ML20087B107

Co Rept 50-219/70-06 on 700923-25.No Items of Noncompliance Noted.Major Areas Inspected:Reported Instances of Malfunctions of Main Turbine Initial Pressure Regulator & Control Linkage Breakage

ML20087B107
Person / Time
Site:	Oyster Creek
Issue date:	11/02/1970
From:	Robert Carlson, Mcdermott R US ATOMIC ENERGY COMMISSION (AEC)
To:
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References
FOIA-95-36 50-219-70-06, 50-219-70-6, NUDOCS 9508070304
Download: ML20087B107 (48)

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{{#Wiki_filter:.. . 70... _ - ~. _. _ _ 3:d j O i / U. S. AIOMIC ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE l / Report of Inspection

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,k CO Report No. 219/70-6 m i d

Licensee: JERSEY CENIRAL POWER AND LIGHI COMIANY 1 oyster Creek 1 T 9. I.icense No. DPR-16 I f Category C ,-) ' { Dates of Inspection: September 23-25, 1970 Dates of Previous Inspec on: May 18-22, 1970 IMI 14 / 70 Inspected by: R. . M'cDermo t, Reactor Inspector [ Delle. Reviewed by : fa dJ du / 2 0 R. T. CarleoK Senior Reactor Inspector 'Date Proprietary Information: None SCOPE Type of Facility: Boiling Water Reactor Power Level: 1600 Mwt Location: Forked River, New Jersey Type of Inspection: Special, Announced . Accompanvine.. Personnel: Mr. W. Farmer, TSB, CO:HQ accompanied on September 24 { g, and 25, 1970, and assisted in the writing of this t a report. I I Scope of Inspection: A special inspection was made at the site to review I reported instances of malfunctions of the main turbine I initial pressure regulator and a control linkage breakage g* that would have affected the turbine bypass valve operation. GE representatives from APED, San Jose, California and Large Steam Turbine Generator (LSTG) Division in Schenectady, New York, were interviewed at the site to determine t.he cause and significance of the malfunctions and t.he generic considerations for other BWR's. l 9508070304 950227 PDR FOIA DEKO,K95-36 PDR I

i ,6...- O 3 s - i

SUMMARY

i i Safety Items - Norte } Noncompliance Items - None l t f Unusual Occurrences - The Oyster Creek facility has exp-rienced sb disturbances to steam pressure control during the period of Septem %r 17-28, 1970. They have j been exhibited as spikes or oscillations in electrical output ard steam flow and j 4 pressure. Tha measured magnitudes of the transients have been in the range of 5 - 70 Mwe. On two occasions, main steam line high flow instruments tave tripped but the combinations of sensor trips was not sufficient to initiate the closure of the main steam isolation valves. On one occasion, however, low preseure (850 psi) i in the main steam line did initiate main steam isolation valve closare and resulted in a reac tor 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 i regulator (IPR) controls, but one of the disturbances that was experienced was j directly related to a malfunction in the feedwater control system. Three scrams have resulted from these transients, and in all cases, all post-scram functions were reported to have operated normally. During the system check-out following one of the scrams, a broken control linkage was observed that would have prevented the turbine steam bypass valves from opening when required. Corrective 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 pressure regulator (EPR) portion of the IFR; (3) changing of two wire-wound rheostats in the EFR control to composition-type rheostate; (4) the replacement of two amplifiers in the EPR portion of the IPR with amplifiers of a similar design; (5) an inspection and check-out of all wiring within the EPR control syst em for solid connections; (6) i eliminating unwanted grounds and assuring zero resistance grounds where appropriate within the EFR control system; and (7) repairing the broken turbine bypass valve operating linkage. GE personnel assisted in the repairs and the checkout of the malfunctions observed. Personnel from the GE, Installation and Service Engineering (1&SE) Group and the GE, LSIG Group in Schenectady, New York visited the plant to review first-hand the obse.rved malfunctiors. ) i Future planned changes for the facil.ity include the replacement of the broken control linkage t. hat was repaired with one of a new design, the addition of cover ) j plates for the control linkages where appropriate, the changeout of the amplifiers l l' within the EPR to a new design with extended service life, and the replacement of the j l turbine control came to provide for more stable steam pressure control for both the l current licensed power limit and the mini-stretch power increase (1690 Mwt applica-tion which hhs been submitted to DRL and is currently pending). JC-GE are currently evaluating the ned to install damping capacitors within the EPR control system to eliminate steam line " noise" from feeding through the control system. The generic considerations for other WR's may be influenced by the results of the planned metallurgical examination of the broken control leakage. This l examir2atien will be perf orvaed by GE, Schenectady, New York in lat.a Oct ot er, 1970 l l l l

d, ~_ _m .-B O 3 I { During the site management meeting which was held with JC and GE representative.s, I the inspectors were informed that GE is also planning to supply r w control valve j cams to the Nine Mile Point plant. Two additional plants (not idertified but thought by GE representatives to be foreign BWR's) were reported to also be under i consideration for new cams. It was also disclosed that. the hydraulic oil filter l installation that is utilized to filter the oil supply to t.he hydraulic control j i valve (Moog valve) in the EPR control system is not of the same d-: Sign at all BWR's. OC-1 modified their oil filter installation as a result of the recent. l f disturbances. Other BWR's utilizing mechanical-hydraulic turbine control schemes, such as Oyste.r Creek, were reported by GE representatives to include Nine Mile i Point, Millstone 1, Monticello, Vermont Yankee, and Pilgrim. GE personnel from j the LSIG Group have stated that it is GE's policy to initiate any required changes to all nuclear power plants of a common design when a correct ive change is made to e any one of these plants. It should be noted that the Dresden II turbine control is of a different design concept (no mechanical linkages) than the Oyster Creek

plant, f

The results of the discussions with JC and GE personnel on the recent disturbances, including the broken control linkage did not discloes any possibility for a more severe transient t o occur than that previoasly analyred, i.e., turbir_e trip-out without bypass valve opering. I Listed below is a summary description of the disturbances and tLe causes which the licensee at tributes the disturbances to: 1. Turbine-Generater Oscillations - Following a " backwash" (tube cleaning) l operation of the main condenser, on September 17, 1970, tbs generator load began te oscillate 10-15 Mwe. The station load was reduced from 530 to 400 Mwe by operator action and the load reduction resulted in a turbine trip followed by a reactor scram from high flux. The turbine trip was caused by an indicated hf gh moisture level in the moisture-separator drain tank and was assumed by the licensee to be caused by " flashing" of the moisture in the drain tank t. hat resulted from the load reduction. Following a checkout of the initial pressure regulator controls by GE and JC personnel, the reactor f was restart ed on September 18, 1970 JC considered the instability An the i initial presst.re regulator to be caused by improperly designed turb:.ae control valve cams (non linear operation) and t nat the backwash operatien may have i aggravated the situatien by changing condenser efficien:y and hence, control valve position and steam flow and pressure. I 2 Turbine -Generator Spike - During operation at 525 Mwe on Sept ember 20, 1970, j the electrical output of the generator suddenly increased % 20 Mwe followed by a decrease of e 30 Mwe. The station load was reduced by operator action to approximately 450 Mwe and control was transferred from the EPR to the mechanical pressure regulator (MPR). The licensee has attributed this spike to dirt in the Moog valve in the EFR control syste% 3. Turbine-Generator Spike - Daring operation at 495 Mwe on September 21, 1970, (1:33 a.m.) the ele::trical output of the generator suddenly in: teased 55 Mwe' and then decreas-d approximately 70 Mwe. The station load was reduced by operator act ion to 450 Mwe and control was t ransferred from the EPR to the MPR. Tha licer_see attributes this spike to dirt in the Moog valve in the - EPR control system. n n-e ---,,,,-,,m, --m+.- -.-.v-

j ~<.. s i ~) b 4. Turbine-Generator Oscillations - During operation at 455 Mwe on Sept ember 21, a 1970 (9:22 a.m.), turbine vibrations were noted and load was reduced by operator action to 390 Mwe. Vibrations returned to normal but daring attempts to recover t.he load, the station output began t.o oscillate slawry at 410 Mwe '{ with a magnitude of 40 - 50 Mwe. Power was reduced by operater action to 350 Mwe, but the load swings continue for approximately 30 minutes before a e stable system was obtained. Control of the initial pressure regulator was 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 4 5 Mwe ( and reactor steam pressure decreased from 1000 to 970 psi. The operator was l instructed to reduce load to 470 Mwe and to transfer control to the MPR. Steam pressure continued to decrease after transferring control to the MPR l and at. tempts were made to regain control of the EPR. Steam pressure continued l to decrease and the main steam isolation valves closed at 850 psi to initiate a reactor scram. The turbine generator was then manually trippsd. Investiga-l tion disclosed dirt in the Moog valve in the EPR syst.em. During a checkout of the turbine controls following the scram, the turbine steam bypass valves would not respond. A control linkage was found to be broken and was repaired before resuming operation. The exact cause for the broken linkage 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 ou Sept e:mber 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 control sye. tem, 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. 'tatus of Previousiv Reported Problems - None S Other Significant Items - None Management Interview - An exit interview was held with Messrs. McCluskey, Ross, j and Carroll at t.he conclusion of the inspection. The inspectors questioned Mr. McCluskey relative to further planned action if additional generator load ditturbances ware observed. Mr. McCluskey stated that if further disturbances i Stere observed, 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 cam position near full generator load and the other being dirt in tha control oil system. In regard to the foriner, the inspector (Mr. McDermott) stated that it would appear prudent riot. to operate near the critical cam position. Mr. McCluskey m_ ____.____._._._____. _ _. -. - _.. - - - -. - - - - - - - - - - - - - - - - - - - - - - - - - - - - - " - - - - - " - - " - " - - - - - - - ^ - - " - - - - " - - - " - - - - - - - - - - - -

,A _ l--.- s () ) l responded by stating that the plant had operated during the summer at full load without IPR stability problems. He further stated that based on this i history, he intended to return the plant to full load. He alec stated that g the backwashing of the main condenser that preceded the instability problem on y September 18, 1970, had also been routinely performed during the summer months at full generator load conditions without instability resulting tut that it was l his plan to reduce load prior to either turbine valve testing or cendenser backwashing to prevent introducing disturbances into the syetem when near the critical cam position. [ The reportability aspects of these recent events were discussed with Mr. McCluskey and he stated thtt he did not consider these events 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 the week of October 4, 1970.* DETAILS A. Persons Contacted: Jersey Central Power & Light Company Mr. T. McCluskey, Station Superintendent, OC-1 Mr. D. Ross, Technical Supervisor, OC-1 Mr. J. Carroll, Operations Supervisor, 0C-1 General Electric Company Mr. P. C. Callan, Controls Engineer, LSTG Division, Schenectady, NY Mr. R. J. Dickinson, Controls Engineer, LSIG Division, Schenactady, 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 1 C. Operations 1. Description of Events Following the May, 1970 rod work outage, the reactor was restarted and j 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 l vhich was caused by a leaking packing on a recirculation pump discharge l valve. Reactor operations resumed on September 17, 1970, and a series l of steam pressure disturbances have occurred since that time.

• Letter, Finfrock to Morris, dated October 8,1970.

A b i n,A .a .. ~.. n-~_ 4: .y O 3 l' { j ] i E . Listed below is a description of the events as obtained from operations logs t 'and discussions with operating personnel: .i September 17. ' 1970 (Event No.1) Time Sequence of Events l 0448 Reactor startup. [ 2215 1007. power - 530 Mwe. l [ 2230 Started condenser backwash. I 2250 Turbine electrical load began swinging 10-15 Mwe. Operator started to reduce load by red sing re-circulation flow. A loud nzmbling noise from the turbine was noted but its sourca could rat. ba established. 2300 The plant was at 400 Mwa and stable. 2301 Turbine trip and resulting reactor seram followed by a main steam line valve isolation. The isola-l tien condensers were placed in service manually to control system pressure.. All control rods scranuned fully except 30-03 which was valved out of service t at position 48. Scram times of monitored rods ranged from 2.53 to 3.06 seconds. The reactor was brought critical at 0522 on September 18, 1970, J 1 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 j operating personnel and Mr. W. Popov, GE I&SE representative. Investigation i disclosed that the level controller for moisture separator drain tank 1-6 was out of adjustment (proportional band) and the instrument department. t corrected and reset this instrument. It was assumed that the cause of the j f' turbine trip was flashing in the moisture-separator drain tank following i j the load reduction. A hydraulic pilot valve (Moog valve) in the EFR ] i portion of the IPR was disassembled and cleaned, control linkage.s were checked from the front standard on the turbine to the control valves, EPR control system and shock absorber (mounted on the torque tube 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 to the start of the oscillations. The turbine cams were reported to have shown a tendency toward instability at the full power position whenever something occurs to swing the load. Mr. McCluskey informed the inspectors i that OC-1 had previously e.xperienced poor main steam pressure control when l 1 j l

4 ), O 3 ! / { the turbine-generator was operating in the range of 530-535 Mwa. He attributed this to a poor design of the four cams that position the four p turbine control valves through the IFR linkages. Mr. McCluskey further stated that the currently installed cams were the third set of cams to be installed at OC-1 and that another set of cams would be installed 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 acconunodate 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 W e when oscillations began again at 2050 on September 18, 1970. Load was increased to 450 Mwe and tha 10-15 Mwe oscilla-tions continued. Control was then transferred to the MPR at 23M on September 18, 1970, and the oscillations vanished. The oil filters for the EPR were cleaned and control on the EPR was re-established at 1125, September 19, 1970. Load was then increased to 520 Mwe by 1500 with no oscillations. Graphs of selected parameters were obtained for this transient 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 - O to 8 feet above O datum - (The 0 datum level was reported to correspond to % 8 feet above the top of the active fuel) APRMS - 0 to 1507. I i Narrow range reactor pressure - 950 to 1050 psi Turbine first stage steam flow - O to 8 x 106 lbs/hr i' Control valve and bypass valve position indication - O to 100% open f September 20, 1970 (Event No. 2) Time Sequence of Events 0240 With the plant operating at 525 Wa, 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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l .i O } s 8-4 t flow alarms (points 15 and 45 in the events recorder), accompanied by a sudden up-spike of Jilr 20 We. on the . }- turbine-generator output. The four turbine contro1 valve positions swung from 81 to 897. 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-e A tion flow to 500.Mwe and then inserted control rods.- [ to reduce the power further to 450 W e. The turbine initial pressure regulator was being controlled by' 3 F the EPR at this time, j Mr. W. Popov (GE) and Mr. J. Carroll (JC) informed the inspectors that they i attributed this spike to dirt in the Moog valve. The small internal i tolerances of this valve, coupled with inadequate filtering of the supply oil, were considered to be the cause for its sticky operation. Sticky operation i prevents the Moog valve internals from immediately responding. When'this' valve does respond, it will then overshoot its control position and result in sudden control valve motion and finally manifest itself in generator power spikes. Load was increased to 500 Mwe on the EPR with no problems by 1208 on i September 20, 1970. Charts of this event are attached as Figures 5 - 8. l sostenber 21. 1970 (Event No. 3) ] Time Sequence of Events l i 0133 With the load at 500 Mwe-the plant experienced an [ uncontrolled steam flow transient due to. turbine 1 control valve movement. The operator received one alarm of main steam line break high steam flow. Indicated electrical output spiked up by 55 he j followed rapidly by a 70 Mwe down-spike. 1 0134 Load reduction was commenced by reducing recirculation j flow. I 0150 450 he 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 oi1~1ines 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 control valve motion (slight ' hunting') would account for the vibrations. 9

I: fc., u_. ~ 27 O ) 9-Graphs of the selected parameters for this event are shown in Figures 9-12 attached. t September 21.1970 (Event No. 4) Time Secuence of Events 0922 Began generator load reduction to 400 Hwe with recirculation flow due to turbine front standard 4 (control system) and control oil system vibrations. EPR in control. 0942 Began increasing generator load to 450 Mwe with recirculation flow. 1000 Generator load at 410 Mwe 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. l 1401 Transferred control to the EPR. 1445 Reached 500 Mwe. 1 i 1803 515 Mwe - experienced swing in feedwater flow l from 5.9 to 4.7 and back to 6.8 x 106 lbs/hr. I Alarm received for feed pump runout, but alarm I did not lock in. Also received a 3% spike on l the APRM due to the cold water injection. l 2200 Commenced power reduction to 500 Mwe. i l Charts of this event are attached as Figures 13-16. ) September 22, 1970 (Event No. 5) Time Sequence of Events f 0828 " Maximum emergency generation" order given to all stations by grid load dispatcher. 0928 " Voltage warning" given to all stations by the grid load dispatcher. 0939 At 500 Mwe, the reactor scrammed from closure of the main steam isolation valves. Reactor steam pressure initially experienced a decrease

(7 L. J ._.1..__. o O 3 1 i \\ ,.t 'l to approximately 970 psi. The oparator was instructed to reduce load to 470 Mwa and t.o ' transfer control to the MPR. Pressure continued to decrease due to the reactor power cut and when j 850 psi was reached, the main steam isolation valves closed. Prior to the MSIV closure, attempts were made to regain pressure centrol whfie 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 tripped following i the scram. Prior to the scram, the reactor was operating at appr tLmately 1520 Mwt. The first indication of a problem was a very maall swing up in electrical load accompanied by a decrease in reactor pressure. The operator immediately ) started to reduce load to reach a more stable position on the ecntrol 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 valve, 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. j 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 oparation (September l/4 17, 1970). During this investigation it was also found that a rubber gasket I was missing from the normal pre-filter which would have allowed some oil to bypass the filter when in service. This dirt would normnlly be caught by a second filer(sintered metaQL but one of the sealing gaskets on theee filters was t found to be pinched in a manner such that it could also have been bypassing i 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 4 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 I increases to the MIR set point or the set point reduces to the steam pressure, the transfer 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 Ifmit incorporated in the set point controls for the MPR that controls the rate of set point change. i b

c..

A. e L O L') ' Other corrective action taken included a checkout of the EFR wystem. All electrical connections were checked for tightness, response times and f control action of the Moog valve were checked at the turbine front standard, some minor repairs were made to correct the mechanical bindit g that was discovered on the DT-1 Moog position feedback sensor and the spiral orifice of the DT-4 pressure transducer was cleaned. The orifice did contain some solid deposits of foreign material, but it was not plugged. Ite output of l the EPR control amplifiers were checked with an oscilloscope sdth varying input signals into the amplifiers. No spiking or unusual operation was ~ noted. During the course of this investigation, it was discovered that the turbine steam bypass valves would not respond. It was discovered that one of the linkages in the control system had cracked.* The crack. in the tube was almost 3600 around the circumference and would have prevented proper operation of the bypa ss valves, but would not impair the normal operation of the control valves. The efact cause of the failure of the rod was not determined. It was thought that the rod may possibly have been damaged by someone stepping on the rod or hanging a chain hoist from it during the construction period of the plant. The rod is mounted horizontally on a span of approximately 12 feet. Metallurgical examination of the rod will be completed as s'oon as the rod is made available to the GE company. It was stated by Mr. McCluskey that the rod will be removed during the October 18-25, 1970, outage and shipped to GE for a metallurgical examination. He further stated that the ree.ults of the examination would be made available to Compliance. The damaged rod or i linkage was repaired by inserting a hollow steel sleeve insida the original bar and fastening this with bolts. It was established during discussions with GE that a failure of this linkage or any other linkage within the turbina controls would result in closure of either the control valves or bypass valves as these valves are mechanically biased closed. Several of the otter control linkage tubes were dye penetrant inspected for cracks without discovering any additional cracking. GE now plans to provide a replacement linkage for the damaged linkage. The replacement linkage will have a thicker tube wall or be constructed of a stronger material. .l Charts of this event are attached as Figures 17-20. 1 September 23, 1970 j Time Sequence of Events 1157 Reactor critical. 2300 Turbine-generator on bus.

• Discussed further in paragraphs C.3 and C.4.

p I.. ! ^ T. Il -) I september 25, 1970 47 Time Sequence of Events 0030 At 510 Mwe the turbine-generator experienced 1 several 10-15 Mwe swings. The load was reduced i 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 system vera replaced 4 with composition type rheostats.- September 26, 1970 Time Sequence of Events Swing Shift Control was transferred to the EER and the generator loading was increased 530 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 Sequence of Events t 1930 450 Mwe, MPR in control - At this time, a flow signal from the B feedwaiter pump was lost.- The feedwater pumps continued to operate and the feedwater flow control valves 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 j two remaining feed pumps (it was thought that the l ~, pump was lost) ro supply the feed flow demand. The reactor level increased from 80 to 85 inches i 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 to 450 Mwe following the repair and checkout of I this failure. September 30, 1970 Sequence of Events i Continued operating at 450 Mwe while inspecting the EPR controls. , ~.,

~ ,,4 y 3 O / i ! October 1, 1970 Sequence of Eventri Load increased to 530 Mwe on the EFR after replace-ment of two amplifiers on the EFR, cle; king all l control system wiring, and eliminating grou.ds from the control system wiring. No additicnal y problems with the turbine controls or the feed-water controls systems had been experienced during the period of October 1 thra October 18, l 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 (copics 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 system supplies the electrical signal to the pilot valve (or Moog s alve) shown on Figure 23 at i drawing location 5 by K. The hydraulic p,ile' valve controls the positioning { of the master hydraulic servo motor shown.djacent to it. The hydraulic filter plugging and pilot valve port pkgging problems reported above occurred in that portion of the hydraulic syst w shown on Figure 23 at drawing location 2 thru 5 K. The hydraulic servo mo.or of the EPR system, through levers, controls the rotation of the tube at drawing location 2 thru 8 by L on Figure 23. The rotation of this tube is transmitted to the turbine inlet valve and bypass valve controls by mechanical linkage. The mechanical linka;;e which has a run length of 20 to 30 feet is shown in the attached drawing labeled Figure 24. The EPR system is backed up by the MPR identified as the 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 MPR actuates the turbina inlet I valve and bypass valve controls through the same linkage and valve actuation controls as the EPR. l The turbine inlet control cams are shown on Figure 22 in drawing location j. 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 I cams, through a mechanical linkage system, position the hydraulic sarvo pilot valve which positions the main hydraulic servo motor which then 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 s is about in the middle of the drawing. This tube moves in a horizontal plane \\ I.. __

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~ ' " ~ ^ ~ ~ ~ ~ " ' ' ' ~ " " " t [h, {} '] t.. h a L" '- 14'- i ,i transmitting motion from the rotating bar-and lever on ona end to the l rotating bar and_ lever on the other end. The tube is always under a ( [~; compressive load. The tube is made of aluminum with a diameter'around'- i 2-1/2 inches, a wall thickness around 0.150 inches and a length around 12 feet. (These were ' approximate dimensions obtained during visual inspection). - Complete failure of this tube would immobilire the turtire a g' bypass valves but have no effect on the linkage to the turbine inlet valves.. j ,w [ 3. Review of Control Problems and Linkage Failure j 'a "l W The turbine control difficulties and linkage failure ware reviwed with representatives of GE and the JC operations staff at Oyster Crmk 1. The GE representatives stated that it was their opinion that the steady cyclic j instability experienced when operating in EPR control was caused by the l profile on the turbine inlet control valve cams not matching tha valve' { characteristics in the high load range. Thus when a slight plant upset occurs there is not. sufficient damping in the system to provent oscillation.: j i The spikes and large amplitude oscillations shown in the operating < events described above are believed by GE to result from particulates plugging up -l the filters in the hydraulic system supplying oil to the hydraulic pilot-l control valve (or Moog_ valve) or plugging the port of the valve. When these j particulates' break loose, the pilot valve is postulated to overshoot its new; j control position resulting in the turbine inlet valve overshooting its a position. The particle size capable of plugging the filter is almost visible a . to the'haked" eye. ' The existing; hydraulic oil supply system has a filtsr j system as shown on Figure 23.. The pre-filter is a cuno cartridge 1/2 micron. size and the after-filter a 10 micron sintered metallic unit. In addition, there are internal filters in the Moog valve sized for 20 to 40 microns. The- .l

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clearances and ports in the Moog valve are extremelyf small so that the valve : will stick if particles accumuhte. When the Moog valve sticks the EPR I control system is unable to activate either the turbine inlet or bypass valves. However, both valves would be closed at slightly highsr steam pres-l sure by the MPR system which operates' independently of the EPR. 'In addition,. f y?; the valves can be' closed by an independent' vacuum or manual trip. j g

• }

j The failure of the aluminum tube representing link 34 in the mechanical l linkage to the bypass valves was discussed next with GE. The tube was stated! 'l to be perfectly straight but had a crack through the wall around almost 3600' I of the circumference..The tube is under compression from the linkage system-at all times. The highest compressive load occurs when the bypass valves 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 i 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 meeting. However, the 1.STG personnel of GE who were present stated that on Icarnim of the failure they completely reviewed their stress calculations. The mechanical design + calculations took into account stall forces which would lead to peak stresses and found no deficiency in the design.

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It was stated that this type of mechanical linkage has been twi on s-nraal hundred turbines supplied in prior years for fossile plants. This was the. first case where a mechanical fracture had occurred. The failure mode of the linkage to both the bypass valva and t:urbina inlat valve was discussed. Failure of a link in either system would result. in t.he valve going to its closed position. The bypass valve would close in about ~( five seconds under these conditions. The mechanical linkage to tha bypau valves is distinct and separate from that to the control valvas, except ier L q 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 operation after repairs had been made. They recommended that the plant be operated under closa supervision 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 consequences of a failure resulting in the simultaneous closure of both the 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 power during the plant test program. This test simulates to some degrea total 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 valve control system. JC personnel stated that the plant had run very smoothly and at full power (530 Mwe) since starting back up after a maintenance shutdown in the spring. The oil filters on the hydraulic system supplying the servo pilot - j valve had been replaced infrequently and no significant dP increasas had been } observed across the filters. There had baen 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 had been conducted during the summer while at full power without " upsets". JC l 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 Oyster Creek 1 startup. Some of the earlier experiences with EPR controller mal-function, dirt in the pilot valve and mechanical linkage binding are discussed in Reactor Operating Exp-riences (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 to Dr. Morris dated November 3, 1969.

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I The amplifiers in the EPR system were giving difficulties eari2cr this year and w ee to be replaced by GE. At the time of the current pr(Mas in t he 1; period September 17-24, 1970, the replacement amplifiers had been delivered but had been returned to GE for correction. 4. Corrective Acttone Durind the week of September 21, 1970, the hydraulie supply syst a filters g were cleaned a d replaced. An additional set of 1/2 mieren filters were installed in series with and after the existing after-filt er. These changes vere made to eliminat a the pilot valve plugging difficult.ies.* t The GE service representative had taken new data on valve travel versus steam flow. New cams are to be supplied by GE and be installed during the next outage. The new cams are supposed t o previde proper cont rol in the higher power range involved in reaching stretch capacity. This will hAd for both the mini-stretet ar.4 maxi-stretch. The JC - Oyster Creek maintenan:e personnei have visually cN.:k ed all of the mechanical linkage in t ha bypass valve and turbine inlet valve control systems. They also dye checked the links for evide.nce of any c ra:ke. GC is planning to make a careful review of their mechanical linkaga design. They plan to install bigger links or use steel in place of aluminum fer added strength. These changes are to be made in the nut extended outage. They also are considering placing covers over the linkage t o protect them from inadver tent damage. (They are ncv opan and running betwaen steam pipes below the front standard of the turbine.) The troken tube is to be. returned to Schina:tady for matallurgical examination in the next sh:tdown. Hopefully, this will enable them to determine the cause of failure. They have already reviewed their mechanical st revs calculations for the linkage. They plan to check these against the strain gage measurements on the troken link. The personnel from the LSIG of GE indicated a report on the results of their evaluation of the If nkage f ailure would be made available to JC and the Compliance inspecter would be permitted to review it. i l gr Recognizing that the same cams were still present, the JC operating staff indicated that thsy wuld drop back load to around 450 Mwe and go on the MPR if they con:Inaed t o have trouble with the EPR. 4 i

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