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ATOMIC ENERGY COMMISSION DIVislON OF COMPLIANCE 201 645 i
- I neoiON :
970 BROAD STREET NEWARK. NEW JERSEY 07101 October 30, 197)
R. T. Carlson, Senior Reactor Inspector Region T., Division of Compliance JERSEY CENTRAL POWER & LIGHT COMPANY (0YSTER CREEK 1)
CO REPORT NO. 219/70-6 I was satisfied with the performance of the facility's management and the review by PORC end.GORB of the events discussed in the report.
I recommend the following:
- 1., C0 should establish that GE does supply modified control linkages, as appropriate, for the BWR's of connon design (identified in the report).
2.
C0 should establish that adequate maintenance practices and procedures are in effect for th0 EPR oil filters to ensure that other BWR's do not experience similar transients, from dirt in the Moog valve.
I intend to, fc,110w up on the metallurgical studies that will be performed by GE on the broken linkage and also follow closely the results that the new cams (planned for installation in late October, 1970) have on steam pressure control and stability.
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hJ.McDermott Reactor Inspector I
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U. S. AIOMIC ENERGY COMMISSION REGION I
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DIVISION OF COMPLIANCE Report of Inspectici CO Report No. 219/70-6 Licensee:
JERSEY CENIRAL POWER AND LIGHT COMPANY Oyster Creek 1 License No. DPR-16 Category C Dates of Inspection:
September 23-25, 1970 Dates of Previous Inspec on:
May 18-22, 1970 s
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/70 Inspected by:'
R.
. McDermo t, Reactor Inspector
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' Reviewed by :
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R. T. Carleo% Senior Reactor Inspector
'Date Proprietary Information:
None SCOPE Type of Facility:
Boiling Water Rea: tor Pcwer Level:
1600 Mwt Location:
Forked River, New Jersey Type of Inspection:
Special; Announced Accompan_ving Personnel:
Mr. W. Farmer, TSB, CO:HQ accompanied on September 24 and 25, 1970, and assisted in the writing of this report.
Scope of Inspection:
A special inspection was made at the site to review reported instances of malfunctions of the main turbine initial pressure regulator and a control linkage breakage that would have affected the turbine bypass valve j
operation. GE representatives from APED, San Jose, i
California and Large Steam Turbine Generator (LSTG) i Division in Schenectady, New York, were interviewed at the site to determine the cause and significance of the malfunctions and the generic considerations for other BWR's.
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SUMMARY
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- S'fety Items - None i
Noncompliance Items - None Unusual Occurrences - The Oyster Creek facility has experienced six disturbances to steam pressure control during the period of September 17-28, 1970. They have
.b:en exhibited as spGes or oscillations in electrical output ar.d steam flow and pressure. Tha measured macnitudes of the transients have been in tte range of 5 - 70 Mwe. On two occastorts, main steam line high flow instruments have trippad
-but the combinations of sensor trips was not sufficient to initiata the closure of the main steam isolation valves. On one occasion, however, low pressure (850 psi) in the main steam line did initiate main steam isolation valve closare and resulted in a reactor scram. The events appeared to be caused by, and were reported by JC
- to be caused by, malfunctions or design inadequacies in the initial pressure ragulator (IPR) controls, but one of the disturbances that was experienced was directly related to a malfunction in the feedvater control systw. Three scrams hcVe resulted from these transients, and in all cases, all post-scram functions were reported to have operated normally. During the system chuk-out following one of the scrams, a broken control linkage was observec that would have prevented the turbine steam bypass valves from opening when repired.
C:rrective measures employed by the licensee to eliminate the malfunctions have included:
(1) the cleanup of the control oil system ; (2) the installation of an additional filter in the oil supply for the electric prussure regulator (EPR) 4 p;rtion of the IFR; (3) changing of two wire-wound rheostats in the EPR control to composition-type rheostate; (4) the replacement of two amplifiers in the EPR port. ion of the IPR with amplifiers of a similar design; (5) an inspection and check-out of all wiring within the EPR control system for solid conne:tions; (6) eliminating unwanted grounds and assuring zero resistanes grounds where appropriate withir. the EFR control system; and (7) repairing the broken turbina hypass valve operating linkage. GE personnel assisted in the repairs and tha che:kout of the
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C01 functions observed. Personnel from the GE, Installation and Service Engineering I
(I&SE) Group and the GE, LSIG Group in Schenectady, New York visited the plant to review first-hard the observed malfunctiors.
Future planned changes. for the facility ir.clude the replacement of the broken control linkage that was repairad with one of a new design, tha addition of cover plates for the control linkages where appropriate, the changeout of t.he amplifiers
- within the EPR to a new de4Lgn with extended service lifa, and the replacement of the turbine control cams to provide for more stable steam pressure control for both the l
current Itcensed power limit and the mini-stretch power increase (1690 Mwt yplica-tion which bhs been submitted to DRL and is currsntly pending). JC-GE are currently evaluating the nesd to install damping capacitors within the EPR control system to eliminata steam line " noise" from feeding through the control syst.em.
The generic considerations for other BWR's may be influenced by the results of the planned metallurgical examir.ation of t.he broken control leakage. This examinatien will be periot'ned by GE, Schenectady, New York in lata October,1970
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..a L During: the site management. meeting which.was held with JC and GE representatives,
.thelinspectorswere informed that GE is'also. planning to supply r w control. valve icanslto the:Nina Mile ~ Point plant. Two' additional' plants (not identifLed but.
thought by GE-representatives to be foreign BWR's)~ vere reported to also be under consideration for new cams. 'It was afso disclosed that the hydraulic oll' filter installation! that is:utilised to ~ filter the oil supply' to t he hydraulic ' control
'OC-1 modified their-oil filter inst.allation as a ras;1t of the recent:
disturbances.- Othar BWR's utilizing mechanical-hydraulic turbina control' schemes, such as Oyster Creek, were reported by GE representatives to include Nine' Mile.
' Point, Millstone 1,'Monticello, Vermont Yankee, and' Pilgrim. GE personnel from
.the LSTG Group have stated that it is GE's policy to initiate any required changes to:all nuclear power plants of a conunon design when a corrective change is' made to any one of these plants. 'It should be noted that the Dresden II turbine control isfof 'a.dtfferent ' design concept - (no mechanical -linkages). t han t.he Oyster Creek plant.'
- The results of the discussions with 'JC and GE personnel on the recent disturbances, including the broken coctrol linkage did not discloss any possibility for a more severe transient t o occur than that previously. ar.alynd, 'i.e., tur bits trip-out without bypass valve opening.
Listed below is a summary description of the disturbances and tie causes which the licensee at tributes the disturbances t o:
'1.
Turbine-Generater Oscillations - Following a " backwash" (tube cleaning) operation.of t he main condenser, on September. 17, 1970, the ges rator load began to oscillate 10-!5 he.
The station load was reduced from 530 to 400 Mwe.
by operator action and the load reduction resulted in a turbits trip followed by a reactor Neram from high flux. The turbine trip was caused by an indicated high moisture level in the moisture-separator drain tank and was
' assumed by tha licensee to be caused by " flashing" of the moisture in the drain tank that resulted from the load reduction. Following a checkout of
..the initial pressure regulat or controls by GE and JC personnel, the reactor was' restarted on September 18, 1970. JC considered the instability in the initial presscre regulator to be caused by improperly designed turbine control
' valve cams (non-1inear operation) and tmat the backwash operation may have aggravated the situation by changing. condenser efficiency and hence, control zvalve position and steam flow and pressure.
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Turbine-Ge % eater Spike - During operation at 525 Mwe on September 20, 1970,
~ the electrical: output of the generator suddenly-increased as 20 Mwe followed by a decrease of e 30 Mwe. The station load was reduced by operator action to approximately 450 Mwe and cont rol was transferred from the EPR to the mechanical pressure regulator (MPR). The licensee has attributed this spike
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to dirt in t.he Moog valve' in the EPR control system.
3.. Turbine-Gene.rator Spike. During operation at 495 Mwe on September 21, 1970,
, (1: 33 a.m.); t he electrical ~ output of the generator suddenly increased 55 Mwe '
and then' decreased approximately. 70 Mwe. The station' load was reduced'by operator action tio~ 450 Mwe and control was transferred from the EPR to the LMPR.::Tha licensee' attributes this spike to dirt in the Moog valve in the-
. EPR, control system.'-
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4.. ; Turbine-Generator Oscillations - During operation at 455 Mws on Sept ember 21, 1910 (9:22 a.m.), turbine vibrations were noted and load was reduced by operatorfaction.to 390 Mwe. Vibrations returned to normal bat daring att empts to-recover the load, the station output began to oscillat.e s10wly at 4 t0 Mwe with a magnitude of 40 - 50 Mwe. Power was reduced.by oparator action to 350 Mwe, but the load swings continue for approximately 30 minates before a stable system was obtained. Control of the initial pressure ts6ulator was 4
i then transferred to the MPR.
'5.
Turbine Trip and Reactor Scram - During operation on September 22, 1970, at 500 Mwe with the EPR in service, the generator load increased suddenly ab5 Mwe i
and reactor steam pressure decreased from 1000 to 970 psi. The operator was instructed to reduce' load to 470 Mwe and to transfer contral to the MPR.
. Steam pressure continued to decrease after transferring control to the MPR
'and attempts were made to regain control of the EPR.
Steam pressure continued, to decrease and the main steam isolation valves closed at 830 psi to initiate a reactor. scram. The turbine generator was then manually trippsd.
Investiga-tion disclosed dirt in the Moog valve in the EPR system.
During a checkout of 4
the turbina controls following the scram, the t.urbine steam bypass valves would not respond. A control' linkage was found to be broken and was repaired before
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resuming operation. The exact cause for the brokea itnkage could not be established. The effect of the breaking of this linkage would not impair normal turbine control, but would have prevented the turbine steam bypass valves from opening when required.
6 Feedwater Control System Malfunction - During operation on September 28, 1970, at 450 Mwe with the MPR in service, a loss of a feedwater pump flow signal was experienced. The feed pumps continued to run but the cont.rol syotem, which' then saw a mismatch of steam and feed flow, called for additional makeup to the reactor. Reactor level increased from 80 to 85 inches before the level input to the 3-element controller overrode the steam-feedwater flow mismatch.
The operator reduced power from 450 to 400 Mwe and following the load reduction, the turbine tripped from a high level in the moisture separator drain tank -
the reactor scrammed on high reactor pressure.
' Status of Previousiv Reported Problems - None 4
Other Significant Items - None 1
Mtnagement Interview - An exit interview was held with Messrs. McCluskey, Ross,
.and Carroll at the conclusion of the inspection. The inspectors questioned 1
Mr. McCluskey relative to further planned action if additional generator load i~
disturbances ware observed.
Mr. McCluskey stated that if further disturbances
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' were obs'erved, the station load would be reduced to a level that would permit stable operation. The inspectors stated that it appeared that there were two types of unrelated problems with the turbine generator - one being a critical
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can position near full generator load and the other being dirt in the control oil j
' system.- In regard to the former, the inspector (Mr. McDermott) stated 'that it
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would appear prudent not to operate near the critical cam position. Mr. McCluskey s,
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responded by stating that the plant had operated during the summar at full load without IPR stability problems.
He further stated that based on riis history, he intended to return the plant to full load.
He aise stated that the backwashing of the main condenser that preceded the instability problem on I
September 18, 1970, had also been routinely performed during the summer months at full generator load conditions without instability resulting 6nt that it was his plan to reduce load prior to althar turbine valvs testing or eendsnser backwashing to prevent introducing disturbances into tne syetam when near the critical cam position.
The reportability aspects of these recent events were discussed with Mr. McCluskey and he stated that he did not consider these ovents reportable by license require-ments. The inspector stated that the reportability requirements were not clearly defined on this issue but encouraged JC to voluntarily submit an information report of these events.
Mr. McCluskey notified the assigned inspe: tor by telephone the following day that a report would be submitted to DRL during tie week. of October 4, 1970.*
DETAILS A.
Persons Contacted:
Jersey Central Power & Light Company Mr.
T.. McCluskey, Station Superintendent, 0C-1 Mr. D. Ross, Technical Supervisor, OC-1 Mr. J. Carroll,,0perations Supervisor, OC-1
_ General Elec'tric Co'mpany Mr. P. C. Callan, Controls Engineer, LSTG Division, Schenectady, NY Mr. R. J. Dickinson, Controls Engineer, LSTG Division, Schenectady, NY Mr. W. Popov, I&SE Group, Millburn, NJ Mr. R. Seimer, Transient Analysis Engineer, APED, San Jose, California Mr. J. Benson, Licensing Activities Group, APED, San Jose, California C.
Operations 1.
Description of Events Following the May, 1970 rod work outage, the reactor was restarted and has been operated continuously with three exceptions. Two unscheduled plant shutdowns resulted on July 11 and August 1,1970, from heavy sea grass accumulation on the main circulating water intake screens.
One additional unscheduled shutdown occurred on September 15, 1970, due to a high moisture accumulation rate (unidentified leakage) in containment which was caused by a leaking packing on a recirculation pump discharge valve. Reactor operations resumed on September 17, 1970, and a series of steam pressure disturbances have occurred since that time.
- Letter, Finfrock to Morris, dated October 8, 1970.
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- Listed belov.is a, description of the events as obtained from operations logs Land ~ discussions with operating personnel:
September 17. 1970 '(Event' No.1)
- Time Secuence'of Events 0448 Reactor startup.
2215 100% power - 530 Mwe.
2230 started condenser backwash.
2250 Turbine electrical load began swinging 10-L5 Mwe.
't Operator, started to reduce load by redacing re-circulation flow. A loud rumbling noise from the.
turbine was noted but its source could rat. be established.
2300 The plant was at 400 Mwa and stable.
2301.
Turbine trip and resulting reactor 6. ram followed by a main steam line valve isolation. The isola-tion condensers were placed'in service mar.ually to control system pressura.
All control' rods scrammed fully except 30-03 which was valved oui of service at position 48.. Scram times of monitored rods ranged from 2.33 to.3.06 seconds. The reactor was' brought critical at 0522 on September 18, 1970, and the plant was maintained in the' hot standby condition while the cause of the turbine trip and turbine oscillations was being investigated by JC operating personnel and Mr. W. Popov, GE I6SE representative.
Investigation disclosed that the level controller for moisture-separator drain tank 1-6 was out of adjustme'nt (proportional band) and the instrument department i
corrected and reset this instrument.
It was assumed that the cause of the turbine trip was flashing in the moisture-separator drain tank following the load reduction. A hydraulic pilot valve (Moog valva) in the EPR j
portion of the IFR war disassembled and cleaned, control linkages were checked from the front standard on the turbine to the control valves, EPR control system and shock absorber (mounted on the torque tuba in the front-standard) response times were checked, and the control oil supply filters were changed. Nothing was observed during this checkout that could be associated with the turbine-generator oscillations ~
. Just. prior to the scram, the main condensers were being backwashed which caused the load to decrease, then increase as each condenser half's flow was reversed.
It is thought that these power swings might have_ contributed Lto the start of the oscillations.. The turbine cams were reported to'have i
shown-a tendency toward instability at the full power. position whenever i
something occurs to swing the load. JMr. McCluskey informed the itspectors that 00-1 had praviously experienced' poor main steam pressure control when e
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the turbine-generator was operating in the range of 530-535 Mws.
He attributed this to a poor design of the four cams that position the four
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turbine control valves through the IPR linkages.
Mr. McClu+kay further stated that the currently installed cams were the third set of cams to be installed at OC-1 and that another set of cans would be installad during the next scheduled outage in October, 1970. The installation cf the new cams is intended to eliminate the poor control characteristics at or near the full load and to accommodate the planned mini-stretch and final stretch power increases.
The operators were instructed to decrease load to a more stable cam position before backwashing condensers and the turbine-generator was returned to service and raised to 380 Mwe when oscillations began again at 2050 on September 18, 1970. Load was increased to 450 Mwe and the 10-15 Mwe oscilla-tions continued. Control was then transferred to the MPR at 2333 on September 18, 1970, and the oscillations vanished. The oil tilters for the EPR were cleaned and control on the EPR was re-established at 1123, September 19, 1970. Load was then increased to 520 Mwe by 1500 with no oscillations.
Graphs of selected parameters were obtaincd for this t ransiett and are included as Figures 1-4 attached. The scales for the variables recorded on the graphs are as follows:
Electrical output 800 Mwe.
6 lbs/hr Totalized steam flow - O to 8 x 10 Wide range steam pressure - O to 1600 psi Feed flow - O to 8 x 106 lbs/hr Reactor level - 0 to 8 feet above O datum - (The O datum level was reported to correspond to av8 feet above the top of the active fuel)
APRMS - O to 150%
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Narrow range reactor pressure - 950 to 1050 psi Turbine first stage steam flow - O to 8 x 106 lbs/hr Control valve and bypass valve position indication - 0 to 100% open September 20, 1970 (Event No. 2)
Time Sequence of Events 0240 With the, plant operating at 525 Mwe, the operator received the following alarms:
turbine excess vibration, APRM high alarm (all channels except No. 8), and two main steam line break high steam
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Li flow alarms (points 15 and 45' in the events recorder),
accompanied' by a sudden up-spike of Jh 20 Mwe on the turbine-generator output. The four turbine control valve positions swung from 81 to 89% open and back and this was accompanied by a steam flow change of 5.6 x 106 to 6.4 x 106 lbs/hr and back.
0245 The operator commenced load reduction with recircula-tion flow to 500 Mwe and then inserted control rods to reduce the power further to 450 Mwe. -The turbine initial pressure regulator was being controlled by the EPR at this time.
'Mr. W. Popov (GE) and Mr. J. Carroll (JC) informed the inspa tors that they
, attributed this spike to dirt in the.Moog valve. The small internal j
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tolerances of this valve, coupled with inadequate filtering of the supply' oil,
'were considered to be the cause for its sticky operation.
Sticky operation prevents the Moog valve internals from immediately responding. When this valve does respond, it will then overshoot its control' position and result 1
in sudden control valve motion and 'inally manifest itself in generator power j
spikes.
L Load was increased.to 500 Hwe on the EPR with no problems by 1208 on September 20, 1970. Charts of this event are attached as Figures 5 -
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September 21, 1970'(Event No. 3)
Time Sequence of Events 4
0133 With the load at 500 Mwe the plant. experienced an
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uncontrolled steam flow transient due to turbine control valve movement. The operator received one
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alarm of main steem line break high steam flow.
i Indicated electrical output spiked up by 55 Mwe
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followed rapidly by a 70 Mwe down-spike.
0134
' Load reduction was commenced by reducing recirculation flow.
0150 450 Hwe reached and the operator held this load controlling on the EPR.
0350.
Excessive vibration was noted on the turbine oil return lines following the transient experienced at
'0133.
~ Start of 8:00 a.m.
Observed vibrations on turbine oil lines and oil to 4:00 p.m. shift tank, as well as on the torque tube and the shock absorber. The vibration was stopped by applying:a firm pressure on the shock absorber weight.
Mr. Popov (GE). informed the inspector that erratic cot. trol valve motion (slight ' hunting') would account'foi the vibrations.
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Graphs of the selected parameters for this event are shown in Figures 9-12 attached.
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September 21, 1970'(Event No. 4)
Time Sequence of Events
.0922 Began generator load reduction to 400 Mwe with recirculation flow due to turbina front standard (control system) and control oil System vibrations.
EPR in control.-
Began increasing generator load to 450 Mwe with 0942 recirculation flow.
1000 Generator load at 410 Mwa and the load began to slowly swing 25 Mwe.
1006-Transferred control from the EFR to the MPR to clean the Moog valve and change the oil filters.
1020 Began raising power with recirculation flow to 450 Mwe controlling the turbine and the MPR.
1401 Transferred control to the EPR.
1445 Reached 500 Mwe, t
'1803 515 Mwe - experienced swing in faedwater flew from 5.9 to 4.7 and back to 6.8 x 106 lbs/hr.-
Alarm received for feed pump runout, but alarm 4
did not lock in.
Also received a 3% spike on the APRM due-to t'he cold water injection.
1 2200 Commenced power reduction to 500 Mwe.
Charts of this event are attached as Figures.13-16.
September 22, 1970 (Event No. 5)
Time Sequence of Events 0828
" Maximum emergency generation" order given to i
i all stations by grid load dispatcher.
0928
" Voltage warning" given to all stations by the
, grid loadidispatcher.
0939.
At 500 Mwe, the reactor scrammed from closure
.of the main steam isolation-valves. Reactor steam pressure initially experienced a decrease s
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- to approximately 970 psi. The oparater was instructed to reduce load to 470 Mwe aed to transfer control to th'e MPR.
Pressurd continued to decrease due to the reactor power cut and when 850 psi was reached, the main steam isolation valves closed. Prior to the MSIV elosure, attempts were made to regain pressure centrol whiLe controlling the EPR but pressure continued to decrease. The power supply to the EPR was turned off in an attempt to force the transfer to the MPR, but the control system did not respond. The turbine generator was manually trippad following the scram.
Prior to the scram, the reactor was operating at approximately 1MO Mwt.
The first indication of a problem was a very small swing up in el utrical load accompanied by a decrease in reactor pressure. The. operator immediately started to reduce load to reach a more stable position on the contral valve cams, but steam pressure continued to drop. The power supply to the EPR was shut off in hopes of effecting a change' over to the MPR.
It was later discovered that it would not have been possible to change to the MPR in this manner.
Upon investigation of the EPR and the Moog val.ve, it was found that the internal filter in the Moog valve was plugged. The plugged filter caused the internal piston in the Moog valve, and hence the control valves, to remain in the position they took just prior to the scram. As this position had opened the control valves slightly, this caused the pressure to start decreasing. The operator, by dropping load, reduced the pressure even further.
The pressure continued to decrease until the 850 psig set point for the main steam line low pressure was reached, at which time the main steam isolation valves closed and scrammed the reactor.
The plugging of the Moog valve nozzle was believed to be caused by dirt which was left in the system from the last pre-filter cleaning operation (September 17, 1970). During this investigation it was also found that a rubber gasket was missing from the normal pre-filter which would have allowe1 some oil to bypass the filter when in service. This dirt would normally be caught by a second filter (sinteredmetal)butoneofthesealinggasketsonthesefilterswas found to be pinched in a manner such that it could also have bean bypassing oil and particulate matter.
The Plant Operations Review Committee (PORC) reviewed both the scram and the mechanical lockup of the Moog valve. They determined that the only way control could have been transferred to the MPR would have been to increase the recirculation flow and reactor power until the steam pressure met the MPR set point pressure, or to lower the MPR set point pressure to the system pressure. As the MPR can only be placed in service when the system pressure increases to the MPR set point or the set point reduces to the steam pressure, the transfe,r could not have been made as the. system pressura was dropping faster than the MPR set point could be reduced. Note: There is a physical limit incorporated in the set point controls for the MPR that controls the rate of set point change.
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Other corrective action taken included a checkout of the EPR syStam. All electrical connections were checked for tightness, response times and control action of the Hoog valve were checked at the turbine fr.'ct standard, some minor repairs were made to correct the mechanical binding that was discovered on the DT-1 Moog position feedback sensor and the Apf ral or ifice of the DT-4 pressure transducer was cleaned. The orifice did contain some solid deposits of foreign material, but it was not plugged.
IP.a output of the EPR control amplifiers were checked with an oscilloscope with varying input signals into the amplifiers. No spiking or unusual operation was j
noted.
During the course of this' investigation, it was discovered that tha turbine steam bypass valves would not respond.
It was discovered that one of t he linkages in the control system had cracked.* The crack in one tube was almost 3600 around the circumference and would have prevented proper ope.rattoa of the byp a s s valves, but would not impair the normal operation of t he control valves.
The ex'act cause of the failure of the rod was not detarmined.
It was thoaght that the rod may possibly have been damaged by someone stepping on the rod or hant,ing a chain hoist from it during the construction period of the plant.
l The rod is mounted horizontally on a span of approximately 12 feet.
Mete.11urgical examination of the rod will be completed as soon as the rod is.nade available to the GE company.
It was stated by Mr. McClu+ key that the rod will be removed during the October 18-25, 1970, outage and shipped to j
GE for a metallurgical examination.
He further stated that the results of j
the exaeination would be made available to Compliance. The damadsd rod or linkage was repaired by inserting a hollow steel sleeve inald.4 tr.e original j
i bar and fastening this w'ith bolts.
It was established durirg diweussions with GE that a failure of this linkage or any other linkage within the turbina 1
controls would result in closure of either the control valvas or bypass valves as these valves are machanically biased closed.
Several of t he otner control linkage tubes were dye penetrant inspected for cracks without discovering any additional cracking. GE now plans to provide a replacement lirkage for the damaged linkage. The replacemest linkage will have a thicket tube wall or be constructed of a stronger material.
Charts of this event are attached as Figures 17-20.
$2ptember 23, 1970 Time Sequence of Eventa i
1 1157 Reactor critical,
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2300 Turbine-generator on bus.
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- Discussed further in paragraphs C.3 and C.4.
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. rv September'25, 1970
' Time Sequence of Events 0030 At 510 Mwe the turbine-generator experienced several 10-15 Mwe swings. The 13ad was reduced to 450 Mwe by operator action and control was transferred to the MPR to check out the EPR system.
The unit remained at 450 Mwe until two wire-wound rheostats in the EPR control systam vera replaced with composition type rheostats.
- September 26, 1970 Time Sequence of Events 4
Swing Shift Control was transferred to the EIR and the generator loading was increased 520 Mwe.
Several small (2-5 Mwe) spikes were experienced and the load was reduced to 450 Mwe and control transferred, to the MPR. While on the MPR, several small spikes in the generator output also resulted.
September 28, 1970 (Event No. 6)
Time
_S_equence of Events r
1930 450 Mwe, MPR in control - At this Ime, a flow signal from the B feedwarter pump was lost. The
- feedwater pumps continued to ope. rate and the feedwater flow control valvas responded to the indicated mismatch in steam-to-feed flow by increasing flow to the reactor. The operator began a load reduction to 400 Mwa to enable the two remaining feed pumps (it was thought that th pump was lost) to supply the feed flow demand.
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The reactor level increased fram 80 to 85 inches before control was re-established. The cause I
for the loss of indicated feedwater flow signal was reported to be a cold solder connection in' the temperature compensation circuit for the B feed pump flow signal. The load was increased j
to 450 Mwe following the repair and checkout of i
this failure.
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September 30, 1970 i
Sequence of Events i
i Continued operating at 450 Mwe while inspecting j
the EPR controls.
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October 1. 1970' Sequence of Eventa Load increased to 530 Mwe on tha EER after replace-i I
. ment-of two amplifiers on the EPR, che; king all control system wiring, and eliminating grounds from the control system wiring. No additicnal problems with the turbine controIs or the feed--
water controls systems had besn experiencad during the period of October 1 thr2 October 18, 1970.
.2 Description of the Turbine Inlet Valve Control System (IPR)
The turbine valve control system is detailed in attached Figuras 22 and 23 which were ' supplied in Amendment 11 to the FSAR for Oyster Creek 1 (copies of which
.are attached).' The EPR electronic module and pressure transducer ara shown on Figure 22 at drawing position 16 by A thru B.
This systen supplies the electrical signal to the pilot valve (or Moog valve) shown on Figure 23 at i
drawing location 5 by K.
The hydraulic p,ilot valve controls the positioning l
of the master hydraulic servo motor shown adjacent to it.
The hydraulic filter plugging and pilot valve port plugging problems reported above occurred in that portion of the hydraulic system shown on Figure 23 at.d,rawing location 2 thru 5 K.
The hydraulic servo motor of the EPR system, through levers, controls the rotation of the tube at drawing lo:ation 2 thru 8 by L on Figure 23.
The rotation of this tube is transmitted to the turbine inlet valva and bypass valve controls by mechanical linkage.
The mechanical linkage which has a run length of 20 to 30 feet is shown in the attached drawing labaled Figure 24 The EPR system is backed up by the MPR identified as tha Forced Restored Regulator on Figure 23 in drawing location 2 thru 5 by F thru H.
The MPR control set point is normally a few psi above the EPR set point so that it will take over on pressure increases. The MFR actuates the turbina inlet valve and bypass valve controls through the same linkage and valve actuation controls as the EPR.-
The turbine inlet control cams are shown on Figure 22 in drawing location 25 thru 27A. It is to be noted that these cams are on a rotating bar at the opposite end of the mechanical linkage from the EPR and MPR controls. The cams, through a mechanical linkage system, position the hydraulic sarvo pilot valve which positions the main hydraulic servo motor which than opens or closes the steam inlet valves through a mechanical lever system..
The cylinder that fractured or cracked in the mechanical linkags system to
+
.the turbine bypass valves is identified as link 34 on Figure 24. This link 3
is about in the middle of the drawing. This tube moves in a horizontal plane a
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transmitting motion from the rotating bar.and lever on ona end to the rotating bar and. lever on the other end. The tube is always under a compressive load. The tube is made of aluminum with a diamster around j
2-1/2 inches, a wall thickness around 0.150 inches ar.d a length around
+
12 feet.
(These were approximate dimensions obtained during vinal inspection). Complete failure of this tube would immobilire tha turt tne bypass valves but have~no effect.on the Ifnkage to tha turbine inlet valves, j.
3.
Review of Control Problems and Linkage Failure i
The turbine control difficulties'and linkage failure ware reviswed with representatives of GE and the JC operations staff at Oystar Creek 1.
The GE representatives stated that it was their opinion that the :,teady cyclic instability experienced when operating in EPR control was canad by the profile on the turbine inlet control valve cams not matching tre valva characteristics in the high load range. Thus when a slight plant up set i
occurs there is not sufficient damping in the system to pt,ve.t on L11ation.
The spikes and large amplitude oscillations shown in the oparating. vents described above are believed by GE to result from particulates plugging up the filters in the hydraulic system supplying oil to the hydraulic pilot control valve (or Moog valve) or plugging the port of the valve. When these i
particulates break loose, the pilot valve is postulated to overshoot its new control position resulting in the turbine inlet valve oversho, ting its position. The particle size capable of plugging'the filter is almost visible to the 'haked" eye. The existing hydraulic oil supply system has a t:iltar system as shown on Figure 23. The pre-filter is a cuno cart ridge 1/2 micron size and the after-filter a 10 micron sinterad metallic unit.
In addition, there are internal filters in the Moog valve sized for 20 to 40 microns. The clearances and ports in the Moog valve are extramely.small so that the valve will stick if particles accumulate. When the Moog valve sticks the EPR control system is unable to activate either the turbine inlet er bypass valves.
However, both valves would be closed at slightly highar steam pres-sure by the MPR system which operates independently of the EPR.
In addition, the valves can be closed by an independent vacuum or manual trip.
The failure of the aluminum tube representing link 34 in the mechanical linkage to the bypass valves was discuss'ed next with GE.
The tuba was stated s
to be perfectly straight but had a crack through the wall around almost 3600 N
of the circumference. The tube is under compression from the linkage system at all times. The highest compressive load occurs when the bypass valvas open. The failure was noted to appear as a brittle fracture. There were a couple of marks on the tube but no evidence of deformation.
Since aluminum is a ductile material and the tube was straight, it did not appear as if it had failed under axial load or through a bending mode.
The GE service personnel had attached strain gages to the repaired tube.
No measurements had been made up to the time of the meating.
Howaver, the
-I.STG personnel of GE who were present stated that on learnirW of the failure they completely reviewed their stress calculations. The machanical design calculations took into account stall forces which would lead to paak stresses and found no deficiency in the design.
9
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. It was stated that this type of mechanical linkage has been ts-21 on serreral hundred turbines supplied in prior years for fossile plants. This was Lt4 first case where a mechanical fracture had occurred.
The failure mode of the linkage to both the bypass valva and turbine inlat valve was discussed. Failure of a link in either system would rasalt in the valve going to its closed position. The bypass valve would closa in about five seconds under these conditions. The mechanical linkaga to tha hypass valves is distinct and separate from that to the control valvas, ex:cpt for the rotating bar connected by independent levers to the EPR and MPR hydraulic controls. The failure of the mechanical linkage to either valve system would not prevent the operation of the other valve system.
The question of safety review was discussed with both the JC and GE personnel.
The events and linkage failure discussed above had been reviewed by the JC Plant Operations Review Committee (PORC) and the General Operations Review Board (GORB).
Both were satisfied that there was not an unreviewed safety question involved and agreed to resuming operat.fon after repairs had been made. They recommended that the plant be operated under class suparvision and the committee be promptly informed of any further or continuing problems.
The GE safety personnel from APED had examined the problems at Oyster Creek and also concluded there were no unreviewed safety questions. The c ensequences of a failure resulting in the simultaneous closure of both tha turbine inlet valves and the bypass valves had been considered in the FSAR, Section IV-2.
The results were still believed by the licensee to be valid. Further, the main steam isolation valve had been closed in 3 to 10 seconds while at pever during the plant test program. This test simulates to some degrea t otal closure of both turbine inlet and bypass valves.
Questions were asked of JC and GE concerning past occurrences of a similar nature to those recently experienced with the turbine inlet valva control system. JC personnel stated that the plant had run very smoothly and at full power (530 Mwe) since starting back up af ter a maintenance shutdown in the spring. The oil filters on the hydraulic system supplying the servo pilot valve had been replaced infrequently and no significant dP in:reasas had been observed across the filters. There had been a few spikes and control valve cycling of a minor nature observed earlier in the year. The recent unstable control events that occurred on September 17, 1970, had been initiated, they believed, by backwashing the condensers.
Backwash operations tad been conducted during the summer while at full power without " upsets".
JC informed us that they intended to reduce power to an appropriate level prior to backwashing the condenser in subsequent operations.
The EPR control system has given trouble at various times since the. Dyster Creek 1 startup. Some of the earif er experiences with EPR controller mal-function, dirt in the pilot valve and mechanical linkage binding are discussed in Reactor Operating Experiences (ROE) 70-9, " Pressure Regulator Tuning."*
- Also discussed in CO Report No. 219/69-9, Section C; CO Report No. 219/70-5, Section H.; Inquiry Memorandum 219/69-B, and Letter from JC t o Dr. Morris dated November 3, 1969.
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The amplifiers in the EPR system were giving difficulties ear:Icr this year and.wsra to be replaced by GE.
At the time of the current pro 51ents in t ne period September 17-24, 1970, the replacement amplifie.rs had teer deltvered but had bean returned to GE for correction.
6.
Correctiva Actions During the week of September 21, 1970, the hydraulic supply system filters were cleansd and replaced. An additional set of 1/2 mL: ten fil:ers wera installed to series with and after the existing af ter-filt er.
These changes were made to eliminata the pilot valve plugging difficulties.*
The GE service ' epresentativa had taken new data on valva t ravel versus steam r
flow. New cams ara to ba supplied by GE and ba ir. stalled d: ring the natt outage. The new c ams are supposed to previde proper control t2 the higher power range involved in reaching atretch capacity. This will hsid for both the mini-stretch a.d maxi-stretch.
Tha JC - Oyster Crsex maintenance personnal have visually eta A ed all of the me:hanical ifnkage in t ea bypass valve and turbina inlet valvs contrel systems.
They al,o dye cha Aed tte links for evid=n:e of ag c ra:ke. GC is planr.ing to make a careful review of their me:banical linkaga design. They plan to install bi6ger links or use steal in place of aismirrn fer added st raagth.
These ch'anges are to bs made in the nc.xt extended outage.
They also are conwider ttg placing covets ovar the linkage t o prot e;t them from inadvertent damage.
(They ara new open and rJaning betvear steam pipes below t'.a front standard of the tutblea.) The broken tube is to be returned to Schna:tady for matallurgical examination in the natt s Ntdmn.
Hope fully, this will enable them to determine the eause of failure.. Thej have already revtewed their mechanical st ress calculations for the linkage. They plan to chack these against the strain gage maasurements on the t rc, ken link. The
)
personnel from the LSIG of GE indicated a report on the results of their
)
evaluation of the linkage failure would be mada available t o JC and the Complianca insp-eter would be permitted to review it.
Recognizing that t he same cams were still present, the JC operating staff indicated that t hey w,uld drop bae'k load to around 450 Mwe and go on the MPR if they con:13aed t o have trouble with tha EPR.
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