ML20024B205

Reviews Current Events - Power Reactors, Jun-Jul 1975.No Action Required

ML20024B205
Person / Time
Site:	Crane
Issue date:	10/23/1975
From:	Toole R GENERAL PUBLIC UTILITIES CORP.
To:	GENERAL PUBLIC UTILITIES CORP.
References
TASK-*, TASK-GB GPU-2482, NUDOCS 8307070323
Download: ML20024B205 (14)

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.. ~.. .-.. 1 ~~ AEC CCCU"ENT REVIEN ^ GL}Q r ~" Plant / Uni t M Tfie attached AEC document has been reviewed for test program and design modification requirements for the above Plant / Unit. DOCUMENT: Operating Experience, dated: l $l Current Events - Power Reactors, dated: 3 v v e d o fv l'f M 1 Other , dated: Review of the attached document has concluded that no action t's required. n ll /V W St rtu esit Manager , A) ate / \\W) to -e s -7c TestSprintencenc. Date Review of the attached document has concluded that action is required by: Problem Report (s) has/have been issued. Startup & Test Manager Date W 06S48 Test Superintencent Date DISTRIBUTION: R.W. Heward, Jr. W.T. Gunn E.D. McDevitt J.E. Kunkel M.A. Nelson R.J. Toole J.T. Faulkner File +M urJb l ~ ~ ~ ~ ~ ~ ~

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$$$$$?_Y N_S wmu= $W$5 $WSi Yh m t cr HE UNITED STATES N CURRENT EVENTS NUCl. EAR l W REGUI.ATORY l E POWER REACTORS C0"33'oN w@ ~ l Ik zvzNrs sztzctra nox arPoRIS snaIrizo to ruz uNItzn srArzS xcI.zAR h 'azcm.Arcar cCMMISSION .s JUNE-JULY 1975 s.N Mk DEFICIENCIES IN DESIGN OF PRIMARY SHIII. DING ~ e _ - > During two recent reactor power ascenaion programs, at a pressurized water sa reactor (?WR) and a boiling water reactor (BWR) facility, radiation in g certain areas was in excess of predicted or design values. Each plant involved different architect-engineers and a different type of radiation. l FitrPatrick 'I j During dryvell entry with tha James A. Fit: Patrick Nuclear Power Plant at 1.3% power, unexpectedly hi;;h radioactivity levels were discovered i at upper elevations in the dryvell. Levels from 5 res/hr neutron to 20 l rem /hr ga==a vere seasured near the reactor vessel level reference leg 1 piping penetration. Investigatio'd of the biological shield around the j reactor vessel disclosed hydro $enous material had not been used for shielding around two instru= enc penetration inspection doors and six reactor vessel veld inspection doors. { 3 It was determined that nine-inches of "Per=ali" (809/f t ) and four-inches of barated (3 weight) "Per=ali" would be required in addition to nine-inches of steel at the inspection doors. Also, long range plans g have been made to shield the recirculation pu=p suction penetrations, the jet pu=p penetrations and the area about the containment spray header. s During an NRC inspection, it was observed that seven of the doors in the biological shielding would, upon opening, strike dryvell piping. Ihese doors provide access to the jet pu=p supply vessel noz:les, and the doors could strike the reactor water cleanup return line or two of the = main steam lines. These pipes are considered to be critical as a rupture j l could lead to a loss of coolant accident. The doors.vich additional shielding to be added will be pinned shut with hardware designed to withstand the pressure transient of a recirculation line break. The doors to six penetrations at the top of the biological shield are to be removed.1 W 06349 i i

. -. ~. I gh - Q-9q$D $Q - - ~ - - Li W W.. k.,_k. h."_.h iN @W G MAEF y5 $4p"N. [h N.y k g n-c n R - WC M .9:w.M1A nr.c,*- W W.5.E rf N @/l 5.9-ra.sww - t l l c: ..~. M w$ Calvert Cliffs-1 When reactor power was increased into che power range (greater than 2% i power) for the first ti=4 the Calvert Cliffs Nuclear Power Plant, Unit 1 at excessive radiation levels '(approx 1=accly the sa=e as predicted for 100: =. power) were noted on the outside surfaces of the containment structure. In addition, if radiation levels taken at 20% power were extrapolated to 100% power, all nor= ally accessible areas inside contain=ent would have 7 had greater-than-design radiation levels. The high radiation levels were caused by neutron and ga==a stres=ing we from an annulus between the reactor vessel flange and the pri=ary shield wall and the annulus around the reactor coolant piping where it penetrates the pri=ary shield wall. To a lesser extent, radiation was g also =easured at the access opening at the base of the pri=ary shield. 9 The reactor vessel primary' shield annulus is approx 1=ately 2.5 ft. ^ wide, and radiation in the vessel cavity, scattered by the vessel vall N and pri=ary shield concrete, was strea=ing out of the large gap. Subsequent M scattering and direct penetration by the streaming radiation contributed to the high radiation level at the 69.-fc. elevation level outside the secondary shield and at the equip =ent hatch. y5 Radiation strea=ing also occurred fro = large openings in the pri=ary i y shield for the six reactor coolant piping-no::le connections. 3Q These openings are conical i= cross-section with an insulation-air gap of j g approx 1=ately 10-inches at the inside surface and 24-inches at the outer surface. The highest radiation levels outside the pri=ary shield pere j @ in the vicinity of the cold legs at the discharge of the reactor coolant j y pu=ps. y l f A high radiation level near the bottom of the pr1~ary shield was caused i J by a 2.5 ft. square personnel access opening that extends through the ! 5 pri=ary shield into the reactor cavity. ! i The access hole is sealed at che inside by a steel door which did not provide a significant a=ount of 5 5 shielding. 43 $?$ Te=porary shielding installed outside the equip =ent hatch reduced radiation levels to less than 0.5 =res/hr at 20% power. Restrictions on personnel access to so=e areas and to the contain=ent structure have been instituted to sini=ize personnel radiation exposure. Te=porary shielding was also installed above the gap between the reactor vessel flange and the ped =ary shield wall and at the personnel access opening at the botto= of the pri=ary shield. Bagged crystalline boric acid (H 30 ) was stacked on the support grating which spans the 2.5 fc. 3 3 gap between the primary shield at the reactor vessel flange. The addition of this shielding reduced ridiation levels in this area by a factor of h. 50-100. The reactor coolant piping no::la shield was insulated with E rectangular sections of polyethylene installed around the reactor coolant g pipes just outside the pri=ary shield walls. 1E W 06S50 E t

i* g ^ .,ss 3-l s 1 l Action was iniciated to design a per=anent shield for the area between { che reactor vessel and pri=ary shield and for the reactor coolant piping no: les.2 sj DAMAGE TO FUEL ASSEMBLIES Hu=boldt Bay During transfer of an irradiated fuel assembly from the transfer basket 3i position in the spent fuel pool at Unic 3 of the Hu=boldt 3ay Power Plant ? co a pool storaga location, the fuel asse=bly was disengaged from the fuel grapple and fell approxi=ately six feet to the spent fuel pool 3 floor. It then tipped over and fell into the ten-foot deep spent fuel ? cask pit in the corner of the pool. An air sample was nor=al, and the grapple was examined and found to be 't functioning properly. It was concluded the fuel assembly had not been grappled properly or properly checked prior to sove=ent of >the fuel 'l bundle. The only da= age to the fuel assembly was that the channel had been forced down over the fuel bundle nose piece and was split in at -l 1 east two corners frem the channel bottom for about eight to ten-inches. a Two days later, when acce=pt was made to recover the fuel asse=bly from the cask pit, and as the fuel assembly was lifted toward the vertical position, the channel came off and fuel rods fell out of the bundle. The remaining portion of the bundle was lowered and the refueling building was evacuated until an air sample showed no abnor=al airborne concentrations. The tie rods and/or tie rod keepers apparently had sheared during the drop, allowing the bundle to separate. It was planned to recover the fuel assembly after the refueling outage at which ctse a complete examination of the fuel bundle would be completed. There were no persennel exposures, injuries or off-site consequences as a result of this event.3 Turkey point-4 1 During refueling at Turkey Point Statics, Unit 4, an observer noted damage to the side of a fuel assembly as it was being lowered to the reactor core. Containment air particulate, gaseous radioactivity and area radiation detectors showed only background radiation levels. The first grid above the bottom nozzle of the fuel assembly was damaged, and the seventh and eighth fuel rods frem the southwest corner of the fuel assembly were distorted. These two fuel rods had been pushed back and out of line with other fuel rods in the outside row. There was no evife me of breach of fuel cladding integrity. However, damage to the gri; a 4 deformation of fuel rods made this assembly unacceptable for furtuer use in the reactor core. W C6951 I 3 --E --g T% 'Y T g .m___. _ __ _ _ _____.__ _ _ _ _ _ _ _ _

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M5 r,m. h e Apparently., when the spent fuel pit (SFP) side lifting fra=e was upended, h the lifting frame struck the fuel asse=bly and pushed it 2 P,/ww frs=e pulley =ounted on the west wall of the STP transfer canal.into the lifting fJf The h location of the pulley was. consistent with da= age to the fuel asse=bly. sm The licensee concluded that procedural deficiencies were the cause of tha occurrence; procedures did not specify the fuel asse=bly =ust be

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lifted to the " full-up" position by the SFP bridge crane before the SFP Bd bridge was =oved fro = the SFP rack position. Procedures did not M specify M that fuel assemblies =nst not be coved over the SFP side lifting frame area until the lifting frame had been upended and ready to receive a fuel assembly. The procedures have been revised accordingly. Q There were no injuries to personnel and no exposure of personnel to g*A radiation or concentrations of radioactive =acerial as a result of this h occurrence, and it was concluded neither reactor safety no'r the health and safety of the public were jeopardized. i: Cuad-Cities-2 l A series of Local Power Range Monitor (LPRM) high alar =s occurred during power ascension of Unit 2 of the Quad-Cities Nuclear Power Station follevin:i = a forced outage. The unit nuclear engineer was informed by celephone of g these alar =s, and each c1=e reco== ended rod position changes that cleared h the LPRM alar =s. After two changes of red configurations, a high offgas alar = indicative of possible fuel-da= age was received and plant load was decreased. The increased offgas was a result of failure of the fuel cladding. 4 The =axi=u= release rate was est1= aced to be 1.5 C1/see, a factor of four higher than the steady state race before the fuel clad failure. .7 ( The cause of the occurrence was a co=bination of in-sequence red patterns J hj that produced abnor:sily high peaking at the botto= of the core because of a low xenon condition followed by a power increase on flow. D The net result was a local power level increase at a rate that would not allow d fuel pellet cladding stresses to relax without cladding failure. Operating tl i{; d personnel could have =ini=1:ed the damage had they more thoroughly understood reactor conditions and inserted enough rods to ce=pletely I1 4 clear the high peaking. Fuel failure frc= rapid power increases have been experienced at both Quad-Cities and Dresden stations as a result of red withdrawal errors. Although differences exist in the circu= stances of these incidents, the cu=ulative experience indicates that significant if the local rate of power increase is excessive. fuel failure =ay result W CGS 52 m f O

I $ @R$0ihb.$5k-h[I?DWYYi$h& ..$(5? $Yh N%'sNWJEND f= bYN. f? $$ Y$ pS$$MNZ W l INAf-- 1 M5h 1 Wys 3;t*.A* W. DeseaW l i$$ To prevent recurrence, the most recent =aximu= power distribution data vill gg be provided to the reactor operator for reference during startup to RM-serve as a guide in deter =ining if the previous cycle =ax1=u= local power Fcs densities are being approached too soon. Written instructions frc= the 5._ nuclear engineer vill be approved by the operating engineer and included f in the daily log for rod =aneuvers which have potential of exceeding the d previous c:ax1=u= power densities. Increased efforts will be =ade to = ore ~ accurately deter =ine when con:rol rod sequence will require =odification to stay within the previous power envelope, especially on xenon-deficient startups. Consideration will also be given to the possibility of reduced rates of power ascension or power soaks following outages of 24 hours or = ore in order to allow buildup of a larger xenon inventory during non-e=ergency load conditions.5 b Surrv-2 During inspection of fuel asse=blies for the first refueling operation of Unit No. 2 of the Surry Power Station, gas bubbles were noted coming fro = one of the outer fuel rods, and it was established the cladding was perforated by local hydriding. The defect area was approxi=ately O.1 inches in dia=eter, with the surrounding hydride area being approx 1:acely 0.25 inches in dia=ecer. Reactor coolant activity levels during the first cycle had showed a 4 j slight increase, typical of a few failed rods, about two =enths after the j beginning of operation. 2 f Westinghouse Electric' Corporation concluded the fuel asse=bly could be operated through its end of cycle 2 design burnup as scheduled. A review of previous operating experience in other reactors with known .i defected fuel revealed no evidence of propagation of s1=iliar failures, and there was no evidence that si=iliar perforations led to gross failure of an effected red. I 3l A review of Quality Control / Quality Assurance records of the fuel asse=bly revealed no deviations or discrepancies that contributed to this defect, and inspection of other fuel asse=blies revealed no other defected fuel rods.* I I Also at Surry 2, when a fuel asse=bly was being re=oved frc= its core location, two adjacent locking fingers on the fuel handling crane failed to engage the top no::le, so the fuel asse=bly was supported only by the re=aining two fingers. When the asse=bly was pulled clear of the core, it was free to pivoc about the axis for=ed by the two engaged fingers. Coolant flow, caused the lower end of the fuel asse=bly to drif: and to bind in the crane =ast. The binding caused an increase in lead, and the crane operator ceased fuel withdrawal before an overload condition was reached. W C6953 A ...-emumh+

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$.m _ _ b b[N ---%n WI? Wai a * ;& [ Unaware of the cause of increased load, the crane operator lowered the fuel assembly. However, the bottom of the fuel assembly had now drifed _&~ over another fuel asse bly.in a different core location, and the bottom a nozzle ca=e to rest on the top not:1e of the stationary fuel asse=bly. Upon contact, the crane operator noted a decreased in lead and stopped the crane. Two of the bottom pedestal feet of the fuel asse=bly were partially resting on the hold-down springs of the stationary fuel asse=bly. Subsequent inspections established the hold-down springs of the stationary fuel assembly had been plastically defor=ed with a per:anent set of 1.0 and 1.1 inches, respectively. No additional da= age to either fuel asse=bly was observed. As a result of a Westinghouse analysis of the possiblity of, adverse consequences of continued operation with the da= aged in-place fuel ) assembly, it was concluded that the re=aining hold-down capability was j !) adequate to prevent the assembly from lifting off the lower core plate during normal power operation. In addition, it was concluded that a p postulated reactor coolant pump overspeed transient condition of 110 f percent er less would not lift the assembly to the extent that further plastic defor=ation of the hold-down springs would occur. 3 The refueling operation was completed and the reactor was returned to service.7 s PERSONNEI. ERRORS Dresden-2 With Unic No. 2 of the Dresden Nuclear Power Station at S5~ power, it was discovered during a =ain steam isolation valve (MSIV) c4 Mg surveillance test that the reactor protection relay was de-energi:ed. This relay is energized by the "<10~ closure" limit switches on MSIV's LC and 2C. Investigation of the 1C limit switch revealed the internal workings of the switch were missing. s During a refueling outage, all MSIV limit switches had been re=oved for inspection and cleaning. Inadvertently, the 1C limit switch was never reinstalled. When the work package for limit switch =aintenance was subsequently reviewed, all signatures were present. The safety-related work request package had required a =aintenance functional test, si two post-naintenance operational tests, MSIV 10% closure tests and a MSIV closure timing test. 1 Absence of the limit switch si=ulated a " fail safe" condition toward a full scram. Therefore, the health and safety of plant personnel and the - l public.were not jeopardized as a result of this occurrence. .W C6954 t ~~. ~ ,em,ses..-.mem. we..,, -.,

. ~. . ~... - -... ~ _... ~... - .. ~...... ,..u... wh.. j . l The unit was operated for four days without the limit switch in the circuit. l Wen the unit was shut down, the svitch was installed and proper operation verified.3 Oconee-2 A quench tank lov level alarm was received in the Oconee Nuclear Station ]. Unit 2 control room with the plant at 100:: power. The alarm was acknow-e ledged. Approximately 20 minutes later, the control operator observed / a low quench tank level of 40-inches. Corrective action was taken and nor=al quench tank level was restored 45 minutes after the initial alar =. Immediately prior to this incident,_the alarm next to the quench tank { low level alarm had been alar =ing intermittent 1v. The ew " ror %=rd \\ iihe aL g u n or cne quenca tank alarm. looked e

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thaum -M s alarm was cne inter =1trent alarm again. The apparent cause of this event was improper identification of an alarm because of proximity of alars panels. j i This incident did not affect the safe operation of'the unit, and the health and safety of the public was not endangered. j / Personnel involved in this incident were reminded of the imoortance of considering each alarm as a new and different alarm.9 Dresden-1 An operator inadvertently started a wrong pump and transfered liquid radwaste from the resin vault to the radwaste contractor's creat=ent facility at Unit 1 of the Dresden Nuclear Power Station. The licensee did not become aware of the event until two days later (contractor persennel did not work over the weekend). The pump ran until the resin vault emptied and liquid waste overflowed the contractor's tank to the ground. An estimated 19,500 gallons overflowed the tank. Surveys of soil samples indicate the surface water overflow was confined within the reactor plant boundaries. The licensee planned to remove.and dispose of approxi=acely 1,000 cubic yards of so11.10 Maine Yankee W CSS 55 During startup of the Maine Yankee Atomic Power Planc, all three safety injection tank motor-operated isolation valves were not opened during plant heacup. These valves are part of the Emergency Core Cooling System (ECCS) valve checklist, and this checklist is to be completed prior to reaching a reactor coolant system temperature and pressure of 210*F and 400 psig. r

_. ~. _ _ _. _... _ _ b 7 [ l f' I Q _A A sG' Iti; fm m W. D'E If The condition of safety injection isolation was detected by the control N room operator during routine review of =ain control board valve position indicators. At this ti=e, the reactor coolant systes was at 370*F and g 700 psig. eb Personnel responsible for completing the checklist had noted the locked handwheels, but sistakenly assumed the values to be locked open when, in fact, they were locked closed. The three safety injection tank cocor-operated isolation valves were i==ediately opened. The entire n ECCS valve checklist was recompleted with no further discrepancias noted. The ECCS valve checklist has been revised to require independent check of =ain control board ECCS valve positions prior to exceeding 210*F and 400 psig. Review of this incident by plant personnel Isd to the conclusion the incident presented no significant health or safety hazard 1, l to the general public.ll Tuikev Point-4 1 ~ Unic No. 4 of the Turkey Point Station was being returned to service after shutdown, and reactor heatup was in progress. A quality control e inspector discovered a disconnected mechanical linkage on the equali:ing valve for the outer door of the personnel airlock. Further investigation showed the valve to be in the open position and establishing a flow path from containment to at=osphere whenever the airlock inner door was open. Containment integrity had been administratively verified prior to plant startup by completion of the prestart check-off list. However, personnel conducting the checkoff were not aware the linxage was disconnected because both valve and linkage were hidden from view behind a vertical steel 1 y cover place. The valve operating handle and position indicators were 8 visable in front of the steel place, so, personnel thought they were checking the valve shut when, in fact, it was staying open.

i A checkpoint was added to the appropriate check-off list to require verification that mechanical linkages for the airlock inner and outer door equalizing valves were connected.

Breach of contain=ent integrity occurred for only a brief period when the inner airlock door was opened concurrent with =cvement of fuel inside containment or heacup of the reactor coolant system above 200*F. The flow path through the two-inch equalizing valve permitted only a small amount of air flow from contain=ent to at=osphere. Therefore, the health and safety of the public were not adversely affected.12 ~ W 06956 i g 4

/ 9-4 FAILID LIGHT SUL3 PREVENTS DIESEL FROM STARTING } With the Yankee Nuclear Power Station at full power operation, the No. 3 Diesel Generator DC control circuit pilot light located outside

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the diesel cubicle was observed to be out. It was fused to its socket. The pilot light located on the control room diesel panel also was out, and an atte=pt to start tha diesel failed. The apparent cause of bulb failures and the failure of the diesel to start was a short circuit within the pilot light, resulting in a blown fuse in the DC control circuit. The defective pilot light and holder were replaced ard new fuses were installed in the DC control circuit. The diesel was successfully test run. I==ediately afcar the failure of the diesel generator to start, two redundant diesel generators, were started and run for 5 minutes to verify their operability. Hence, this event did not jeopardi:e the health and safety of the public.H ABNORMAL DEPRISSURIZATION OF PRIMARY SYSTEM AND RELEASE OF GASECUS ACTI7ITY Unit 1 of the Zica Station was in a hot shutdown condition, and valves of the excess letdown system were being lined-up to service a relief valve to the pressurizer relief tank (PRT) that had been weeping. When the reactor coolant drain valve was opened to establish excess letdown flow, high seal water flow and high outlet temperature indications were noted on two reactor coolant pumps. Reactor vessel flange leakoff temperature increase (. rapidly, and pressurizer level and standpipe level alarms were received from three reactor coolant pumps. The licensee thought a pump seal had blown, so safety injection was manually initiated and the reactor coolant pumps were deenergized. Closure of the reactor coolant loop isolation valves caused al=ost 1::snediate stabilization of reactor systems. In the twenty minute excursion, reactor pressure decreased from 2235 to 1560 psig, containment pressure reached approximately 1 psig and a containment humidity change of 10:: was noted on one detector. Approx 1=ately 3 to 4 inches of coolant had accu =ulated on the containment floor, and the rupture disc (100 psi burst pressure) on the pressurizer relief tank ruptured. A manual drain valve in one loop of the reactor coolant system that had been inadvertently left in the open position. This caused the abnor=al conditions and leakage. W C6957 ,w .. ~ em+ '+-N

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-m g ?; hM38b'Ihs' n~ amy n --r N i ,t uwh VEAT w. YM m 2D n;e n'd The position of the valve had not been listed in the abnor [h^ line-up when opened, or added as a temporary change to the d

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F <.i reclosed. After the loop was refilled, the valve was not rain or to a. c M Release of gaseous activity originated from the auxiliar '~~ from the contain=ent sump to the auxiliary building flo y pumped tank. of safety injection.The contain=ent su=p valves had not been repositione rain analysis a Ed o reset The maxi =us release rate was calculated to be 69 000 uC1/ Specification limit is 60,000 uCi/sec) for a total relea ec (Technical 0.5 curies. The safety of the public was not endangered because of se of less than short release duration, the direction of release, the short half lif the total magnitude of the release. N e and Permanent ci.2,ees have b'een =ade for the check-list of v l preclude recurrence of chesa events. W,15 reset of safety inj a ves for o C 9

ri Point of

Contact:

apr% y Theodore C. Cintula !"g 5 Office of Management Information and Program Control R U. S. Nuclear Regulatory Co= mission m h,2 E nc.. o 1 .i I E m ,,,an en 0 tz a 1 5 a s. Vi 06958 -Wume

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e =? S REFERENCES 7 l. Letter, R. R. Schneider (Niagara Mohawk Power Corporation) to K. E. Coller, USNRC, Division of Reactor Licensing, January 24, 1975. Unusual Occurrence, Docket No. 50-333. 2. Letter, A. E. Lundvall, Jr. '(Baltimore Gas and Electric Conpany) to J. P. O'Reilly, USNRC, Office 'of Inspection and Enforce =ent - Region I. Unusual Event, Docket No. 50-317. 3. Letter, P. A. Crane, Jr. (Pacific Gas and Electric Company) to R.L Engelken USNRC, Office of Inspection and Enforce =ent - Region V. June 11, 1975. Docket No. 50-133. 4. Letter, A. D. Schmidt (Florida Power and Light Company) to 3. C. Rusche, USNRC, Office of Nuclear Reactor Regulation, May 6, 1975. AOR No. 75-7, Docket 50-251. 5. Letter, N. J. Kalivianakis (Com=envealth Edison) to USNRC, Office of Nuclear Reactor Regulation, June 5, 1975. ~ Docket No. 50-265. AOR No. 75-17, 6. Special Report, SR-52-75-01, Report on Fuel Asse=bly N-10, Surry Power Station, Virginia Electric and Power Company, June 2, 1975. Docket No. 50-281. 7. Special Report, SR-S2-75-02, Report on Fuel Assembly N-20, Surry Power Station, Virginia Electric and Power Company,' June 2, 1975. Docket No. 50-281. 8. Letter, 3. 3. Stephenson (Commonwealth Edison) to J. G. Kappler, USNRC, Office of Inspection and Enforcement - Region III, May 29, 1975. AOR No. 75-29, Docket No. 50-237. 9. Letter, A. C. Thies (Duke Power Company) to N.C. Moseley, USNRC, Office of Inspection and Enforcement - Region II, April 30, 1975. UE 75-4, Docket No. 50-270.

10. Office of Inspection and Enforcement Notification of an Incident or Occurrence No. 150. May 14, 1975.

l 11. Letter, D. E. Moody (Maine Yankee Atomic Power Company) to USNRC Office of Inspection and Enforcement - Region I, June 25, 1975. AOR No. 75-9, Docket No. 50-309. W C6959 ~ ^

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w s <3 Edi !id W-w w; Uh-12. Letter, A. D. Schmidt (Florida Power and Light Cc=pany) to 3. C. Ik Rusche, USNRC, Office of Nuclear Reactor Regulation, June 20, 1975. AOR No. 75-9, Docket No. 50-251. 13. Letter, Herbert A. Autio (Yankee Accaic Electric Co=pany) to J. P. O'Reilly, USNRC, Office of Inspection and Enforcement-Region I, April 23, 1975, AOR No. 75-5, Docket No. 50-29. 14. Office of Inspection and Enforcement Notificatien of an Incident or Occurrence No. 152, June 10, 1975. 15. Letter, J. S. Bitel (Co==cnvealth Edison) to J. G. Keppler, USNRC, Office of Inspection and Enforce =ent - Region III, June 16, 1975. AOR Nos. 74-13, 14 and 15. Docket No. 50-295. F l A .e L_ J

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