ML20087A736

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Co Rept 50-219/69-10 on 691007-08,21-22 & 1104-06.No Items of Noncompliance Noted.Major Areas Inspected:Scram That Occurred on 691002 & Reported Increases in Control Rod Insertion Times
ML20087A736
Person / Time
Site: Oyster Creek
Issue date: 12/10/1969
From: Caphton D, Robert Carlson
US ATOMIC ENERGY COMMISSION (AEC)
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ML20086U000 List: ... further results
References
FOIA-95-36 50-219-69-10, NUDOCS 9508070146
Download: ML20087A736 (36)
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h.[ %.. (, t s U. S. ATMtIC ENERGY COMMISSION i REGION I DIVISION OF COMPLIANCE Report'of Inspection CO Report No. 219/69-10 Licensee: JERSEY CENTRAL POWER & LIGHT COMPANY Oyster Creek 1 License No. DPR-16 Category B Dates of Inspection: October 7-8, 21-22, 1969 and November 4-6, 1969 Dates of Previous Inspections: September 29 and 30, 1969 '/ /69 Inspected by: D. L. Caphton, Reactor Inspector Date. el* I N Reviewed by : R. T. Carlson, Senior Reactor Inspector Date Proprietary Information: None SCOPE Type of Facility: Boiling Water Reactor Power Level: 1600 Mwt Location: Lacey Township, Ocean County, New Jersey Type of Inspection: Special 4 Scope of Inspection: Investigate a scram that occurred on October 2, 1969, and obtain information about reported increases in control rod insertion times. The October 7-8, 1969 visit obtained preliminary facts regarding the scram

  • and reported long control rod scram times.

The long scram time problem is covered in this report. The October 21 and 22 visit provided followup for the long

  • Reported in CO Report No. 219/69-11.

9500070146 950227 PDR FOIA DEKOK95-36 PDR

.s ,.4 m z* D'= ( . scram time problem and endeavored to ascertain the JC-GE startup testing program for re- ~ qualification of control rods. The November 4 thru 6, 1969 inspection reviewed and observed I first hand the requalification testing of control rods. Information relative to other items of inspection not directly related to-control rods are covered in separate reports. Mr. D. L. Pomeroy, CO Technical Support, accompanied the inspector during the October 21 and 22, 1969 visit. Reactor Inspector R. T. Dodds participated in the inspection concerning control rods on November 4 thru 6, 1969. Mr. Dodds inspected and provided the information incorporated in this report concerning GE testing at San Jose, California.

SUMMARY

Safety Items - During a scram on October 2, 1969, control rods failed to meet performance specifications regarding scram times listed in the technical specifications item 3.2 B.3.* GE determined the cause to be crud (nr90% iron oxide) pluggage of the control rod drives inner filters.** Inner filters were subsequently removed from all drives, except 8 left in for R&D, to correct the problem. A re-qualification testing program for all drives was successfully completed. An intensive surveillance program of drive scram times and water chemistry is being followed. Noncompliance Items - None. Management Interview - Due to the nature of the October 7 and 8,1969, visit a normal CO exit interview was not held. The inspector met with Mr. McCluskey at the conclusion of the October 21 and 22,1969 visit to ascertain JC's position on several matters. Messrs. Ross, Hess and Hetrick (part time) were present for this meeting. Mr. Pomeroy was also present.

  • Inquiry Memorandum, 219/69-E, dated 10/8/69.
    • Reference to Attachment 1 for a description of control rod dri a filters.

i 1

. c si C

. Mr. McCluskey stated that JC did not consider any further reporting to be required per the linense regarding the abnormal scram times during the October 2, 1969 scram.* I Mr. McCluskey reaffirmed his. previous commitment to the inspector that he would provide at least a 24-hour reactor startup notice to-I Co, either to inspector Dodds or Caphton. Mr. McCluskey stated that he would provide the JC-GE control rod j drive requalificatien program to the inspector on October 23, 1969, by telephone. On November-6, 1969, the inspectors discussed an error found by the inspectors invelving transposing of scram time data for control rod 50-27. The error was acknowledged by JC. The inspectors emphasized the importance of data verification. DETAILS A. Personnel Contacted: Jersey Central Power & Light Company (JC) T.-McCluskey, Station Superintendent D. Hetrick, Operations Supervisor D. Ross, Technical Supervisor I. Finfrock, Maintenance Supervisor S. Daalgard, Chemistry Supervisor (Consultant) General Electrje (GE) W. Hess, Site Operations Manager W. Bibb, Operations Superintendent D. Die fenderfer, Principal Test Engineer F. Smith, Manager, Engineering Design Unit, Reactor Components F. Brutschy, Chemist C. Hills, Chemist

  • Ahnormal scran times problem discussed in report from Kelcec (JC) to Morris, dated 0:tober 14, 1969.

Problem also discussed in JC-GE/DRL nceting hcid at Headquarters on October 29, 1969.

.,'.e s,- b ) _4_ E. Recognition of Problem i Mr. Diefenderfer stated to the inspector that he and Mr. Bibb recognized during the five pump trip test scram

  • that the control rods scrammed abnormally slow.

He observed the long scram times by i observation of the control room's graphic digital display panel.** He further stated that he and Mr. Bibb immediately proceeded to investigate the print-cut of the control rod scram time recorder. It was thereby verified that a problem was apparent. Mr. Hess stated that the magnitude and cause of the problem at that time was not known, only that a problem was indicated. Mr. Hess stated that GE personnel were aware that scram times had been 1 increasing from their review of previous control rod scram recorder printouts,**

  • however similar experience enccuntered at the KRB reactor in Germany, had shown that the increase in scram times at KRB**** reached a plateau and did not result in further increases.

Mr. Hess stated that since the KRB scram times never became a problem, ** * ** it was not expected to be a problem at OC-1. 1 Mr. Hess reviewed with the inspector on October 7,1969, a 1 tabulation of average scram times obtained from the scram time recorder.****** The times were for two scrams prior to the five-pump trip scram on October 2, 1969, the five-pump trip scram, and for one scram en the folicwing day. The following is Mr. Hess' tabulation of average times obtained from the scram recorder: t

  • Occurred on October 2, 1969 and discussed in CO Report No.

219/69-11, Paragraph C l.

    • The graphic panel presents a digital readout for position of

[ each control rod. Observation of the rate of digital change l provides an indication of the speed of control rod movement. There are 53 position indicator switches in each control rod l drive to provide input to the digital readout. l

      • 26 rods monitored continuously.
        • The GE drives at KRB dif fer from OC-1 drives by a 16" shorter stroke than the 144" OC-1 stroke, per Mr. Hess.

The KRB l control rod was also stated to not have the velocity limiter as does the OC-1 control rod.

          • Mr. Hess stated that 30% of the KRB drive scram times were not affected, 30% slowed down (5.1 see was slowest 90% time),

30% varied from fast to slow. I

            • The scram time recorder is a 30 point recorder that is arranged to automatically start upon initiation of a scram.

Only 27 points are cperative on the installed recorder. One point is used for timing, providing 26 points for monitoring 26 indi-vidual control rods. (See Attachment 2 for the core position of the rods monitered.) t

j ' Co. i.: -~ j.l l (6. ] ? Average Insertion Time (Seconds) Scram Date 10% 50% 90%- l 9/9/69 .32 1.39 2.81- .l 9/23/69 1.97 4.29 9/24/69 Scram Recorder Malfunctioned 10/2/69 .44 2.51 5.79 10/3/69 .57 2.71 6.0 i The inspector independently reviewed scram times recorded by the control rod scram recorder for five scrams, including the four tabulated by Mr. Hess. The 26 rods monitored were the same for each scram except the October 3, 1969 scram. A modified group of 24' rods were monitored for the october 3, 1969 scram as part of the GE invsstigation program. The inspector's tabulation follows: i Reactor Scram Pressure Insertion Time (Seconds) Date (psig) Comments 10*/. 50*/. 907. 9/9/69 1000 Scram from test: Average 0.321 1.393 2.807 High Scram Dump Tank Maximum 0.35 1.48 3.16-Volume. Reactor Minimum 0.29 1.24 2.34 Power 30 Mwt No. Meeting Tech Spec 26 26 26' Avg. Times 9/23/69 1000 Scram from test: Average. 0.315 1.968 4.292 High Scram Dump Tank Maximum 0.42 2.62 5.0 i Volume. Minimum 0.27 1.71 3.47 No. Meeting Tech Spec 26 21 24 Avg. Times 9/29/69 990 Scram from low Average 0.336 Recorder Failed power Maximum 0.37 Minimum 0.28 No. Meeting Tech Spec 26 Avg. Times'~ 10/2/69 988 Scram at low Average 0.436 2.509 5.793 power following a Maximum 0.52 3.36 7.8 five-pump trip test Minimum 0.39 1.98 4.25 L from 800 Mwt No. Meeting Tech Spec 26 2 3 Avg. Times 10/3/69 1000 Scram from hot Average 0.57 2.71 6.0 standby. (Note Maximum 0.64 3.70 8.08 only 24 drives Minimum 0.50 1.86 4.78 No. Meetin monitored during this scram.) Tech Spec 24 2 5 Avg. Times

(, J (~ ) . 7 A review of the control room operating log book for the 8:00 a.m. to 4:00 p.m. shift on October 2, 1969 made no mention of long scram times. The log noted at 0908 a.m. " Reactor scram from flow bias", at 0955 "RNM normal and permission to startup granted from T. McCluskey" and at 1003 " Started withdrawing control rods." A 12:47 pm i entry stated " Reactor critical on ed200 second period". A 4:00 p.m. turnover entry stated " Reactor critical, startup mode, A/ 13 Mwt." A notation at 5:00 p.m. stated " Proceeded to perform scram test... ". Messrs. McCluskey, Hetrick and Hess were subsequently questioned regarding the startup with an apparent indicated control rod scram time problem. The reason stated for the startup was to provide reactor conditions appropriate for scram testing such that meaningful test data could be obtained. It was also pointed out that the average scram times for all rods (only 26 rods out of 137 rods were recorded) at that time had not been demonstrated to be outside of the technical specification average insertion times for all operable control rods.* It was emphasized by Mr. McCluskey that the 10% average scram times of the monitored 26 rods were well within the technical specification limit. Upon subsequent review of the facts by the inspector, it was apparent that technically no noncompliance had taken place relative to the license and technical specifications. C. Investigation Program 1. Program Outlined The initial objective of the GE investigation program, as stated by Mr. Hess to the inspector on October 7, 1969, was to determine the magnitude of the problem. It was the program's aim to determine if only operating control rods (those positioned other than fully inserted in the core during reactor operations) were affected or if all rods appeared to be affected. Mr. Hess stated that 62 different control rod ** drives were scram tested, some of the 62 drives were tested repeatedly. All 62 drives were tested at hot standby conditions. The conclusion drawn by GE was that the long scram time problem was general in nature. The test indicated that all control rod drives regardless of their service or usage had relatively slow scram times.

  • Technical Specifications 3.2.B.3.
    • The OC-1 reactor has 137 control rods total.

'.i, ) l < 2. Drive Testing Two of the slowest scramming control rod drives, 30-27 and 30-19, were friction tested. This test showed no frictional problems with the drive mechanism or reactor inte'rnals which could cause the slow scram times. Six drives were tested with a five minute out drive signal to provide time for the drive water to flush the drives.* These drives were then scram tested at pressure.- No significant change in scram times took place. These rods were part of the group of 26 rods that had been monitored by the scram recorder. The coordinates were 22-11, 30-19, 38-27, 14-19, 06-27 and 14-35. One drive 06-19 was scram tested for eight successive times to investigate the effects of multiple scrams. This test was also at hot standby conditions. The following are the results of this test. 06-19 Insertion Time (Seconds) 10% 50% 90% i 1 0.46 2.96 5.60 2 0.50 2.94 5.68 3 0.47 2.87 5.73 4 0.46 2.76 5.43 5 0.45 2.84 5.64 6 0.43 2.78 5.58 7 0.50 2.90 5.78 8 0.40 2.90 5.68 The results of this test GE concluded, indicated typical test data scatter; the test did not indicate any detectable trend. The 10% times were well within the 0.7 second tech spec time for average rod insertion time, however, both the 50% and 90% times were over the 2.05 and 5.00 seconds, respectively, tech spec times for average insertion. It was noted that on the 7:04 p.m. October 3, 1969 scram from hot standby (all 137 rods were mass scrammed), 06-19 scram time increased as follows:

  • Flushing occurs because of seal leakage and designed leakage passages in the individual control rods.

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~z~.. e --_.s ] QQ N -m. ~ v..,. 4 J - 4 4 l s ' j 06-19 ' Insertion Time-(Seconds) 10% 50% 90% fr 8.08* l [ 0.64 3.70 o ^ 'i i 4r. - The noted difference in the system between the individual }: scram' test.and the 137 rod mass scram is the effect thatt 3 the control rod drive systems hydraulic'chargingLpump has. j i This pump provides, in parallel, makeup water to all 137 : j 4 control rod drive ~ accumulators.** During a mass scram,xthe 1 makeup is relatively negligible'to any one accumulator ~ compared to when only supplying-to one of the1 individual. j ~ drives accumulator, which is the case during individual _ j scram tests.' The pressure-in the accumulator directly.affects-scram times, i.e.,. the greater the pressure the shorterL the. scram time. It isLtherefore expected that the individual j drives scram test times will be. biased, and some amount of 1 time shorter than mass scram times..This fact'probably accountsforsomeofthedifferenceobservedin~06-19'insertionn,{ times.on the-individual scram test versus that of the mass scram. * *

  • S '

Four drives. (06-11, 06-43, 14-19, 06-27) were tested with l the control rod drive charging pump supply to the individual accumulators valved off. No abnormal hydraulic irregularities { .were detected as a result of this test. Sixty-six additional individual scram tests beside those-already discussed above were made at hot standby conditions % on october 3, 1969. A review made by GE and subsequently' l by the inspector of this scram time data indicated that -the i 10% insertion times were all within the technical specifica-tion 0.70 s6cond time..The inspector determined that_only l 11 of the 66 tests give results (scram times) that could j 1 meet the 50% tech spec time of the 2.05 seconds for average l l insertion times. However, there were only.16-of the 66- ~ j tests that had results over the 90% tech _ spec time of s5.00 i

ecohds for average insertion' times.. Only13 tests had 90%'

, imes less than 4 seconds,. however, 38 had scram times j i

  • This was the longest scram time observed for any recorded drive.

j

    • The accumulator stores hydraulic energy (water at high pressure) to impart the initial momentum to control rods when scramming.
      • GE is conducting a test to determine the effects of the pump assist.

on scram times. This will be covered in a subsequent Co report.. ]

k,; G P 2.~ - Q" + + (f. ) 9-greater than 4.50 seconds, and one drive (06-11) had a S. . scram time of 7.16 seconds. The averages for these 66 i individual scram tests were determined by the inspector to be as follows: 'l Insertion Times (Seconds)* 10% 50% 90% Averages for 66 0.42 2.34 4.72 scram tests. Tech Spec. average 0.70 2.05 5.00 insertion times A'high speed pressure transducer was connected to the scram discharge volume to determine if any blockage existed.in or to the scram discharge volume.- This test, completed during a mass scram, determined that~the scram discharge volume was not the problem. The /kP indicated no pluggage. A different group consisting of 24 control rods was comected.to the control rod scram time recorder for the scram on October'3, 1969, from hot standby conditions, to provide a different sampling during-a mass scram of all 137 control rods. The results of this test were as follows: ~ Insertion Times - (Seconds) 10% 50% 90% Average of 24 0.57 2.71 6.00 drive times Minimum Drivo-Time 0.50 1.86 4.78 Maximum Drive Time 0.64 3.70 8.08 Tech Spec Average 0.70 2.05 5.00 Insertion Times Upon coeling the reactor down af ter the october 3,1969 scram from hot standby, four drives (38-35, 06-19, 38-19. e and 06-43) were scram tested. The reactor was at zero pressure. All of these scram times indicated normal when

  • Determined at hot standby conditions and with the control rod drive pump providing hydraulic assist.

E

t V., s. (4 1 [ compared to previous cold zero pressure times. This test t further confirmed that there were no apparent friction problems and that the long scram time problem was related to scrams when at a pressurized condition, i.e. the condition at which large 21 P's can be experienced across i the inner filter.* 3. Inspection of Drives GE next decided to remove and inspect some drive units. Two drives (06-43 and 06-19**) were removed and completely dismantled. The inner filters of both drives were found to have censiderable pluggage. Mr. Hess stated that the internals of these drives were found to be in good. condition and without any abnormal presence of crud. Four other drives (42-27, 14-43, 14-23 and 14-39) were removed and the filters only inspected. The inner filters were found to have considerable pluggage. The problem therefore appeared to GE, per Mr. Hess, to be the pluggage of the inner filters. All of the drives that were inspected were reinstalled with various combinations of filters in preparation for scram tests to.further verify the required "fix" for the long scram time problen. 0 The reactor was pressurized to 1000 psig atae100 F using the control rod drive pump. All of the drives were scrammed separately and individually. Two drives that had not been ~ reworked or disturbed in any way were also scram tested with the reworked drives in order to provide a control reference for the test. These test times were determined by the inspector to be slightly longer than their previous test times. The following data was obtained from the "fix verification" test.

  • The scram time of a control rod drive is proportional to the thP across the inner filter; the larger the d P the slower the drive The inner filter passes and filters approximately two scrams.

gallons of water during each scram. The filter is rated for a 2 psi pressure drcp at 40 gpm.

    • Had longest 90% insertion time of 8.08 seconds.

~ ,3.<.. _m s Drive Test Insertion Time (Seconds) No. No. 10% 50% 90% Remarks 18-31 1 S0.43 2.38 5.42 This drive was not re-2 0.44 2.37 5.43 worked or bothered. It I, 3 0.44 2.22 5.16 was tested to provide a j control reference. 14-19 1 0.51 3.44 7.08 Same as above comment. 2 0.49 3.06 6.58 3 0.47 2.97 6.58 14-43 1 0.40 1.41 2.57 No inner filter. 2 0.40 1.42 2.56 Upper outer filter was 3 0.40 1.41 2.56 not changed (dirty). 14 -23 1 0.43 1.50 2.59 No inner filter. 2 0.43 1.48 2.56 New upp'er outer filter 3 0.43 1.48 2.56 installed. 06-19* 1 0.40 1.48 2.91 New inner filter and 2 0.40 1.43 2.75 new upper outer filter 3 0.40 1.43 2.71 installed. 06-43*- 1 0.38 1.27 2.32 No inner filter. 2 0.38 1.32 2.36 Upper outer filter was 3 0.38 1.32 2.36 not changed (dirty). 14-39 1 0.38 1.36 2.48 No inner filter. 2 0.38 1.36 2.41 New upper outer filter 3 0.39 1.36 2.43 installed. Mr. Hess stated that it was GE's conclusion, based upon the above tests and investigation program, that removal of the inner filter from each control rod drive would correct the long scram time problem. The decision was therefore made [ to remove the inner filters ** from the control rod drives. This work was begun on the morning of October 8,1969. Mr. Hess estimated that it would.'take approximately two weeks l }

  • Drives that were completely disassembled and inspected.
    • The inner filter with the 32 micron wire mesh was actually replaced with a new inner filter heuring without any mesh.

The housing is perforated with 5/32" diameter holes. ) )

, e, '(1 ) 1[ : t6 complete the actual work of changing the filters.- The work actually took 2-1/2 weeks. The inspector, accompanied by Mr. Hetrick on October 8, 1969, 1 3-l inspected the inner and outer filters removed from a control' ~ [ rod drive, coordinate 14-39. The inner filter was a dark l reddish-brown color.* There was approximately 1/4 inch depth ] of crud settled on the inside bottom of this inner filter. The inspector observed what appeared to be several small particles of weld splatter on the inside walls of the filter's wire mesh and also a fine coating of the crud covering the inside surfaces.- Mr. Hetrick stated that both the inner and outer filter's mesh had 1.2 mil openings.** The' wire mesh was located on the inside of the perforated backup structure. The outer filter was observed to have the same reddish-brown color as the inner filter, however the outer filter did not have a visible accumulation.of particles of crud on its surfaces as did the inner filter. The inspector reviewed the scram test times for the drive inspected in order to obtain some degree of perspective i relative to the amount of crud and the length of the scram times. The scram times for this drive were obtained from a test conducted on October 3, 1969, at 1000 psi and 1000 F with the control rod drive pump assisting during the test. Coordinate 10% 50% 90% 14-39 0.45 2.45*** 4.89 The inspector also observed the work that was in progress beneath the reactor. Drives that were selected to have only their filters changed had this work completed in the room beneath the reactor. If the drive was selected to be dis-assembled **** for inspection and/or for example, seal replace-ment, the drive was removed to the area outside the drywell i

  • New filters have the appearance of shiny machined stainless steel.
    • GE and JC later described the size as 32 rderon wire mesh.
      • Exceeds the Tech Spec average insertion time of 2.05 seconds.
        • Mr. Bibb stated that a total of 23 drives (es17%) were disassembled I

during the entire job. Mr. Hess stated that any drive that had collet seal leakage of greater than 3 gpm, the seals were j designated for replacement.

L, -. ~.. f,(I ') . air lock door and the reactor building railroad air lock door for this work. During the visit on October 21 and 22, 1969, the inspector, accompanied by Mr. Ross and Mr. Pomeroy, again inspected the i work progressing on the control rod drives. The inspector observed work being p,erformed on one disassembled drive unit. .f During the inspection of the seal piston rings, it was noted that several cf the rings were apparently wedged in a contracted position.rather than the normally expanded positien, indicating to the inspector that foreign material may have been the cause for the observed position of the rings. A statement by a GE worker doing the actual work, was that considerable foreign material had been found in some of the drives he had disassembled. He pointed to some mate-rial on the work bench that he indicated had been removed from drives disassembled. The inspector observed that the material resembled that seen on the filters. Subsequent discussions with Mr. Hess indicated that some piston rings had been observed by GE to have been wedged in a contracted position as a result of foreign materials. Mr. Hess emphasized that the worse problem that might result from stuck rings would be so much seal leakage that the drive would not drive out. He stated that this condition would not prevent the rod from scramming into the reactor.* The inspector observed that dne general work area where the drive disassembly work was being done appeared to have no extra housekeeping requirements, the area was roped off and posted as a radiation area. The inspector observed distorted inner filters that had been removed from the reactor. The bottom plate of the filters were observed to be bulged ** outwardly indicating the effects from a large t1P across the filter. Mr. Hess stated that all drives that had been in the normal withdrawn pattern, approx-imately one-half, had their inner filter ends similarly bulged outwardly. A removed upper outer filter was also in-Spedad and indicated the effects of an abnormal 4k P across the wire mesh located on theinside of the support housing.

  • GE states that the area of the "in" drive piston relative to the "out" drive piston is so.much greater, that the rod will not fail to scram if seal failures occur.
    • San Jose test results determinec that a 2h P of 200 psi across the inner filter weuld bulge the end.

g s v.., we i=>a .+.3s se see -emu r..+ . The housing was not observed to be distorted, however, the wire mesh was. The mesh was pulled away from the inside [ diameter of the housing in what appeared to be an axially corrugated form with the maximum diameter of the corrugated [ wire mesh still in contact with the inside diameter of the housing. The appearance of thB filters color was similar { (dark reddish brown) to that of other filters removed. i There was no loose material observed on the surfaces of the filter, however, some pluggage was visually observed. The inspector, assisted by Mr. Pomeroy, took color pictures of distorted filters. Pictures *were also taken of the work area and parts of disassembled control rod drives that were being worked on. 4. Effects of Filter Pluqqage GE conducted tests at their San Jose, California test facility to determine the effects on drive scram times of actual _ complete filter pluggage of both the inner and outer filters. Reactor Inspector Dodds reviewed the data from the inner filter test and actually witnessed and reviewed data for the outer filter test. (Reference to Attachment 3 for details of inspector Dodds review). GE welded shim stock over the filter mesh to completely prevent any liquid passage through the wire mesh. Scram tests were then conducted and the test drive continued to scram with the available leakage passages serving to permit passage of water. With both the inner and outer filters plugged (the worse case) and pressure conditions comparable to BNR normal operating conditions, the longest 90% insertion time was 14.840 seconds.*

  • With just the inner filter plugged the test drives maximum 90% insertion time was 12.055 seconds.

With just the outer filter plugged the test drives maximum 90% insertion time was only 2.401 seconds. The San Jose tests showed that pluggage of the inner filter has large effects on lengthening scram times while pluggage

  • Pictures (slides) filed at CO:I Office
    • OC-1 Tech Spec specifies 5.0 seconds for average insertion time.
  • /

h ) . of the outer filter has only nominal effects on lengthening scram times. Leakage passages around the inner filter are quite small, however there is clearance between the drive uncoupling rod and the inner filter guide tube. Some clearance could exist where the inner filter attaches to.thel coupling spud. Leakage passages around the inner filter when the. mesh t becomes plugged would total only a very small fraction of an inch in total area. Information about the pluggage of' leakage passages of OC-1 inner filters is not known. Mr. Smith stated that there were four passages for possible leakage around the outer filter. These passages are the guide cap to the index tuber filter to the thermal sleeve, thermal sleeve to the reactor, and filter to the guide cap. It was stated that the total leakage area of all of these 2 clearance passages would be between 0.6 in and 1.0 in2, It can be seen that from the large leakage passages available, outer filter pluggage has very nominal effects on scram times. The pluggage of leakage passages of the OC-1 outer filters is not known. Filter R&D Program 5. GE installed eight inner filters with wire mesh.* Mr. Hess described this program as "a modest R&D program" to aid in determining an acceptable wire mesh size for the inner filters. The location and wire mesh size of these eight filters are: Reactor Location Wire Mesh Size Comments 06-19 32 micron original design 18-23 100 micron 18-47 100 micron 18-07 150 micron 30-51 150 micron 10-39 250 micron 14-43 250 micron 30-03** 250 micron

  • 129 control rod drives have no wire mesh on the inner filter The housing is perforated with 5/32 inch diameter holes.

housing.

    • Drive was subsequently declared inoperable because of failure to latch onto the control rod.

Rod has been positioned full out and drive valved out of service.

~ O ' These drives are included in the drives that are continuously monitored by the control rod scram time recorder. D. Requalification and Surveillance 1. Rod Tests Mr. McCluskey stated to the inspector on October 22, 1969, following considerable and repeated urging on the part of the inspector, that JC would provide on October 23, 1969, the details of their program for requalification of the control rods

  • before JC-GE proceeded with their planned startup test program.

The program was obtained from Mr. McCluskey by telephone on October 23, 1969, however it was stated to be "only preliminary thinking". A meeting was requested by DRL at Headquarters to review the long scram time problem. The meeting was held on October 29, 1969, with both JC and GE cognizant representatives in attendance. This inspector was also present. Out of this meeting ** came an agreed upon rod requalification and acceptance program; also, additional surveillance guide lines were outlined. The DRL letter requested JC to inform CO if: Any drives measured scram time for 90% insertion "does a. not fall within the range of 1.9 to 3.6 seconds". b. "The average scram time of all 137 control rods for 90% insertion is not within the range of 2.4 to 3.1 seconds". "All control rods will be scram tested under cold c. pressurized conditions." d. " Control rods will then be withdrawn for nuclear heating to achieve hot standby conditions... and the rods used to achieve this condition and those which will be used to achieve 25% power will be scram tested. All other control rods shall also be scram tested before they are used, such that each control rod shall have been scram tested in the hot pressurized condition before it is with-drawn for any purpose except its own scram test."

  • Reference to Inquiry Memorandum, 219/69-H.

Peter A. Morris, Director, DRL to George Ritter, JC j

    • Letter:

dated November 6, 1969 (Docket No. 50-219). I

..+ (; )- . L i. JC-GE had already proceeded under their own initiative with scram tests of each drive at cold pressurized conditions prior to receiving the DRL letter. This work was completed 5 on October 29, 1969. JC-GE elected to scram test all drives at hot standby condi-tions before proceeding with their planned startup test program. This work was completed on November 4, 1969. The writer and reactor inspector Dodds jointly on November 4, 1969, reviewed the data from the test. It was determined that the reactor power level for the hot test was approxi-mately 160 to 180 Mwt.* The following tabulation gives both the cold and hot test data results: Scram Time (Seconds) Cold Hot Cold Hot Cold Hot-10% 10% 50% 50% 90% 90% I Average ** 0.397 0.397 1.398 1.380 2.493 2.449 Minimum 0.35 0.35 1.29 1.23 2.26 2.21 Maximum 0.50 0.46 1.61 1.57 2.81 2.79 Rods above avg. 64 72 68 70 61 71 Rods below avg. 72 64 68 66 75 65 The above insertion times were all within the guide lines of the DRL letter. During the cold pressurized scram test, control rod drive j 50-19 had a 90% 4.8 second scram time. Two additional tests l were made and the times were 4.9 seconds and 5.36 seconds. The drive was removed and found to still have a dirty inner filter (with wire mesh). The filter was removed and the rod retested. The 90% time from this test was 2.38 seconds. The hot time was 2.28 seconds. l

  • Hot standby for this reactor had been noted to be acl00 Mwt in previous observations.

JC stated that the power level obtained provided for stability during the rod testing.

    • Average is for 136 rods.

Drive 30-03 inoperative.

~ s -e (i ) ; The inspectors on November 6, 1969, noted one control rod, 50-27, that had a positioning error on the control room graphic digital display. The raw scram data was audited for this rod and the inspectors determined that 16 out of 48 reed switch indications were missing on the scram time recorder chart for the hot test of this rod completed on. November 4, 1969. The cold pressurized test data was good and the 90% hot test scram time appeared good. It was, however, not possible to accurately obtain the 10% and 50% times, yet the data for these times appeared on the summary data sheet. The inspectors determined that the engineer making the original data analysis had written a memo concerning the matter but the information was lost in transposing the data. The 50-27 position switch problem was corrected on November 7, 1969. JC-GE, in keeping with the guide lines of the DRL letter out-lining surveillance of the control rod scram times, conducted a scram test on November 6, 1969, at 50% power on the 26 rods that are monitored continuously by the scram time recorder. The following tabulation lists. the scram times from the above test and an actual scram that followed from a turbine trip test from 50% power, (part of the scheduled startup test program). The inspectors witnessed the turbine trip scram and audited the d6 duction of the scram time data by JC engineers. The scram test was performed by individually scramming rods. The turbine trip was a mass scram of all rods. Both tests were conducted at 800 Mwt. The hot test data is listed also for reference. All scram times were within the DRL guide lines. (See Tabulation on Following Page)

j ',, .a (' . ) '/; 90% Insertion Times (Seconds) Turbine Coordinate Hot Test Test Scram Trip Scram a I; 06-43 2.30 2.35 2.34 06-27 2.54 2.54. 2.52 a j 06-19 2.39 2.42 2.38 1d-39 2.35 2.42 2.48 14-19 2.72 2.74 2.77 14-27 2.56 2.59 2.62 14-35 2.40 2.45 2.44 14-43 2.66 2.68 2.67 18-07 2.56 2.52 2.50 18-23 2.70 2.69 2.70 18-47 2.70 2.81 2.67 22-27 2.41 2.43 2.38 22-35 2.51 2.53 2.50 22-43 2.37 2.38 2.34 30-19 2.45 2.50 2.44 30-27 2.56 2.57 2.50 30-35 2.34 2.42 2.39 30-43 2.45 2.52 2.54 30-51 2.51 2.55 2.50 38-11 2.45 2.50 2.46 38-27 2.42 2.46 2.50 38-35 2.40 2.53 2.52 38-43 2.44 2.40 2.46 46-11 2.35 2.40 2.37 46-19 2.46 2.50 2.57 46-35 2.45 2.45 2.43

e ( ) J < One of the above listed drives, 18-23, was reported to CO by JC on November 10, 1969, as being over the DRL guide line of 3.6 seconds for an individual drive. This drive is one of the 8 GE-R&D drives and contains a 100 micron size inner } filter. The following is a summary of 90% scram times for this drive:* 11/9/69 4.02 seconds Test 11/9/69 4.34 seconds Scram 11/13/69 4.02 seconds Test 11/14/69 5.52 seconds Scram The 10% time and the 50% time for the November 14, 1969, scram are 0.37 and 2.32 seconds respectively. JC reports that all other drives are continuing to perform within the DRL guide lines. No abnormal trends have been reported. The inspectors will continue to followup this program. 2. Inoperable Control Rod, 30-03 Control rod 30-03 would not latch to its drive. The drive was changed out before reactor startup and the rod still failed to latch to the drive. GE's investigation program concluded that the control rod's coupling shaft was bent. To correct this problem would require removal of the control rod from the reactor. Mr. McCluskey stated that the condition would be corrected at the first refueling outage, scheduled-for approximately 18 months from November 1, 1969. The rod and drive were declared inoperable ** and both positioned in the full out position. The drives hydraulic valves were locked and tagged closed to prevent any movement of this rod and drive.

  • Information obtained in telecons with site subsequent to these visits.
    • Reference to Tech Spec 3.2.B.4.

O ) . E. Crud Problem 1. Crud Analysis t Eleven grams of crud removed from one inner filter were j analyzed by GE and found to be 75% iron (or v 90% iron oxide) and less than 2% organic materials. In addition to iron, a spectrum analysis indicated traces of chromium and cobalt. Other analyses have indicated similar amounts of iron. The iron particle size in the inner filters was stated by GE to be as small as between 3 and 5 microns, many particles were considerably larger. The crud material has the appearance of being a reddish brown rouge. 2. Sampling for Crud Both GE and JC were taking water samples for crud determina-tions of reactor water, feedwater and condensate up to the time of the control rod problem. The sampling technique being used by JC and GE differed and the analysis results reflected a difference; the JC crud analysis were usually higher than GE 's. A meeting of the JC and GE chemists was held on October 31, 1969, and a standard sampling technique was agreed upon for future crud determinations. It was also agreed upon to record sample results in only one official sample log. 3. Locations Sampled GE undertook a sampling program to determine if there was an abnormal concentration of crud in any part of the total system. GE sampled large quantities of water by straining j the water through pillow cases and visually observing for l crud. Smaller samples were collected in sample bottles for I analysis. The largest concentration of crud was found in the condenser hot wells. Mr. Hess stated that six wheel barrow loads of crud were removed from the three condenser hot wells. Mr. Hess further stated that the hot wells were the expected crud collection point and crud concentrator in the overall system. Mr. Hess further added that some of the material removed from the hot wells was construction type debris, other was mud and iron oxide.

9, ~~ - y mya 7 -(.. -;,a L;1.,.. . m.i. - L,-a * 'x.. , ;.. r - . O l i 22 - U? Mr'. Hess stated that-a 2 inch' drain lins rrom the bottom a ( of the reactor was sampled andlonly a small trace of crud' i was detected using the pillow-case technique for sampling.. ' -[ A spare instrument tap line that. runs from the' top;of.the l l0 core support. plate in'the reactor was sampled.using the- 'l [ -pillow case technique. 'only small traces of the crud'were j detected from this~ sampling. { Samples of water going Lto the cleanup system were taken._ Small amounts of crud were detected. ] Mr. Hess stated that a special Jacking device was designed j to unseat

  • control rods when their drive unit was removed.

.) ~ The unseating permitted water to drain.out of the-guide tube j ~ 1n. order to enable sampling for crud..The water was-routed ] via a hoselthrough a pillow case. Mr. Hess stated that small quantities of crud estimated to be generally less than.

)

the amounts'found in the inner-filters, were detected as a l result of this check. Twenty drives were sampled usingEthis. j technique. The drives were selected-to provide.a representa - tive sample across the reactor. Mr. Hess stated that other. locations sampled were: re-circulation loops, reheaters, feedwater piping, moisture separators, heat.exchangers and condensate domineralizers. j The control rai drive water supply and filters also were j sampled and no evidence of abnormal crud concentrations.were . detected here. Mr. Hess stated that it was concluded from the sampling program that there were no significant concentra-1 tions' of crud-in any locations except the condenser hot wells and the control rod drive-filters. j Mr. Hess stated that the chemical cleaning of the feedwater. system was completed in September and October of 1967. No cleaning had been done since. He stated that an inspection of'the feedwater chain was completed in January or February. of 1969, however, no, cleaning was done at the time nor abnormal conditions observed.

  • Control rods back seat against the bottom of the guide tube to provide a reasonably water tight seal when the drive is removed.

./.,- -..~# = ..-,,,....~.--m._.~, 9' (-) ') 4 /.- , Initial critical of the reactor was achieved. in May of 1969. [. ' The reactor.has operated' intermittently in the startup testing ) 4 program since May. Extended outages have occurred which may l have affected the crud levels in the feedwater chain on the reactor side of the demineralizers and thereby increased i the crud quantities entering the reactor. J J r-l Mr. Hess postulated one possible mechanism for concentrating crud in the area of the control rod drive inner filters. He stated that crud may be depositing on the control rod velocity limiter, then when the rod scrams, the crud would be washed off. Mr. Hess stated that GE had estimated that the flow that passes the velocity limiter during a scram was approxi-- mately 1500 gpm. 4. Water Chemistry-The inspector reviewed sample results of water analyses taken both before and after the Oct.ober 3, 1969, shutdown for the filter work. The chemistry surveillance program'was determined to have been considerably improved as a result of ~ the filter pluggage problem. Both the GE and JC chemists were questioned regarding the program. Both chemists stated that they were given new responsibility to determine that the feedwater and reactor water was within specifications prior to changes in reactor power -levels and before and after-transient tests. The results of this change in operating philosophy provided the chemists enough time to make and complete their appraisal of the water conditions before changes were made to the system. Mr. Ross stated that sampling limits and sampling. frequency guide lines had been established to control water chemistry for operations af ter the control rod filter work was completed. The following guide lines are for reactor water and reflects minimum frequenciess v. (See Next Page) 1 l i: J -r-

g .' ;h t m A. y g"'"" ~ ~ ~ ~~ ~ ~ ~ ~ /p',3-m.. t- .) y t - Reactor Water- ?. 't i sv Imposed Limit Sampling Frequency -,,y 4 j Ph 5.0 - 9.0 Dailyf l L 4 y conductivity 4 1 uho/cm Daily t T Chlorine <100 ppb Daily d Silica 4 5 ppm Daily-i Baron c.50 ppm Weekly Filterable Solids ' < 3 ppm Daily (Crud)'- Metals: Fe Every 2 Days-Cu Every-2 Days Ni Every 2 Days-Cr Every 2 Days 3 Mr. Ross stated that the self-imposed minimum guide line for l feedwater was to sample once to twice per shift. The filterable solids (crud) daily mean is to be 4100 ppb and - the maximum to be 4 250 ppb..-A metals analysis-will be run-1 on feedwater every two days and timed with the similar sample for reactor water. i The minimum guide line for sampling demineralizer effluent 'l for crud is daily. Each individual demineralizer is to be sampled every 3 to 4 hours using'a reflectometer. J i The minimum guide line for sampling condensate crud is daily.. 1 Mr. Daalgard stated that reflectometer analyses'for. crud 1 are now being made on a nearly continuous basis for de-l mineralizer effluent and;feedwater whenever power changes- + or transient tests are made. .Mr. Daalgard stated that round l the clock chemistry coverage is being provided. Mr. Daalgard stated that a GPU chemist, Mr. Robert Stoudenour, is I providing weekend relief for himself. 1 j 1

~~ F-q? a ~ } i ;

shows the results of reactor water analyses i

starting from May 3, 1969 through November 6, 1969. The l audit was a sampling audit; results are shown where avail-able. The maximum crud sample, item 5 in the table, observed [ by the inspector was for August 22, 1969 and was 4,950 ppb;- i the minimum was for August 6,1969, and was 5 ppb. Mr. Hess stated to the inspector that the crud average was approx- [ imately 200 ppb for reactor water. 1 i Sample results for condensate, demineralizer effluent, and feedwater were mu,ch less frequently than for reactor water. The attached table lists data obtained from a sampling audit of chemistry records. (Attachment 5) Mr. Hess stated that feedwater crud levels had averaged between 20 and 30 ppb prior to the control rod filter problem. The inspectors review of 6 feedyater sample results after reactor startup following the control rod filter work, determined the feedwater crud to be <C 15 ppb and averaged e-ll ppb. The inspector will continue to follow the chemistry program. 5. Feedwater Chain The entire condensate and feedwater piping chain up the reactor vessel's safe ends, per Mr. Hess, is carbon steel piping with the exception of stainless steel tdbes in'the condensate-feedwater heaters. The only equipment'in this chain for crud removal is the condensate demineralizers. OC-1 has seven demineralizers connected in parallel. GE operated the demineralizers with half beds ( ~ 1/2 the designed resin capacity of the demineralizers) through mid September before filling the beds to their designed capacity. Mr. Hess stated that GE believed that the " half beds" would do an adequate job during the initial phases of getting the three separate feedwater heater chains in operation and particularly when it was expected that the beds would require frequent backwashing. It was anticipated that the condensate-feedwater chain would pz. luce more than normal amounts of crud, per Mr. Hess, when the chains were initially placed in operation and subjected for the first

' ~ x; i 1. ...~-. _, a. _ _. ~.. . p: O 26 - l i time to flows and' temperatures not previously experienced. } Mr. Hess stated that GE's past experience had shown that j " half beds" would do the job. Subsequent discussions with 3 Mr. Ross determined that the_ cost of resin was a prime l reason for using " half beds" instead of full bed demineralizers. + f Mr. Hess stated that the normal set up of the demineralizers had been to keep four on line.* Mr. Hess stated that GE did not know whether there had been crud carrythru in the i i demineralizers. Only two of the condensate-feedwater chains had been placed in service at the time of the October 3, 1969 shutdown. One of the two had relatively minimal service per Mr. Hess. l t I i i i p r

  • Each demineralizer is rated to 2650 gpm.

Six demineralizers are required to handle all three feedwater pumps ( ~ 15,000 gpm total) when operating (this condition had not yet been attained by OC-1). l w e..m-7 4

- - -. ~ (E h + JERSEY CENTRAL POWER & LIGHT CO, CO Report No. 219/69-10 ATTACHMENT NO. 1 The following descriptive information about the inner and outer ) filters was obtained from a GE manual, GE 1-92808.- The reference numbers 1 and 2 refer to the attached sketch from the same manual. i The inspector measured the two filters and determined the approximate dimensions: Outer Filter O.D. E 3-1/2" Length Overall a 9" Active Filter Length

  1. 7-3/8" Inner Filter O.D.

E l-3/4" Length Overall E 3-3/4" Active Filter Length & 3" 1.3.9 - Filters "The external filter 1 and the internal filter 2 are installed near the upper end of the drive mechanism. Both are provided to filter reactor water flowing into the drive, removing foreign particles or abrasive matter that could result in damage and excessive wear to the drive mechanism." 1.3.9.1 - External (Outer) Filter 4 The external filter assembly 1 consists of a ring, an outer cylinder, and interior cylinder subassembly, a screen, and a guide welded together. The external filter is installed on the drive mechanism by three lock-wired screws which secure the lower end of the filter guide to the guide cap on the drive mechanism. This filter removes foreign particles from reactor water entering the annus between the outer tube of the drive mechanism and the thermal sleeve of the drive housing. The outer cylinder is made of perforated stainless steel and has approximately 40% of its surface open for the entry of water to the wire cloth screen, which is installed on its inside surface. The ring welded to the upper end of the outer cylinder provides a close fit with the inside wall of the thermal sleeve. A pocket is formed by the guide at the lower end of the filter which provides a close fit with the outside wall of the index tube. Particles filtered from the water accumulate here.

s (.) ) 1.3.9.2. - Internal (Inner) Filter The internal filter assembly 2 consists of a cylinder which encloses a wire - cloth screen, and to which a guide tube is welded. The i internal filter is installed on the drive mechanism by three screws which secure it to the lower end of the coupling spud inside the index tube. It prevents the entry of foreign matter with reactor { water flowing to the interior of the index tube. The filter cylinder is made of perforated stainless steel, and has approximately 46% of its surface open for the entry of water to the i wire cloth. The guide tube is welded to the lower end of the cylinder and furnishes a recess for the accumulation of foreign matter and a guide for the rod and tube assembly. The inner and outer rings, welded to the upper end of the cylinder, provide support for the filter and form the flange by means of which the filter attaches to the bottom of the coupling spud. The wire cloth portion of the filter is a stainless steel, dutch weave filter cloth. ATTACHMENT 1 Page 2 of 3 Pages

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  • I GEI-92808 f

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p m i Ma ,CottFT Pt! TON i h ,j l ik j[m,%n i,'I i,...nio%]{]hj 4 r surru notts /h Ih {' f o"st !" ~ ,H ;'m v ~ .I b 6 i g ; je g 9e j i; ENEEtttN i nousins rtanct G ,y 4 1 l ! g7 ga u y f [ jfj g],hi ,,g.g,ita mvt n y .atactoa watta rmast ' Q"Q-lj g @M 4 e E 3) s +;7 t / enws.murras sotti di"" % r,umlr. l..io..tu .s hW-uu mtaty .!5 h' 1 e ty (xf imsiu ict a \\ 1 nec rtanct uouariasicars sj .=h 2150-1 ( ATTACHMENT 1 Page 3.of 3 P2 ges s t \\ Figure 1-1.' Schematic Diagram - Model 7R-Dh-144-Al :ont ol Rod Drive Mechanism o h

j 0 J,- i ep-EAST A _ _ _ _ _ _ _ _ _ _ _ st I I A A A 4 _ _ _ _. h3 5 1 t, 1 n;4 A A A A A . _ ~35 ri i! E h A A A A . _27 a -\\ r-H \\ i l! A) A A. A-A A _ _ 1, i i 1 I l i l I i ^ l ! I l Ar A A A _l __ _ n .i I 1 l i I l ,~ I i i A _ _ _ _ii_ _ _ _._ _i_ _ _ 3 1 4 i I i i i l I 1 l l 02 10 18 26 4 h2 50 j h e- ~---- i / f I ' JERSEY CENTRAL POWER AND Co mFW 329/69-10 CORE ARRANGEMENT CONTROL CELLS TOP VIEW (Mt hoS CM $ % 6 MY boWT\\we DMCAC s A% ckmeC

.s ner-otniuv ] RUE 0AHA T-710-995-92 { "' R 280Dios OCT 69 ~ i FM USAEC R T-00DDS' BERKEL'EY CALIF TO RUEOAHA/USAEC R T CARLSON 970 BROAD STREET NEE) ARK NJ .i j &EN/BY3 USAEC J P O REILLY DIV 0F COMPLIANCE dASHDC, -AE v BT { UNCLAS JERSEY' CENTRAL POWER AND LIGHT' COMPANY / OYSTER CREEK / CMM l 1 t DOCKET NO. TPArt00H GENERAL ELECTRIC CMM APED CMM SAN JOSE CM.M ' CALIFORNIA WAS VISITED BY R. T. DODDS REACTOR INSPECTOR CO-I CMM.ON OCTOBER 24-2S CMM l'969 FOR THE PURPOSE'OF OBSERVING TESTS OF A 4 i CONTROL ROD DRIVE AT - THE CONTROL ROD DRIVE TEST FACILITY. THE ? DRIVE HAD BEEN MODIFIED TO DEMONSTRATE SCRAM CAPABILITY WITH COMBINATIONS OF PLUGGED INNER AND OUTER FIL.TERS. i THE FILTERS WERE PLUGGED BY WELDING A SLEEVE OJER THE TOP OF THE ' SCREEN BACKUP PLATES. ONLY TESTS OF THE PLUGGED OUTER FILTER WERE t OBSERVED SINCE THE T.ESTS OF THE PLUGGED INNER FILTER ASSEMBLY HAD, ) ALREADY BEEN CONDUCTED. HOSEVER CMM THE SAND 30RN TRACES OF THESE SCRAM TE'STS WERE SCRUTINI&ED.' INSTALLATION OF THE PLUGGED OUTER I' ) FILTER ASSEMBLY ON THE DRIVE AND INSERTION OF THE DRIVE IN THE J } TEST PRESSURE VESSEL WAS OBSERVED-ON OCTOBER 24'CMM 1969. COLD t \\ [- 4 i* / i 1. l ) 1 / sTERSEY CENTRAL POWER AND LI'GHT COMPANY l ~' CO REPORT NO. 219/69-10 5E san a,_ se . <-i er l es ,.s ~ /% l rk fi D4 W1dd j f 'Page 1 of 4 Pages .2

a- ' ~ f S"PAGE'2RHhlNEAQO27QNCLAS () . f o. ~ h TESTS'WERE ALSO OBSERVED ON OCTOBER.24. HOT TESTS AT FULL PRESSURE A /1030 PSIG/. OBSERVED ON OCTOBER 25. THE RESULTS OF TESTS CONDUCTED - ON OCTOB'ER - 26 WERE 0BTAINED BY PHONE FROM MR. FLOYD SMITH CMM MANAGER ENGINEERING DESIGN REACTOR COMPONENTS., i ( ~ IN GENERAL CMM AT LEAST THREE TESTS idERE RUN AT EACH OPERATING i CONDITION. RESULTS OF THE TESTS WERE AS FOLLOWS CLN 1 INNER SCREEN PLUGGED / 1030 PSIG REACT 08, VESSEL PRESSURE ? j ..J? / i 1510/87S PSIG /1/ ACCUMULATOR PRESSURE s f 10 PERCENT INSERTION SCRAM TIME - 0 745-0 980 SEC 11 782-12 055'SEC I 90 PERCENT INSERTION SCRAM TIME s. /1/ NOTE CLN ACCUMULATOR GAS PRESSURE TO 875 PSIG THEN BROUGHT TO 1510 PSIG WITH ' WATER CHARGING PUMP. j i

2. OUTER-SCREEN PLUGGED TEST A t

. REACTOR VESSEL PRESSURE O PSIG i ACCUMULATOR PRESSURE 1510/875 PSIG i 10 PERCENT INSERTION SCRAM TIME O.400 SEC + 'l 90 PERCENT INSERTION SCRAM TIME 1 795-1 805 SEC i TEST B i, i REACTOR VESSEL PRESSURE 1030 PSIG t .! ^ s 4 .g ., + Q* j (. i 1 . \\ ATTACHMENT 3 Page 2 of 4 Pages l 7

,7,y4 aiw-et. - -. fa-j' "P AGE 3 RHi1NEAQO20 0NCLAS O j 1510/875 PSIG g! . ACCUMULATOR PRESSURE: t'I 10 PERCENT-INSERTION SCRAM TIME 0 440-0 485 SEC v

90. PERCENT INSERTION SCRAM TIME 2 394-2 401 SEC-J y

TEST.C 7 -i. REACTOR.NESSEL PRESSURE 1030 PSI.G 1 .y ISOLATED j -[ ' ACCUMULATOR PRESSURE 0 572 0 582-' I 10 PERCENT INSERTION SCRAM TIME y 2 910-2 972 l 90 EERCENT INSERTION SCRAM TIME f j y' INNER.AND OUTER SCREENS PLUGGED i 3 ,4 I f TEST A i 1030 PSIG .l t-RE4CTOR VESSEL PRESSURE -t 1510/875 PSIG ACCUMULATOR PRESSURE 90 PERCENT INSERTION SCRAM TIME '11 454-12 140 SEC j . 10 PERCENT INSERTION SCRAM TIME 1 164-1 330 SEC b TEST B f a } REACTOR VESSEL PRESSURE ~ 1030-PSIG l f. ~ ISOLATED. k AC'CUMUL'ATOR PRESSURE i 90 PERCENT INSERTION SCRAi, TIME 14 646-14 840 SEC L i 2 230-2 290 SEC 10 PERCENT INSERTION SCR M TIME i i' I, j TEST C 850 PSIG 6

j REACTOR WESSEL PRESSURE

.i 1 l l j 'q. y = - - - - 7,..y i 7 ~ ~ j 3 i i i* i .V 9 ~ ? .g f l i -l ATTACHMENT 3 Page 3 of 4 Pages f

M - s.. ~ \\- ' AS

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] i'PAGE 4 RHb!NEAQO273 U r 1510/875 PSIG iJ, ~ AC'CUMULATOR PRESSURE l t 9 654-9 806 SEC g. 90 PERCENT' INSERTION SCRAM TIME ' O.983-1 106 /10 PERCENT INSERTION SCRAM TIME r[j 4 TEST D 850 PSSIG REACTOR UESSEL PRESSURE [ ISOLATED ' ACCUMULATOR PRESSURE t 17 280-17 553 SEC I I 90 PERCENT INSERTION'SCRA'M TIME .) 10 PERCENT INSERTION SCRAM TIME 2.S95-2 668 SEC / REFERENCE DATA / [ NORMAL DRIVE - INNER SCREEN REMOVED 4. / ll TEST A 4 1030 PSIG r REACTOR UESSEL PRESSURE 1510/875 PSIG ACCUMULATOR PRESSURE s 2 393 SEC 90 PERCENT INSERTION SCRAM TIME 0 380 SEC l 5 PERCENT INSERTION SCRAM TIME TEST B a 1 1030 PSIG i REACTOR-WESSEL PRESSURE ISOLATED x> ACCUMULATOR PRESSURE 2 700 - 90 PERCENT INSERTION SCRAM TIME ' s 'O.443 .h 5 PERCEN INSERTION SCRAM TIME END. REk CO CLN U CLN RTD 2061 t, BT I .f I 0273 f l E p 's ,r . ~ MC198/23 3 fI. i l' NAVY PHILA s / i Q / i ,,.,,- y,,,,,,, .....,4..... ._..ms. a e s + ~ k ATTACHMENT 3 Page 4 of 4 Pages t +-.... .-m

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, g. fy 79 w- _' l ),. .s J ~ u. ^ ~ ~ }:- JERSEY CENTRAL POWER & LIGHT 00. ATTACIMENT 5 -{. Date 5/9/69.. 6/6/69 9/4/69_ 9/11/69 9/24/69 9/25/69 9/26/69 9/27/69 9/28/69 11/3/69 11/4/69 '11/5/69 I Systest C DE C DE C DE C N C. DE N C DE N C M N C E N C W W W N N [.i 1. Ph 6.3 6.3 6.05 5.64 6.9 6.2 6.25 6.25 6.05 0.21 0.13

2. Conductiv- 0.860.88 1.821.3 0.18 0,19 0.92 0.69 0.72 ity (unhos) h -[
3. Oiloride *0.02 4.02=0.02=0.02 25 20 0.047 40.02 0.03 25 '

O.010 4 20 i b ppm I {

4. Silica 0.14 0.14 0.38 0.345 41.0 41.0 0.57 0.44 0.46 f

(SiO ) ppm 320 1000** 320 410 50 150 230 50 30 410 300 270 140 99 130 9 9 9 to.15* 2

5. Insolubles ppb 23 8

20

6. Iron ppb me 0 10 1.25
7. Copper ppb

~0-0.5

  • 0.5
8. Nichol ppb 15.8 1.0
g..
9. Oiromium ppb Condensate C

= DE = Domineraliser Effluent M= Feedwater

  • Results from four samples.
    • Possible bad sample.

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