ML20024B203
| ML20024B203 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/15/1975 |
| From: | Toole R GENERAL PUBLIC UTILITIES CORP. |
| To: | GENERAL PUBLIC UTILITIES CORP. |
| References | |
| TASK-*, TASK-GB GPU-2481, NUDOCS 8307070317 | |
| Download: ML20024B203 (15) | |
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AEC DOCUMENT REVIE'.-I
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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: hew /f2[
Other
, dated:
Review of the attached document has concluded that no action is required.
Startup & Test Manager Date Test Superintencpnt Date Review of the attached document has concluded that action is required by:
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G FEDWATER SPARGER MOVDIENT KsE is Unit 3 of the Hu=boldt Bay Power Plant was shut down for a regular d
scheduled refueting outage when it was discovered that five of the eight hold-down U-bolts on the feedvater sparger had failed where the bolt enters the upper lock nut; the sparger and ther=al sleeve had shif ted approxt::ately 1.7-inches from the vessel inlet nozzle.
Both 0.5-inch dia=ecer type 304 stainless steel legs rere oraken on three U-bolts; the other.two had one leg broken. There were no missing g
parts.
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A metallographic examination of two bolts led to the conclusion of
@y f ailure from high cycle fatigue; the U-bolts vibrated in response to i
the feedvater pump or to recirculation flow.
It was postulated the j g sparger was not vibrating or that the vibration amplitude was very i
small because there were no unusual wear = arks, d
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c The sparger was reinstalled in its original location with redesigned
[y and substantially stronger sparger restraints which clamp tightly j
around the sparger. Because the restraints are in firm contact with stE the sparger, the new restraints should be less susceptible to vibration.
g The failure of the bolts and movement of the sparger did not impair I
the ability of the :parger to perform its intended function during either n,or=al or accident conditions. There were no indications of any I
changes in the thermal-hydraulic operating characteristics of the
,j feedwater or reactor systems. Since the ther=al sleeve is an inter-j, ference fit and the nozzle inside diameter is uniform for the entire i
length of the noz=le, no increase in feedvater flow around the sleeve g_
would have occurred as a result of any shift in the sparger. 12 This g
event did not af f ect the health and safety of the public.
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E R$$iiEBMiBME$WBBGERGEMG 2-REACTOR STARTUP ON INCORRECT POWER RANGE RECORDER 2 was being increased the Oconee Nuclear Station Unit Power level at following a shutdown the previous day, and was at 3% of full power as Following a control room indicated on the Reactor Power Range Recorder.
the shif t change, the operator noted from redundant instrumentation that The Power Range R
reactor power level was actually at 15% of full power.
Recorder was in the O to 125% full power range instead of the O to 25%
The selector switch for the power range recorder had full power range.
not been changed to the expanded range af ter the reactor shutdown.
The operator normally uses the power range recorder during startup 4tt#24 Because of the assumed e c' -
because of the instrumant's expanded scale.
lower power level, the control operator did not utili:e other redundant low power R
instrumentation for comparison because of their limitations at levels.
The Trip Recovery Procedure has been revised to specify that the lower range on the power range recorder be utili:ed whenever reactor power is In addition, personnel were reminded of the below 20% of full power.
instrumentation.
importance of utilizing redundant E
The Reactor Protection System is totally independent of reactor power the unit even indication devices, and would have functioned to protect This incident did not with the power range recorder in the wrong range.
affect the health and safety of the public.3 UNPLANNED RADI0 ACTIVE RELEASES - PERSO:CTEL ERRORS Dresden - 1 With Unic No.1 of the Dresden Nuclear Power Station operating at a steady state power of 143 5'e, an operator was given instructions to line up the valving to discharge the "3" Waste Holdup Tank.
The pro-cedure for this operation is specified on a discharge card which con-tains all infornation necessary to discharge the tank, and includes a The operator misinterpreted the card to be for the "3" check-off sheet.
Laundry Holdup Tank. As a result, the unsampled liquid from the Laundry Holdup Tank was discharged for a period of 45 ninutes.
Upon realitation that the wrong tank was being discharged, the action was terminated and the water remaining in ene Laundry Holdup Tank sampled for radioactivity. The gross Beta Gamma analysis of this sample was 1.2 x 10-4 uCi/cc. An estimated release rate of 20 gallons per minute led to a calculated release concentration of 20 uuCi/cc af, r iu o.
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It was concluded the health and safety of the general public was not
.c affected by the unplanned release of liquid from the "3" Laundry Tank; at no time was the technical specification allouable linic for liquid releases exceeded.
The c.orrective action to prevent the recurrence will be to color code the discharge cards with a different color for each tank.
Each dis-charge card will contain instructions and valve check-off list only for the specified tank.
In addition valves required for discharging tanks will have color coded identification tags matching the discharge card f or the specific tanks.'
Palisades With the Palisades Plant in a cold shutdown condition, the south filtered waste tank containing laundry waste was placed on recycle for analysis to determine if the contents could be released. The technician per-forming this analysis had calculated the release, and transf erred the data to the batch release form. The following day, an operator released the. north filtered waste tank as authorized by the batch release form.
The release should have been for the south filtered waste tank; when transposing the data' to the batch card, the technician wrote down the wrong tank identification.
No safety limits were exceeded by the release; none of three monitors sampling the release alarmed. A sample of water remaining in the north tank was analyzed for activity, and a comparison was cade with the south tank analysis. No significant diff erences were found. The total release was esti=ated to be 12 nillicuries.
1, This evenc was reviewed with the responsible technician and operator.
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i With the Oconee Nuclear Station at 99" power, preparations were being
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nade to replace the letdown filter. The valves necessary to isolate the filter were properly positioned and tagged. However, =aintenance per-sonnel accidentally disconnected the vent piping from the filter (a quick discennect fitting). This action resulted in release of reactor coolant to the letdown filter room and adjacent hallways in the auxiliary building. The saintenance men i= mediately evacuated the area.
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Vr-spray from this incident discharged approxi=a:ely 3500 The resul:an:
The gallons of wa:er, and a total gaseous release of 16.5 curies.
gaseous release was 3.7". of the one-hour release ra:e persi::ed in :he technical specifica:1ons. The water was analy:ed for particulates, iodine and tritiu=; only negligible quanti:ies were found.
It was concluded that the heal:h and safety of the public was not affected by this incident.
The apparent cause of this occurrence was the failure of main:enance personnel :o identify the fil:er they were to replace. The filter had no identif ying =arkings.
x Identification of the letdown fil:ers will be =ade permanent, legible, and easily recogni:able to prevent future occurrences.
In addition, a program will be developed to assure that all equipment is iden:ified by unit and componen:.6 Oconee - 1 Uni 1 of the Oconee Nuclear Station was in a cold shutdown condi: ion when an opera:or discovered three feet of water standing in the pump Electrical power was isolated from the pumps and a submersible
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pump was used to pump the water to the lov ac:ivity waste tank.
Elec-trical and functional checks of :he low pressure injec:1on and reac:or building spray pumps were perfor ed to verify operability.
Samples of water were taken from the pump room; chemical and gn==a spec: rum analyses indicated a radioactivity level and boron concentra-tion consistent with water f reu the Low Pressure Injection System.
l The apparent cause of the flooding was personnel error. A utility
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opera or incorrectly assumed : hat isola ion of the low pressure in-jection header had been ec=placed by the control room opera:or by j gggggll closing the renotely operated valves; he did not visually verify that j
- he valves were closed. The header drain valves were opan.
I The water drained from the header to the floor drains and the automatic
- ggggggg cycling su=p pumps apparently tripped because of pu=p overload. The g
draining water collec:ed in the pump : con.
J The draining of the pump room and subsequent verification :es:ing of the j p fl affected componen:s was completed ten hours af ter initial discovery. Since "g
the radioactive water was not discharged to the environment, i: was j
thehealthandsafetyofthepublicwasnofaf#arad06936 concluded that j
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b To prevent a similar recurrence, a su=p pump =enitoring alars was in-stalled to detect early pu=p failure, and the necessity for attention to detail in all station operations was stressed to personnel.7 Also at Unit 1 of the Oconee Nuclear Power Station, with the reactor operating at 30% power, several radiation monitors alar =ed in the auxiliary building. An instrument line for the unloading valve on the gaseous waste separator tank was found disconnected. The loose tubiis was reconnected.
Apparently, during a station =cdification performed by the day shift, the piping was not fully connected. The contents of one of the gaseous j
waste decay tanks e=ptied into the auxiliary building.
The total gaseous activity released was 25.3 C1, which was 0.0f:
of the annual release limit. The total iodine released was 2.!* x 10
C1, which was 0.06% of the annual limit. The =ax1=u= release rate averaged over a one-hour period was not exceeded, and personnel did not receive any significant radiation exposure. The health and safety of the public was not af~ected.
f At a =eeting of the station canager and all supervisors the necessity for attention to detail and co=pleteness of =aintenance action was discussed.8 0uad-Cities While Units 1 and 2 of the Quad-Cities Nuclear Power Station were in operation, discharge of a known quality of liquid fro = the "A" floor drain sa=ple tank (FDST) to the river was initiated. Later in the day,
. radvaste operator started processing the controls of the floor drain collector tank to the "A" FDST rather than to the correct tank, the "3" FDST. The incorrect transfer ccn'tinued for about fifteen =inutes during which about 1000 gallons of water of unknown quality was =oved to the "A" FDST.
The radwaste operator noticed the level increasing in the "A" FDST and checked the valve line-up.
Realizing the =istake, the dis-charge to the river was stopped. Approximately, 90 gallons of uncon-trolled, but =onitored effluent was discharged to the river in this fifteen minute period.
l The =istake was caused by operator error. Between the time the "A" FDST
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discharge was initiated and the pu= ping of the floor drain collector tank there had been a shif t change. Following the shif t change, the new operator =ade a valving error while processing the water frc= the floor drain collector tank.
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UST # t vicy ef fluent concentration was 1.2 x 10-3 uC1/ce.
- tE analy:ed f or the original batch release was 2.9 x 10 3 y
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occurrence the activity of the water coming from the floor drain collector was less than that being discharged.
seven months earlier when an There had been a similar occurrence about to the"A" FDST instead of Q
operator had accidentally opened the inletthe "B" TDST while processing water from the floor drain collec:or :ank.
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The "A" FDST was being discharged to the river at In both events there were no adverse ef f ects on the heal:h or saf ety of the public or plant personnel. The applicable li=its in 10 CFR 20 were not violated because of ghe inherent safety limitations designed into the discharge procedure.
PlPE WALL EROSION 1 of the Calvert Clif f s Nuclear Power Plant in a cold With Uni: No.
shu:down condi: ion, a pinhole leak was discovered in piping immediately
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An downstream of a but:erfly valve in the sal: vater return sys:em.
ultrasonic test measured a =aximum thinning of the 0.375-inch pipe wall the 12 o' clock position, approxi=ately one and a half to 0.110 inches at inches from :he valve flange. Also, a general reduction of pipe wall thickness had occurred in the top portion of the pipe in an area abou:
four inches downstream of the valve.
The pipe erosion was assumed to have occurred from prolonged operation l
with the butterfly valve partially closed, thus acting as a throttle This valve is throttled :o control temperature in one of the valve.
service water subsystems; a si=111ar erosion of the pipe wall had occurred previously in another sal: vater subsystem.
The pipe erosion had been a slow localized process and would not have resulted in a catastrophic loss of the service water or saltwater sub-There was no i= mediate safety hazard to the plant, its per-systems.
sonnel, or to the general public.10 LEAK IN LOW PRESSURE INJECTION SYSTDI A leak vas discovered in the low pressure injec: ion system piping in l
the decay heat removal room at the Oconee Nuclear Station Uni: 3.
The leak j
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originated from a ce==en sa=ple line for the A and 3 low pressure in-jection discharge headers. The def ect was found two inches from the A header isolation valve. The 3/8-inch stainless steel piping failed from vibration of the low pressure injection discharge headers.
The leak was discovered when the reactor was in a cold shutdown code.
The small si:e of the piping did not have any effect on the decay heat re= oval operation; the leak did not affect the health or safety of the public. 1The line was repaired and a coil was added to absorb vibrational stress.1 p/
DIESEL GENERATOR FIRE At Unic No. 1 of the Three Mile Island Nuclear Station, a small fire occurred in the lagging around the diesel generator engine exhaust cani fold adjacent to the turbocharger exhaust gas inlet. The fire was
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promptly extinguished and the diesel engine was shut down.
The cause of the fire was oil leaking from the engine inspection cover place into the exhaust =anifold lagging. The oil-soaked lagging was ignited by the heat frem engine exhaust manifold. Since there was no apparent damage to the engine, it was restarted and the operational surveillance procedure was completed satisf actorily.
A temporary oil catch tray was installed under the engine cover plate to prevent oil contact with the exhaust manifold lagging. The engine manufacturer was contacted for a per=anent solution to the oil leakage
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problem.12 VALVE FAILURES - SEPARATED DISCS Kewaunee-1 Unic I of the Kewaunee Nuclear Station was preparing for reactor startup when a valve developed a packing leak. The affected valve was isolated, repacked, and the isolation valves were reopened, but flow through the hot leg portion of the resistance temperature detector (RTD) bypass loop could not be reestablished.
X-ray revealed the stem had separated from the disc of a valve; the d
direction of flow through the valve then caused the free disc to act as a check valve.
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us 8-The def ective valve was a Rockwell-Edwards F. stainless steel Univalve, General Assembly Number 3624-7-316J.
The apparent cause of failure was excessive closing torque. An impactor handle is designed into the valve to aid in opening or closing.
Failure occurred from continued impacting after the valve was either in the fully closed or fully open position.
A plant directive was issued specifying the number of turns required to open and close the valve, and that the i=pactor handle was not to be used to force the disc against the backseat.
The bypass loops were provided with flow indication and temperature sig-nals, so a reduction or stoppage of bypass flow at operating conditions wouldbeadequatelysensedandcorrectivemeasurescogdbe taken.
There was no danger to the public or plant personnel.
Prairie Island 1 A similar event occurred at Unit 1 of the Prairie Island Nuclear Gen-erating Plant. A low flow condition was indicated in one of the RTD manifolds. Nor=al flow was observed in the redundant loop. 3y isolating the hot and cold leg manifolds, it was determined the obstruction existed in the hot leg RTD manifold. X-rays of valves in the hot leg showed separation of the valve stems from the discs. Again the, separated discs were acting as check valves obstructing flow in the manifolds.
Rockwell-Edwards considered excessive backseating to be the probable cause of failure. However, they believe that backseating with the r
impactor handle would not cause valve da= age unless a sledge or cheater is utilized. The two valves at Prairie Island 1 were located under floor plating in a position where it would be almost i=possible to utili:e a sledge or a cheater.
To prevent recurrence, the valves are to be seated gently. Rockwell-Edwards is analyzing one of the damaged valve seems and discs to deter-mine the cause of failure.14 Point Beach 1 In early 1973, Unit 1 of the Point 3each Nuclear Plant was experiencing problems with the seventeen Rockwell-Edwards valves in the RTD bypass line. The most coc=ca were associated with valve packing Leakage.
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b' hen packing leakage was discovered, it was dif ficult to disassemble or j
remove the valves. Galling and lack of access space, together with high radiation levels because of crud accumulation, created re= oval difficulties, b' hen attempting to back flush the valves to reduce the radiation level, some of the valves appeared to be acting as stop check valves. X-ray inspection revealed two of the valve discs had become separated from their stems; because of the orientation of these valves, they were acting as check valves.
l Because of radiological and disassembly problems, the entire piping was cut intact from the system.
Three valve discs were found to have separated from their tems.
No parts were missing. The valves were replaced with similar valves with a modified method of retaining the stem to the disc.15 MALFUNCTION OF MAIN STIAM LINE TRI? VALVES During periodic testing of the Mair Steam Trip Valves (MSIV's) at the Surry Power Station Unit 2 while the reactor was at 58% of rated power, three MSIV's did not respond to a signal. On the second test of Valve B, all the air was bled frem its actuating cylinders; the valve closed fully before limit switches could operate solenoid valves to restore air to the cylinders. Closure of the valve prevented normal steam flow, and a reactor trip occurred as a result of "B" steam generator Lo-Lo level. During the unit trip, Valve A closed correctly, but Valve C re=ained open.
Following the trip, the reactor was brought to a cold shutdown. An investigation of Valve C revealed a slight crud buildup and mechanical binding of the rockshaft in the stuffing bex bushing. This binding apparently was caused by a minor bend in the rockshaf e as it passed through the stuffing box bushing. Repair involved relieving the oilite bushings in the area of the splined portion of the rockshafc en eliminate interference and removing the crud buildup.
Each of the main steam lines has a main steam trip valve and a non-6 return valve.
These six valves, in total, prevent bIovdown of one or more steam generators regardless of a break location even if one valve fails to close.
During this occurrence, one valve remained open even though the other two experienced some binding.
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Theodore C. Cintula John J. Ri::o q
Office of Operations Evaluation U.S. Nuclear Regulatory Cer:: mission I
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h REFERENCES y
Letter, F. T. Searls (Pacific Gas and Electric Cocpany) to K. R. Coller.
1 27, 1974 USNRC, Assistant Director for Operating Reactors, November
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AOR No. 74-4, Docket No. 50-133.
l l
2.
Letter, P. A. Crane, Jr. (Pacific Gas and Electric Company) to K. R. Goller, USNRC, Assistant Director for Operating Reactors, i
1 January 10, 1975. AOR NO. 74-4, Docket No. 50-133.
l Letter A.C. Thies (Duke Power Company) to N.C. Moseley, USNRC, Office l
3.
1 of Inspection and Enforcement - Region II, October 30, 1974. AOR No. 74-5, Docket No. 50-270.
4.
Letter, 3. 3. Stephenson (Coc=onwealth Edison) to J. G. Keppler, USNRC, Office of Inspection and Enforcement - Region III, December 19, 1974. AOR No. 74-19, Docket No. 50-10.
5.
Letter, R. B. Sewell (Consumers Power Company) to USNRC, Office of Inspection and Enforcement, December 26, 1974. AOR No. 74-28, Docket No. 50-235.
I 6.
Letter, A. C. Thies (Duke Power Company) to N. C. Moseley, USNRC Of fice of Inspection and Enf orcement - Region II, Nove=ber 26, 1974. AOR No. 74-16, Docket No. 50-270.
7-8. Letters, A. C. Thies (Duke Power Company) to N. C. Moseley, USNRC, Office of Inspection and Enforcement - Region II, October 8 and 25, 1974. AOR No. 74-16 & 17, Docket No 50-269.
9.
Letter, N. J. Kalivianakis (Cc==onwealth Edison) to J. F. O' Leary, USNRC, Office of Nuclear Reactor Information, November 26, 1974 AOR No. 74-38, Docket No 50-254 10.
Letter, A. E. Lundvall, Jr. (Balti= ore Gas and Electric Company),
to James P. O'Reilly, USNRC, Office of Inspection and Enf orcement -
Region I, December 11, 1974 AOR No. 74-13, Docket No. 50-317.
11.
Letter, A. C. Thies, (Duke Power Company), to N. C. Moseley, USNRC, Office of Inspection and Enforcement - Region II, November 12, 1974 AOR No. 74-6, Docket No. 50-287.
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Telegram, J. C. Herbein (Metropolitan Edison Cempany) iggs O'Reilly, USNRC, Of fice of Inspection and Enforcement, to J. P.
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- Region I, Dece=ber 24, 1974 ACR No. 74-31, Decket No. 50-239.
!ke 13.
Letter, E. W. James (Wisconsin Public Service Corporation)
Case, US:PC, Office of Nucelar Reactor Regulation, Oeceber 17, to E. G.
197+.
Docket No. 50-305.
14 Letter, F. P. Tierney, Jr. (Norrhern States Power Company)
J. G. Keppler, U%'RC, Office of Inspection and Enforce = enc -
to Re? ion III, November 1, 1975, Decket No. 50-282.
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OIZ Inspection Repor: Nos. 050-266/72-10 and 73-02, January 29, 1972 and February 14, 1973.
16.
Letter, C. M. Stallings (Vi"ginia Electric & Power Comoany)
N. C. Moseley, trSNRC, Office of Inspection and Icforcement to Region II, October 22, 1974 ACR No. 74-06, Docket No. 50-231.
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