium as semblies Typc VII - Pu -1). There is no axial variatica in fuel enrichment or poison concentration in any of the 96 new assemblies.

As a result of high crud buildup on the Type I and II Cycle 6 tripod orif t:es, all Cycle 7 nosecones were modified to a ningle bail design prior to installation in the core.

This results ia lA and 2A (codified) numbers for Cycle 7 orifices.

Sixty-fcur (64) of the new Type VII fuel assemblics have a central region orifice (lA), 26 have a peripheralorifice 2A, and 6 have a 2B orifice. The 23 crifices are single bail type tha t were installed as a design test.

J

.A

33 12.

Inspec t ions a.

Irradiated Fuel Insocetion Inspection of irradiated fual was performed in the fuel butiding by both United Nuclear Corporation and General Electric Company representa tivo s.

Tabic 6 lists the pertinent infornatien of the fuel inspected.

Durin3 the ccurse of United ::uc1 car's inspection, tuo Type VI assemblies (U" 022 and UK 063) cuf fered damage to the top fuel rod spacers.

The damaged assemblies were left cut of the Cycle 7 fuel loading.and returned to storage, b.

Metal Surveillance On September 29, 1963, six specimen containers fron tbc reactor core periphery and fcur from the tarning vane vare renoved to the fuel handling building for analysis of indivi;ual spechnens by General Electric representatives. The follouing tab 1c summarines the lccation and number of the containers.

Reneved frca Part No. cnd Name Returr,ed to Core Icsitica 67-25 42 Danny Fuel 94 67-25 5S-23 43 Der.=y Fuel 43 53-23 30-02 44 D2:=y Fu:1 45 56-C2 67-02 48 Dana;- reel 17 67-02 51-18 Duncy Fuet 68 31-18 74-18 Durry Fuc1 #9 74-15 Turnine Vcne Part Un. and "ame Returned to RSp 168-109 45 Saddle Bag #2 Rsp 168-169 R3p 177-173 Saddle ac7,#3 RSp 177-178 RSp 1622 Scudic Bag #4 RSp 162 S RSp 175 N 49 Turning, Vanc IIanger RSp 175 N The inspectica work included photegraphin; the specinens and removal of 47 corresien samples and 21 ruptured capsules for exami.natica at Vallecitos.

Six ner cocrosion specimen racks, two new stress bend racks and cue neu Charpia iv.pa c t capsule tn.rc cd;hd to the d==y fuci element

1. 5.

These vere the only neu stecin. ens installed.

34 o

TALLE 6 FUEL ASSE!CLICS INSPECTED BY UNITSD NL' CLEAR Assembly Cycle VI Evoosure Cvele Core Location

!C 'UD / ?.y Londed Descrintion G22 1665 17.3 4

Type IIIF Normal Assembly G40 1467 16.5 4

Type IIIF Normal Assembly G50 1259 16.6 Type IIIF Normal Assembly DU63 1660 9.9 5

Type V High Gd DU72 1460 9.7 5

Type V High Gd UN022 2157 5.2 6

Type VI With Gd Poison Rod UNO30 0961 6.2 6

Type VI Rith Gd Poison Rod UN053 1761 7.2 6

Type VI With Gd Poison Rod UNC64 1153 7.1 6

Tyrc VI Uith Gd Pois e: Red UNOS7 1359 7.1 6

Ty?3 VI With Instrunant Channel UN092 1363 6.7 6

Type VI With Instrument Channel FUEL ASSEM3 LIES INSPECTED BY GENERAL ELECTRIC CC'IPANY DUS2 1456 9.3 5

Type V High Gd DU69 1864 9.2 5

Type V High Gd DU100 2064 9.2 5

Type V High Gd DU102 1070 8.4 5

Type V High Gd G40 1467 16.5 4

Type IIIF Normal Assembly G42 1100 17.5 4

Type IIIF Normal Assembly G61 1255 12.9 4

Type IIIF Normal Assembly G87 1661 13.9 4

Type IIIF N'ornal Assembly G100 1459 17.2 4

Type IIIF Fowdered Fuel XE103 1858 15.3 4

Type IIIB Loaded BOC 4 l

E120 0562 15.0 3

Type IIIB Leaded BOC 3 DU46 1060 9.7 5

Type V liigh Cd

• nren/r = vrn/? v in3

35 There wre five retal surveillance samples which ware not removed for inspection. They are as listed below:

l 1.

Lcng rack at core position 75-13.

2.

Stress bend rack at TV RSp 172E.

l 3.

Pigeon ladder at TV RSp 163S 4.

pigeon 1 adder at TV RSp 173W.

5.

Steam drum - right, specimens G-4, G-5 and G-9.

g c.

Fuel Transfer Carrier Inspection Just prior to the 1969 refueling outage the fuel transfer carrier was inspected while the pool was emptied and cleaned.

The transfer basket was removed from the carrier and moved to the storage pool. The basket was picced on beams that vere across the stcrage pool.

In this position the bottom was under water and the bellous available for inspection. The bellous was dye penetrant c

inspected by Ce=nonwaalth Edicon's Operational Analysis t

A '

Department. No defects uere found.

s The carrier was removed to the decontaminatien pad and decentaminated.

The four pillcw block bearings were taken apart, cleaned and inspected. 7feasuromants uere taken and no signifiennt wear was found.

s' b

i *[\\s When the carrier was replaced, the tunnel sheaves and cable were inspected. All cere found in sood order except the south cable 3 -

((

t, g

which van ttmporarily replaced with carbon Oteel ccble until

\\,'

Ig stainless steel cculd be ordered.

1 s

\\.

0 t

\\

'*'3

'$ d. P,eactor Vessel Inarection s.

v s

w -

On Fiovember 20, visual inspection uas made of the reacter y'

N vessel internals with all fuel assemblics and control rods

.\\

t t*

7,,,,., l

• u 12 }

~ removed from the core.

4 Typical areas observed during the inspections included; the

' 's hold doua lugs and lifting eyes on the upper grid, the "T" handles on the hold dcun lu3s for the bottom core support 5

' plate, and the thiLbbii and guide tubes in the bottom of the c

ves se l. Frca the conditions cbserved, no pr,0bicms exist in

~

the integrity of the, vessel internals, g

e.

Prina rv Scs te :' D_16 Insnec tica 1)

Reac ccr _Flanc, Tnnr ection s

,4

['

\\}, h' The reactor flange was dye penetrant and ultrasonically w

inspected by Tittsburgh Testing Labcratories en Decenher 8, s,

1969. The tasis were the same as those conducted during s,"'

h the 1968 refualing cutage inspections, using the j

\\%

iis s,

4 e

i e

i S

9 p *m

.k 4

36 same calibration standards and methods enployed at that time.

Testing was perfonced under t he supervision of Comnon-wealth Edison's Operatienal Analysis Departnent and uas witnessed by representatives f rom Travelers Insurance Company.

The ultrasenic and dye penetrant examinations were performed at the inside diameter and between the vessel stud holes on the cladding on the face of the flange. The dye penetrant inspection was conducted thrcugh tue quadrants of the

flange, whereas the ultrasonic inspection included an area 360 around the inside surface of the flange and at 15 various locatiens between the vessel stud holes. No indications of defects were found.

2)

Reactor Finnee Stu_d Bolts In addition to the weld inspections, ultrasonic tests were conducted on 25 vessel stud bolts by rittsburgh Testing Laboratories.

"o indications of defects vere fcund. Tests were rade on December 8, 1969.

3)

Reactor Thichle Weld Tasting Twenty-five of 28 control rod drive housing thimble welds were ultrasonically inspected by Pittsburgh Test.ing Laboratcries following removal of the control rod drives for everhaul during the refu ling nutage. The inspections were concuctec using tne ulcrasente rixture aw; reference standard constructed for the 1967 refueling cutage inspections.

Of the 25 thirble welds ultresonienlly inspected, no indicatiens of discontinuities were found.

Ultrasonic testing of thimble J-welds was initiated during the 1957 refueling outasc. A summary of the lccaticas of the thinbles inspected during the 1967 and 1968 outages is shcun in Figure 5.

4) Erimarv Fining a)

Exaninations The 1969 pricary system stainicss steel piping ueld inspectiun on Dresden Unit #1 vaa coupleted during th2

' sixth partial refueling, maintenance, and inspection outage. The inspection dupliccted ultrasonic testing methods of welded areas performed during the 1967 and 1968 excminations with use of dye penetrant techniques on the reactor vessel and steau drum Bi-metalic no :le welds.

37 Figure 5 REACTOR THDBLE ITELD INSPECTIONS OCTOBER, 1969 X

X X

X 10 X

X X

X X

X

+

+

9 X

X X

't X

8 10/69 10/69 X

X X

7 10/69 10/69 10/69

+

10/69 10/69

+

X X

X X

X 6

10/69 10/69 10/69 10/69 10/69 5

10/69 10/69

[.<

X x

+

10/69 X

x-x X

4 10/69 10/69 10/69 10/69 10/69

+

X X

3 10/69 10/69 10/69 X

X X

X X

2 10/69 10/69

+

i X

X X

X 1

A B

C D

E F

G

.H J

K X

Thimble welds ultrasonically inspected during the 1967 and 1966 refueling outages.

,a

+

Drive housings (7) installed with liners

F 3G Ultrasonic examinations were performed en 100% of all accessible velds en piping 10 inches and under in diameter and on 25% of 'he walds on primary piriac 16,18 and 22 inchas in size.

In addition, ultrascaic and dye penetrant examinations were perforced on nine 16" sttaa outlet not. les, tuo 10" unloading heat exchangar (capped) noasles and one 22" recirculation inlet nozzle on the reacter vessel.

Lesults of the inspection revealed no new indications on smaller diaceter piping as reported frca previcus inspections and only-two minor surfacc imperfecticus en the larger dicreter piping. All indicaticas fcund uerc well within,al]occhle tolerance as specified in the AS:3 Boiler and Pressurc '!cssel Ced:, Nu: lear Vessels, in paragraph N 322, and no repairs were necessary.

The sccpe of the inspecticn, as compared to the previous 1967 and 1968 inspections, is summarized in Table 7 b) Fvdrostatic Test The integrity of the entire pricary syc tam was confirmed by a 1000 psig hydrcstatic test following the refua1.ing cutage. This ;:cs: vas delayed by approxirately fear dcys, when during an initial att npt to prescurine the system, n leek was discovcred in a forced stainless steel tedacat ou the reaceut veut p ipin3 The defect ura 1ccated in a f^ur i.nch-to-tuo inch re92cer en the recctor yea; fl.cn;e, and originct:M Err.= tuo pin-hole indienticus 1/4" cpart and apprexinctelj 1/2 '

fren the reducer to flanga weld.

The fic was ccepletely rc=oved fro:a che vnll of the reducer and vas repnired in accordance with ASLI Ccd2 /.290 for total arc pasacs in filler metal on 304L stainless steels.

Following the repair, c hydrostatic test at 1600 poig was perfetted en the flanga and no leakage.:as evident.

Radiographic, ultrascnic end cye penetrcut exaninntiens were performed follouing the test, and rev;aled no evidenca of further crac':ing. An inservice inspection of this flan:;e end all similcr f orgings *:111 La incor-l perated into the annual prir.cr; system weld inspection to insure the integrity of these forgf.ngs.

I L

39 TABLE 7 ULTRASONIC INSPECTIO*1 OF PRI11ARY SYSTE'1 PIPINO UCLDS 1969 Neld Previous Weld Total Ucids Inspections **

Inspections *

lelded Plate Field Shop Field Shop Field Shop Risers (16")

.56 36 18 13 18 9'

Downcomers (16")

50 20 8

2 18 5

Suction Piping (22") -

38 32 8

10 18 11 Return Piping (18")

32 22 8

6 13 8

l Return Piping (22")

4 4

1 0

5 0

Ultra-conic inspections conducted during 1967 and 1968.

• • Nunber of indication: (by U.T.) during 1969 = 2.

Scamless (10" and under)

Total Total 1969.; eld **

Previous t-?e l d

• 10/67 Total Pi e Icntch si,.

c.1d c, insr.:: t ter.:

Ih:n:ctiene M Inrpected ' *' " " N ! * " * *M 1967 1968 10" 12 10 83 100 0

8" 29 29 100 100 97 6"

96 96 100 100 94 4"

60 60 100 100 68 3"

55 45 100 100 32 2"

34 29 68 85 65 1"

15 15 100 100 0

• Ultra-sonic inspections conducted during 1967 and 1968, except for the 1965-1966 inspection of 30 welds on the six inch by-pass loops

• • Nu=ber of indications (by U.T.) during 1969 = 3 h

2

w 9

13.

Erc.er~ancv Condenser a.

_ Sand 31n s tine and Marainted The entire interior surface of the energency condenser, including the vent stack, uas cand blasted and painted during the refueling outage. The surface was blasted to a white metal finish and was brush painted. When the entire vessel had i

i one brush coat, a second cent was sprayed on.

The raterial used,uss plasite 7155F.. An average thickness of.00S"to

.010 vas measured.

A 24 inch 150 pound weld ncch flange cas walded to each man-way.

This, when closed with a 125 pound blank flenge, provides a testable volume for checking the can-vay clesure uithout a sphare test.

b.

Vanhecd Instc11ntion crd Tents The north and south emerger.cy nanheads vere leak tested on Novenber 17, 1969, and Decceber G, 1969,respectively.

The leah tast uns perforced to insure the in:egrity of the primary containment fclicuint; the recent closure of tha manheads. The ranhcads had bcen opened to c11ou.ccess to the emergency condenser interior for peinting, in orde r to racilitate tne tes tir.3, a finnga was const.ructed around the exterier of the nanhead to allou prescurizing ca the sp h2 re side.

The test procedure included installing c test gauge on the flance, presnurizing uith service air to 20 psis, ar.d soap testing all accessible connections. Tne test uns conducted fer thi.rty minutes and leak rete data were collected.

The results of the test are as follous:

Penetratien

% of Allcwable Lerknee of 37 ps.~

4 South Manhead 0390 North Manhead 0139 TOTAL 0529 14.

LJ. and I.P Turbinc Overhnyl a

n.

In general, the results of the inspection shcued the 1.P. and L. P.

turbine to be in very good condition. Radiatica levels were much icwcr than expected. Uindnu open raadings at the I.P.

inic; were apprc::imately 35 MR/3R.

I

41 The I.P.

turbine spindle was removed while the L.P.

turbine spindle was not removed. The condition of the blading en both spindles was found to be excellent.

Blades vare checked by nagnetic particle examina tions.

The. I.P. Horizontal joint diaphragu ledges from the 17th through the 22nd stage vare built up with veld and ground to s ize.

Both upper and lover halves were repaired. Support i

struts on the lover half of the I. P. vere padded. A seal area on 03 packing was welded.

Several turbine gland scaling surfaces care machinnd and matal-sprayed. Four sections of number four gland and three sections of number six gland were repaired.

Both of these glands had been machined and metal sprayed in the 1962-63 cutage. The spray metal used then was Metco #2.

The spray metal was contained in.040 inch curbs machined into the shaf t sealir.g faces.

Eacause of the differential expansion, the sprayed retal buckind up and rubbed tha scal strips. This cid overlay uas machined out, a coat of bcnd uns applied, and then Mateo f 5, which has a louer coefficient of expansion, was applied.

The stean seal r2gulator controls were cica-cnd adjusted.

The General Electric test section in the 15th stage moisture removal pipe was removed. Af ter inspection by General Glectric, it was replaced.

The main steam clactro-magnetic valve pilot valves were c1 caned and lubricated.

The I.P.'and L.P. bearings and the thrust bearing vore inspected. They all were in good condition with satisfactcry clearances and good wear patterns.

All I. P. and L. P. bolts were sonic checked and found to be satisfactory.

The lubrication system was in good repair and was cicaned.

The hydrogen system was not checked since there was only very low leakage in the system.

b.

Several turbine ecmponents ucre replaced during this cutase and some neu ltems were added. These items are explained below.

The 20th, 21st, and 22nd diaphragns were replaced in the I. P.

turbinc. This uas a planned replacement. The old diaphrages were cast iron while neu ones are cas t s cel.

42 The old carbon steel subway grating stean flow straighteners, which were installed in 1962-63 and removed in 1968, vera replaced with stainless steel subuay grating during this outage.

The ADS (Autenatic Dispatch System) was added to the speed governor.

c.

The primary ctop valves were opened during the outage and the discs vere roved to the shop and sandblasted cican.

The discs were then lapped into the seats. New paching was installed and the valves were tested by an isolation valve test and found to be leakage free.

The high pressure turbine casing pressure sensing line stop valve was replaced, d.

An inspection was rade of the turbino c:ctraction pipin6 Tha horiscatal runs of both crossunders ucre inspected. The usual wear pattern uns evident. The pattarn dccs not shou strd.ght erocion tarks but is in a non-directional pattern.

Scme crogfon type uear was noticeable at the 90 elbcv.

The ucar, in general, was not appreciable and no pockets were nora than 1/C inch into the 1/2 iach thict: wall.

Up under tha turbine (!I. p.)' a t the 30 to 24 inch reducers, no special wear was noted. There was loss of metal jist beloi tLe casing and will be built up on tha future turbine everhaul.

"D" extraction tras inspected in three areas; one carben steal; one copper bearing carben steel; and one 2-1/4 chrcne uoly.

All cf the areas showed an even coffee-brown color uith no sign of wear.

"C" extraction was inspected and fcund to be in good condition.

43 15.

Main Condenser Retubing and Feedwater Heater Replacement a.

Condenser Retubing Retubing of the Unit #1 main condenser was a result of numerous forced plant outages during Cycle VI operation caused by a rash of failures in the Admiralty metal tubes.

During Cycle 6 operation, s ixteen CL6) plant outages were attributed to condenser tube leak repairs. The retubing operation was initiated concurrent with plant shutdown for refueling on September 6, 1969.

The condenser work was performed in conjunction with the outage and was essentially completed by the end of November.

In general, the operation entailed erection of a temporary building to act as a pull space for the tubes, removal of concrete shield walls and condenser water boxes for access, removal of all 16,438 tubes from both the east and west halves of the condenser, and installation and fitting of new 304 L stain-less steel tubes through modified single tube sheets.

The retubing operation and an initial tube icak check were completed during November. Upon completion of turbine overhaul work, a vacuum was pulled on the condenser and a final tube check was cade.

The check uas satisfactory.

b.

P.criacement cf "n" Saccr :ar- ?caduatnr Mrater The "D" secondary feedwater heater tube bundle vac replaced during the refueling outage. The original bundle was a Foster Wheeler bundle with 500-5/8 inch nickel-copper,18 gauge "U" tubes.

This heater has had a history of tube leaks, approximately 88 plugged to date, and was selected previously as the first heater to have a neu bundle installed.

The new tube bundle was manufactured by Yuba and consists of 470-5/8 inch, 20 gauge, 304 L stainless steel "U" tubes.

Upon completion of the tube bundle installation, a hydrostatic test was performed on both the shell and tube side of the heater.

On December 5, 1969, the tube side was hydroed at 225 psig and on December 8, 1969, the shell side hydroed at 1250 psig.

Both tests were witnessed by plant personnel and representatives of Travelers Insurance Company. The test results were satisfactory.

14 16.

Tests a.

Schere Inte rity Test Fronrnn As part of the sphere inta;rity test program, the sphere ventilaticn va lves, the primary stes:a is01s tion valva emergency condanser manheads, transfer tube covar, and the reactor containment vesr.el w2re leak tasted durin3 the outage.

1) Reacter Containment Vassel Leak Tes t The sphere was leak tested at 20 psi;~on October 4-5, 1969.

Tuc test consisted of a sin;1e 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Lcst follouing the same procedures used in the 1965 sphere tests. The calculated lenhage rate at 20 psig was.14527./21 hours of the totc1 quantity of air in the sphere. 2::trapolating this valut to 37 poig using :he laminar flev ccuation results in a leakage rate of.21737./24 hours (L3. 57, of license 11 nit).

2). Sphere Ventilation Valves The sphere ventilatic, supply rad exhaust valves ucre tested for leakage on December 17, 1959.

The test was ccudocted at 20 psi'd and the data extrapcicted to 37 psig.

Cal.cula ted l

leahage retes indicata a leahase rate at 37 psig to be 2.497, cf the clicrable license limit of.53/24 hours for the supply valves and.35% of the limit for the exhaust vtives.

3) Primary Stean 1solatien 'Jr.lves The inciation valvcs en the pt L-cry steam lLnes were leakage chect:ed on December 10, 1969. The tes t uas pe rCc=.d at 37 psi; and the resulting leat: rate was.347,/24 hours of the total quantity of air in the sphere at 37 psig.

(35.17. o E liccate limit).

4)

Transfer Tube Cover Leak Test On December 2S, 1969, th-2 transfer tube betueen the isolation valve and bolted cover uas pr ssurized uith air to 20 psig.

The arca above the bolted cover teos flooded with water to locate visually any lea > age which might occur.

The transfer tube bolted cover teos tight.

The test uas conducted at 20 psig and extrapoleted to 37 psig using the laminar ficu relaticeship. Equal 12akage was assured in each directien. The contribution of the transfer tuba to the total alleuable sphere lea' rage was calcula ted to be 2. 367.

of license limit.

,)

43 j

5)

Air Locks All air lecks, ventilatini valves, and process isclation valves were tested periciically during the year and found to be within the licensed allowable lecinge rate.

b.

Pr iru rv Steam Drun Safetv Va lve s -

Four spare primary safety valves were installed on the pri=ary steam drun on I;cvember 10, 1969, replacing-four of the safety valves previcusif in service. All four spare safety relief valves were tested for relief pressure in December, 1968, and were set at their respective design pressures i 10 psi. Relief prescures vere checked by repeated popping. The valves installed were cleaned, set at leak-checked in the shop facility, c.

Ter' era ture Coe f ficient of 'anctivity Check The codcrater temperature coefficients were neasured at several points during the reactor startup on February 2,1959. On this date, the core exposure accu =ulated during the cycle uns approritately 130,000 ! Cat. For purgeses cf relating reactor period to reactivity (J), cur"'.a tive Ore burnup uns assuned to he 7,000 IND/ ton.

In figure 6, the temperature coef ficient is plotted versus node-ter tempera ture.

For couparisen, the tenpercturc coef ficients reasured Juna 1, 196S, a t the besinning of the Cycle VI. are alte pletted.

d.

Fuel Sinping Fuel sipping in the reactor vcscel con =enced September 12rh, six days af ter Unit #1 was shutdot.n.

Fuel sipping too!. precedence over all other refueling activities. All Cycle 6 fuel, uith the exception of assenblies /.463, P710 and SA-1, was sipped at the cana l.

Fuel cipping at the canal uas concluded on Septenber 19tii.

Sipping in the fuel building was conducted en the three cssemblies not sipped at the canal and ca eight other assemblics

• with bordarline indications frc: canal samples. Final analysis of all sipping saeples determinec 33 fuel assemblies to be defective.

Table 9 su==arices the fuel determined to be defective by sipping and discharged at the end of Cycle 6.

e.

?:inirur Critical Tests Fuel loading commenced on Nevanber 23, with the loading of the sourec ossembly "N 192.

Critical chechs uere undertaken af ter each cell of four assemblies was loaded and its respective control rod ar:cd.

Plots' of.:.ultiplication cera rade and used to deternine how many additional assemblies could be loaded safely before the next check. Four asscnblies cere Iceded at a time until there vere 24 asse-blies In the core.

Ti'e text chech var nsde with 26 csserilien nnd thereafter fuel was added one ascenbly at a time until criticality was achieved.

• G1, G14, Gl9, G28, G64, G90, E139, and DU36

iG FIGURE 6 MODEFJ. TOR TEMPERATURE COEFFICIENT OF REACTIVITY agg --

+5.0..

+4D

+ 10 a.20

+LO e,3.._. _

c Ie

~

_1n "

~

1.i 2D --

s' 6/1/o8 5

- 10 2/2/6:

u

-49 i

y l

L k

-50 6

v

-uD O

1 e

f;

'.0 Ec H

-80

_ 93 --

-ICD

-110

-120..

-110

-;40 h

5 0

100 200

>U0 400 50(

L00 Mc ierau r Tempe ra t ur-i,'

o

47 TABLED {

SIENARY Op DE?ECTIVE FUEL ASSEM3LICS (ECC 6)

Assembly Cyc l e VI Exoosure Cycle Description Cor e Location Dit?D/T*

Loaded G1 1461 16.8 4

Type IIIF Normal Assembly Gli 1061 15.1 4

Type IIIF Normal Assembly G14 1065 11.7 4

Type IIIF Normal Assembly G15 1055 14.6 4

Type IIIF Normal Assembly G19 1071 15.8 4

Type IIIF Normal Assembly G20 1263 17.0 4

Type IIIF Normal Ascenbly G27 2065 16.0 4

Type IIIF Normal Assembly G28 1069 11.2 4

Type IIIF Normal Assembly G42 1160 17.5 4

Type IIIF Normal Assembly -

G51 1465 17.0 4

Type IIIF Normal Assembly G61 1255 12.9 4

Type IIIF Normal Assembly G64 0869 11.2 4

lype IIIF Normal Assembly G68 1257 14.5 4

Type IIIF Norm 11 Assembly G73 0964 14.1 4

Iype IllF Normal Assembly G79 1659 13.2 4

Type IIIF Normal Assembly G30 1265 15.1 4

Type IIIF Normal Assembly G35 1861 12.8 4

Type IIIF Normal Assembly GS7 1661 13.9 4

Type IIIF Normal Assembly G95 1463 16.6 4

Type IIIF Powdered Fuel GiOO 1459 17.2 4

Type IIIF Powdcred Fual Gill 0865 12.3 4

Type IIIF Special Gd-U Rods XE45 OS63 15.4 4

Type III3 Loaded E00 /

XE103 1858 15.3 4

Type IIIB Leaded BOC 4

• KEID/ T MiiD/T X 10

=

40 TABLE 8 (cen t. )

Assembly Cycle VI Exposure Cycle Description Core Location KM!JD/ T

• Loaded E139 0956 15.4 4

Type IIIB Loaded BOC 4 DU36 1464 9.2 5

Type V High Cd DU46 1060 9.7 5

Type V High Gd DU50 1062 9.6 5

Type V Iligh Gd DU51 1264 9.5 5

Type V High Gd DUS1 1058 8.7 5

Type V High Gd DU62 1456 9.3 5

Type V High Gd DU89 1864 9.2 5

Type V High Gd DU92 1064 9.6 5

Type V IIigh Gd DU93 1262 9.1 5

Type V liigh Gd

• CulD/T mlD/T X 10

=

E

49 The first critical was achieved on h'ovember 24, at 10:20 p.m.

uith 33 fuel assen51ies. There were nine control rods ava il-able for uithdrawal and a pe: icd of 98 saccnds vas attained on the fourth notch of the ninth cod.

The period of 98 seccnds indicated an excess reactivity to be apprcxicately.06&%

4 K/K uith eight rods, four notches withdraua.

The second phase of the test censisted in re=oving a neu Tyre VII assembly frc= the lattice and replacing it uith a Type VI assembly which had an exposure of approxima tely 6700 M'Ja /t.

At the sane rod pattern, the period was measured to be ISS secends.

t Af ter ccmpletion of the minimum critical chechs, fuel looding was resumed according to procedures f.

Shurdmm Mar ~.in Chechs Reactor shutdown nargin chec'es ucre cenducted on Decerbor 5,1909, to dc=anstra;e that the refueled Cycle VII ccre wet licen u rer uitecents wi_th regard to the " stuck red" criteria and thct s

the cargin is in excess of cne percent throughout the core.

The core was shutdown in excess of 1.09*/. on the periphery ned in excess of 2.097. in the central regicns.

Both values mot the license limit of a minimum shutdo.:n margin cf.l.u7,.

l l

l

50 B. License DPR-2 Table 10 lists the amendments to our license requested and/or.uthorized during the year.

Pertinent correspondence pertaining to thes< requests are listed in the Correspondence References.

S I

51 TABLE 10

SUMMARY

OF LICENSE AMENDMENTS PENDING DURIMG 1969 Date Request _

Authorization Request to a:aend License DPR-2 to permit 4/15/69 9/11/69 loading of up to 85 Type VII and 11 Type VII-Pu plutonium fuel assemblies.

(Change No.18)

Request to amend License DPR-2 to permit 11/12/69 irradiation of four cobalt-59 capsules to canceled provide data on cobalt-60 production in 12/23/69 BWR' s.

(Change No. 19)

Request to amend License DPR-2 to establish 11/26/69 12/31/69 ccmpatibility between DPR-2 License and the proposed Technical Specification of the proposed License for Dresden 2, DPR-19.

(Change No. 20) 9 e

d n

su'

52 Correspondence References - 1969 (1) Letter to AEC, dated May 15, 1969, applying for an amendment to DPR-2 to allow operation'with a replacement fuel batch, consisting of Type VII and Type VII-Pu fuel.

(Change No. 18)

(2) Letter to AEC, dated August 20, 1969, submitting cdditional information pertinent to Change No. 18.

(3) Letter from AEC, dated September 11, 1969, authorizing Change No. 18 to the Operating License DPR-2.

-(4) Description and Safety Evaluation Report of Proposed Change No. 19 to

-the AEC, dated November 5, 1969.

(5) Letter to AEC, dated November 12, 1969, requesting an amendment to DPR-2 to permit irradiation of cobalt, during Cycle 7 and to correct Table 1.

(Chauge No. 19)

(6) Letter :.o AEC, dated December 23, 1969, concerning Proposed Change No. 19, dated November 12, 1969, and the decision not to install the cobalt-59 during Cycle 7 on Dresden Unit 1.

(7) Letter to AEC, dated November 26,1969, concerning Proposed Change No. 20 to establish compatibility between DPR-2 and DPR-19.

_(a) Letter frca AEG, cated December 31, 1969, ~ autnorizing Unange No. 20 to the Technical Specifications of Facility License No. DPR-2.

e

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