ML20024B615
| ML20024B615 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/15/1978 |
| From: | Otoole R GENERAL PUBLIC UTILITIES CORP. |
| To: | METROPOLITAN EDISON CO. |
| References | |
| TASK-03, TASK-06, TASK-3, TASK-6, TASK-GB B&W-0172, B&W-172, NUDOCS 8307110034 | |
| Download: ML20024B615 (13) | |
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NRC 00CLWEt;. REVIEW i
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Blant/ Unit - //--
I The attached fiRC docume#*ns has been reviewed for test program andi
- =cdification requirements for the above Plant /Ucit.
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CCCUMENT:
Operating Experience, dated:
f Current Events - Powr Reactors, dated:
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, Other
, dated:
e Review of the attached docment has concluded that no actica is required.
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8 1 -/S~-78 Stay & i'est Manager Date est superintancent.
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Review of the attached document has concluded that action is required by:
Problem Report (s) e has/have been issued.
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l Startup & Test Manager Date 1
W CS733 Test Super 1ntencent Date STRIBUTI0tt:
R.W. Heward, Jr.
W.T. Gunn E.D. McDevitt i
J.E. Kunkel M.A. 4elson R.J. Toole J.T. Faulkner File 8307110034 780115 DR ADOCK 05000289
i UNITED STATES CURRENTEVENTS NUCLEAR REGULATORY POWER REACTORS COMMISSION THIS COMPILATION OF SELECTED EVENTS IS PREPARED TO DISSEM INFORMATION ON OPERATING EXPERIENCE AT NUCLEAR POWER Pt.
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TIMELY MANNER AND AS OF A FIXED DATE. THESE EVENTS ARE S PUBLIC INFORMATION SOURCES.. NRC HAS, OR IS TAKING CONTINUCUS ACTION ON THESE ISSUES AS APPLICABLE, FROM AN INS?ECTION AND ENFORCEMENT, LICENSING AND GENERIC REVIEW STANOPOINT.
1 SEFlPIER - 31 Ctiutx 197 i
(PU8LISHED DECEMBER 1977)
OPERATOR ERROR on January 11, 1977 while the Fort Calhoun Station Unit I was operating, water from the Refueling Water Storage Tank was pumped
.into the containment through the containment spray header due to an operator error.
During the performanca of a and containment spray pumps,quartarly test of the safety injection the operator noticed an increase in the containment sump level,approximately tan minutes after the low pressure safety injection pump had been startavi.
3300 gallons of water had been pumped to the contatnment.Approximately About one
..;;ygg minuta later the ventilation isolation actuation signal was received.
At this time the operator realized he had failed to follow the sur-veillanca procedures and had left the discharge valve of the icw head safety injection pump open. He immediately secured the pump.
The Reactor Coolant System was checked for leakage and containment entry was made approximately one hour later.
Inspection revealed that a discharge from the containment spray nozzles had occurred.
A few minutas later power reduction was startad. A second containment entry was made about an hour later, after containment air samples confirmed that a full face mask would provide adequate respiratory protaction for'the levels of radioactivity in the building. A M
detailed inspection revealed no serious deficiencies and no electrical
?] 2 grounds; the power redue:1on was terminated at a power level of 83t.
Although the operator had not followed the procedure and the discharge valve was open, the containment spray header isolation valve (HCV-345)
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. and the low pressure safety in cross-connect valve (HCV-335) jection to containment spray header should have prevented the event. The electric / pneumatic ccnverter on HCV-345 had failed and both red and partially open. green position indication lights are on, indicating the valve wa Operster had takan local control of the valve in an atte completely close the valve.
After about 1/2 inch of stem travel, the position as demanded by the valve positioner. operator remove The third valve ~ (HC/-335) but no cs...ctive action had been taken.in the incident had a The pneumatic relay on valve HCV-345 was replaced and valve HC/-335
.w repaired.
Valve HCV-344 and HC/-345 are now required to be placed in the test mode prior to operating the icw pressure safety injection pump or contain spray pumn for testing.
This mode along with verifi-cation of an annunciator will ensure that both of thage valves.ars in the fully closed position prior to pump operation. '
VALVE MALFUNCTIONS 1.
PMaary System DepressuMzatier.
On September 24, 1977, Davis Bessa Nuclear Power Station Unit
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No.1 experienced a depressurization when a pressurizar power relief valve failed in the open position., The Reactor Coolant System (RCS) pressure was reduced from 2255 psig to 875 psig in approximately twenty-one (21) minutas. At the beginning of this event, steam was being bypassed to the condenser and the reactor themal power was at 253 MW, or 9.5%.
Electricity M
was not being generated. Tha following systems malfunctioned duM ng the transient:
Steam and Feedwater Rupture Control System (SFRCS).
a.
b.. Pressurizar Pilot Actuated Relief Yalve.
No. 2 Steam Genentor Auxiliary Feed Puso Turbine Governor.
c.
The event was initiated at 2134 hours0.0247 days <br />0.593 hours <br />0.00353 weeks <br />8.11987e-4 months <br />, when a spumous " half-trip" occurred in the SFRCS, resulting in closure of the No. 2 Feedwater t
Startup Valve and loss of flow to No. 2 Steam Generator. Approxi-W, mately one minute later, low level in the No'. 2 Steam Generater p
caused a full SFRCS tM p, closing the Main Steam Isolation Valves l
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(MSIV). The loss of heat sink for the reactor caused the RCS l
tamperature, pressure, and pressurizer level to rise.
The RCS pressure increased to the pilot actuated relief valve setpoint (2255 psig) and the valve cycled open and closed nine Meanwhile, the reactor operator observed the pressu increase and manually tripped the reactor about one minuta after MSIV closure (two minutes into the transient).At this point the RCS pressure was approximately 2000 psig and decreasing while the pressurizer level had reached its maximum initial rise of about 310 inches. The RC3 pressure continued to decrease due to the open relief valve and upon reaching 1620 psig approxi-mately three minutes into the transient, actuated Safety Features
. including high pressure (watar) injection and containment isolation.
Approximately five minutes into the transient the rupture disc on the pressuM zer quench tank, which was receiving the RCS blowdown, burst.
by the actuation of containment isolation, which had 1solatedBu the quench tank cooling system, resulting in expedited pressuM-zation of the quench tank.
The RCS continued to blow down through the open pressurizer-power relief valve and the quench tank rupture disc opening until pMaary coolant saturation pressure was reached, about six minutas into the transient.
The femation of steam in the RCS caused an insurge of water into the pressurizar.
This insurge and the high. pressure watar injection then restored pressuMzar level to about 310 inches after nine minutas into g
the trsnsies.t.
Approximately thirteen minutes into the transient, the secondary side of the No. 2 Staam Generator went dry. About fourteen minutes into the transient,. the operators noticed the low level condition and found that the auxiliary feed pump was operating at reduced speed.
Manual control of the auxiliary feed pump was started and water level restored to the No. 2 Steam Generator.
At approximately 27 minutas into the transient, the operators j
discovered that the pressurizar power relief valve was stuck open. Blowdown via this valve was stopped by closing the block valve, thus taminating the reactor vessel depressurization. The l
RCS pressure recovered to normal and cooldown of the systam followed.
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The reason for the spurious " half-trip" of the SFRCS has not yet been detemined.
An extensive investigation revealed several loose connections at taminal boards, but nothing conclusive.
Investigation into the failure of the pressurizar pilot actuated relief valve revealed that a "close" relay was missing frem the control circuit.
This missing relay would nomally provide a.
" seal-in" circuit which would hold the valve open until the pressure dropped to 2205 psig. Without the relay the power relief l
valve cycled open and closed each time the pressure of the RCS went.above or below 2255 psig. The rapid cycling of the valve caused a failure of the pilot valve stas, and this failure caused, the power relief valve to remain open.
It was detarmined that the auxiliary feed pump did not go to full speed because of " binding" in the turbine governor.
The transient was analyzed by the ff533 vendor and datamined to be within the design parameters analyzed for a rapid depressurization.
With exception of the above noted malfunctions, the plant functioned as designed and there was no threat to the health and safety of the general public.,3 2.
Feedwater Isolation Valves On two occasions in July, at tin Tr'ojan nuclear plant, a hydraulic feedwater isolation valve failed to close upon receipt of a close signal.
All other equipment required to operate, functioned nomally.
The first failure Jul.y 6,1977, had been attributed to an improperly assembled solenoid in the hydraulic actuator.
Investigation of the second failure indicated that both events were due to a lack of sufficient hydraulic pressure.
FaiTure of the valve to close was caused by the pressure regulator leaking and failing to close down to regulata the pressure. This caused the hydraulic system on the valve to be drained down to a point that the valve would not operata.
Inspection of the regulator revealed that a locking screw on the regulator adjusting knob was loose and would allow the knob to vibrata to any position.
s With the regulator improperly set it would not close down to I
regulata pressure and would allow the hydraulic fluid to drain before the hydraulic operator could
. as discovered on two other valves, function. A similar probles w
although the maladjustment was not sufficient to prevent these valves frem operating.
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. i All of the regulatcrs were reset and the adjusting knobs were locked in place so that they could not vibrata loose. Ths isolation va adjustments.{ves were tasted satisfactorily following these i
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Off-Gas System Valves At the Oystar. Creek nuclear generating station on August 27, 1977, the reactor building ventilation system isolated and the standby gas treatment systam (SGTS) automatically initiated.
Investigation revealed that at approximately 1850 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.03925e-4 months <br /> a station employee perfonning housekeeping duties in the main control room accidently caused the augmented off gas (ACG) itd4 switch to move from " isolate and bypass" to the "isolata" position. This resultad in the off gas valve and the off gas drain valve going closed, and since the A0G was not in service the gas flow was stopped. The isolation of the reactor building ventilation system and initiation of the SGT3 occurred at 1905. The two off gas valves were opened four minutes latar and the SGT3 was secured. The reactor building ventilation system was returned to normal at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.
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The off gas drain valve did not seat properly and was not leak tight. This condition allowed the gaseous radioactivity within the isolated off gas system pioing to travel *up through the stack sump in the stack base and fill the air space in the ventilation tunnel.
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When the radiation level in the reactor building l
ventilation duct reached a level of 17 mr/hr the monitors located next to this duct initiated the SGT3.
The safety concern Associated with this event is the possibility of a submergence dose a person would have receivef from the radio-active gaseous atmosphers if they were in the tunnel area. The atmosphere in the tunnel area is processed through the radwasta ventilation systam, which contains both roughing and absoluta filters, prior to axhausting through to the stack which is mnitored. The maximum radiation level sensed in the tunnel was 26 mr/hr.
No personnel exposures or reiseses to the environment resulted from this event. The licar.see is investigating the feasibility of installing an alans to alert operations personnel to the closure of the off gas valve when the A0G is out-of-servica.5 O
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SPALL PIPE BREAX ANALYSIS On June 9,1977, an orderly shutdown of the Yankee Nuclear Pcwer Station (Yankee Rowe), a pressurized water reactor, was initiated i
by the licensee because of an error discovered in the Emergency Co Cooling System (ECCS) perfomance analysis.
Yankee Atomic Electric Company (YAEC), the licensee, notified the
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Nuclear Regulatory Ccmaission (NRC) that an error had been discover in a particular small break loss of coolant accident which pemitted reactor operation with Core XII in a m(anner) analysis, LOCA conservative than assumed in the original analysis.
less.
While performing a review of the analyzed small break accidents for t
the Core XIII reload, the YAEC Safety Analysis Group detemined on June 7,1977 that an incorrect fluid flow resistance calculation was made in the safety injection Ifne break analysis.
characteristics study had taken credit for the 2-1/4 inch safety The fluid flow a tank which supplies borated water to the reactor of a reactor coolant system pipe break.
sleeve should not have been includM in the flow calculation, as aThe flo new worst case pipe break was identified in a 4-inch diameter line section.
'The recomputed decreased flow resistance allowed increased accumulato flow to be caTeislated for the break reflood capability of the ECC3. pressure to less than had been a This corrected flow resistance assumption was used for the accident analysis of the present core, Core XII, which was operating at 79% of rated power in a coastdown program prior to the June 9,1.977 shutdown.
Operation of the reactor with Core XII cemenced in December 1975.
Upon discovering the error, the licensee reduced power level to 300 megawatts themal (50% rated power), which was believed to conservatively accannodate the analysis error.
During subsequent analysis, however, the licensee was unable to assure himself that the 10 CFR 50.4 on peak fuel cladding temperature could be mainta'ned for the postulated small break.
Therefore, the faciltty was shutdown pending resolution had been previously scheduled to cannonce on July 2,197 G
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p The licensee subsequently perfomed an approximata best estimata analysis of the postulated worst case small pipe break, which included Core XII operation. assumptions based on actual facility equipment ava The results of this analysis indicated that the calculated peak fuel cladding temperature was well below 10 CR 50.46 limits.
The more conservative 10 CFR 50 Appendix X reanalysis of Core been exceeded in the event that the safety injectio actually occurred.
prior to returning the Core XIII the licansas: plant to operation after refueling of
- 1) performed flow measurements tasts j
to determine the actual flow resistance through the safety injection piping; 2) changed the flow resistanca in the safety injection lines, by an ECCS modification; and 3) analyzed appropMata pipe break i
accidents in accordanca with 10 CFR 50 Appendix X cMtaria.
i The changes and results of tests and analysis were submitted to the NRC and were approved prior to restart of the plant after the refueling.6-7
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DIESEL GENERATOR TRIP i
j During a loss-ofwe test on August 2'6, 1977,*the E-4 diesel of the Peach Bottom Atomic Power Station Unit 2 started properly as a
- i result of the undervaltage condition,'but tripped irmediately.
This tMp was caused by the overspeed mechanism.
The circuitry was reset, an adjustment was made to the mechanical governor to ifmit the diesel speed duMng a start and the unit was started successfully.
Because-the exact cause of tha tMp was not firmly established, surveillance tasting of the diesel was increased from once a week to once per shift.
l During one of these tests, on August 27,1977. the diesel tripped again.
Another adjustment was made to the mechanical governor, the load capability was checked and several successful starts were perfomed.
l Onca per shift surveillanca was continued.
On August 29, 1977, the diesel again tiMpped on overspeed and was declared inoperable.
The diesel was then operated in excess of synchronous speed in order to detarmine the exact speed at which the overspeed mechanism would function.
This test determined that the j
diesel would trip at 940 rpm instead of the desired setpoint of 990 i
The tMp mechanism was adjusted to 985 rpm by,a manufacturer's i
rpm.
representative and diesel was started twice, successfully.
,I Investigation into the cause of the change in the trip setting y
datarmined that during the diesel maintenance in June 1977 a camshaft e
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In order to replace this camshaft the overspeed mechanism had been removed.
When the overspeed mechanism was replaced, seme necessary shims were not installed.
Although this was the only diesel requiring this maintenance during the annual check, the other diesels were operated up to a speed of 945 rpm to verify proper operation.
None of these diesels tripped cn overspeed.
Analysis of this event revealed that a deficiency exists in the maintenance procedure associated with the diesel yearly inspection and the post-maintenance testing procedure. These procedures will be revised. to correct the deficiencies.8 ELECTRICAL FAULT Cn July 13, 1977 while the personnel at James A. Fitzpatrick nuclear power plant were conducting refueling operations a short in a cable caused 600 volts AC to be introduced into a 115 volt circuit.
The 600 volt AC supply for the refueling bridge and the 115 volt AC circuit for refueling interlocks are both located in the same cable.
Flexing Of the cable with bridge motion over the core caused the cable to short internally.
The introduction of the 600 volts into the 115 volt circuit caused ninetewn relays in tNE rod manu'al control systaa to burn out. All of the refueling operations were haltad until the interlocks were repaired. The rod worth minimizer and rod sequence control sysems were also checked for damage.
.A modification is being prepared that will remove the 1T5 volt AC
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interlock circuit from the ca l
This will prevent recurrence.gle carrying the 600 volt AC supply.
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'i PIPE CRACK I-The Brunswick Steam Electric plant Unit 2 was in het shutdown and 1
preparations were underway to startup the unit when the Shift Foreman y) noticed a small leak of the recirculation loop suction piping. This
'l discovery was made during the closecut inspection of the drywil.
Q Investigation revealed the leek was from a crack in the socket weld
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l on a three-quarter inch test connection 900 elbow that was nonisolable and the plant was placed in the cold shutdown condition. The cracked 7
I pipe was cut out of the system and the connection was capped. Similar connections on both Units 1 and 2 were dye-penetrant checked with no other indications of cracks.
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the internal and external diameters of this section of pipe no other cracks.
weld joints showed that a proper gap was present b e
I and the pipe end.
Based on a stress analysis and the observed condition of pema i
deformation of the failed area, along with th in the weld fillet area.
result of workmen (during constructionIt is believed that this deformation ii This use of the pfpe for this purpose p)lus vibrational stres using the pipe as a step.
j in the failure.
l A visuai inspection of similar piping on the other loop of Unit 2 j
and both loops of Unit I revealed no deferination as was observed on the failed pipe.
remaining pipes is such that they are not. likely to be l
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or support because of physical interferences.
i be supported to protect them from experiencing excessThese three pipes leading and vibration, or will be removed and capped.gegternal i
Point of
Contact:
1 Joseph I. McMillen Office of Management Infonnation
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and Program Control U.S. Nuclear Regulatory Consission 4
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_ REFERENCES 1.
LS 77-2, Occket No. 50-285, January 31, 1977.
2.
LER 77-16, Occket No. 50-346, Octcher 7,1977.
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3.
Supplement to LS 77-16, 'Occket No. 50-346, November 14, 1977.
4.
LS 77-23, Occket No. 50-344, July 29,1977.
5.
LS 77-21, Occket No. 50-219, September-23, 1977.
j 6.
LS 77-30, Occket No. 50-29, August 3,1977.
7.
Sumary of June 17 Meeting, NRC-YAEC, June 22,1977.
8.
LER 77-37A Occket No. 50-277, September 9,1977.
9.
LS 77-43, Docket N3. 50-333, August 11, 1977.
i 10.
LS 77-7, Oceket No. 50-324, February 28, 1977.
j 11.
Supplanent to LS 77-7, Occket No. 50-324 Septamber 30, 1977.
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