ML20024B200

Current Events - Power Reactors,Aug 1974

ML20024B200
Person / Time
Site:	Crane
Issue date:	08/31/1974
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
References
TASK-*, TASK-GB GPU-2479, NUDOCS 8307070309
Download: ML20024B200 (24)

Text

.

GN MM J-

h. dwacrsect=mdsmsg:EMEEENs?WsE

$$INNNNN6N!$NbNE w$#r he #

5 M-w pf hkh 2-T vCURRENT EVENTS F)f^ J 5W Mr b'f $

1' bg N

7 3 c c / h g $ 9 A c / q A O g..

POWER REACTORSp %

se 1s : ppu~-L z

.............. 2 m o POTCITIALLY SIGNIFICANT EVOITS SELECTED FRCM REPORTS SU3 MINED TO THE UNITED STATES ATOMIC ENERGY COMMISSION AS OF:

E, i

AUGUST, 1974 I

n LOSS OF PLANT STACK MONITORING SYSTEM With the Vermont Yankee Nuclear Power Station operating at 80 power,

?

direct lightning strikes to the top of the plant ventilation stack were observed during a severe electrical storm. Both plant stack sonitoring systems became inoperable. The Technical Specifications require at least one of the two plant stack monitoring systems to be operable at all times.

Approximately 30 seconds later, another lightning strike resulted in a generator / turbine trip that caused the reactor to shut down automatically.

In addition to the loss of the stad. sonitoring systems, the peri =eter j

fence monitor, the stack base fan, the process computer, miscellaneous j

annunciating and indicating circuits, four auxilary equipment circuits, j

and the site meteorological tower instrumentation were disabled.

h, While the stack gas monitors were not working, grab gas samples were g

$'e, taken at the stack on a periodic basis. The =axi=um release rate was j

11,119 uCi/sec; a release rate of less than 10% of the Technical f

Specifications.1 h

j At least two direct lightning strikes hit the top of the ventilation stack last year.

In each case, the lightning caused a detonation of the off gas system; the back pressure fractured the rupture disc of the air

{

J ej ect or.

After the first strike, lightning rods were extended approxi-mately 15 feet above the stack to increase the probability of intercepting l

electrical discharges.2 l

1 8307070309 740831

~

DRADOCK05000g9 W

06cS4 1

1

w..aih- ~ x -

5Q*g

~

c.

h?~"5 W

W i

Wb i

aw'-w l 5.y

\\,

-m i

5 ai.w 1

a TUBE DEGRADATION IN STEAM GENERATORS During the second refueling outage at Unit No. 2 of the H. 3. Robinson Nuclear Pcwer Plant, three steam generators were inspected for tuce degradation.

"A" steam generator. Prior to the outage, there was a leak of 0.05 gpm in the A leaking tube was found in the inlet side of the generator in the U-bend approximately two inches above the top tube support.

Another tube in the same general location had an 38" through wall penetration Both tubes were plugged.

-r

-N.'~

a There were 274 cubes with wall thickness reduced 20% to greater than 50%.

[

The 35 tubes with greater than 50% wall thinning were plugged.

L This was increase over the number of tubes with wall thinning in p

a significant j

1973; most tube degradations then were located in the U-bend and tube y

sheet location.

Q t'

Carolina Power & I Lght Co. of ficials believe the lon's periods of opera-3 4

tion in 1973 with a March-Halstead sodium-co phosphate ratio of below 2.18 led to the problen.

Tube thinning at this facility has a minimal potential safety impact because tests have shown that a wall thickness of 25% (75% penetration) s i

in the region of the tube sheet will withstand the loading conditions imposed by a loss of coolant accident. 3 r

At Unic No. 2 of the Surry Power Station, there were indications of a minor primary-to-secondary leak in the "B" steam generator.

Mille the reactor was shut down for an unrelated problem, a one gpm primary-to-2AJ secondary leak was found in one of the tubes in the hot g

leg side of the steam generator.

This tube and five others with some degradation were plugged.

The unit is able to operate safely with minor steam generator leakage, and there was no violation of the Technical Specifications.

There was no adverse affect to the health or safety of the general pub lic.

Control of the problem is expected through use of a new sodium-co-phosphate ratio.4 I

i 3

?

W C6FS5 t

ON wmwn.we.~

• W^

~ ~

_a

". r.". _...*

T

~ ': '

~ ~ ~ '

.9 n +

,~~ ~ = ~ m - u

- ~ N. O 'e @ l n $ YhM2yth--

bM M

'h1

- N$ M E h -w k.._%gxvw- @= M'. e w - -

- ~,

--w -

7

'9

~

3-RADIATION EXPOSURE TO PLANT PERSONNEL After plugging the steam generator tubes at Unit No. 2 of the H. 3.

Robinson Nuclear Power Plant, a supervisory contractor employee, in his desire to complete the task, decided to aid the work directly by vacuum cleaning the channel heads of the steam generator cuoes. After properly dressing to work in the steam generator environment, but before he put on a respirator, the man attempted to start the vacuum cleaner.

%* hen it did not operate, he opened it to fix it, but was unsuccessful. He then donned the respiratory equipment and entered the steam generator for approx 1=ately five minutes. After exiting the containment vessel, he found he was contaminated. Extensive showering appeared to re=ove the contamination, and he successfully cleared hi=self through the control point. Shortly thereafter, the employee returned to the control point and found he was still contaminated. After further efforts to lower the radiation level were unsuccessful, the suspected area of contamination was enclosed in gauze and tape and he was sent home.

The next morning, use of a skin cleansing agent did not appreciably reduce the contamination level. This, with the now generalized radi-ation level throughout the abdominal cavity and circu= stances of the exposure, led to the suspicion that the contamination was probably internal rather than external. Later analysis of urine and fecal sa=ples proved conclusively that significant levels of internal radioactive contamination were present. A whole body count indicated that the man had lung burdens of 1.4 uCi of cobalt-60 and 1.6 uC1 of cobalt-58, meaning he had inhaled 13 uCi of cobalt-60 and 15.2 uCi of cobalt-58.

Lung burdens of these magnitudes will result in a lung exposure of 16.6 rem in one year and a lifetime exposure of 17.4 rem.

Although well trained and experienced with nuclear systems, the employee failed to consider the radiological hazards of opening a vacuum cleaner identified as being contaminated. The potential for this type of accident occurring from radioactive dusts was discussed with all plant employees and locks have,been placed on all vacuum cleaners to prevent repetition.5 W

066S6 s

w m

e

• O..

w-v he.

A

_ < > - =

p.-.

While Unit No. 3 of the Turkey Point Nuclear Station was at 75 power

=

and flux mapping of the core was in progress, one of the flux capping detectors ceuld not be =oved by the drive motor. Two people with a valid work permit entered the centaic=ent to repair it.

The problem was caused by a defective electr&-dynamic brake. The only solution was to move the detector by rotating the cable reel by hand.

Both men rotated the reel initially, and then one man rotated the reel slculy while the other man moved away approx 1=stely 30 feet to co=municate with the control recm. The drive motor again was energized but the detector still would not move.

The cable reel was then rotated by hand at a high rate of speed; the detector ca=e out of the tubing and fell on the floor beneath the reel.

The man at the cable reel i==ediately j oined the other man at the cc==unications station and they lef t the contain=ent building. Both pocket dosimeters (0-200 =R) were off scale; the indivi-duals involved assumed the dosimeters were faulty.

M The following day, thermoluminescent dosimeters (TLDs) for both employees IU$EEUI were processed and the results revealed that although no overexposure had occurred, the TLD exposure of the =an at the cable reel was 2730

= Rem y (whole body) *and 660 = Rem B (skin dose only); the individual at the ec==unication point had exposures of 910 = Rem y (whole body) and 180

= Rem 8 (skin dose only).

The exposures were caused by failure to ce= ply with the Radiation Work y

Permit, the Operating Procedure for repair of the flux detector drive j j mechanism, and the Radiacion Protection Manual.

Specifically, survey j

i instruments were not used in accordance with the work permit, a health

[

physics survey was not requested when the detector was fully withdrawn, and the pocket dosimeter was,not read frequently as required by the

,C Radiation Protection Manual.'

^

s 00 5 j I

4q W

06ES7 e6W eM*

= _.,. _ _ _..

.o.

~ hE1-

[Ed y

bN~

Nh

-~

. UNPLANNED REL3SES OF RADI0 ACTIVE LIQUID While the Duane Arnold Nuclear Plant -as at 30% power, an unplanned release occurred in unich radioactive liquid waste was discharged to the Cedar River.

The incident occurred during a backflushing of the con-densate desineralizers and was suspected to have been caused by the f ailure of a valve in the system to close completely. The valve mal-function caused the condensate backwash receiving tank to overflow, and the plant waste su=p was unable to direct the entire flow to the rad-waste system. The collection of water in the vaste su=p was pumped to the storm sewer and was eventually discharged to the river.

The reactor was shut down when the backflush system could,not be isolated.

It was estimated that a maxi =us of 1650 gallons of water was released to the river. The activity concentrations of backflush water in the building and the point of discharge follcwing sewer dilution were 0.0001 aci/cc and 0.00001 uCi/ce, respectively. The activity concentration of the release to the river was within acceptable limits as set forth by 10 CFR 20.

The Duane Arnold Energy Center Operations CWtree concluded that the release did not present a hazard to the health and safety of the public.7 3

3 The contents of a vaste distillate storage tank at Unit No. 1 of the Indian Point Nuclear Power Station were released to the discharge canal before final approval for the release was obtained.

Involved was approxi-mately 290 cubic feet of liquid with a specific activity of 0.00202 uCi/ce.

The total amount of activity released was 0.017 Curies; the major constituents were cesium-137 and 134, cobalt-60, xenon-133 and iodine-131.

This release of radioactive waste was well within the Technical Specification limit for the site and this incident was not considered to be safety significant.

The cause of the release was failure of an operator to adhere to pro-cedures.

In addition to reprimanding the individual responsible for the occurrence, all cogni: ant personnel were reinstructed in the need for strict procedural adherence.9 W

06888

. e-o a w www.asa w

Fact: rsnam -*m.= =~~.w..-- Or= s@c ~.+L. m.?-c~;v;;oa av-.w'--wme grem<w; =-

EG'-- w-m:w::

r.w-4 %e

..r

,-2 ww,, L.a::a; z: - :sien:::.a W.:w

. cwpW-ar-r NT

<w

~ew..-; w-m._v,.,s v.

i

=

y.s

.w

.e w

s

. :: w ce. ;.s.

m

• w -~

m

.-gW -4. c,,- C f,.,;

a W

4-"~r W N'$. m \\.%,'-

-c

.w.-.-

-c~~

-~&s. "_V~.n %q *- %-a.r

.J~'

r* s., y. ~.-

r

.,. R Qq b,*_y%g' gr,-

--- r _v

~.. n. w w w g m e.- w g y

. s.4 * * ** DiGNM

. -s s w -

pe 5*

nwg,. sa:,a.w--.a.w,.w.-~%. r..m.. sw%..,. ?s.ON.e.WYM, WE2mR. w.sm, !v**u**~M'

- 2EM.'

2#:SkOfh

+e 5 k

'?--**~

~~~ hm-

-e e.w. 2 n y ~ n -rtn y.-e n yc

-p.

-s

.,,es m w.,. m,. w 4,m w

-~a n.u m s-~- ny w e m. ~~s g

4

- -- n--.: s s.gy.n.s,.y_. ; ym.;;*-m%...,,

s5 g%am -

,,,;r::WM'Sy.

.,.r-

-e. W-Wh._ _

ng'~

-ge v.m.

7-h, -

l=2 :--

l~L~ASES TO AUXILIAZ - -.,,,,,,

aimum

.4

,n.

-,. n,.iuc, ear Genera-e --

pe a

- - --- a ru started to divert reE-

~ D.a'n letdown water r,

- -- --.,,- 3, ra p.:e,-

to the e-

<.. building drain a hour later, the aux

<a

.4.u4 ta -

-c

.ng vent system

-, --, ors alarmed; 20 - -.. as acer the charging h

m__-.

s p-

- - - a,

,- e,n. one hour late s

-.a. e tunnel =enitor r

a _,. - -

_ ci-

,- c ease in the c-2<

ent purge exhaust gas low to the auxile a

a-.

4-n u. m ---

Wmt hain

'l ca: ->

- :: zaseous ac eviey bega: to decline shor 1.I I

amJ _-

-- had functioned <-

r

%=

- e - t... the The prope-r ca,,:re, gaseous release

..as

= - = - - - -. -c f

q error.

u csm

.-..- ank rather than

.w.e aux 4 liar / building n

a or diverting g _-

_e-.,

,.c

-, : -.: ----r m

-m M

-- s een f ollowed -

--. _- -a. drain was not sea a, and the gas escaped

...e hat.h cover of the y_

~-.

.,..,e au:r.

a 8~^'M-

na -  ? "-* r1 a =as 12,900 uC1/sec or 41.3,~. oc the Technical

- - e-.

s a e cubicle monitor a'a -...

- - =

--m

.".i,',,a g_. [-],

,-.4 *

'nstantanecus gaseous activity.

r i--

. ate was '1082 '.Ci/sec-The release rate of

-- --a t

- - ---. aec, a minor frac-< n e

-he instantaneous

-,. _.:=

g g

.a s,,c.4vity was background.

s

=i

--:a 4:

- F "ed was 76. 7 C4

-m of xenon-133 and 1.46 uCi of

- - =- -a ease rate was lov and the release was of 4 -..

=. -. -- c - =. a.

af e ty oc. the public was not affected. M

- - ~ -..

i.

3 I

I ed the Oconee ' luc 1 ear Power Pl

'l

=

q a ---- - -- s t,ined a sa=p le

'a ant vere operating

. a-3 r:= ---,-. - on4 ors in the Arom the letdown storage s

.-c

.-u.

it 2: ~ ~ ~ ' ~

a ed tec tr=al radiation leve us a e

g was e e s:.+ 2:

1~=3 =

a '~:it 1 waste disposal sa=pl

.-------a.

e hood were Tu aars;- a radiation leve1s returned to normal and

- se<

-- or entry.

+ '-

C'- ---- I st ling b e c aus e g as

'e

.4 t a the sampling system,

J -. u --- - a,aseous was te disposal sys tem, s

leaked C:

=a ute gr.s exhauster iuto the Auxilia 2

r.

ry Building.

.a,,..., _ a : r 3.

3,. of the magg,um release race

., m -

. - :- - a

a.,, acions.

The max.4 mum exposure to one W

06889

_m

-w-

.-e 7+

-.-..w--

y g--e--

.--.-.e y

=

Nbkf Jh k

~b C N N kNkNNYiE b

L tig!iEGjgl2!$1$E@e%3IY3Mh $$5M E4 W 1G M GFS E4M6 aweieA M=sm%$stemessweaMwbumsw individual was 25 mrem as =easured frem the technician's dosimeter.

This level of exposure is less than occupational limits. The event did not represent a problem of safety si the health and safety of the public was concerned.1gnificance as f ar at e

D W

CSSS9 I

4 m wn,

+

J

$5kfb^i5hb?4.' #wyt-c.oSM-kW>8W-~ *5@cf S

5 Y

Mi LYl%ggg.r$hymne,,ckl$ n

-r n?

t.

~-~~:. g;. :n>ww=tw: ww&OML%.gm%$4tiWE::.;:,%.

TL5 M t

?

,Q WSM~5%~,LLi=v=?s;p 3 y

~

~

?

Q s

&MMm M

3-121 y

FAII.URE OF DIESEL GENERATOR TO START a

&N On two occasions while attempting to test the No. 11 emergency Diesel h

Generator Engine from its local operating station at Unit No. 1 of the Calvert Cliffs Nuclear Power Plant, the diesel would not start. An eM attempt to scare it from the control room was successful.

A solenoid valve energized from the local station in the air starting system had failed to open. This prevented compressed air from turning the engine. The defective solenoid valve was dismantled; it contained rust particles.

This malfunction was of minor safety significance because the tests were performed before the plant had reached criticality. The No. 12 e.:ergency Diesel Generator Engine, a redundant system, was tested from the local station and the centrol room after the No. 11 engine failed to start M

anc, in each case, operated satisfactorily.

The Baltimore Gas & Electric Co. replaced the existi-ng solenoid system J

with a new unit, including special air filters upstrea= of each solenoid.12 a

O UJ I M*

6 d & r##

F G

~

i;

{w

+

e 1

.4~.4 ap 454'-

i e

p,,a a-e-s

, nn a-

,fn.s, Ai<

f9/ Tem k5%He/ uj' S fes?>n d S o l e J L 'u e s.

W csass i

.d M L.-M.w w :--.. $.,,,-.#._-s$$$w g w w$$ h N $s-Y~~ ?5Th.

$$h

=. % - - -. _m.w-- wem-E Y_ Y 5

hh??s?~es$

h w$sw'NSYAS N

$$ww-l%gt=w?, genm:W2%kYW>yp??l$YT@i"Y%'$ w.a%.es e=ve w-aq we mes m $, h wamm umee M'zwga&mm e.

s 25> w

• M

-<M L

vgt LEAK IN PRESSURIZED LEVEL INSTRUMENTATION bnile conducting reactor coolant boundary integrity tests at the Kewaunee Nuclear Power Plant, a steam leak into the contain=ent was discovered near the pressurizer. The cause of the leak was failure of a bellows and of a gasket in the pressuri:er level sensor. The instrument was isolated and re=oved from service to replace both the bellows and the gasket.

The reactor was shut down at the ti=e, but some fission products in the coolant were released to contain=ent.

The release was within AEC limits, and there was no da= age to any ccmponents and no injuries to individuals. The consequences f rom a standpoint of public health and safety vera non-existent since all activity was confined to the contain=ent vessel.13

~/,

O e

a f

[M

' '&S.

W C6892

.=i---

... = _.

m W

w,

.$"ML5'DSG M M M S%*3% ws.Yw~1% W-a wzx,YO M : $n?b..d n

=-

h s w:'r-w " & & w~b$2$$$ n'BI N

e'Il$k -

???

W;t= $$U5Ae'NN?52iWdif5$$

N d $ f N W W $ N M, - t N M M Yh5 @~ EI N-m 1-r m.. a n25MN w

nu w

= %.~,

t~:ww t ~.-x.m$s yc.

D w-:m -

~

hEMf w~

WB Ma bM W

M:f;;

_ 10 -

'.h VALVE CUIDE FIN FAILURE i

While Unit No. 2 of the Oconee Nuclear Pcuer S tation was shut dcun and, with the reactor coolant system partially drained, the icw pressure injection flow in the decay heat removal mode could not be reduced belcw 4000 gpm.

The valve stem of the outlet valve to the icw pressure injection cooler was not traveling full stroke, the extended throttle ficw guide pin at the base of the valve plug had broken of f and the g

valve seat was cracked. The 10-inch, 200-lb. rating, cast alley steel gicbe valve was manuf actured by Crane Co., catalog No. 151 1/2 LU.

The e

3/4-inch dia=eter by 4 inches long cylindrical 304 stainless steel guide pin extended below the valve seat into the inle portion of the valve y

body.

t The guide pin could not be located in the vicinity of the valve piping, but its loss did not af fect the valve shutof f capabilities.

The apparent cause of the guide pin f ailure was ficw cavitation resulting

~~

c frcm Icw volu=e flow in the decay heat removal mode. The Crane Co. and

.g Babcock & Wilcox confirmed flow rates of less than 3000 spm could cause cavitation with this valve design.

The integrity of the guide pin was verified in January 1974. Hcwever, gg since then, the system had operated at ficw rates of less than 3000 gpm in the decay heat removal mode f or considerable periods of ti=e.

I i

An extensive search for the guide pin was unsuccessful. Mcwever, there is no indication that the missing valve guide pin is interfering with reactor operation or that the pin is loose in the reactor vessel. Duke I-Power Ccmpany determined that continued operation did not represent an undue risk to the health and saf ety of the public. W gg bhe$

f.,

%&k Q

. ; ].

f L ( (.S H W / ? W CSES3 p', vses Conn / cm,aiya _ 'QQ //glVL,# L l3 A 3 /0 /3 W - (O GW-W/? Q (, ) w h r e. f f *T l achd fn. ffe,e &,. 79,;, ppoj/am j, ,,,.t-i e mf ay 4u.. G ,e we m

=.; Nf- '4 .%.es er:htv ;il., w-. -[;EM.$d5$FN,ce*s::e%:7-M N ="r-tmoerse p m w e _S. J r _we m,- e %$ y m ic. k. U M d EE: M'EM ;w :,w%.4 N _%wmp-,-e. gg.%s. ;;vy.42:- -g - 2 L.-e._. w - c e. 2 a. c 2 w 7 ye e.. y u :.Cf NbN i M $ h j IE N h 1 11 - OPEN VENT VALVE While heacup of the Kewaunee Nuclear Pcwer Plant was in progress, water and steam were discovered in the contain=ent base =ent which appeared to originate from a vent valve left partially open. One of the checklists used during filling and venting of the reactor coolant system had not been completed. The reactor was not critical; the onlf fission products present were residual products from previous reactor operation. There was no damage to any system, cc=ponent or structure, and no personnel injuries were sustained. Personnel exposures were minimal. There was no adverse affect on public health and saf ety because the activity release frem the vent valve was confined to the contaie=ent, and air samples indicated that the level of radioactivity in contain=ent was within allowable naximum permissible concentration limits.15 4

1

/.+<. 4 w esas4 t v r m'r

m. - MA-TCZ w @.~{ @*[ M 2%= M N-5~ N $ f & $$? h'k V ? ':O**= Q*" Q ) 5 % Y$ 2 & ~1 T c - 7 1 m e W .We=5 _ m,w _.4.- %..t %,=c.m.i_ W:.-G w,qn'indy;tecf:Ci35 5-G'& W.L..qq%_ _W-.R.. =t=iM.._ru..n.

2. m%RS:

+ e g.4'- Ri1+2W it e,e._zz_:,2._,?9t_ w..&y.t.,. w e e~ 6iC.y kh m m. MS ', es AG FAILURE OF LCOP ISOLArION VALVE PACKING -P k'hile Unit No.1 of the Surry Nuclear Pcwer Plant was operating at N* approx 1=ately SO: pcwer, a reactor vessel leakoff high ce=perature alar = was received. A check of the reactor coolant system indicated leakage of approximately 30 gpm. The reactor was placed in the hot shutdown cendition and a steam leak from the valve stem of loop "C" hot leg isolation valve was discovered. The leakage rate was reduced to less than 1 gpm by back seating of the valve and, because the leakage rate was reduced by back seating, it was g concluded the reactor coolant system leakage was caused by failure of Si the valve packing. No other i==ediate corrective action was considered sa necessarf. p ad$ During the period that the valve was leaking the loop could have been isolated frem the reactor vessel, but despite the degradation of the g packing, the loop "C" hot leg isolation valve remained capable of per-e forming its intended function. Therefore, this event did not represent g an undue risk to the health and safety of the public.16 E Es erw -E= w aw: =m E x M 3= S' if l ur += 3 W_ Bf W 06SSS Y ^ 6 46e _ ~., y

_c ; -,. z.n m a.-.- w.g -- - .m...mt -..n..=, m zy = $k f-w d 5 bY <@g%g$ n=~Gyffdfg$' f$~-~'=- 4 nnw%Qf; aA^ ~g hy.w wm wy=y-:-g~~3A=4> MNFfENNEbNfiNDANE}Ed r khh%d4'fdsst$p*jw,:%s. ~]Mg gtiMQ$ s-ig MM? KINE . SAFETY RELIEF VALVE MODIFICATIONS A design review at Unit No. 2 of the H. 3. Robinson Nuclear Power Plant indicated the 2-inch safety relief valve inlet pipe on the safety injection system accu =ulators could be overstressed from the reaction force of a blowdown. It also was found a siphon head could develop in the level sensing line in the event of a safety valve discharge, with water in the level float chamber being drawn into the safety valve, creating a water ha==er. For correction, the 3-inch elbows at each safety relief valve discharge were to be replaced by 3-inch tees. This will provide opposing discharge forces at each tee and reduce the bending and subsequent stress on the safety relief valve inlet pipe. A 1/4-inch diameter orifice will be installed above the nor=al water level in the line between the level float chamber and the safety relief valve connection to reduce the volume of water siphoned into the safety relief valve. This also will minimi:e oscillation in the water level during a relief valve discharge.17 The design review was performed in accordance with the W.stinghouse publication " Criteria and Guidelines for the Design and Safety and Relief Valve Installations on Westinghouse Pressurized Water Reactors." Y e e W 06SS6 _.= ww-

.... -_.....~.:..... ... ~... _ - 3 1-W QCW 'f,f ~ J [~ f&fff;ff=$fghQ h ~ hhh':. g p5I b g,.we ~$$ k-. b E55 N M_ M.~yc.-fg,h?y.A=2nD W.?.m.xE N Y$ A,a. NM.c.mmem ~ re.w - w.. v ~.,.,.r ce . :~~e.w_m -~ eg p: E ;- ' w~~~m MALFUNCTION OF 30RCN INJECTICN TANK OUTLET VALVES L'hile operating at 96 of raced power, Unit No. 2 of the Indian Point Station was shut dcun by a high steam line ap safety injection signal. All safeguards systems operated correctly with the exception of the boron injection tank outlet valves. t There are two series and two parallel valves on the outlet { g inj ection tank. of the baron Normally, the series valves are open and the parallel j valves closed. If safety injection is initiated, the parallel valves i are auto =atically opened. In this instance, the parallel valves started to open, but closed along with the two series valves. 3 Closure of all four valves should occur only follcwing discharge of the tank when icw level is indicated by at least two of three boron injection tank level channels. At the ti=e of this occurrence, one of the level channel transmitters was being supplied by its backup. electrical supply from a lighting bus. A low level bistable in one of the other channels had been incorrectly wired so it, too, was supplied by the lighting bus, b' hen the safety injection system actuated, the lighting buses were stripped from the 480 V bus, and the temporary loss of the lighting bus resulted in closure of the four valves. b' hen the lighting bus was reenergi:ed as part of the normal procedures following safety injection, the boren injection tank outlet valves opened correctly. At the ti=e of the malfunction, the boren was not needed for reactor shutdevn even under accident conditions. The Boron Injection Tank provided excess shutdown in the event of a break in a steam line, hence the safety implications of this incident were of sinor significance.13 e W CSES7 .qD. __.___.____m_

. _ ~ ..s. 6 0 -4% %% M W3MEEwe.55z w w kn % %3& W ~ M F $$>Y M S%N%KM ?W43EM~NAELf5:W M O4 &,?h && $ r W5'E 5WV2W-+.x?- G Mi233?$b7M5Ei?21Q2rs#N!?@54s Ris %YW N ,MMISSYW<~Mai36tM f5fl'EUc ~3OT5~05h* WIDE *Y'"85 ^ -.'5'OWMEC 8 h _%4 ,,, 3.r.. n d. : > i&. w ~. ~-~--- ^ ^ - - ._=~,s.---= P. ~ [. E 13 - is9 7 MALF0;iCTION OF RHR VESSEL INJECTION VALVE e [ Ouring cooldown of the Cooper Nuclear Station, the Residual Heat Removal b P.HR) vessel injection valve could not be opened f rom the control room. I The valve limit switch mounting block screws had loosened, apparently 3

s a result of thermal cycling and/or vibration, and the mounting block 4

ild not make proper contact with the operator cam. This affected opera-

ion of the Limitorque operator. Valve operation was reestablished by I.

jumpering the limit switch interlocks. The time required to reestablish operation of the valve resulted in only a minor delay in the normal cooldown sequence. The redundant E=ergency Core Cooling Systems were operable. Therefore, this event did not affect the health and safety of the public.19 s I ) W C6S98

~.....-.- f $hS k SN5M Y$h h2 s=hw$$f$ nw=2fi EM E M 5F# 6t s is Fi @g &s. p?S $q MwbfF2 $d$n.bsi$wsm6Eih tn wwf @m M E W S E M= : 2 J au s M& 16 - Sr m:E $E e@ FAILURE OF NON-RETURN VALVES While Unit No. 2 of the Surry Power Station was approaching 4% of rated de power, instrumentation indicated that the =ain steam non-return valve "C" (Rockwell Manuf acturing, Model 607 MO, had not opened. The reactor ~^ 418 8 was shut down to free the valve. After the valve was operational, and g as reactor power was increased, non-return valves "A" and "3" exhibited y similar problems, so the reactor was again shut down and all three non-

rs return valves were dismantled.

In_ each case it was f ound that the disc had separated from the disc / piston asse=bly of the valve. M There was a crack on the inside surf ace of the piston asse=bly above b the weld that joins the disc to the piston. The crack was propagated by Q cyclic fatigue stresses experienced during prolonged startup' periods. 52 The valve is constructed so steam builds up beneath the disc until g4 sufficient pressure has been produced to lift the disc / piston asse=bly r off its seat. The lift force on the disc is greater at the point of F steam exit than at any other point on the disc. The steam produced at startup has a high moisture content and this low quality steam would impinge on the disc with a greater impact force than normal high quality steam. N**M During startup, this cyclic pressure buildup and release, or perking, may continue for several hours, and this perking of the valve disc, together with thermal gradient buildup, caused the f ailure. M{ 3 Each main steam line has a fast closing trip valve and a non-return g valve. These six valves would prevent blowdown of more than one steam RE' generator for any steam break location, even if one valve f ailed to close. The occurrence could have rendered a portion of the Engineering g Safeguards System incapable of performing its intended function. How-ever, the main steam trip valves were operable and would have precluded b-the blowdown of more than one steam generator. Therefore, there were no g safety i=plications associated with this occurrence.20 it"6 i EE DN l 'D.' h g g d A jI' V M OM S M k %d Mk N g mQ^ s ~ w 'Q, % N, N k w ~ _.D(ww -*f, Gm ~$ s c a w e, i= gwL-U a.- m W CSSSS r 6f'dt m f L-a. S i. 6 n i z..+ n d w c a 7 M m ~f .2 i, W f L.O*(.t, , q g*!sM n.

x. f 3

N s. M e ~ e t.' a u

._-..--i-~------------- ~ L --- ~ ~ - ~ ~~ ~ $f3-bY b M + i DEFECTIVE CHECK VALVE I During leak race testing at Unit No.1 of the Quad Cities Nuclear Power Station, the volume of an 8-inch diameter pipe between the reactor core isolation cooling (RCIC) turbine exhaust check valve and the manusi valve could not be pressuri:ed. The check valve was disassembled and the check valve disc was found approximately 6-inches downstream of the check valve in the RCIC turbine exhause line. The fracture had occurred at the check valve disc hinge joint. The Crane Co. did not have a valve disc replacement. Because the RCIC turbine exhaust valve is infrequently operated, the extrp hardness of the stellite seat was not required, so the disc was replaced with an identical disc without the stellite seat. Backward leakage through the check valve and exposure of the suppression chamber atmosphere to the exhaust line could have been isolated by a downstream manual valve. Backward leakage through the check valve would act render the RCIC system inoperable; it would 'be capable of performing its intended design function. Equipment dauage and personnel exposure would not have resulted from operation with backward leakage through the check valve.21 i I t i I \\ 1 W 06900

...- ~..__._.... -... &;pg&mgp.+%E$N;bg.,+&.co-J3h.~$E$,.$.1 Y A$ n. N M'M'Y T th.fE5&& _h pw% +w w n~ gg -. gr ge n -... m. ET P. m, 13 - PROBLEd.S VITH MAIN STEAM ISCLATION VALVES During testing of the Main Staam Isolation Valves Od5IV's) at the Menticello Nuclear Generating Plant, the leakage rate past an inboard MSIV was 73.4 sefh; the limit is 11.5 scfh. The main poppet was not mating properly with the valve seat. The leakage was reduced to 3.35 scfh by lapping the main seat and pilot seat and truing the main and pilot peppets. The Northern States Power Co. believes low spots on MSIV seats have been the result of seat warpage which occurred when stresses in the valve body were relieved at operating temperatures. Similar problems were experienced in the past with three other MSIV's, but since there were fewer MSIV's with this problem during the 1974 refueling outage, Northern States believes that in the future, the MSIV's will demonstrate satisf actory leak tightness. ili Earlier this year at Mencicello, two outboard MSIV's f ailed to close i during testing. The two redundant MSIV's on the "3" and "C" steam lines j were closed and the reactor was brought to a hot standby condition. The AC solenoids on the MSIV a'ir operators were not venting properly when f deenergited. Metal chips, which could have prevented the solenoid plunger from repositioning properly, were found in the AC solenoid on one manifold; no significant quantity of =etal chips were found in the f AC solenoid on the other sanifold. The chips were the result of inade-j quate deburring and cleaning of the manif old during sanuf acture. t j Upon examination of the solenoids, it was found that the viton seats in 5 the solenoid plunge'rs were per=anently def or=ed. Leakage through the j exhaust port could have created a low pressure area across the top of { the plunger, possibly af f ecting proper plunger operation. I{ The soleneids and sanifolds for all MSIV's were disassembled and cleaned and new plungers were installed in all AC solenoids, e i l l 1 Another problem reported by Northern States Power Co. was the closing of % l two MSIV's in six seconds and ten seconds, respectively, instead of the i required 3-5 seconds. In both cases, the valves were slow because of i excess friction between the spring support yoke and yoke rods. A pair I j of rollers attached to the top of the yoke were not riding on the yoke i rods. This caused metal-to-metal contact during the valve closure stroke. To prevent recurrence, the rollers were adjusted to obtain clearance between the yoke and yoke rods, and the contact areas between the rollers and yoke rods were lubricated. W 06901

.~_s_.---- - -... -. - - .W3. .%f9290P' FmfKMp: Cr-EN2MVM-hMWw 4t &yaf~==wwwasy v=m -Kg5c;-&g:&:.g::re-weest M fen.,c"cfswd=c:cw;i;i@ifif626N}$e.gy.c e -w \\ q _gW~r 32:W mit ?m: c n.wty m a -w g L- -w. .Nh'3[N.s. i "MJ4YM-l W =!? w e.as.,Ac,vM+w.e.m,vre.,M., ~mrwwwc pi%e 5 % e. %, w - Mw -- A.- m ww Az m: c;y:.nn.m t'ny Lew-r u.c.c e. w -r-ww e-7 m: w rw?f-;,m:= h-www Av

3 3 Q 9 & s w+-. w_- =~- - w- ~ _ m e _- w

- ~ - - - - - -- - -- l i t t, j Although a nu=ber of problens have been experienced with MSIV's at 3 Monticello, corrective action prevented any possible adverse ef fect frem these events on the safe operation of the plant.22 I i p-i W CSSC2

-... _.... - -.. ~.. -...~..- - -. - - - - - - - ~ - - -' " ~ ~ " ~ " ' ' ~ ' M $2&h,$JR.Nb,a ?Wb5.:$$$$?kh5W v. id MEF:'2ih R.'L v% $?l==y.& csCf&*j s w:s m % a. Es = _w% -90 ~;?G$?W, 2Gv. ,5Q1 35;3 &-Mp yg __h .g bW 55? eune m* VALVE STEM FAILtRE .t During a 1973 cooldown of the Unic No. I reactor at the Surry Poser p@$ Station, there was a reduction on "3" reactor coolant loop to approxi-si mately 707. of rated flow. It was caused by the stem failure of the N_ reactor coolant system isolation valve. dkh The valve was a 34-inch stainless steel gate valve designed for 2500 psi service at 600*F manufactured by Darling Valve Company. The valve stem Eliij was approximately 10 feet long and 4 inches in diameter with one end R forged to a 6-inch diameter for a 7-inch length. The stem material was 14-4 PH stainless steel. Visual examination revealed the failure occurred at a sharp ' change in cross section directly above the 6.25-inch diameter collar which was iP i j contoured spherically to the stem to form the stem backseat. The failure was essentially straight across the stem. The fracture surfaces were g-clean and free of corrosion oroducts or contamination. Investigators of the failed stem, 'Jestinghouse Electric Corp. and E 3accelle-Columbus Laboratories, concluded that the fracture was trans-j;. granular and that chemical and physical hardness properties were within s original specifications. They agreed the fracture resulted from several g cracks initiated in a very sharp fillet (0.03-inch radius) just above g the backseat ring, and that the crack grew in stages, rather than being y the result of a sudden co=plete failure. However, the investigators did n not agree on the failure mechanism. '4estinghouse believed stress corrosien s-cracking and hydrogen induced 9 tress cracking (hydrogen embrittlement) were not a potential cause of failure; they describe the failure mechanism je as " quasi-cleavage." e-h The analysis by Battelle was not conclusive in identifying the failure mechanism, but their results indicated secondary aging" had occurred, I resulting in a higher tensile strength and lower ductility. The increase 1 of tensile strength in the fracture region could iave increased the F sensitivity of the steel to environmentally induced cracking =echanisms, -{ such as stress corrosion cracking or hydrogen stress cracking. = g- = r -r 1 ?-5 = = s ~ { W C6903 m L -_[ h?

... ~...... .-~, -- s C A .N eemm4m*-2.sZr-se$. %CY;inL%'SEM.e,$rN~~sr#4w.t:W 'A e%mw>:M= ore-.YSWM.~IEST35h: WW~.' 5+ De s sund"sf

arr'cr e n e M ee 21 -

The reactor was in the inter =ediate shut down condition at the ti=e of the valve stem failure and there would have been no significant safety 1.p11 cations associated with a transient resulting from the reactor coolant system isolation valve f ailure. At no time was there any ha:ard to the health or safety of the public.2 3,24 Theodore C. Cincula John J. Ri::o Office of Operations Evaluation U.S. Atomic Energy Cc= mission f I i l 4 W 06304 4 1 1 .me. .*.-e. l

..-..~....-.. _- -.... ~ ~ E Ef - ". j a ,.fE.3% y g_m.>Q~ g.m.m%:,.w.M,,g. , -_ m. 4Lw.wy c cz 4 -~, % .M M.%j$0h. " -'M e * * *k [" M M M 3 N' 534N .MTNlJr% m E mg 1-v mr m-i.~w. --aar nc- ~ S + r REFERENCES 1. Letter, B. W. Riley (Vermonc Yankee Nuclear Pcwer Corp.) to USAEC, Directorate of Regulatory Operations, Region I, July 8, 1974 ACR No. 74-11, Docket No. 50-271. 2. ROE 74-7, Off-Gas Explosions, March 22, 1974. 3. Letter, E. E. Utley (Carolina Power & Light Ccmpany) to K. R. Coller, USAEC, Assistant Director for Operating Reactors, July 10,1974 AOR No. NG-74-797, Docket No. 50-261. 4. STEAM GENERATOR PRIMARY-TO-SECONDARY LEAKAGE, SURRY POWER STATION, VIRGINIA ELECTRIC AND POWER CCMPANY, July 8,1974. AOR No. S2-74-05, Docket No. 50-281. 5. Letter, E. E. Utley (Carolina Power & Light Co.) to J. F. O' Leary, USAEC, Directorate of Licensing, July 15, 1974. ACR No. NC-74-873. Docket No. 50-261. 6. Letter, A. D. Schmidt (Florida Power & Light Company) to J. F. O' Leary, USAEC, Directorate of Licensing, August 1, 1974. AOR No. 74-8, Docket No. 50-250. 7. Directorate of Regulator Operations Notification of an Incident or Occurrence No.121. 8. Letter, B. R. York (Iowa Electric Light and Power Company) to J. Keppler, USAEC, Directorate of Regulatory Operations Region l III, August 10, 1974. AOR No. 74-25, Docket No. 50-331. 9. Letter, W. J. Cahill, Jr. (Consolidated Edison Company of New York) to J. P. O'Reilly, USAEC, Directorate of Regulatory Operations, 6 l July 5*, 19 74. AOR No. 4-1-16, Docket No. 50-3. i 10. Letter, J. S. 31cel (Commonwealth Edf son) to J. F. O' Leary, USAEC, Directorate of Licensing, June 19, 1974. AOR No. 74-13. Docket No. 50-295. 11. Letter, A. C. Thies (Duke Power Cocoany) to A. Giambusso, USAEC. Deputy Director for Reactor Projects, July 1,1974. AOR No. 74-10, Docket No. 50-269. 12. Letter, J. W. Gore, Jr. (Baltimore Gas and Electric Company) to D. F. Knuch, USAEC, Directorate of Regulatory Operations, July 24, 1974. AOR No. 630, Docket Nos. 50-317 and 50-318. W 06905

.-.. ~. ~ _... ,4 TM+ W E M N.-_.. E L~. 2_ L 5.-. 4 9 1 R R. $,d ii t,W i n L "

• Q W'T+ p f 5-.-Q1 G l M"~f %,%.SL $ %-cWk Q WA WE

%gG MMW%QWyf $3%W %' 1 x_ ; a..,uv. smrg m 3:. .e , f.c

MvM sM Esi M

13. Letter, E. W. Ja=es (Wisconsin Public Service Corporation) to h J. F. O' Leary, USAEC, Directorate of Licensing, July 19, 1974. .Z AOR No. 74-9, Docket No. 50-305. p 14. Letter, A. C. Thies (Duke Pcwer Cc=pany) to A. Giambusso, USAIC, T Directorate of Licensing, July 18, 1974. AOR No. 74-3 Docket No. 50-270.

15. Letter, E. W. Ja=es (Wisconsin Public Service Corporation) to J. F. O' Leary, USAIC, Directorate of Licensing, July 19, 1974.

AOR No. 74-8, Docket No. 50-305. 5-g 16. FAILURE OF LCOP ISOLATION VALVE PACKING, SURRY POWER STATION, VIRGINIA ELECTRIC AND P0bH COMPANY, June 11, 1974. AOR No. m~-- S1-74-01, Docket No. 50-280. 17. Letter, N. 3. Bessac (Carolina Pcwer & Light Ccmpany) to J. F. _~ O' Leary, USAIC, Directorate of Licensing, June 20, 1974. ACR No. NG-74-745, Docket No. 50-261. -k p 18. Letter, W. Stein (censolidated Edison Company of New York, Inc.) { to J. F. O' Leary, USAEC, Directorate of Licens' ng, July 3,1974. i AOR No. 4-2-21, Docket No. 50-247. =- 19. Letter, L. C. Lessor (Nebraska Public Pcwer District) to E. M. Ecuard, USAEC, Directorate of Regulatory Operations, RO IV, July 1,1974 ACR No. 74-40, Docket No. 50-298.

20. MAIN STEAM NON-RETURN VALVES, SURRY P0kE STATION, VIRGINIA ELECTRIC AND POWER COMPANY, July 8,1974 AOR No. S2-74-04, Docket No.

50-281. 21. Letter, N. J. Kalivianakis (Cc=monwealth Edison) to J. F. O' Leary, USAEC, Directorate of Licensing - Regulation, July 18, 1974 AOR No. 74-8d, Docket No. 50-254. 4.. &E 22. Letter, L. O. Hayer (Northern States Pcwer Cc=pany) to J. F. l 5-O' Leary, USAEC, Directorate of Licensing, July 9,1974 Docket 8 6 No. 50-263. E.E 1 = 23. REACTOR COOLANT SYSTDi ISOLATION VALVE FAILURE, SURRY POWER STATION, [UE VIRGINIA ELECTRIC AND P0kE COMPANY, February 20, 1974. AOR No. 73-06, Docket No. 50-280. 24. SUPPLEMENT TO REACTOR COOLtNT SYSTDi VALVE STDi FAILURE, SURRY POWER STATION, VIRGINIA ELECTRIC AND P0bH COMPANY, July 12, 1974. AOR No. 73-06, Docket No. 50-280. = yo...... W 06SCG F

I e .N' b*:- "'ag.e. dh $4W bD7M'N"JM.

• -WJ**T,e*CbTCi;.~db-: q.W.DWu f'in v",:%g%. r.Lp>W~y&..gC'*^.%WY'I,SXf.El.W-@aq*,-(,? +@7:

5 5MWQWM 9 htGW. M M h' ~* m w d--

• yt't'>&^' 4%y %

n.i W

• M**

a &agy; '-' %..<r c?3 T-t .kn7 c w','. k M > &-w % f. d.:xY4c4 %% r' 4K .~ W,.. g,. - ede r%'t%n.

• v&.C w+ r:w._ d. 4 % y.&.**%s f es y*sr "*t' hc g s-Y Y.

h w t 4 '?. %,w.."' a 5-- m% s L r. m3% _ te 4..::e p M_..g_ Qty-R.; +'*f %C

• 7>. e ser

&p72'Y s s fY W ?TY h57*2'. -Y_ n_ '.?i$_._,,._._.. _~, _.WL?5. _ _ m.,&m.?5,.% w $ bk $..b$ts .u t.d_ YM 3 gg 5$.,g 7eOh. $$h. t g, egagg 7g x k. I UNITED STATES p t; ATCMic ENERGY CCMMISSION WASMmGTON O.C.19S4S C*

• evea6ty pom peivatt ust snoe L

J fl0 P.'.: Att D I CK I i:S O:t ROE % GPO STARTUP BOX 480 M I D D L E TO'.' lit. P A. 17057 O er 5" e W C6907 M _g e. g -h a v -~ ~ g e,,, -, - -}}