NG-75-1559, Forwards Responses to NRC Questions Re 750613 Rept on Wetted Insulation & Lagging Following Reactor Coolant Pump C Seal Failure.Summary of Piping & Insulation Insp to Be Performed During Next Refueling Outage Encl

Forwards Responses to NRC Questions Re 750613 Rept on Wetted Insulation & Lagging Following Reactor Coolant Pump C Seal Failure.Summary of Piping & Insulation Insp to Be Performed During Next Refueling Outage Encl

ML20086F652
Person / Time
Site:	Robinson
Issue date:	10/01/1975
From:	Utley E CAROLINA POWER & LIGHT CO.
To:	Moseley N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20086F657	List: ML20086F660 ML20086F674 NG-75-1559, Forwards Responses to NRC Questions Re 750613 Rept on Wetted Insulation & Lagging Following Reactor Coolant Pump C Seal Failure.Summary of Piping & Insulation Insp to Be Performed During Next Refueling Outage Encl
References
NG-75-1559, NUDOCS 8401030451
Download: ML20086F652 (8)

Text

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Carolina Power & Light Company Raleigh,N C. 27602 S pg ,

G October 1, 1975 '$

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g Og,sd ,/ ,, 7 Serial: tiG-75-1559 4:4 tc,j 2 .

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File: XG-3513 (2)

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Mr. .iornan C. :.oscley, Directorc,ulatory Co-raission ,

iJ. S. :uclear %. -

Region II, Suite 313 N.W. ,

230 Peachtree Street, 30303 Atlanta, Georgia *

Dear Mr. Moseley:

~1 b. !

• U. LIC B. ;iSE ROLIJSCN IT31T lio. 2 110. DPR-23 AILLT.E IHSULATIO;i NG PIPING AFFECTLD BY pay 2, 1975 FOLL REACTOR C001rT PC'P SEAL, '

OP "C

questions raised by 12C Thic follo.-up report is 13, in response 1975 on tovetted Ainsulation sumary of and rcrardinr. the initial report of Junel purp seal failure.d during the next lar, gin , following the "C" reactor coo an.tthe piping and insulation inspecti a

refueling outasc is also attached. rns and responses to I'RC The following is a sunnary of the conce questions:

Iten 1 does not coincide conpletely vith Corporation letter CPS-75-07S Data in Table 2 of the initial reportdata contained in dated June 3, 1975. .

Iten 2 f CPS-75-073 is erroneous stated. in at lea Corrected plant data in Tatle II oused as basic data for conversion three instances if Table I was Response to Ite:rs 1.and 2 Ucstinghouse Electric Corporat Also a corrected Corrected Tabics I and II fron theletter CPS-75-07 1975 are enclosed.i h the correc Light Conpnny letter of June 13, Graph 1 is include centioned above.

8401030451 1001 a nt PDR ADOCK 05000261 +

S PDR 3 '

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- i October 1, 1975 Mr. Norman C. Moseley Item 3 Graphs 1 and 2 of !Ir. 'Jtley's letter do not specify units used for '

(

plotting the curves and respective data.

Response to Iten 3 The units (PP.') used for plotting the curves and respective data on Graphs 1 and 2 are the sana as for Fl yre 1 shown in Regulatory Guide 1.36.

Item 4 CNarts showing the field location of the sanpled Piping insulation are illegible. .

'tecnonce to Iten 4_ ,

Drewincs are attached which more cicarly indicate the a.ffected pipe and insulation.

The insulation and pipe incpection scheduled for the Jovember 1975 refueling outare vill rrovide data to confire the results of the initial inspection. Inculation shall he re:,oved from each of the pipes that was suh2r;ed, and the pipe and insulation shall be e:almined for the effcets of chlo*ide present. 'Ihis additional inspection will deternine if any pipe cc:yradation has occurred due to chleridcs. A procedure to acconplish thia is Lel:_g initiatcd ana vill be available on site for istC review prior

. to the upconing refueling outage.

Yourc v. cry truly,

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E. kl. Uticy~

Vice-President Eulk Fover Supply CC:Jvh cc: Massrs.11. it. I;anha -

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Pipo ,

Line !! umber Chloride Fluoride Silica Sodium 4

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In::ulation -

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Carolina Power a Light Company u Haleigh, N C 27602 July 28, 1975 Ng&

W l-File: ::C-3513 (R) Serial: IiC-75-1139 So-a ,

Mr. 1;or=an C. Mosaley, Director U. S. .iuclear negu12 tory Corr:ission Region II, Suite G13 230 Peachtree Street, N.W.

Atlanta, Geor;;1a 20303 Daar :1. Mose197-H. D. pMINSON U;iIT :0. 2 LICZNSE NO. DPR-23 CORRECTIVE ACTIO:iS AND INSPECTIONS SUBSEQUEhT TO "C" PJACTOR C00LA'IT PUMP SEAL FAILURE In response to an infor=al request by a medoer of your staff, the following su m ary of actions taken following Abnor al cecurrenca 50-261/75-9 is provided fer your infor ation. Includad are sn w rics of repairs and inspections perfor:ed during the forced outage in order to return the plant to safe operation.

The event which caused this outage was the failure of "C" reactor coolan: pu=p (RC?) seals and subsequent flooding of the contairmant lower icvel. This nas reported via Abnor=al occurrence Report 50-261/75-9. An account of this occurrence is included below.

On Thursday, May 1, 1975, the plant was operating norcally at 100I po.ier having just completed a two-week caintanance outage. At about 1700 the saal leak-off floi on RCP "C" began to oscillate a bit. At 1811 the seal leak-off flow failed high, indicating a failed ::o. I seal. The plant was reduced to 387; power at 10% per ninute by control of tha turbine (can operate at 40.' cn two loops) . At 1813 T.CP "C" was stopped. At 1819 Reactor Trip h'o.192 occurred due to "Turbina Trip" caused by high stea= generator leval.

At 1332 LC?'a "A" and "B" vare stopped when the Component Cooling Return Valva from the RCP's (CCU-626) shut on high flow and would not stay open. Later this proved to be a result of flashing of the component cooling vater in' RCP "C" thernal barrier cooling coils which caused sur;;es on the return line and on flo mater FIC-626 (0-150 gna) which shuts CCW-626 on hi S h flow. There was not continuous high flo in tais line, no increase in component cooling surge tank level *ias noted, and radiation conitor ?.-17 showed no increase. ?cr theaa reasons the integrity of RCP "C" therral barrier cooling coils vae cancluded to be intact. A hu;h standpipe level on RCP "A' was also received at this tI=e. Thara was cen:arn that saal flov vould be lost on RCP's "A" and

'r' since there was flashing aad high temperature (300*F) on the seal water return line for these p"=ps, thus the pumps vare secured.

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12. Corman C. Moseley July 28,1975 Following tha shutdown of RCP's "A" and "B", the plant remained in a

' ot shutdown with temperature and pressure nor=al for that condition. At 1915 seal lesk-off flov was Icst to RCP "A" and at 1928 to RCP "B". Tha objective at this time was to get the plant cooled down so that replace =ent of the seal on RCP "C" could be made. However, no RCP's were running, and at least one was needed to circulate the reactor coolant to equalize te=peratures and boron concencration in the RCS (some boric acid had been charged by this tiza f rcs the RWST through the charging pumps) and to give correct temperaturea indication on the loop bypass line RTD's. At approxi=ately 2130, to try to establish seal leak-off flow on RCP's "A" and "3", four cen entered Contain=ent and =ade an unsuccessful attempt to rotate RCP's "A" and "B" by hand. Pressure on the RCS at this time probably prevented the pumps frca rotating. Adjustments were =ade to seal injection flew, bearing lif t oil lineup, and seal leakoff and bypass lineup, but flow could not be established through the No.1 seals.

The decision was cade to start RCP "C". It was started at 2242 (con-ponent cooling water had been restored to all three pumps by this ti=e). When the pu=p had run for a minute, No.1 seal leakoff was shut to cake the seal

" film riding." The pu=p seemed to operate well. Seal leak-off te=perature pe3 Fed hi S h at 300*F i==ediately, but notor stator te=perature came up nor= ally and all bearings in the pump and =otor had nor=al temperatures.

Main steam isolation bypasses were opened and drawing of a vacues was begun to cool down the plant. At about 2300, with the core at about 480*F and RCS pressure about 1700 psig, pressuriser level began a slow downward trend. The one-half foot level alarm in the reactor su=p vas received about this time. At 0015, May 2, af ter receiving a high standpipe level and with pressurizer level at 12 on the strip chart, the decision was cada to stop RCP "C". Enan RCP "C" was stopped, pressurizer level turned downward at a rapid rate. At 0016 SI Pump "A" was started and the hot leg injection valves opened. Pressurizer level on the strip chart had reached zero by the ti=e, but level indicator LI-462 (cold cali-bration) read 6% and never bottomed out during the transient. Two ninutos af ter the first SI pu=p was started, the other two pucps were started. Pressurizer level began to increase and reached zero on the strip chart by 0023. Thereafter, pressurizer level was maintained between about 20,~, and 30,5 by cy: ling of SI pu=ps. By 0036 RCS pressure had reached 1150 psig and core te=perature was below 400*F.

The objective at this point was to reduca te=perature and pressure aufficiently to go on pla and to depressurize as ouickly as possible to slev the leak. Valve 311 (auxiliary spray) uns used to reduce pressure, and at 0048 pressure was SCO psig. Use of the RER System was delayed due to letdown valve 4603 being shut with the air line to the operator brokan. This was repaired allowing pressurizatien of the RER System. At 0145 vacuus was broken on the condenaer and use of steam was secured. At 0341 the RHR Systen was warned up and placed in service with the RCS at approaimately 285"? and 400 psig. Cold shutdown conditions were reached at 0443. To reduca the leah rate, it was desirable to' reduce system pressure. This was attempted by opening valve 311 (auniliary spray) to collapaa the pressurizer bubble. Whan valve 311 vas opened the RCS pressure did not drop noticeably, but the pressurinar level increased rapidly (nore rapidly than charging and SI would raise it). Uhen valva 311 van shut, praasurizer level decreased rapidly as tha bubble refor ad

9 g Mr. .iornan C. ioaeley July 28, 1975

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in the presaurizer. Core thermocouple temperaturas showed the reactor te=pera-ture to be stable and Ister when the reactor head vent was opened, little gas or staam escaped. At 0514 P.CS presaura reduction below 100 psig was initiated.

At 0715 charging was secured and system draindown ec==enced with pressurizer level at 555. By midmorning, the leak vaa terminated by draining to belov +50 inches reistive to the reactor vessel flange. At 1530, May 2, RCS level was 39 inchea below the reactor vessel flange.

When the leak vaa terminated there was about 12 inches of water on the Containnant ground ficor and temperature was at approvi-ately 100*F. RCS baron concentration was 1694 ppm at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, :Sy 2.

Following, the incident plans were developed to clean up and decontemi-nata the affected areas, to perform numerous inspections to assure co=ponent integrity, to co==ence reactor coolant pump repairs, and to perform evaluations of the effects of the event regarding cooldown and associated stresses. A listing of individual areas of concern and the resultant actions taken are provided below:

Reactor Coolant Purp Repairs and Inspections Westinghouse f.ssistance and Evaluations Evaluation of Vetted Insulation and Piping Evaluation of Containment Flooding System Flushing Filter Inspections Electrical Inspections Banger Inspections Reactor Coolant Pu=p Repairs and Inspections Repair work was initiated with the renoval of "C" Reactor Coolant Pump from its casing. Upon rs= oval it was determined by survey that radiation levels were excessive, (30 R/hr) for prolonged work on the i=peller end. A special procedure was written for decontamination of the impeller end of the coolant purp. Two solutions were used to remove the contamination, an alkaline per=an-l qanate and an onalic acid. Three cycles were used in the effort to reduce the radiation levela. Honetheless radiation levels at the pump remained high.

It was decided to continue with the disasse=bly with the pump in a -

i highly contaminated condition. After removal of the impeller and thermal

barrier, the shaft was recoved. There was obvious shaf t damage on the bearing

! and seal surface area. The damage occurred after the seal failure and during the period in dhich there was upward flow of pri=ary coolant across the bearing.

This overheated the bearing, resulting in its failure and eventual shaft da-age.

The labyrinth seal vaa inspected and also found to ba da= aged due to the failed l bearing and seals. Since the labyrinth seals are an integral part of tha thermal barrier, this necesaitated co=plete replacement of the thernal barrier.

With the exception of the thermal barrier, all replace =ent ite=s ver2 in stock on site. A new ther=al barrier was obtained from another utility.

During reassachly of the purp it was necessary to nachine the seating surface

o o 1.r. Tornan C. Moceley July 28, 1975 of flange to provide a proper fic of the charral barrier. After conpletion of the nachining, the affectad surfaca uas liquid penetrant tested. The surface was found to be in satisfactory condition. lhe pump was reassechled and the inpeller lock nut was seal welded in placa using approved procedures and a qualified weldar.

During the installation of nov seals in "C" RCP, frag =ents of the old Jo. 1 seal were observed to be inbedded in the seal housing wall. These were in a nonseal surface araa and were ground out to prevent tha possibility of future seal damage. The surface was then liquid penetrant inspected and released for.

installation. All seals were replaced and the puup was reinstallad in its casing.

'~ nile work on "C" purp was progressing, valva CVCS-303C, seal water return isolation valve for ' C" pu=p, was recoved and inspected. During the seal failure incident, a flov transmitter downstream of this valve exhibited indication that the valve did not fully close. Though the indication was erratic and could not be identified as true indication, it was decidad that an inspection would be in the best interest. The inspection revealed to visible waar or dansge to the valve. The valve was reassenbled uaing new gaskets and returned to service.

The motor on "L" reactor coolant pump was renoved to permit access to the pump seals for inspection. After seal removal, the seal asse=blies and runners were inspected for wear and damage. The Mo.1 seal ring and runner were inspected and found suitable for continued use. The So. 2 and lio. 3 seal rinss showed normal wear, but were replaced based on the a cunt of time remaining in this cycle. Since seal rings and runners coma as catched pairs, the runners vera also replaced.

Prior to reinstallation of the notor on "3" pu=p, the oil lif t system orifices were rc=oved and cleaned as part of preventative saintenance. No other problens were identified on "B" T.cactor 1oolant Purp.

The motor was removed on 'A" Reactor Coolant Fu:.p to allow access to the shaft seals. These seals were disassenblad and inspected for wear. Co. 3 seal ring exhibited evidence of wear and was replac2d for tha sane reason as these vare on 'B" purp. All other seal rings and runners were suitable for continued operation. The orifices of the oil lif t systen on ' A" were renoved and cleaned. _

All work was performed under the supervision of a 'Jesringhouse repre-sentative and a CP&L job coordinator. The job coordinator also provided QA surveillance coverage.

17estinthouse Assistanes and Evtluations 7cedor assistance, e; ployed as a result of the event, was provided by Festinghouse 21ectric Ccrporation. Uestinghouse reactor coolant pu=p and notor service representatives were utilized in perforning punp repairs and inspection. Westinghouse also provided evaluations of pire nd reactor 4

O O July 28, 1975

. Hor =an C. Mosaluy vassal stressas that resulted from the excessive cooldown following the ECP seal failure. This study indicated that no significant stresses resulted that would jeopardise plant safety.

Westinghouse also provided recommendations regarding action to be taken in respect to wetted insulation and piping resulting from the seal f ailure and subsequent flooding. These recommendations and final action are addressed in the following section.

A meeting was held with M2C ONRR on !!ay 12, 1975 in Sathesda, Maryland to provide them with first-hand inforcation of the details of the event.

Ueatinghouse assisted Carolina Power & Light in relating the incident, and as a result of questions arising during the meeting, provided an evaluation of the leah rate which occurred as a result of the failure. Original estimates vers that a loss of pump seals would only result in a nazi =um leakage of about 100 -

gyn. A much higher leakage was experienced in this case. In order to explain tha higher leak rate, Westinghouse prepared a report defiwfng the leakage and estimating its r:agnitude. This study was submitted to ONRR and posed no further concern. A report was also prepared and submitted regarding the loss of No. 1 1 caal flow on the "A" and "B" RCP's. Uo firm conclusion as to the cause was i for=ed, but the study established that the most probable cause was small particles which interrupted the flow across the seal face.

Evaluation of Wetted Insulation and Piping As a result of the containment flooding, piping near the containment base mat was submerged. The stainless steel piping, which was submerged, con-sisted of the seal water injection line to RCP "A", excess letdown lines, normal letdown lines, residual heat re=cval line to "A" 1 cop, and nafety injection line to "A" loop cold leg. This constitutes approximately 650 feet of piping. Portions of this piping were flooded from early Friday morning. -

tay 2,1975 to about 0400 Sunday corning, May 4,1975. The chloride concen-tration of the water was 0.25 ppa as measured on 167 2, 1975. Therefore, there was a concern for the saturation of the insulation with chlorides, possible leaching out on piping, and the potential for chloride stress corrosion.

Westinghouse was contacted to provide recommendations for corrective action.

The initial Westinghouse reply was received on !by 6,1975.

Recocmendations were to renove insulation from all subnerged stainless steel piping, cican the pipe, and reinsulate. It was also suggested to remove insu-lacion from the botten of the reactor vessel and to rinse the vessel and clean the incore thimble penetrations. The specification, which was referenced for usa in the evaluation, was ',;estinghouse PS S4351 UL, Revision 2, " Determination of Surfaca Chlorides and Fluoride Contamination on Stainless Steel Material."2 l

This specification requires a naximus chloride concentration of 0.0015 mg/d2 l for insulated surfaces, i

Initial plans were then cade to recove all the insulation and proceed as suggested. However, a sa pling program was begun on loop piping that was not wetted by the incident and the chlorides on this piping were found to be higher than the ref erenced acceptance Ifmits. Based on this as-found condition and a concern for minimizing radiation exposures of workers required

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Mr. !ior=an C. Moseley July 78, l')75 i to remove insulation and reinsulate the sukcerged piping, Westinghouse was again contacted to provide clarification and/or justification for the proposed I

acceptance criteria. It was at that time that Westinghouse reco== ended sa=pling the piping and piping insulation with respect to the acceptance criteria in 4

Westinghouse PS 83336 Ki, " Requirements for Thermal Insulation Used on Austenitic Stainless Steel Reactor Plant Piping and Equipment." This was done and the resulta indicated that the insulatien and piping were acceptable.

Westinghouse was then requested to provide a justification for the applicability of the process specification for insulation to the piping. A report was then received on May 21, 1975 justifying the acceptance standards. This report concluded that the data f alling within the acceptance curve of the subject specification indicated that sufficient neounts of sodius silicate were presant

in the insulation to inhibit chloride stress corrosion. Further, there is no concern for the contamination if the surface is covered with sodium silicate inhibited insulating natorial when it is ascertained that there is sufficient silicate inhibitor to prevent halide stress cracking. This mechanism for in-hibition of halide stress corrosion was established by H. F. Rarnes in his report, "The Corrosive Potential of Wetted Thermal Insulation," presented at the AICHE 57th National Meeting, September 26-29, 1965. With this final i'

justification, it was decided that the wettad piping and insulation were acceptable without replace =ent.

Other work related to the flooding consisted of rc=oving insulation

! from the reactor vessel bottom, sanplins for chlorides on.the incore penetrations, and rinoing the vessel. This was conpleted and no chlorides were detectable following the sa:pling. The entire vessel, loop piping, or the vessel safe ends were not sub=erged. The piping in the area of the spill was sprayed with pri=ary coolant but was not submerged. The chloride content of the primary water was not auch as to present a stress corrosion problem and none of the insulation was replaced. A detailed report has been submitted regarding the insulation vetting and concern regarding chlorido centamination.

Evaluation of Containment Flooding In order to determine the effects of flooding on the floor of the containment vessel, it was first necessary to approximate the quantity of accumulated water. Three methods of approx 1=ation were used, i.e., calculating the vetted volu=e cf the containment vessel; calculating the quantity of water recoved frcm the containment vessel; and calculating the quantity of water emptied into the containment vessel. The quantities of water determined by these methods (See Appendix A, calculations 1, 2&3) were 135,105 gallons; 133,000 gallons; and 129,000 gallons, respectively.

For conservatism, Zbasco was requested to investigate the effect of ficoding the contain=ent floor by 200,000 gallons of water. Their review concluded tnat there had been to significant effect cn tha containment floor which would varrant further reviev prior to continued operation.

Subsequent to the failure of "C" Reactor Coolant Pump seal, the water 12 vel en the floor of tha contain=ent veasel reached a haight no greater than 12 3/4". This was determinad by measuring high water narks" on the contaic=ent vallo at various locations., The avarage height of water was approxtnately 12 1/2".

O O July 28,1975 dr. Morman C. i.osaley The elevation of the ccatainment vessel floor is 223'0", and the elevation of the reactor vessel nozale centerline is 242' - 2 11/16". There-fora, no = ora than 14' - 1 25/32" of the reactor vessel was submerged with the canimum water level at 13' - 1 15/16" belou the noz:le centerline and 10' - 7 3/32" below the no==le supports.

These dinensioca are str-ari::ed in Figures 1 and 2.

Systan 71ushing A portion of the CVCS systes was flushed to remove any foreign particles which night cause damage upon a return to operation. Filter cloth was placed over the openings of drain and vent cocnections. The volute control tank was filled to approxi=ately 503 and pressuri cd to provide a driving head for flushing water.

Lines to the charging punpa and charging line flow control valve were flushed. .

The flush was perfor=ed in accordance with a special procedure, and the flush continued until the flushing unter condition was within the acceptance linits.

These acceptance limits were as follows:

1. In each flush path, af ter approxicately 50 gallons of water has been flushed through, a filter cloth was placed over the end of the exit valve or pipe and approxinately 5 gallons flushed through the cloth.

Two successive flushes =ust neet the acceptance criteria.

1

2. The general appearanca of the filter cloth shall be that of a clean uhite vet cloth showing no more than slight speckling and no more than slight soiling or staining from rust or dirt.

3. There shall be no particles on the cloth larger than 1/32" in any dimen-clon except that fine hair-like slivers or thin flakes (much less than 1/32" thick) cay have a najor dimension up to 1/16".

4. Readily apparent quantities of unusual inpurities in the exit flush water or on the cloth such as resin particles, abrasive grit, oil, or other foreign catter shall be reason for nonacceptance of the flush.

The flush was successfully cc=pleted prior to filling of the Reactor Coolant System after repairs to "C" purp.

A flush of the Residual Heat Recoval systems was perforced to remove vater from the lines which case fron the containment vessel floor when water

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was pumped to the refueling water storage tank. The water in these lines was renoved by eiths; draining to floor drains or pumping to floor drains. These drains terminate at the auxiliary building sump. After draining, the lines vara rinsed with primary water to renove any re:naining material. This was natisfactorily co=pleted dering the repair outage using a special procedure approved by the pMSC.

Filter Insnactions In addition to tha flushing previously discusaed, it uns decided to inspect the seal water injection supply filters and the reactor coolant filter to determine if any fra~nents potentially generated by the punp seal failure

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?.r. :;orman C. Moseley July 28, 1973 were entrapped. Tnere are two saal watar injection supply filter housings and one reactor coolant filter. The saal vatar filters are in parallel and are used alternately rather than concurrently.

The saal vatar filters were replaced during the April steam generator outage prior to the pu=p failure. The seal water return filter had been bypassed earlier due to a s=all leak. This allowed the return water to enter tha voluna control tank unfiltered. Therefore, any particles returned would be entrained on the injection supply filtars. 4 A radiation curvey of the filters indicated that the filter that vaa in service during the event read 1507./Er on contact of the housing. A survey of the spent cartridge indicated approx 1=ately 500 2/3r on contact. Tae filters were black with no removable surface particles. Smear samplas were taken for analysis. The results of the rmalysis rcvaaled the materials were corrosion products.

The second housing which had not been in service was opened and all cartridges were clean. In both housings the filters were intact and properly installed. New filters were inserted in both housings at the coupletion of the inspection.

The reactor coolant filter was changed during the week of "ay 19.

It was not expected that caterial othar than nor.al systen corrosion products would be entrained on the filter, and no fragnents that would possibly have been generatad by the seal failure were detected when tha filters were replaced.

Electrical Inspectiona Subsequent to the flooding which occurred in the containment vessel, an investigation was performed to determine the extant of moisture damage to instru=entation/ electrical co=ponents.

Random checks were perfor=ed on approximately 60% of the valve limit switches located inside of containment. Tais consisted of all limit switches accessible without constructing scaffolding. ;!o water or noisture uns notad in any of the limit switches checked.

• Transmitters were checked by re=oving covers and parforming a visual inspection for water da= age. The following instrumenta vera included:

Instrunant Functice.

PI-403 2eactor Coo 2 ant Systan Pressura LT-484 S/G 2 Ch.11:2r ov Range Laval LT-435 S/G 2 Ch. 2 ;: arrow nnnge Lavel LT-486 S/C 2 Ch. 3 Jarrcw Langa Laval LT-437 S/G 2 Wide Range Level FT-484 S/G 2 Ch.1 Staan l' low

- FT-433 S/G 2 Ch. 2 Staan Flew  !

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Mr. a.orman C. Poseley July 28, 1975 Instrument Function

?T-135 no. 1 Seal A P RCP No. 2 PT-123 RCP Loop 2 Thernal Barrier a ?

7I-491 RID Lypass Line Flow Indicator LT-462 Pressurizer Leval Cold Calibration PT-444 Pressuri=er Pressure Control PT-445 Pressurizer Pressure control PT-45SB Pressuricer Pressure Calib DP Cell LT-459 Pressurizer Level Ch. 1 PT-453 Pressurizer Pressure Protection Ch. 1 FIC-678 Rod Control Drive Cooler Cooling Water Flow Indicator LT-460 Pressurizar Level Ch. 2 .

PT-436 Pressurizer Pressure Ch. 2 Fr-932 SI Flow to RCS Hot Leg FT-933 SI Flow to RCS Hot Leg LT-461 Pressurizer Level Ch. 3 PT-457 Pressurizer Preasure Ch. 3 LT-494 S/G 3 Ch. 1 Harrow Range Level LT-435 S/G 3 Ch. 2 Marrow Ranga Level LT-496 S/G 3 Ch. 3 Narrow Range Level LT-497 S/G 3 Wide Range Level FT-494 S/G 3 Ch. 1 Stean Flow PT-154 11 Seal A P RCP #3 PT-125 RCP Loop 3 Thernal Barrier S ?

FT-495 S/G 3 Ch. 2 Steam Flow PT-403 RCS Pressuro Narrow Range PI-404 Reactor Coolant Systen Prassure PI-1543 No. 1 Seal A P RC2 3 FI-492 RTD Bypass Line Flow Indicator PT-138 Tzeess Letdown HX Outlet Pressure Indicator TT-1058 Reactor Coolant Drain Tank Tamperature PT-1004 Reactor. coolant Drain Tank Pressure LT-1003 Reactor Coolant Drain Tank DP FI-490 RTD Bypass Line Flow Indicator FT-475 S/G 1 Ch. 2 Steam Flow LT-477 S/G 1 Wide Range Level LT-476 S/G 1 Ch. 3 Uarrow Range Level LT-475 -

S/G 1 Ch. 2 Marrow Range Level LT-474 S/G 1 Ch. 1 Harrow Rcnge Level PT-131 RCP Loop 1 Therral Barrier A P PT-155 #1 Seal A P RCP #1 77-474 S/G 1 Ch.1 Stesa Flow LT-922 Accumulator A Lavel LT-920 Accu =ulator A Level LT-CMS-4 Condensate Measuring System Level

?T-414 RCS Flow Loop 1 FT-415 RCS Flow Loop 1

, FT-416 RCS Flow Loop 1 PT-921 Accumulator A Pressure PT-923 Accenulator A Pressure

Mr. Noruan C. Moselay July 28, 1975 Icatrument Function FIC-156 no. 1 Seal Bypass Flov RC? il FT-1563 .io. 1 Saal Leakoff RCP #1 Lo Range FT-156A Ho. 1 Seal Lesioff RC? J1 Ei Ranga PT-915 Accumulator 3 Fressura PT-927 Accu =ulator 3 Preasure LT-926 Accumulator B Level LT-924 Accu =ulator 3 Level LT-CMS-3 Condensata Maastring System Laval FT-424 Raactor Coolant System Flow Loop 2 17-425 Reactor Coolant Systen D?

FT-426 Reactor Coolant System Flow Loop 2 FT-155L Mo. 1 Scal Leakoff RCP #2 Lo Ranga -

FIC-155 No. 1 Seal Bypass Flow 7.C? J2 LT-CMS-2 Condensata Measuring System Level FIC-154 1*o. 1 Saal Bypass Flow RCP J3 M-154B No. 1 Seal Laakoff RCP #3 Lo Range FT-154A No. 1 Seal Leakoff RCP #3 Hi Range FT-434 RCS Flov Loop 3 FT-435 RCS Flav Loop 3 FT-436 aCS Flow Loop 3 LT-CMS-1 Condensate Measuring System I.evel LT-9"3 Acetraulator C Level LT-930 Accunulator C Level PT-929 Acet=:ulator C Pressure PT-931 Accu =ulator C Pressure Evidence of misture was found only in FT-154A and FT-154B. Additionally, FT-154A vaa found to be stuck in the extrene hiS h position apparently as a result of high seal vater flow. Both rotoneters were cleaned, recalibrated and satisfac-torily tested.

Rod drive cables were neggerad and found in good working order. The rod position indication cables were resistance chacked. Moisture was detected in two connectors betwean contaicsent penetrations and cable inside of containnent. These connectors were disassanbled, dried, and reassemblad.

Uater level indicators in the ste:p vere checked and evidence of

=oisture was observad. These indicators were disassenblad, dried and reinstalled.

Tha unter level detector in the reactor coolant punp bays operated properly when tested.

All instru:r.entation and electrical ce=penanta checked were lef: in satisfactory operating condition.

Eangar Insoections All pipe hangers and supports in "C" Punp Bay and on the Safety In-joction lines used during the event were inspected. The follo-ring diacrepancies i vere observed:

_ -- _ -_- . _ = . _ _

July 28, 1975 dr. .:orman C. roseley

_ Lina ;iumber Location Discrepancy 2-SI-56 Downstrean of valva Missing "U" Holc i 1.

8663 on SI line to loop 2 hot lag

! ear "C" Accumulator Loose Pipe IIanger

2. B-31-37 upstream of check valve S76A

3. 8-SI-35 Downstrean of valve Two spring hangers 4 744B on line from show no weight RER Pu=ps to loop 2 cold leg l

4. 14-AC-9 Downstream of valve Spring hanger shows l

751 on RiR loop supply no weight l

5. 1-nc-1512 No. 2 Seal Leakoff 2 Itissing "U" Bolts in "C" Punp Bay The "U" bolts were replaced on itens 1 and 5. The pipe hanger on itan 2 was tightened. The spring hansera in iteca 3 and 4 were adjustad to the p;cper tension.

Uo other discrepancies were noted. There is no indication that the hanger deficiencies were a result of the incident. Ihe hangers are on portions of the syste not included in the inservice. inspection scope, and the problems apparently existed prior to the pu:sp failure, i The above action was completed with all plant concerna as' to safety rcoolved and the plant returned to operation on Ifay 27, 1975.

Very truly yours,

.~hf.. . p /

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b:.E.Uticy4 . 4}

1 Vice-President -

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9 9 Calculntiran ? I l

Calculating the Quantity of Water Removed From the Containment Ves::e1 ,

Tnicked Off-Site 27,900 gal ,

Duptied into A CVCS 49,000 gul Holdup Tank (0% to 96%)

Dsptied into Refueling 54,500 gal Water Storage Tank (67% to 83%)

Haptied to Waste 1,600 gal Holdup Tank (39.5% to 45%)

Total Water Removed From Containment 133,000 gallons Calculation 3 Calculating the Quantity of Water Emptied Into the Containment Vessel From the Refueling Water 86,000 gal Storage Temk (92% to 67%)

318,000 gal

-232.000 c a

• 86,000 gallons From the lloric Acid Blender 23,238 gal 21,986 gal. primary water  ;

+ 1,2'i2 gal _ boric acid l 23,238 gallons q From the Reactor Coolant System g 19,762 gal l Operating Level Compensgted to 200 F=48,186 gal Drain Down Level at 200 F' -28.424 Amount spilled onto floor 19,762 gal Total Quantity Spilled 129,000 gallons A.4

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