ML20024B219

Current Events - Power Reactors, Oct 1974

ML20024B219
Person / Time
Site:	Three Mile Island
Issue date:	10/31/1974
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
References
TASK-*, TASK-GB GPU-2489, NUDOCS 8307070362
Download: ML20024B219 (18)

Text

y l

-.. :..=:

G DU.8 mer swem

.%.,%.m

%m.m@m$558sn?d!!m$4Mi!5MW@W@,.nW+M

• S ESBE 5 ins g.gw..mAGCE$,9.niw.., su.3x.-

s km. x c.7 sh4

? Usa t

9

-zui"J

.en am-

=.

m w

m.x_

n._wn-,.-

W y

3 CURRENT EVENTS Eif e ~~ -

v

~

3,

.U

...m h:.2 POY1ER REACTORS iid h

m s FOTENTIALLY SIGNIFICANT EVEWS SELECTID FROM REPORTS SUEMITTED TO THE UNITED STATES ATOMIC ENERGY CCMiISSION AS OF:

j-g-j

OCTOBER, 1974 iP.. -

b FAILURE OF SCDIUM THICSULFATE TANK W

k'hile filling the sodium thiosulfate tank at Unit No. 1 of the Three Mile Island Nuclear Station, the upper 40lll of the tank buckled inwar t B7#EE The tank, made of stainless steel, is 53 feet high and 84 inches in diameter. It is vented with a combination vacuum breaker / relief valve and a valved vent line. The tank is entirely hea't traced.

When the reactor was shut down, the sodium thiosulfate storage tank for m

['

the reactor building spray system had been partially filled with 8,000

'j gallons of liquid. The filling operation was discontinued af ter exhausting j

the supply of chemicals. When additions 1 chemicals were received, the required solution was mixed in the supplier's tank truck, and remained overnight in the unheated truck tank. During the night, the air tempera-ture dropped to below 40'F.

The next morning, the proper valve lineup

{

for transfer vas made; the vent valve on top of the tank vas opened, and the blank flange that covered the end of the vent pipe was loosened bec not completely removed. The tank truck pu=p was started and, vich discharge pressure at 85 psi, the tank fill valve was opened.

I= mediately af ter the fill valve was opened, the upper portion of the sodium thiosul-face tank buckled inward; the loosened blank flange on the vent valve was sucked tightly against the gasket surface.

The vacuum breaker / relief valve was in proper operating condition.

The temperature of the sodium thiosulfate solution pumped from the tank truck was 40*F; the solution in the sodium thiosulfate tank was 110*F.

The transfer line entered near the top of the tank directly into the vapor space.

The tank failure was caused by the rapid pressure reduction when the relatively cold (40*F) solution contacted and condensed the warner (110*F) vapor at the top of the partially filled sodium thiosulfate i

i i

W O'7CS8 8307070362 741031 gDRADOCK 05000289 mx.

.+

M.1. @,I$E DTkN5i53$hE[IkNNI NP58d 3 EE@MEMY M5NU

,64 E 1 6.$ Q h.iO*:#T Q M k jgd2iM E^@A[ N ?Mhh N NN55 ikddN'kCr;G% %.M?.%:@ &;$;&T47 L d % +g"l:7tTAl 2 X2

T.%~. 5 f

iY i if l

EO h SW?gSQW Y

5 T-b t.Y?:&+'d~ji!" K!MNY * - * :. 2*'

w#

I e-4_=k W"M

% r.

,s..

h 2

z N

$w tank. The nanual vent valve, although open, could not equali:e the

,y y pressure because the blank flange had not been re=oved.

In addition.

the vacuum breaker was sized to equalize pressure during functional

.c

{

draindown and was incapable of venting a rapid depressurization.

%p -g The da= aged portion of the tank was re=oved and replaced with a prefabricated new section. To prevent recurrence, circuitry design

%g Q changes were made to the upper heat tracing to permit its actuation as required during severe cold weather conditions. The lower section

~ ~ '

of heat tracing is sufficient to maintain a tank temperature of approximately 75'F.

1,2 There were no personal injuries as a result of the incident and there were no ras'iological consequences because the reactor was in a cold

^

shutdown condition. Therefore, this event did not affect the health and saf ety of the public.

3..

u$w 1

i

.-Ec ir i

+

l l

l l

l l

W G7CS9 l

...... _. ~. ~ -

. _. ~. -.

r t

.... wu emM-**===*.w

==

PIPE CRACKS Browns Ferry 2 While conducting an inspection in the drywell area at Unic 2 of the Brcwns Ferry Nuclear Power Planc, a small stream of water was observed flowing from the pipe insulation. Further investigation revealed a circumferential crack approximately 3/4-inch in length in a one-inch diameter Type 304 stainless steel instrumentation line. This line is welded to a half-coupling which, in turn, is welded to the 28-fach reactor coolant recirculation line.

As a result of this crack, the utility owner, Tennessaa Valley Authority, examined 58 similar lines, which included all vents, drains, and instru-mentation lines connected to the reactor coolant system. Dye penetrant av==4 nations were conducted ou each weld, and the supports and restraints for each line also were av==4ned.

The investigation revealed one additional circumferential crack in a 3/4-inch Type 304 stainless steel vent line located at a welded joint between the line and a 45-degree elbow fitting.

TVA has attributed the pipe failures to excessive vibration of small

" field run" piping. Therefore, additional supports or restraints were installed. There was no safety problem because lost reactor coolant water from the 1-inch instrumentation lines could have been replaced from the coolant makeup system.3 Quad-Cities 2 With Unic No. 2 of the Quad-Cities Station, at approximately 200 We and increasing power at a race of 75 We/ hour, the feedwater vibration alarm actuated (the switch is mounted on a 4-inch feedvatar bypass valve).

The operator verified that the bypass piping was vibrating and that the valve was oscillating from the 60% open position to the 100% open. One pipe hanger was broken and steam was coming from a crack near a weld between the valve and pipe.

Approximately one hour later, the instrument air line to the 4-inch feedvater bypass valve failed and the valve went to the full open position. This caused the water level to increase in the reactor vessel. The operator manually tripped the feed water pump and vessel water level was controlled with the reactor core injection coolant i

(RCIC) system.

W 07C70

~

we y-t m-m.. - -wl %

.---.wm.sm.,..-_, _.a-.

- =. #. m 1 u-

'.$W s.+~ff$$&)ci~m.m.

m..

1w.

.w.~

+-w -wmm L-:.myy.m:-myts

}**' EIf 33;4y,a. u c~irw=mu~-: e,; c4. v.,?{p[.J-~.n?:*1?y=O!'iI-

. a.=.,. y~ m$ L Y % m :.n m.n,-~~.-.: m w

.a Si N $5M5iC 5.5.i,k ?u".*Q. ic %f%b Q w,. w ~.,'ts.S t.

j:5.$?A*S

~.W ll:.?

. ;u:-m. ~. wh.;.m-;b.%;w;L..n:. ;e.T'~%

_._.1

. - - w.., - mw.,.

m.

,..~~c:1 n-t-

g/c?ggy-[- L]Q.19ya-, rgy,._jg n

.r e

e

& m.

. < ~.

'_;['T.y [!g"*7.-j; v~

-w.-w at

_..Y2[55_1.M d F- 0. Q ~-'*P.T.f~{ @ S:

p

=w

~

J'.,

4 Eqqd

s. W.'M

D=.

.N

,/ays-:, irXMi

• %1; E

A circumferential crack about 3 inches long and less than 1/32-inch in vidth was found at the bottom of the 4-inch feed water valve in the area of the veld to the pipe fitting. Magnetic particle testing identified several other cracks in velds of the bypass valve and pipe reducers.

The schedule 120 reducers were replaced by heavier schedule 160 reducers.

A second water leak acce=panied by pipe vibration also was observed in the heater bay. One heater was isolated and the pipe vibration stopped.

The water leak continued; it originated at the discharge valve of the equalizing line of one of the feedvater heaters.

This broken 3/4-inch equalizing line for the feedvater heater isolation valve did not have support to minimi:e deflection and the break apparently was caused by fatigue from vibration.

It was esti=sted that the total loss of fluid frem the system was less than 6000 gallons; all water was collected in the floor drains and

-D processed by the radvaste systen. This event did n,oc affact the health and safety of the public.4

=

M E

=

E e

~

e W

07071

a:

.-u._..

\\

hen f &1& hav 5 %;nc.:.4 % s-s e hka. T - @p W

= -- - - = = -

w n_-.

-. h...;a2

.m ;s :. q.yy. h D*..,.v:; yw. n_s u.ig.1. y.nrg..--:w y g w-g-

j

_p

_m y g7.

s m

._.m.-

2_.

s s

-5~

WELD DEFECTS s

Oconee 3 With the Oconee Nuclear Station, Unit 3, in a hot shutdown condition, a leak test of the Reactor Coolant System at a pressure of 2,285 psig revealed a weld failure. The leak was located on a 1-1/2" socket veld joint through a small pinhole on the surface of the veld.

The apparent cause of weld failure was lack of fusion (cold lap) between the veld filler material and the elbow fitting. The lack'of fusion was near the top of the veld and was covered with a weld pool. Prolcnged pressure on the small surface opened a pinhole size leak. The system previously had passed a dye penetrant test and a successful hydrostatic test of 4,575 psig.

s The defective portion of the veld was ground approximately one inch on either side of the pinhole, and the veld was repaired.

Six welds, similar to the defective weld subsequently were dye penetrant inspected; five of these velds were =ade by the velder responsible for the defective veld, and were inspected by the same inspector. None were defective.

This event occurred prior to initial criticality. Therefore, leakage did not result in any contamination, and the health and safety of the public was not affected.5 e

l l

l W

07072

=

. &.3.gy?%gM'R.agg:w*s A-Wr5Pd5&O$3$5+55$-M WuM.T M5Li2

&W Y. ~.

$**V

3 $ m w-N,~,O h. ~W 5.. J = Q.. WhN$5 y;a?i. 3555%C$ NO5WkY&k$$5$.$h?Y$4

,.=--=w~----

- ~ ~

w EQ~j =:?MW4WMEn%%%mWW%We??ww+% Eiwl. h

?

d m

p q

W4Fj

.k.. hb

,e.m

%s M

FUU. GRAPPLE MAL 7UNC'" ION

~

While transferring a spent fuel bundle from a fuel preparation =achine to a fuel rack in the spent fuel pool at the Millstone Nuclear Power Station, Unit 1, the bundle fell from the =ain fuel grapple to the floor of the pool. It ca=e to rest with its lower tie place approximately five feet from a fuel rack, and was observed to be bowed over its entire length.

7 Personnel operating the fuel grapple, at the ti=e of the event, noted bubbles escaping from the fuel bundle. They evacuated ;he refueling area and the entire reactor building. Air sa:ples throughout the building did not detect the presence of radioactive iodine. The max 1=um level sampled was 6 x 108 pC1/cc of cobalt-60 en the refueling floor, and this was attributed to feedwater sparger work in progress and not S-from the dropped fuel bundle.

Investigation of the latching action of the =ain fuel bundle grapple revealed it was possible to jas a fuel bundle into the grapple so that the "J" hook was not fully closed. The bundle could be sucessfully lif ted, but subsequent =otion of the bundle (jogging of the platform while traversing) would permit the bundle to fall frem the grapple.

This =alfunction of the grapple could occur culy when the bundle bale was partially inserted into the grapple at an angle of approxi=ately 20 degrees.

No visual damage was discovered in the spent fuel pool and no fuel pool 8

leakage was observed. Continued air sa=ples indicated no release of activity; the health and safety of plant personnel and the public was i

not jeopardized by this event. Analysis of activity of the fuel pool water remained constant and showed no signs of fuel pin failure.

General Electric was requested to make a design change for the grapple which would give a positive indication to the operator of complete grapple closure.6 in W

07C73 i

g M

-ev-

a

.hW QM"%

.E T W* I e,pg. g w - ds.- < m gh.:7 - SYW$N555kh.w-Ng, ENW

%dw CWesq e

pj&ip$C@$59$$m,$&Nb.&$YW:

?5hi fR&

y$.

Ec T

y%W

?

2 M

Ebh85%gSch 1g%.:q-@QkK;;$4}%!$gnypn?rms. 55RK d

vn=2g ;

5n.:

r 0W i

i

! DISCHARGED EATIF3.Y Unit No. 2 of the Quad-Cities Station was reducing power from 100.We because of a failure of the air supply to the feedwater low flow regulating valve. After an automatic reactor shutdown occurred, an attempt was made to start the High Pressure Coolant Injection (HPCI) System, but the associated valves could not be operated. Investigation to determine the cause of inoperability of the HPCI valves led to the discovery that the 250V de battery had discharged to 70 volts.

When the reactor had been shut down the previous day, the Unit No. 2 breaker in the line to the bactery charger had tripped, and a correct alarm indication of the trip was received at the control room. A control room operator visually checked the breaker, and incorrectly determined that the breaker was closed; the breaker trip alarm was thus considered faulty. Equipment which should have been secured when the reactor was shut down oui.the previous day, was lef t operating and discharged the battery. This condition went unnoticed by maintenance personnel and operators.

The safety implications of this event were minimized by the fact tha'e division of the loads on the 250V fuses for the two-unit station assured availability of emergency core cooling capacity for both units when required.

The reactor was shutdown in a normal manner. The safety implications of this event were minimi:ed by the Unit No. 1 safety systams; all Unit No.1 emergency core cooling systems (ECCS) were operable.7.

O e

t e

W 07C74

.w-

.--o e--.

~ - * - ~ ~

N Y

N hN

'.W.x,:k

' Y k

b T f.4 l ;g;~5 %p;h - n ;~~.E.

ip

&n$-~ 2 -

2%- qQ&Ln-

=~

- %G.h:f,.:XY,5 R c4.:: f E( I+ Y,Z, W,,.U E f [ % g % W u?- p: > q '.n w i - 2' R =N__. N bT.ff$.N N N_M_ E. N f O $_ E'h~/ S E i-9 % s.~ T, S, b~%3p.iti.;t t'%pe:

m +: us 2

~LK:fgme

}'NNkd-YTETNiNYWI~IN;%'if I.

Ye=

-- m

.X,,MC' M.

L -- r w a7"M.. M_

W#4db**

>.<Mm y Q_.

. E C'iPE MQJ SPRAY VALVE CIRCUIT FAILL2E C-4w EbN During quarterly surveillance of the reactor building cooling system at the Three Mile Island Nuclear Station, Unit 1, it was found that a spray

~

valve did not open in response to a signal. The failure was caused by an inproperly installed wire in the logic panel associated with an engineered safeguards actuation system in conjunction with a spurious ground on one of the batteries. This resulted in a current path to a

-7 ground which caused the control fuse associated with the spray valve to open. The licensee concluded that the inproperly installed wire was the

'E Mg result of a nanufacturing error.

This failure did not represent a threat to the health and safety of the

_ J public because a redundant system was available at the ti=e.8 E

t E

1 e

I b

=

W 07G75

- _ WM

. ~

........ _. ~

h,- %yg ~--

7 w;s,;sxw yNR%gpp g-h 7-md--

-M

.53T/4r 4:p%,NO

' ' t j

my r

.Q%1&W d~ -

M.& Mb ?S MSN W M wok -? h*)?~ [ C % Nxr$ M ikiFI5i$ $ h.w e 2

4 9-l RADI0 ACTIVE CAS RELTAiES i

Oconee 1 While Oconee Nuclear Station, Unit I was at 100lll power, a small quantity of radioactive gaseous vaste was released during routine draining of i

accumulated moisture from a vaste gas decay tank. The operating procedures posted on the tank required closing the drain valve as soon as water cleared the sight glass. The operator failed to notice that i

che sight glass was free of liquid, and the tank vented to the high activity waste tank. Radiation alarms were received in :he Spent Fuel Building and in several locations throughout the Auxiliarf Building.

f The control operator noticed that the pressure in the waste gas decay tank decreased from 73 to 13 psig and that the drain valves were closed.

The volume of gas released to the atmosphere was calculated to be 4,300 cubic feet, but the amount of iodine-131 released was insignificant.

Personnel exposure was less than permissible limits, and it was concluded

].

that this occurrence did not affect the health and safety of. the public.9 i

Maine Yankee During fuel inspection at the Maine Yankee Atomic Per.rer Plant, service personnel requested dominera11 zed water be made available in the contain-i ment for cleaning operations. They were advised that a valve lineup would have to be made by operations personnel before water would be available.

Upon entry to containment, plant services personnel opened the designated faucet, but there was no water, so they closed the valve and toured the j

building to review the work to be performed.

Later they returned to the sink, and once again opened the designated faucet, varying it between a wide open and barely open position. There still was no water so they left the faucet open and, upon calling the main control room, were advised that operations personnel were coming to align the system. They then closed the faucet.

Shortly thereafter, containment monitors detected an increase in radiation levels, causing the containment purge system to close down automatically.

As=soon as alarms were received in the control rnemi. a a n aa= = 1 * *a - a

---l 4

i W

07076-l 1

&me.s

'8

t. %.~. g-~~? m f.,22 7r2=.w.d3 c. w a m x.y =.z,z. w y w~ M m. y y max::%w W.-e2dgyg,.

A N

..rk

t =

e

nw c -a- :g -n

':.3 s

yy Q L%w,KDn;ri&m t.Qsw.~,.x:: :Wi=my.pg-gggQ:

..A 6%.lp%fq;~ '3?y%n= %,;JDgp h y:' r:-g = 3 3 ~. y }QG y'&~g:

s. va:OU.= :*J."

p:

W

{=

53 Q [g Q q.h, Q. p 5 --@I N

r6s r_

~

_ 10

~

r..

!5

-w47 While plant service personnel were in the contain=ent, the =ain control p,M room operators observed a slowly decreasing pressure frem 5 psig to 2.5 g

psig, and were atte=pting to deter =ine the source of the pressure decrease.

B,bythetechnicalspecificatio The increase in radiation level was caused by gas escaping frcs the collecting tank back through the primary water header while the faucet was opened during the premature nanipulation of this faucet. P.axi=um esiculated release race was 1.33 higher than the release rate permitted g

Personnel film badges indicated that no overexposures were received.

Also, the health or safety of the public was not affected.

A check valve and a remotely operated isolation valve were installed between the manual isolation valve and the demineralized water header to preclude inadvertent gas backflow frem the collecting tank.10 tr Indian Point 1 At the Indian Point Unit 1 Nuclear Power Plant a rupture disc in the makeup coolant system failed, causing the excess pri=ary coolant makeup water being returned to the clean water storage tanks to be diverted to a

che vaste collection system through an open su=p.

This resulted in the release of entrapped radioactive gas that was detected subsequently by p

local sonitors.

The alternate makeup coolant system was placed in However, the rupture disc on this line also failed although service.

M, the design rupture pressure was never reached.

4 The release consisted of approWeitely 5. 7 curies of gaseous radioactivity, the major constituents of which were xenon-133 and -135.

Total radio-g active iodine released was approx 1=ately 86.5 microcuries.

t There was no health we safety significance associated.rith this event since the release rates were less than 1,1 of the technical specifications limit.11 9

W 07C77 ww.'.mp--

em-4

. ~ - - -.. -

...... ~.

"4 8

2.=~&~ ;..,

m.-

.s1

.Kr._,..--,_.,e---

.r;_,

-2W

-w.

-->J $~2 W -w w w ~* &w p=w 'm-x pw4 1

_w A"n:h.*> a. -

, p d,:-~h;'C~1. Ma s. M s+v Q wa 7en M. w.r.~5 bT Q$1Wg5 Qq# spgo*? 'M f.'f?@F225@%g..~Es$rCif,Q.~y.c,.u&sSca

yue n

s -

e.-w w

~-a

+

~

fy'BSNM wan &

WW

~f STni2 sG25 9 2n U36

= % ;c y g g.

- Tre%*.!g;;;Th ans2'$b t%

11 -

FAULTY ARRIST0h3 IN CORE S? RAY SYSTEi During inspection of two snubbers at the Oyster Creek Nuclear Power Plant, Unit 1, it was found that neither contained oil in their accunu-lators. Futharnere, the piston rod in one unit was discovered frozen in position; the piston in the other unit would not operate snocchly. Both of these two Bergen Paterson Hydraulic Shock Arrestors were located on j

the South Core Spray Systen. The two snubbers were replaced with units that contained rebuilt seals.

Since the loss of two shock arrestors represented a partial loss of the seisnic restraining ability of a section of the core spray systen, there could have been a potential safety problem if an earthquake had occurred.12 i

1 I

W 07078 l

j i:

t

>.q-dbar@/.pdW59+]"c%--glowpe$59TG'4T I5 v.#19e.gy,@, M~ I?q$

d*% % &$

$74E F'iR$EitW7':iM W N *d.2 T C:[;gyys:She.5% S&1.s95fG.pv ?M'Q5.;j % ~M

~b E

9

.>pW2-15.$4qib =y:gMj

$ $ g $ h.. E N G 3 m f g g,5 E f q g &-

sW-f gW '$.3..: $. u: rY$?..,:Nj'fihY 5 G $

hff5WW--

WEWE?U1 W

g

n..

. a. ~

?:O" T,:

i n B.-

WA,:,.

~

A' p-O HEAT TRACE CIRCUIT FAILURE L'ich the Indian Point Unit 2 at 100:: power, periodic surveillance tests of the heat trace circuits on the safety injection system containing boren solution detected a discontinuity on the primary winding. The heat trace circuit consisted of two separate strip heaters. I=nediately upon discovery of the discontinuity, the alternate vinding was placed in service.

The rubber insulated wire failed.c a point located beneath the pipe insulatio.. Apparently the pipe insulation had contributed to tempera-tures in acess of the design rating of the wire.

The low tenperature alars had not annunciated; the safety injection pun:o tested satisfactorily on recirculation, indicating the boren solution contained in the suction elbow of the punp had not cooled to solidifica-tion. The ti= elf discovery of failure prevented any adverse effects fron this event.

The rubber insulated heat circuit had been installed for about three nonths. Because the potential for a repeatable failure was high, the defective vire was replaced with a wire more adequately suited for this

_a application.13 t

-i l

W 07G79 l

l

...~.._.___u..w..-._:.-.m.

r g " W Q.ii*."a;+,,.ce M -C e rw v

^ws GW49-

%-C "

@N TA-sap.t W g'9

=&2*,+'.f4

~

~ '"

'#~

• = Al%'#SE98#dS W i-W r'_

EO~phrlVpW;?:~2 mg -- %d~%n.

m,.dka..

~

M*;fe.d'.2%e.cip--mS}.5"dNM%RM.R%,..,MW.'Qygg g,, ___ y %rW_1T?WI -

Q.Nn, :nw Ce W~

a 1

CRACK INDICATIONS IN SEISMIC RESTRAINT COMPCNINTS The AEC was infor=ed by Virginia Electric and Power Cc=pany that there was binding in a swivel end coupling associated with one of the two steam generator seismic restraint asse=blies for Surry Nuclear Power Plant Unit 2.

Examination revealed a crack in the end coupling.

There was no evidence of lubrication in the coupling. This apparently resulted in excessive friction and binding. Inspection of other restraints using ultrasonic techniques indicated flaws in two other swivel end couplings on the steam generators.

The steam generator seismic restraints are designed to per=ic a safe amount of hori: ental, vertical and angular =otion during seismic events while permitting nor=al operational motion. Since the entire weight of the steam generators is supported by vertical support members only, the restraints should not have oeen subje:ted to vertical leadings. Inasmuch as there are some questions regarding the safety of continued operation under these conditions, Virginia Electric and Power has shbt down the plant until the problem is corrected. Corrective action may involve redesign of the restraint. system.14 1

e W

07080

_e -

~

a m..

.......~ - -..

......c.~

i

Nh=g;fM2M---.M?hNM[YiM'YNNW bYb mud f

wmn -~

%a GkM. ~i9BMW=~

ap:::-D. :r= ua=,pkey.,u;i& r.~uemw

$ g.,J,,$ E h. _f G p.g. a.vs g N 5 Y f $ 1 D 3 M E M d b l $ b

.~ - w z

-M1 f

grre-- -

y :- % s m +~,S p m.

e sc

%..M.._k'.$_ 'y C.C'S?ff*=

- - ^ " - -

%~

w w -+

a.

m L*49

.M % C

" A W.:&

F.2sm

.W.c.ar

~

o D.

• 4 r.:.

W e W

(;-

MALFUNCTION OF SArr..I RELIEF VALVE CAUSES DEPRESSURIZATION E *kt The AEC was informed by Philadelphia Electric Cc=pany that a safety c

i relief valve at the Peach Bottom Aconic Power Station Unit 2 lifted spuriously causing depressurization of the reactor coolant systes.

The valve, manufactured by the Target Rock Company lifted during routine operation even though no overpressure condition occurred. Atte= pts to the valve were unsuccessful and the reactor was shut down c:anually.

seat t.' hen reactor pressure decreased to about 400 psig, the valve reseated.

The condensate and feedwater pu=ps supplied sufficient vater to maintain reactor water level during the occurrence.

No radioactivity was released to the environ =ent; there was'no danger to the health and safety of the public. During shutdown, the valve was dissembled, inspected, repaired and successfully tested.15 fl I_

i i

s J

]

M S

W G70S1

.h M

w na' w..

.e

-~

• o kNS

(

h fgh'55$h

~

~~&

?

h-[hdQWWM.mm-m=@.v_xc?nNEsid.w,'--_vyn.__J.,g' _

r e

- - mmw_w w - N. is s

I s

~_mg

~ ~s-fit FAII.URE OF MAIN STOP VM.VI While preparing to return the Indian Point Station Unit 1 to power, the main steam stop valves for three steam generators were tested by re=ote control from the control room. Two of the valves operated nor= ally.

However, one valve did not respond to the test signal. The reactor was shut down and the malfunction was investigated. It was found later that the motor winding insulation associated with the valve operator drive had wurned, resulting in a short and causing the power supply fuses to open. Inspection of the defective =otor revealed that the current drain was caused by application of a motor operator brake over an extended period o# ti=e as a result of failure of the solenoid that operates the brake.

h-brake is normally applied only when the occor is de-energited to prevat coasting of the =otor operator at either end of the valve travel.

There was no safety problem associated with this event because the valve failed in the closed or safe position. Even if the valve had failed in the partially open or fully open position, pri=ag loop isolation would have been available to protect the reactor against an uncontrolled blevdown resulting from a steam line break downstream of the valve.

Corrective action involved the installation of a new operator =otor and brake solenoid.16 Theodore C. Cintula John J. Ritto Office of Operations Evaluation U.S. Atomic Energy Co= mission I

P e

w 07082

l5".}.T_T.*."*:

L^

'l r. ~ ~ r ~-~ ~ ~ -"~~" ~ ~ ~ - ~~~~~~~~"~

~~~" ~~~ ~ '

~

REFERENCES 1.

Letter, J. C. Miller (Metropolitan Edison Company) to D. F. Knuth, USAEC, Directorate of Regulatory Operations, April 11, 1974, Docket No. 50-289.

2.

Letter, R. C. Arnold (Metropolitan Edison Ccmpany) to A. Giambusso, USAEC, Directorate of Licensing, June 17, 1974, Docket No. 50-289.

3.

Directorate of Regulatory Operation Notification of an Incident or Occurrence, dated 10/18/74.

4.

RO Inspection Report No. 050-265-74-08, September 27, 1974.

5.

Letter, A. C. Thies (Duke Power Company) to N. C. Moseley, USAEC, Directorate of Regulatory Operations, Region II, September 17, 1974, Docket No. 50-278.

6. " Letter, W. G. Council (Millstone Point Cc=pa$y) to A. Giambusso, USAEC, Directorate of Regulatory Operations, Docket No. 50-245.

7.

Letter, N. J. Kalivianakis (Co=monwealth Edison Company) to J. F.

O' Leary, USAEC, Directorate of Licensing, September 10, 1974, Docket No. 50-265.

8.

Letter, R. C. Arnold (Metropolitan Edison Co.) to A. Giambusso, USAEC, Directorate of Licensing, September 20, 1974, Docket No. 50-289.

9.

Letter, A. C. Thies (Duke Power Company) to N. C. Moseley, USAEC, Directorate of Regulatory Operations, Region II, August 30, 1974, Docket No. 50-269.

10.

Letter, D. E. Moody (Maine Yankee Atomic Power Ccmpany) to J. P.

O'Reilly, USAEC, Directorate of Regulatory Operations, Region I, July 1, 1974 Docket No. 50-309.

l 11.

Letter, W. J. Cahill (Consolidated Edison Company) to J. P. O'Reilly.

USAEC, Directorate of Regulatory Operaticus, Region I, July 29, 1974, Docket No. 50-3.

12. Letter, J. T. Carrol (Oyster Creek, Power and Light Ccmpany) to J. P. O'Reilly, USAEC, Directorate of Regulatory Jperations, Region I, September 16, 1974, Docket No. 50-219.

W d'7C83

,4 M

~

• se.w :wr#

.puW%t&WM" ~ ~ ~

~ T8745[4i 6

Y PA bb w -

~p

{

1. ~

~,

G M,. g M -t._., m w-:-m m e $_ N,1 W A-m m s. T~~, w. ~ p1__m,-w e m e

= ~

2&

s r?

GhMe,l W:Gr 2 --

13.

Letter, W. Stein (Consolidated Edison Coupany) to E. G. Case USAEC, Directorate of Licensing, August 30, 1974, Docket No. 50-247.

14.

Directorate of Regulatory Operations Notificatien of an Incident or Occurrence, dated 10/25/74.

15.

Directorate of Regulatory Operations Notification of an Incident or occurrence, dated 10/18/74.

M

16. Letter, W.

J. Cahill (Consolidated Edison Cc=pany) to J. P. O'Reilly, USAEC, Directorate of Regulatory Operations, Region I, August 22, 1974, Docket No. 50-3.

3 r

1

=

1 W

07084 m....,3,

~~~~'~~"'~' ' ' '

~~~ ~'~ ~~~~

~

9

~

M

_ _ _-r-i userto starts h;WM ATOMIC ENERGY CCMMISSION CC W AS.*eNG TO'*. O-C. 20S49 I

1 s

p g g-g OFFICIA4. auseNES3 is.s. aro==s saaae' ce=='esion

- ' " ~

Praeaa.fy r.3m me,varg ygg, 93,,

ty s, yang, L

-J c.=

.d^

k a

fl0R.'.!AN D I CX i flS0ft ROE

% GPO STARTUP l\\l ga BOX 480 MIDDLETOWN PA. 17057 t

W 07C85

-. - ~.

-