ML20024B214

Reviews Current Events - Power Reactors, Feb 1975.Action Required

ML20024B214
Person / Time
Site:	Crane
Issue date:	06/26/1975
From:	J. J. Barton, Toole R GENERAL PUBLIC UTILITIES CORP.
To:	GENERAL PUBLIC UTILITIES CORP.
References
TASK-*, TASK-GB GPU-2486, NUDOCS 8307070352
Download: ML20024B214 (14)

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AEC 00CU"ENT REVIEil

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f Riant/ Unit 24-OYb The attached AEC document has been reviewed for test program and design modification requirements for the above Plant / Unit.

DOCUMENT:

Operating Experience, dated:

lX Current Events - Power Reactors, dated:

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Other

, dated:

Review of the attached document has concluded.that no action is' required.

Startup 4 Test Manager Date Test Superintencent Date Review of the attached document has concluded that action is required by:

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o Problem Mr. John Barton of GPU Stirtup 'for.rarded an AEC Docu=ent Review for= to me for follow up action regarding diesel fuel oil s torage t'anks.

In the Februarf issue of Cur' ent Events - Pe ser Reactors issued by the NRC an incident at the Edwin I.

Hatch Nuclear Plant was described wherein their emergency diesel generators did not start due to water accu =ulation in the diasel day can.c.

I was requested to inves tigate the water tight condition of the Unit's2 diesel fuel tacks.

In conjunction uith Messrs. Harper and Lunson of S&R an investigation of this iten uas conducted with the following results. Since neither the diesel storaga tanks nor the day tanks are buried tanks, the only credible way for water to enter the tanks is through the tank vents. The vents are enternal to the build-ing and. point vertically into the air. Mounted on top of the vents is an open flame arrester which would act as a rain water collection bucket and allou vater to enter the tanks. ~3&R (Lunson) is to issue an ECM by

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6/20/75 directing the field to add e U-bend pipe on the vents such that the fla=e arrestors will now be pointed d.::invard and will not collect rain uater.

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5j.W EVENTS SELECTD FRCM RE?CRTS SiJEMITTD TO THE UNITD STATES NUCLEAR REGULATCRY COMMISSION AS OF:

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FUEL RCD CLADDING FAII.URES N'

On October 30, 1974 at Dresden Nuclear Power ation Uni: 3, while che position of the control rods was being chan ed, excessive pcwer

,s peaking was observed in the lower region o' the core. Additional control g.

rod novements to reduce peaking caused th peaking :o increase. This was accompanied by a high off-gas radia on alarm. The control rods l

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were then inserted to reduce both peakids and power, and the radiation l

f alarm ceased.

1 The esti=sted stack off-gas releas rate during the transient was jh approximately 300,000 uC1/sec. T/e off-gas release race for the day of j

the occurrence has been es:i= aced' by :he licensee :o be an average of j

45,000 Ci which is less tnan b,ilf :he technical specifica: ion limi:;

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the 48-hour technical specification = limi: probably was exceeded for less a

chan 10 minutes.

Prior :o control red novemept, reae:or power was 440 MWe; the power increased to 520 MWe during control rod movement and stabilized at 370 MWe after clearing thi high off-gas radiation alarm. Design power is 800 MWe.

The apparent cause of 7.e occurrence was personnel error. A xenon transient occurred, wi,th a known ecndi:1on of very icw fuel exposure at the core bot:cm. The! rapid changes in power occurring low in :he core probably resul:ed in some fuel r,cd perforation. Failure of fuel red cladding resulted in. che high of f-gas release race.

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R 24 Uni: 3 had a known flux dis:r1bution peaked :oward the cop of the core, 7

with relatively unexposed fuel at the lower sec:1on of the core. The A

control rods were being moved to change the axial power dis:ribucion of the core at the ti=e of xenon peaking, and when xenon started burning out, the bottom power peak, which had jus: for ed by control rod move-increased rapidly. Because the bottos of the core had relatively

ment, 11:cle exposure, the power peaking problem was exaggerated.

Since.:he occurrence, che undesirable race of off-gas radiation indicates C.

several fuel rods with eladding failure.

The high off-gas radiation levels have resul:ed in a station imposed 50% derating of power.

At a derated load of 400.We, the stack gas release race has'been about 24,000 uCi/sec, about one four:h of the technical specification limi:.

The cause of the suspected failure is evidently pellet clad in:eraction.

f No fuel safety lisi:s were approached in :he occurrence.

This = ode of fuel f ailure has occurred previously in the Dresden units during nor=al reactor operation.

l An isotopic analysis of the off-gas composition indicated 11::le or no holdup in the pellet or fuel column for fission produe: gases. This J

would i= ply that few relatively large cladding perforations occurred during the peaking transient. There was no adverse effect on :he public health and safety.

As a result of this occurrence, = ore trnning for all nuclear engineers have been initiated; procedures for 3kA control red sequences have been i

reviewed and new procedures approved; and the General Electric Ccmpany has recoc= ended certain operational procedure changes to reduce the number of fuel failures during normal power redue:1on operation.I I

CONTROL RCD L'ITc'DRAk'AL PROBLDiS Dresden-2 While Dresden Nuclear Power Station Uni: 2, us.s in a refueling outage with the mode switch locked in :he shutdown position and control red drive (CRD) overhaul was in progress, a repaired drive was installed at posi:1on 10-35 and an at:empt was made :o remove drive 18-11.

The uncoupling tool for drive 18-11 f ailed :o indicate if :he drive had properly uncoupled from :he blade, so the drive probe was reinstalled and indicated the drive had uncoupled.

Shif: personnel then issued a temporary procedure change to allow the General Elec:ric naintenance crew to skip drive 18-11 and proceed :o the nex:

step, the withdrawal of drive 6-35.

At this time drive 10-35 had been valved in service and not reinser:ed to posi: ion 00.

of service, Drive 6-35 was then withdrawn and valved out y

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d The operator and shif t personnel on duty did not notice that control red 10-35 was still at position 48 before withdrawing adjacent control rod 6-35.

When it was noted, approximately an hour and a half later.

drive 10-35 was immediately reinserted.

The temporary procedure change to allow the General Electric main-tenance crew to proceed to the next step viclated the intent of the 1

orginal procedure. The procedure for CRD replacement had been written such that if a step in the procedure were skipped, the next CRD to be removed would be adjacent to the CRD vichdrawn previously. The pro-cedure also stated that the two CRD's could be pulled only if separated by two or more control cells. All personnel involved failed to notice that the two adjacent control rods would be withdrawn with theo bnplementation of the temporary procedure change.

The safety of the plant and public was not in jeopardy. Neither of the withdrawn control rods were the highest

• worth rods; the reactor was subcritical by greater than 1.34%.

There was no da= age to any systen or structures. No personnel received injuries or exposure, and no radioactive =aterial was released.

A temporary procedure change was subsequently issued to instruct the

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operators to verify that procedure changes will not result in a control rod being pulled within four cells of a withdrawn control rod.

In the future, station nuclear engineers will review procedures involving control rod movement. The control rod drive procedure will also be revised so that several steps may be skipped without violating separation cr iteria. "

Brunswick-2 Prior to fuel loading at Brunswick Steam Electric Plant, Unit 2, and during performance of a periodic test it was discovered that the de-select interlock on four control rods had failed. The refuel interlock e

check required withdrawal of.one control red to position 02 and verification that the centrol and " Withdrawal Permissive" light remained energized.

Before performance of the 'next step, the operator was able to select an adjacent control rod. The " rod block" alarm was not received and the operator was able to withdraw the second control rod, so all rods were fully inserted and management apprised of the situation.

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Inves:igation revealed that one pin of a connce:or had lost its mechanical connection with the copper bus in a relay module of the reae:or manual a

control system. The failure could have been caused by manufae:uring, P'

shipping, or installation error.

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and The relay module was removed, repaired and re:urned :o service, the refuel interlock check was revised :o include a func:ional :est of

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Because the malfunction of :he relay module was detected prior to fuel j

leading, there were no adverse effec:s on the plan: or' o the health and general safe:y of the public.3 l

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A: :he Quad-cities Nuclear Power Station, Uni: 1, prior to reactor star:up the red worth minimiser (R'd) was operable. '41:h reac:ce power i

level of approx 1=stely

$*.', a control red was withdrawn that should

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have resul:ed in a R'41 stall which should alarm in :he control room and block rod wi:hdrawal; net:her event happened, u

4 The operator was unaware of :he R'ai f ailure and continued w1:hdrawing d

cont Cl rods; a violation of the technical specifications. Approximately a hour and a half later, the reactor opera:or mistakenly selec:ed and y

withdrew an out-of-sequence control rod. Ten minu:es later the error was noticed. The out-of-sequence rod was reposi:ioned and reac:or star:up was halted un:il :he source of :he R'M f ailure could be determined.

The f ailure was at:ributed to an error in the star:uo procedure for :he

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This procedure was modified during a spring 1974 refueling ou: age, and a step requiring the computer rocm console mode swi:ch :o be :urned to the "of f" position was omi::ed. As a resul:, wi:5 :he console sw1:ch in the local posi: ton, the con:rol room did not receive indication of a ecmpu:er stall and no rod blocks f ollowed :he s:all.

the pos:ulated rod drop accident :ha: forms :he basis for the R'41 :echnical s

specification would require a high worth rod :o become uncoupled. the rod to stick when the drive is w1:hdrawn, and :he rod :o f all bef ore the rod was detec:ed to be uncoupled. There were no uncoupled control reds, no rod denp, and no abnormal : ore ::ansients as a resul: of :his occurrence.

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• f 5-CRACKS IN FEDUATE SPARCES Two cracks were discovered in the f eedwater spargers during a General Electric Company recommended f eedwater sparger inspeccion at Uni: 2 of :he Dresden Nuclear Power Station. One crack was located on the upper par: of the righ: side header pipe :o :he junction box weld area of the southwest quadran: sparger.

The crack was rela:ively s:raight and extended about 90 degrees around :he pipe circumference.

The other crack was loca:ed on the upper part of :he lef: side header pipe to :he junc:1on box weld area of the nor:heast quadran: header.

This crack was more jagged, near the weld ares and extended about 200 degrees around the pipe circumference.

The bearing bars (preload spacers) on the right side of :he sou:heast and northwest quadrant spargers were not in contact with the vessel wall.

T*/ inspection and cold flow tests led to the conclusion that the primary cause of cracking was flow induced vibration.

Tests by CE have shown that sparger vibration will occur even if only leakage flow between the inside of the :hermal sleeve and the outside of the feedwater no::le is present. The :ypical service conditions of changing feedwater flow race, :he :empers:ure difference between feed-water and the water in the vessel, and movement of water wi:hin the vessel also contribute to the problem by producing thermal cycling and thermal stress.

The four spargers were tc. be replaced during a refueling outage wi:h spargers of a new design, a design based on :ests where there was no vibration under flow condi: ions and of schedule oO rather :han schedule 40 stainless steel pipe.

Af ter an evaluation by General Elec:ric and Commonwealth Edison it was concluded :har the consequence of sparger cracking when sparger posi: ion is maintained is not of saf ety significance. The event did not cause any personnel injuries, exposures, or release of radioac:ive material.

Ie was concluded that :his event did not endanger the heal:h and safety of the public.

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tihen insulation was removed from the core spray lines at the Dresden I

Nuclear Power Station, Unic 2, cracks were discovered in each line.

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• The core spray lines are 10-inch diameter stainless steel pipe welded to the reactor vessel no::les. Short pieces of welded stainless steel pipe (Dutchmen) connected the core spray line with the safe-end of the reactor vessel.

• • • 1 Two longitudinal cracks one 0.75-inch, and the other 0.5-inch, occurred at the heat affected zone of the Dutchmen to the safe-end weld.

One 0.4-inch circumferential crack occurred at the heat affecte'd zone of the Dutchmen to the inlet pipe. Two circumferential cracks occurred at the heat aff ected zone of the other core spray loop Dutchmen to the pipe weld.

Each crack was 0.12-inch in length. All five cracks were weeping, indicating each was a through-wall penetration. All other welds in the core spray lines were to be examined.

Engineering drawings indicated the saf e-ends and Dutchmen were type 316 g

furnace sensitized stainless steel. The inlet piping was type 304 y

stainless steel.

The reactor was shutdown for refueling and replacement of the 4-inch recirculation line bypass piping; there was no danger to the health and safety of the public.'

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• iacer was discovered to be spraying from a crack in the suction pressure indLcator piping of the charging pump at Unit 2 of the Zion Station.

The leak was estimated to be 1 gal / min, so the pump was stopped and isolated.

Normal flow was maintained without interruption with an alternate charging pump.

The crack developed in a 3/4-inch pipe immediately adjacent to a socket wold which had been repaired one month earlier. The failure was aggre-g vated by high vibration;

..e charging pump is of positive displacement Q

de.i.;n eind nearby piping is subj ect to high vibrations.. Material adjacent to the crack was submitted for metallurgical examination.

Vibration aggrevated fatturos have been a c mmon occurrence on the char.: tnc p ipes a t " ton-2.

A study is underway to determine the feasibility of installing pressure pulsation damping in the positive p mp discharge u

piping.

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. The released pri=ary coolant was collected in the floor drains and e

routinely processed with other liquid radioactive vastes. There was

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no danger to the plant or to the health and safety of the public.7 UNFI.MGED REI. EASES Oconee 1&2 i

A sample from a vaste gas tank at the Oconee Nuclear Station Units 1 and 2 had been analyzed and a release rate of 40 CFM had been determined from the analysis. However, during discharge, an alarm was received from the vent gas monitor; release of the vaste gas tank was eersinated.

The vaste gas tank had been sampled with the correct 100 21 container, but when entering infor=ation to the computer to calculate the release race, a volu=e of 3300 m1, used in several procedures, was used instead of the 100 mi sample value. Subsequent recalculation with the correct volume indicated the correct release race should have been 8.6 CFM.

However, the actual release rate was still only 9% of the one-hour release rate limit. Because there was no radiation exposure and the release was within the limitations of the technical specifications, it was concluded that the health and safety of the public was not affected.8 Surry 1&2 With both units of the Surry Power Station in the cold shutdown condi-cian, a tube rupture occurred in a component cooling heat exchanger.

The rupture resulted in the release of approximately 8000 gallons of low level radioactive component cooling water to the service water system and ultimately to the James River. The leak rate was esti=ated to be 170 gal / min.

Investigation of this failure and resulting large leak revealed that the component cooling system had been undergoing minor dilution for a 66-day period. Officials est1=ated that the dilution over this period would have been the result of a leakage of at least 33,500 gallons of low level radioactive water. This release, too, was to the James River.

Only one tube had failed in the heat exchanger.

It was presumed that the slow leak was caused by gradual degradation of this tube; the tube was plugged.

It was the first instance of a tube failure in the component cooling heat exchangers.

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Af ter the exchanger was returned to service for a short time, component i

cooling system leakage was again discovered; there were pinhole leaks l

in two tubes. The two leaking tubes resulted in an additional 2445 gallons I

of low level radioactive water being released to the James River.

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failed tubes in the component cooling water heat exchanger were plugged.

Eleven days later, a radiation monitor that senses service water leaving j

the component cooling heat exchangers alarmed.

~he heat exchanger was l

i==ediately isolated and pinhole leaks were observed in two tubes. They l

had been leaking for more than 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, which meant there had been an additional release of 349 gallons of low level radioactive component i

cooling water to the James River.

h Preliminary results of an analysis of a previously failed tube e.v eacted a

from the heat exchanger indicated the failure was =echanical in nature ll j

and not corrosion related.

? i i l Procedures have been instituted to prohibit the use of component cooling system pump and heat exchangu combinations which could sesess the heat f

exchanger tubes and cause further failures. Service water radioactivity i

is being logged on two hour intervals to ensure that trends are detected in a timely manner. Two tubes have been removed for further inspection.

An insignificant amount of radioaccivity was released to the environ-ment and all radioactivity concentrations were within 10 CFR 20 limits.

This occurrence did not aff ect the safe operation of the station or the hesich and safety of the general public.3 STE.W GENERATOR TUBE DETERICRATION In a continuing program of eddy current testing of the steam generator tubes at the Surry Power Stations, Units 1 and 2, a :otal of 10,381 hot leg side tubes and 1689 cold leg side tubes have bet a tested;195 tubes had wall thickness deterioration of greater than 50% so these tubes were plugged.

Tube wall deterioration of the steam generators at Surry 1 and 2 is believed to have been caused by sheet sludge deposits and deleterious ef f ects of sodium-phosphate chemistry control. However, this phenomenn is a generic problem and its solution is not ccepletely understc7d.

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changed from phosphate treatment to all volatile treatment. An all volatile creatment specification has been provided by the nuclear steam supply system manufacturer, the Westinghouse Electric Corporation.

Plugging of tubes in the steam genera: ors resulted in only a negligible reduction of available heat transfer area. There were no safety implications associated with the tube plugging, and this occurrence did not af f ect the saf e operation of the s:stion or the heal:h and safety of the general public.13

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DIESEL GENERATOR PROBLEMS A number of problems were experienced with the :hree die el generators at the Edwin I. Ratch Nuclear Plant, Unit 1.

One diesel generator would start and operate a,short time, but would not come up to speed because a einer (Agastat) was cutting out before the diesel reached i

raced speed. The timer, which had been set for 7 seconds, was :ested j

and found to trip in 4 seconds.

In addi tion to.:he timer problem, the i

booster for the governors of both diese, generators was rusted on the air side. Without the booster, the diesel could not rotate fast enough to allow the shaf t-driven pumps to supply sufficient oil :o :he governor.

The vendor, Fairbanks Morse, recommended cleaning and increasing the l

port si:e from 0.025 inches to 0.050 inches.

The larger por: allowed I

air to enter the booster faster and open the fuel rack so the diesel generator would come up to speed before cimer closure.

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During a loss of off-site power :est, one diesel star:ed automatically and picked up its load on the bus. However, after one minute of opera-cion, the diesel shut down.

Investigation revealed that :here was approximately 50 gallons of water in the diesel day :ank. The day tanks for ths other two diesel generators were checked and found :o contain less than two gallons of water.

The day tanks are supplied from underground storage tanks located ex:ernal to the diesel generator building. Water had accumulated in :he access area co the storage tanks where'the transfer pumps and fuel oil sample penetrations are located; the cap on :he sample penetration pipe, located below water level, was.only hand :ight.

Water also could have leaked into the storage :ank through :he pump seals which were under wa:er.

In addi: ion, the storage :anks

cntained a manhole which could leak. I: was found that :hree gallons of water had leaked into each storage tank.

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grade. In addition, the hatch cover vill be =ade water tight.ll jprn!

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and Program Control U. S. Nuclear Regulatory Co=sission tr4 h3 s

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4 REFERENCES 1.

Letter, B. B. Stephenson (Co==enwealth Edison) to J. G. Kappler, USNRC, Office of Inspeccion and Enforcement - Regf on III.

January 17, 1975. AOR No. 74-38,;acket No. 50-249.

2.

Letter, B. B. Stephenson (Commonwealth Edison) to J. G. Keppler, USNRC, Office of Inspection and Enforcement - Region III, February 3, 1975. AOR No. 75-10, Docket No. 50-237.

3.

Letter, E. E. Utley (Carolina Pouer & Light Company) :o N. C. Moseley, USNRC, Office of Inspection and Enforcement - Region,II.

January 29, 1975. Docket No. 50-324.

4 4.

Lecter, N. J. Kalivianakis (Commonwealth Edison) to J. F. O' Leary, USNRC, Office of Nuclear Reac:or Regulation, December 26, 1974.

AOR No. 74-39, Docket No. 50-254.

5.

Lecter, B. B. Stephenson (Commonweal:h Edisdn) to J. G. Keppler, i

USNRC, Office of Inspeccion and Enforcement - Region III, January 27, 1975. Dockee No. 50-237.

6.

OIE Notification of an Incident or occurrence No.143, January 30, 1975.

7.

Letter, J. S. Bical (Commonwealth Edison) to J. C. Keppler, USNRC, Office of Inspection and Enforcemenc - Region III, January 9, 1975. AOR No. 74-56, Dockee No. 50-304 8.

Lect e, A. C. Thies (Duke Pewer Company) :o N. C. Moseley, USNRC, Of fice of Inspeccion and Enf orcement - Region II, January 9, 1975. AOR No. 74-8, Docket No. 50-270.

9.

Lee:ers, C. M. Stallings (Virginia Elec:ric and Power Company) to N. C. Moseley, USNRC, Office of Inspection and Enforcement -

(

Region II. January 10 and 23,1975. AOR Nos. 74-13 and 74-16, Dockee No. 50-280.

10.

Letter, C. M. Stallings (Virginia Electric and Power Company) to N. C. Moseley, USNRC, Office of Inspection and Enforcement Region II. January 23, 1975. AOR No. 74-15. Docket No. 50-280.

11.

Letter, D. P. Shannon (Georgia Power Company) to N. C. Moseley, USNRC, Office of Inspection and Enforcement - Region II, January 16, 1975. AOR Nos. 74-16, 18, 19, and 20, Docket No. 50-321.

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