ML19208C067
| ML19208C067 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/14/1977 |
| From: | Kirk I NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA) |
| To: | NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA) |
| Shared Package | |
| ML19208C060 | List: |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7909240615 | |
| Download: ML19208C067 (12) | |
Text
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ATTACHMENT 1
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NUCLEAR REGULATORY COMMISSION
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Those on Attached List COMPUTER LISTINGS OF LICENSEE EVENT REPORTS SORTED BY FACILITY The enclosed computer listing provides information concerning licensee event reports entered into the file during the month of November.
If you desire additional information or special searches, please do not hesitate to contact us.
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e I". A. Kirk, Acting Director Regulatory Info. Systems Division Office of Management Information and Program Control
Enclosure:
As stated V
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020 09 1977 LER POMIHLY OuiPUT 50RIED 97 FACILITY PAGE 18 PROCESSED OuRif4G NOVEMBER FOR PowfR REACTORS F AC IL IT Y/ 5YST EM/COMPuNENI /
OCCEFT N0./
EVENT DATE/
CDM Pora E N T %USCCLE/CAuSE CODE / ts4 NO./
REPORT DAIE/
EVENT DESCRIPTION /
C405E SuuCOLE/"ANUFActudE4 CON 14GL NO.
REPORI fYPE CauSE DESCRIPil0N DAVIS-dESSE-t 05C00246 010177 INP-51-77-F64 at 1540 HR5 DN 9/t/7F. REACTOR COOLANT PUMP 1-2-1 TRIPPED CCOL ANI dECIRC SYS + CGNTRot 5 FF O lt t004F7 00E TO LOW SEAL INJECfl0N & COMPONENT COOLING WATER FLOW.
IN ACTION STA PurP5 019876 10-uAY IEMENI 0F T.S.
3.4.1, WHICH REQutRES 4 RCPS IN MODE 1.
REACTOR C00LANI teu 50nCOMPONENI PRudlDEO PUMP 1-2-1 RF-5 TARTED At 1524 HOUR $ ON 9/t/F7 PER50aNEL ERROR CauSE SuGCODE '4GT PROVIDio ITEM NOT APPLICauLE PERSONNEL ERROR.
AN OPERATOR MISTAKENLY CLOSED WALVE I A 608 WHICH $UPPL IES INSTRUMENT AIR TO NORMAL MAKEUP FLOW CONTROL VALVE.
04vl5-BE55E-t 03000346 090F77 (NP-ll-F7-748 AT 0810 HR5 ON 9/F/FF, SWITCHING FROM NORM AL TO ALIERNATE C00LANI RECIRC SYS + CONTRut5 FF O lt 100477 POWER SUPPLY FOR DEC AY HE AT SUCTION V ALVE DHt 2 v&LVE CLOSE0. REACTOR 0
((((() VALtES 019447 10-04Y PERAIOR IPMEDIATELY RE-OPENED VALVE & REESTABLISHED FLOW A80VE 2800 GPM.
NO Su8 COMPONENT PR0vl0E0 OESIGN/ FABRICATION ERROR CAuSE Su8000E NOT PHOVIDED ITEM NOT APPLICABLE DEstGN DEFICIENCY IN IN T E RL OC K/ CON T R OL CIRCulf CAUSED THIS DCCURRENCE.
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- [7JEs OAvis-1t55t-1 03000346 09tFFF INP-33-FF-733 Ai 1800 H45 ON 9/t7/77. MAKEUP PUNP t-2 WAS REMOVED FROM
- E REAC COOL CLEANUP Sv5 + CONT FF u it 10127F 5ERvlCE TO ALLOW MAINTEt4ANCE TO REPAIR AN OIL LEAK ON OUT80ARD BEARING E tD PLATE.
IN ACTION STATEMENT OF TECH SPEC 3.1.2.4 SINCE 2 MAKE UP PUM k)U3B PUMP 5 Ot9574 10-04v 9
NO SU8 COMPONENT PROvlDED
- P5 REQutRED IN PODE 3.
p=ene OEstGN/FABRICAlluid ERdOR CAU5E SURCODE NOT PR0v10ED a lteGH AM PUMP CO DESIGre UEFICIENCY.
BUNA-N 0-RING ON REARING FND PLATE DION'T FORM ADEQu ATE SEAL At BEARING HOUSING.
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C 9tvi5-ut55E-l 05000146 0924FF H8LP IM I P OF' S TE AM C FEE 0hAIER RUPIURE CONTROL SYSTEM CAUSED al5E IN RE A OTHR INSI Sv5 NEGO FUA SAfffY FF-Ol6/utt-0 LOOFFF CI0a CHOLANT SYSTEM TEMP C PRES 5URE.
CAUSED PRES $UnilER POWER RELIEF VA
-ShD IN5tRUMENTAllON e CONikOL5 089180 2-wEEh Lv[ 80 HPEN C VALVE FalLED 10 CLOSE. CAUSING REDUCTION IN RCS PRE 55uRE.
l OTHtR LC05 wsRE ExCEE0fD FOR 5 T.S..
3.4.1, 3.4.5, 3.4.6.2 3.6.1.4 C 3.7.1.2 OTHtH pgy NOT APPLICAutt O
CON 50LIDATED CCatROL5 CORP.
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HALF IRIP CON 0lilON FROM 5FRCS CHANNEL 2e WHICH CAUSED VALVE FWSPFA 10 C LOSE.
CAUSE OF THIS HALF TRIP H A5 NOT BEEN P05tilvELY DETERMINED ALTH00 LH EEIEN51vE INVE5flGAfl0N HL5 REVEALED LOOSE CONNECil0NS Af IERMINAL 80 ARDS (POSSIBLE CAuSEl.
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ATTACHMENT 2
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UNITED STATES CURRENT EVENTS NUCLEAR REGULATORY POWER REACTORS commission THIS COMPILATION OF SELECTED EVENTS IS PREPARED TO DISSEMINATE INFORMATION ON OPERATING EXPERIENCE AT NUCLEAR POWER PLANTS IN A TIMELY MANNER AND AS OF A FIXED DATE. THESE EVENTS ARE SELECTED FROM PUBLIC INFORMATION SOURCES. NRC HAS, OR IS TAKING CONTINU0US ACTION ON THESE ISSUES AS APPLICABLE, FROM AN INSPECTION AND ENFORCEMENT, LICENSING AND GENERIC REVIEW STANDPOINT.
W 1SEPTETER-31 OCTOBER 1977 (PUBLISHEDDECEMCER1977)
OPERATOR ERROR L
On January 11, 1977 while the Fort Calhoun Station Unit I was operating, water from the Refueling Water Storage Tank was pumped W
into the containment through the containment spray header due to an operator error.
During the performance of a quarterly test of the safety injection and containment spray pumps, the operator noticed an increase in the containment sump level approximately ten minutes after the low pressure safety injection pump had been started. Approximately 3300 gallons of water had been pumped to the containment. About one minute later the ventilation isolation actuation signal was received.
G At this time the operator realized he had failed to follow the sur-veillance procedures and had left the discharge valve of the low head safety injection pump open. He imediately secured the pump.
The Reactor Coolant System was checked for leakage and containment entry was made approximately one hour later.
Inspection reve: led that a discharge from the containment spray nozzles had occurred.
A few minutes later power reduction was started. A second containment entry was made about an hour later, after containment air samples confirmed that a full face mask would provide adequate respiratory protection for the levels of radioactivity in the building. A detailed inspection revealed no serious deficiencies and no electrical grounds; the power reduction was, terminated at a power level of 83%.
Although the operator had not followed the procedure and the discharge valve was open, the containment spray header isolation valve (HCV-345) n ;;.:
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. 7 and the low pressure safety injection to containment spray header cross-connect valve (HCV-335).should have prevented the event.
The electric / pneumatic converter on HCV-345 had failed and both red and green position indication lights were on, indicating the valve was SENfp' partially open. Prer to the event the auxiliary Building Equipment Operator had taken local control of the valve in an attempt to completely close the valve After about 1/2 inch of stem travel, the operator removed the valve pin and the valve went back to its previous position as demanded by the valve positioner. The third valve (HCV-335) in the incident had a leakage problem that had been previously identified but no corrective action had been taken.
The pneumatic relay on valve HCV-345 was replaced and valve HCV-335 repaired.' Valve HCV-344 and HCV-345 are now required to be placed in the test made prior to operating the low pressure safety injection pump or contain spray pump for testing. This mode along with verifi-cation of an annunciator will ensure that both of thege valves are in the fully closed position prior to pump operation.'
i VALVE MALFUNCTIONS 1.
Primary System Depressurization On September 24, 1977, Davis Besse Nuclear Power Station Unit No. 1 experienced a depressurization when a pressurizer 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) minutes. At the beginning of this event, steam was being bypassed to the condenser and the reactor thermal power was at 263 MW, or 9.5%.
Electricity was not being generated. The following systems malfunctioned during the transient:
a.
Steam and Feedwater Rupture Control System (SFRCS).
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b.
Pressurizer Pilot Actuated Relief Valve.
c.
No. 2 Steam Generator Auxiliary Feed Pump Turbine Governor.
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 spurious " half-trip" occurred in the SFRCS, resulting in closure of the No. 2 Feedwatre Startup Valve and loss of flow to No. 2 Steam Generator. Approxi-mately one minute later, low level in the No. 2 Steam Generator caused a full SFRCS trip, closing the Main Steam Isolation Valves w.
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(MSIV). The loss of heat sink for the reactor caused the RCS temperature, pressure, and pressurizer level to rise.
The RCS pressure, increased to the pilot actuated relief valve EN8 0
- setpoint (2255 psig) and the valve cycled open and closed nine times in rapid succession, failing to close on the tenth opening.
Meanwhile, the reactor operator observed the pressurizer level increase and manually t'*ipped the reactor about one minute after MSIV closure (two minutts into the transient). At this point the RCS pressure was app oximately 2000 psig and decreasing while the pressurizer tel had reached its maximum initial rise of about 310 inck.s.
The RCS pressure continued to decrease due to the open relief valve and upon reaching 1620 psig approxi-EffI- -
mately three minutes into the transient, actuated Safety Features including high pressure (water) injection and containment isolation.
Approximately five minutes into the transient the rupture disc on the pressurizer quench tank, which was receiving the RCS blowdown, burst. Bursting of the rupture disc was aggravated by the actuation of containment isolation, which had isolated 6
the quench tank cooling system, resulting in expedited pressuri-zation of the quench tank.
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The RCS continued to blow down through the open pressurizer power relief valve and the quench tank rupture disc opening until primary coolant saturation pressure was reached, about six miautes into the transient. The formation of steam in t'.e iS caused an insurge of water into the pressurizer.
This r
insurge and the high pressure water injection then restored pressurizer level to about 310 inches after nine minutes into the transient.
Approximately thirteen minutes into the transient, the secondary side of the No. 2 Steam 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 21 minutes into the transient, the operators discovered that tue pressurizer power relief valve was stuck open. Blowdown via this valve was stopped by closing the block valve, thus terminating the reactor vessel depressurization. The RCS pressure recovered to normal and cooldown of the system followed.
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The reason for the spurious " half-trip" of the SFRCS has not yet been determined. An extensive investigation revealed several loose connections at terminal boards, but nothing conclusive.
Investigation into the failure of the pressurizer pilot actuated b'I relief valve revealed that a "close" relay was missing from the control circuit. This missing relay would normally provide a
" seal-in" circuit which would hold the valve open until the pressure dropped to 2205 psig. Without the relay the power relief 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 stem, and this failure. caused the power relief valve to remain open.
g It was deter.nined that the auxiliary feed pump did not go to full speed because of " binding" in the turbine governor.
The transient was analyzed by the NSSS vendor and determined 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.2-3 2.
Feedwater Isolation Valves On two occasions in July, at the Trojan nuclear plart, a hydraulic feedwater isolation valve failed to close upon receipt of a close signal. All other equipment required to operate, functioned normally.
W The first failure, July 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.
Failure of the valve to close was caused by the pressure regulator leaking and failing to close down to regulate the pressure. This caused the hydraulic system ~on the valve to be drained down to a point that the valve would not operate.
Inspection of the regulator revealed that a locking screw on the regulator adjusting knob was loose and would allow the knob to vibrate to any position.
With the regulator improperly set it would not close down to regulate pressure and would allow the hydraulic fluid to drain before the hydraulic operator could function. A similar problem was discovered on two other valves, although the maladjustment was not sufficient to prevent these valves from operating.
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All of the regulators were reset and the adjusting knobs were locked in place so that they could not vibrate loose. The isolation va adjustments.{ves were tested satisfactorily following these g pMEe 3.
Off-Gas System Valves At the Oyster Creek nuclear generating station on August 27, 1977, the reactor building ventilation system isolated and the standby gas treatment system (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 performing housekeeping duties in the main control room g4pp7 accidently caused the augmented off gas (A0G) mode switch to move from " isolate and bypass" to the " isolate" position. This resulted 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 _ven'ilation system and initiation of the SGTS occurred at 1905. The two off gas valves were opened four minutes later and the SGTS 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 />.
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 piping to travel up through the stack sump in the stack base and fill the air space in the ventilation tunnel. When the radiation levei in the reactor building ventilation duct reached a level of 17 mr/hr the monitors located next to this duct initiated the SGTS.
The safety concern associated with this event is the possibility
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of a submergence dose a person would have received from the radio-active gaseous atmosphere if they were in the tunnel area.
The atmosphere in the tunnel area is processed through the radwaste ventilation system, which contains both roughing and absolute filters, prior to exhausting through to the stack which is monitored. The maximum radiation level sensed in the tunnel was 26 mr/hr.
No personnel exposures or releases to the environment resulted from this event. The licensee is investigatirg the feasibility of installing an alarm to alert operations personnel to the closure of the off gas valve when the A0G is out-of-service.5 e
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SMALL PIPE BREAK ANALYSIS On June 9,1977, an orderly shutdown of the Yankee Nuclear Power Station (Yankee Rowe), a pressurized water reactor, was initiated by the licensee becattse of an error discovered in the Emergency Core Cooling System (ECCS) performance analysis.
Yankee Atomic Electric Company (YAEC), the licensee, notified the Nuclear Regulatory Comission (NRC) that an error had been discovered in a particular small break loss of coolant accident (LOCA) analysis, which permitted reactor operation with Core XII in a manner less conservative than assumed in the original analysis.
While performing a review of the analyzed small break accidents for M~
the Core XIII reload, the YAEC Safety Analysis Group determined on June 7,1977 that an incorrect fluid flow resistance calculation was made in the safety injection line break analysis. The fluid flow characteristics study had taken credit for the 2-1/4 inch safety injection line thermal sleeve to retard spillage from the accumulator --
a tank which supplies borated water to the reactor core in the event of a reactor coolant system pipe break.
The flow resistance of the sleeve should not have been included in the flow calculation, as a W,
new worst case pipe break was identified in a 4-inch diameter line section.
The recomputed decreased flow resistance allowed increased accumulator flow to be calculated for the break, and decreased the ECCS supply pressure to less than had been assumed, thus decreasing the core reflood capability of the ECCS. 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 gq program prior to the June 9,1977 shutdown. Operation of the reactor with Core XII comenced in December 1975.
Upon discovering the error, the licensee reduced po.ter level to 300 niegawatts thermal (50% rated power), which was believed to conservatively accommodate the analysis error.
During subsequcnt analysis, however, the licensee was unable to assure himself that the 10 CFR 50.46 limits on peak fuel cicdding temperature could be maintained for the postulated small break. Therefore, the facility was shutdown pending resolution of this matter and to proceed with the Core XIII refueling outage which had been previously scheduled to commence on July 2, 1977.
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f The licensee subsequently performed an approximate best estimate analysis of the postulated worst tase small pipe break, which included assumptions based on actual facility equipment availability during Core XII operation. The results of this analysis indicated that the
- @er calculated peak fuel c? adding temperature was well below 10 CFR 50.46 limits. The more conservative 10 CFR 50 Appendix K reanalysis of Core XII operation, however, indicated that 10 CFR 50.46 limits might have been exceeded in the event that.the safety injection pipe break had actually occurred.
Prior to returning the plant to operation after refueling of Core XIII the licensee: 1) performed flow measurements tests to determine the actual flow resistance through the safety injection pa piping; 2) changed the flow resistance in the safety injection lines, by an ECCS modification; and 3) analyzed appropriate pipe break accidents in accordance with 10 CFR 50 Appendix K criteria. 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 DIESEL GENERATOR TRIP 54*m During a loss-of-power test on August 26, 1977, the E-4 diesel of the Peach Bottom Atomic Power Station Unit 2 started properly as a result of the undervoltage condition, but tripped immediately.
This trip was caused by the overspeed mechanism. The circuitry was reset, an adjustment was made to the mechanical governor to limit the diesel speed during a start and the unit was started successfully. Because the exact cause of the trip was not firmly established, surveillance testing of the diesel was increased from once a week to once per shift.
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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 performed.
Once per shift surveillance was continued.
On August 29, 1977, the diesel again tripped on overspeed and was declared inoperable. The diesel was then operated in excess of synchronous speed in order to determine the exact speed at which the overspeed mechanism would function. This test determined that the diesel would trip at 940 rpm instead of the desired setpoint of 990 rpm.
The trip mechanism was adjusted to 985 rpm by a manufacturer's representative and diesel was started twice, successfully.
Investigation into the cause of the change in the trip setting determined that during the diesel maintenance in June 1977 a camshaft M
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In order to replace this camshaft the overspeed mechanism had been removed. When the overspeed mechanism was replaced, some necessary shims were not installed. Although this was the only diesel requiring this maintenance during the annual check, the other diesels N'
were operated up to a speed of 945 rpm to verify proper operation.
None of these diesels tripped on overspeed.
Analysis of this event revealed that a deficiency exists in the mainttnance procedure associated with the diesel yearly inspection and the post-maintenance testing procedure. These procedures will be revised to correct the deficiencies.8 LI W ELECTRICAL FAULT On 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 l'15 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 W-volt circuit caused nineteen relays in the rod manual control system to burn out. All of the refueling operations were halted until the interlocks were repaired. The rod worth minimizer and rod sequence control systems were also checked for damage.
A modification is being prepared that will remove the 115 volt AC interlock circuit from the cable carrying the 600 volt AC supply.
This will prevent r.ecurrence.9 YN PIPE CRACK L
The Brunswick Steam Electric Plant Unit 2 was in hot shutdown and preparations were underway to startup the unit when the Shift Foreman noticed a small leak of the recirculation loop suction piping.
This discovery was made during the closecut inspection of the drywell.
Investigation revealed the leak was from a crack in the socket weld on a three-quarter inch test connection 900 elbow that was nonisolable, and the plant was placed in the cold shutdown condition. The cracked 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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Further investigation revealed that the crack was contained in the weld metal and intergranular stress corrosion in the heat affected zone of the base metal was ruled out. A dye-penetrant inspection of the internal and external diameters of this section of pipe revealed M
no other cracks. The inspection of the internal diameter of the socket weld joints showed that a proper gap was present between the socket and the pipe end.
Based on a stress analysis and the observed condition of permanent deformation of the failed area, along with the location of the crack, it is concluded that the initial crack was caused by stress concentration in the weld fillet area.
It is believed that this deformation was the result of workmen (during construction) using the pipe as a step.
4 This use of the pipe for this purpose plus vibrational stress resulted in the failure.
A visual inspection of similar piping on the other loop of Unit 2 and both loops of Unit I revealed no deformation as was observed on the failed pipe.
It was also noted that the location of the thr.e remaining pipes is such that they are not likely to be used as a s*,ep or support because of physical interferences. These three pipes vill be supported to protect them from experiencing excessgxternal loading and vibration, or will be removed and capped.
Point of
Contact:
Joseph I. McMillen Office of Management Information and Program Control U.S. Nuclear Regulatory Comission Ms.
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REFERENCES 1.
LER 77-2, Docket No. 50-285, January 31, 1977.
2.
LER 77-16, Docket No. 50-346, October 7, 1977.
3.
Supplement to LER 77-16, Docket No. 50-346, November 14, 1977.
4.
LER 77-23, Docket Mo. 50-344, July 29, 1977.
5.
LER 77-21, Docket No. 50-219, September 23, 1977.
6.
LER 77-30, Docket No. 50-29, August 3, 1977.
7.
Summary of June 17 Meeting, NRC-YAEC, June 22, 1977.
8.
LER 77-37A, Docket No. 50-277, September 9, 1977.
- ff2s 9.
LER 77-43, Docket No. 50-333, August 11, 1977.
- 10. LER 77-7, Docket No. 50-324, February 28, 1977.
- 11. Supplement to LER 77-7, Docket No. 50-324, September 30, 1977.
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'T ATTACHMENT 3 TOLEDO G amaw October 7, 1977 L77-312 FILE: RR.2 (NP-32-77-16)
Docket No. 50-146 License No. NPF-3 Mr. James G. Keppler Regional Director, Region III Office of Inspection and Enforcement U. S. Nuclear Regulatory Cocaission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
Reportable Occurrence NP-32 '7-16 Davis-Besse Nuclear Power Station Unit 1 Date of Occurrence:
Seotember 24, 1977
.nclosed find three copies of Licensee Event Report NP-32-77-16 with a supple-mental information sheet, which is being submitted in accordance with Technical Specification 6.9 to provide 14 day written notification of the subject occur-rence.
Yours truly, i
J /re tack Evans Station Superintendent Davis-Besse Nuclear Power Station JGE/JRL/ljk Enclosures cc:
Dr. Ernst Volgenau, Director Office of Inspection and Enforcement Encl: 40 copies Licensee Event Report 40 copies Supplemental Information Sheet Mr. William G. Mcdonald, Director Of fice of Management Q
Infemacion ind " w ram Control hO h b I
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3 copies E.usce Event Report 3 copie:s.+yplemental Information Sheet 2 copies Telecepted Report j jj
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gI causing a rise in Reactor Coolant System (RCS) temperatitie and pressure. t g LThis resulted in the pressy izer power relief valve to open and this i
o s i valve failed to close, causin6 a reduction in p'.S pressure.
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parj r Conditions Foc Operation were exceeded for five Technical Specifications,,
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3.4.1, 3.4.5, 3.4.6.2, 3.6.1.4 and 3.7.1.2.
(NP-32-77-16)
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ev <r ets:= erie. a.c c:.=asen. r :.:.s (:y i #O The cause of the occurrence was a half trip condition from SFRCS i
y ; Channel 2, which caused valve FUS,P7A to,close. The cause of this half a r 7-j, trip has not been positively datermined although extensive investigation,
L 1.2j has revealed loose connections at terminal boards (possible cause).
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,u e J November 14, 1977 L77-380 FILE:
RR.2 (NP-32-77-16)
Docket No. 50-346 License No. NPF-3 Mr. James G. Keppler Regional Director, Region III Of fice of Inspection and Enforcement U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
Supplement to Reportable Occurrence NP-32-77-16 Davis-Besse Nuclear Po'.ier Station Unit 1 Date of Occurrence:
September 24, 1977 Enclosed find three copies of Licensee Event Report NP-32-77-16 Supplement, which is being submir.ted in accordance with Technical Specification 6.9 to provide addi-tional information of the subject occurrence.
Please note this report also satisfies the special 90 day report requirement of Technical Specification 6.9.2 for the Emergency Core Cooling Actuati n on Septem-ber 24, 1977.
Yours truly, Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Statich TDM/JRL/ljk M
,.i, ra Enclosures N
G cc:
Dr. Ernst Volgenau, Director
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- 40 copies Supplement
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'tr. '.Jilliam G. Mcdonald, Director Of fice of !!anagement O
v Informacien and Program Control t DUPLICATE DOCUMENT Enc 1:
3 copies Supplement Entire document previously entered l into system under:
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No. of pages:.
67 s4 101,1 93o"
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RETu".ilTO ElillATORY CDfTRAL FG oFWtill e
R0kl0L MEMORANDUM FOR:
L. V. Gossick, Executive Director for Operations FROM:
G. R. Mazetis, Section Leader, Systems Group Bulletins & Orders Task Force
SUBJECT:
CORRESPONDENCE TO THE PRESIDENTIAL COMMISSION ON TMI A memo from H. Berkow dated July 11, 1979 requested that information copies of material requested by the Presidential Comission on THI should be sent to EDO. Enclosed find a copy of material requested by an investigator on August 3, 1979.
Included are:
1.
Memo Ma:etis to Crutchfield dated 05/21/79 on TMI record of actions.
2.
Memo Ross to Seyfrit dated 10/20/77.
3.
Personal notes at TMI.
4.
Personal notes at 0B-1.
5.
TECO letter dated 05/18/79.
6.
Handwritten trip report on DB-1 09/24/77 event.
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Gerald R. ' a: is, Section Leader Systems Group Bulletins & Orders Task Force
Enclosures:
As Stated cc:
D. Ross T. Novak S. Israel L
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WASHINGTON, D. C. 20555 5^
N Docket No. 50-320 MEMORANDUM FOR:
D. Crutchfield, Program Support Staff, NRR FRCM:
G. R. Mazetis, Section Leader, Reactor Systems Branch, DSS
SUBJECT:
RECORD OF ACTIONS, TMI-2 By memorandum dated 04/24/79, H. Denton requested a record of personnel engaged in the NRC response to the TMI accident. Since I am not certain that my actions and/or communications are logged in the IRC, the enclosed is a summary (to the best of my recollection) of tha period 03/29/79 through 04/01/79 with personal notes attached.
Gerald R. Mazetis, Section Leader Reactor Systems Branch.
Division of Systems Safety
Enclosure:
As Stated cc:
D. Bunch, P-433
Contact:
Jerry Mazetis, NRR
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ANo No. of pages:
/O 1014 221