Text

To Oyster Creek Nuclear Generating Station: Reactor Water Level Transient of 800717
	ML20063K841
Person / Time
Site:	Oyster Creek
Issue date:	03/19/1981
From:	; GENERAL PUBLIC UTILITIES CORP.
To:	;
Shared Package
ML20063K823	List: ML20063K820; ML20063K841;
References
	TDR-239, TDR-239-R01, TDR-239-R1, NUDOCS 8209080504
	Download: ML20063K841 (104)
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{{#Wiki_filter:_ __2,m._. =m ~ W M Tt-R NO _ _ l PEvlStON NO. 1 239 [ MT !,3cf 1 OF P '3 JECT NO.1100/G0103, TECHNICAL DATA REPOf T ~"~# ~ ^ ' Systems Eng. Misnt Analysis PROJECT: DEPARTMENTIGCCT;Oh Report to GORB 3-19-81 Action Item #375 RELEASE DATE REVISION DATE Reactor Water Level Transient of July 17, 1980 DOCUMENT TITLE: Oyster Creek NCS: '\\0P IGIN A' RE DATE AcPRO'/AL(Si SIGNATURE. DATE QArwW ihvwut 'ame As A ! %A, T T I i l f l O \\ h/ f]Fk.. c.73r.N , ATE D D'S TRIBUTION I O di.-~_- 17P. t y DISTRIBUTION ' A3STRACT: the 81st meeting of the General Office Review Board 1 At T. G. Broughton p a presentation was given by Mr. J. L. Sullivan, Manager J. T. Carroll, Jr. R. J. Chisholm j of Operations at the Oyster Creek N.G.S. on a startup that occurred July 17, 1980. In an effort K. O. E. Fickeissen y[ transientto obtain more information regarding this matter, GO I. R. Finfrock, Jr. q}ActionItem375wasissuedtoMr.R.F.Wilsonforr D. N. Grace The response to these items have been sponse. R. W. Keaten completed by the Plant Analysis Section of System J. Knubel All available information and data J Engineering. transient have been reviewed M. Laggart [ relating to the subject are brief su=maries C. Pwu The following then, and analyzed. b of the results of investigation of each of the Action A. H. Rone J. L. Sullivan i Item questions. N. G. Trikouros b P. S. Walsh

1. Several possible situations have been postuisted j

R. F. Wilson g" in this report which might have occurred during the P. R. Clark with more serious consequences than what 0,l These situations are safety-bounded, transient J. R. Thorpe (12) l" cook place. J. A. Camire (2) however, by the most limiting loss-of-coolant K. R. Goddard ll inventory transient - a complete loss of feedwater from the " Analysis of the Loss of flow. An excerpt

j Feedwater Transient with Isolation Condenser q'

Actuation" developed by EXXON Nuclear Co., Inc. is included in response to the triple-low level i l

Briefly, transient that occurred on May 2, 1979

k it calculated that if feedwater flow goes to zero 3.5 seconds after a scram from 100% reactor power, and a level of 12.34' TAF, approximately 19,300 lbm I

inventory is lost by the time the MSIVs of coolant (Main Steam Isolation Valve) are signaled to close by a " low-low" indication. A minimum downcomer 8209080504 820831 PDR ADOCK 05000219 PDR p

1. ,(Continued) water level of 5.36' TAF occurs at 35 seconds. It is worth noting that about 4 seconds after a full power scram the power level is close to 7%. This is the maximum power the isolation condensers are capable of handling. As is discussed, the plant.was operating'at-close to 7% reactor power during the worst period of the July 17' transient, and the water inventory loss was approximately only 11,600 lbs. This occurred. just prior to. the reactor scram on " low" level, when the bypass valves were full open. Since feedwater was. operable, the level was immediately raised to the normal point. 2. We believe that the G. E. analysis concerning the effective coolant density is the major contributing effect in the cause of the triple-low level signal received during the event. However, the data illustrates a strong contributing ef fect f rom the partial separation of annulus and core areas, and some possible effects from the presence of a pressure wave. 3. We agree with members of the plant operations staff that the torus level indication should b'e upgraded. Since increasing the physical size of the meter to a point where accurate readings would be obtained is impractical, we see the use of a digital indication for this application as the best overall solution. However, it-is suggested that until any instrumentation changeout can be made, the time between readings taken from the level indicator be spaced as far apart as possible, so that any level difference is as large as possible, and will be noticeable enough so as to be read, and not " lost" due to_the accuracy with which the operator can read'the scale. As is evident from these summaries, several areas-related to operational control of the plant are in need of upgrade. Each has safety related considerations which must be treated to preclude similar events from happening, possibly more severe than those that occurred during this transient. A series of curves of the various plant parameters important to an understanding of the transient have been generated to illustrate their behavior during this period. This inf ormation was obtained from that supplied from the plant as part of the PSMS digital trend record. Additionally, notations have been made on some of the curves showing points of interest. Corrective Action '(1) Reactor water level indication Although the sudden opening of the bypass valves may not happen again as it did,.there might be a' failure at some point in the -future which would cause a similar set of conditions of occur. In this case a quick and early automatic scram would prevent a more a a-a ii

l i t We understand that at present a wide range l severe transient. instrument has been installed which measures the core area level, This is due to the l but only when the RCR pumps are not running. l fact that the core _ area level is effected by a number of hydraulic v forces present during plant operation which will not allow the Althougn actual reading and the real core level to be the same. this indicator ma'y not provide an absolute measurement of the true level; it might be possible to correlate readings obtained, with those of the annulus, thus developing a data base for determining l l -what the normal range of level operation is with respect to the f Tracking of this core level vs. the annulus level would annulus. in the measure of communicability between the l additionally assist two regions..The primary purpose of this indicator, however, would be to transmit rapid changes in core area level as would be It is expected that any experienced during water level transients. i change in indicator level should be accurate, to a high degree of a It is this data that would i .si_milar level change in the core area. [ be of paramount importance during future transient analysis where l annulus / core area separation is a factor. Investigation into the t necessary engineering design inf ormation associated with this arrangement should be initiated to determine the feasibility of in this application for i employing an indicator / recorder instrument [ full time operation. I ~ i (2) RBCCW System Isolation t As a result of the spurious RSCCW isolation that occurred during the reexamination of the design l initial phases of the July 17 transient, It appears,that if the bases for the initiation should be reviewed. then RBCCW system can be isolated without an actual loss of coolant, the possibility exists for several temperature related problems to develop in the drywell before automatic initiation of containment Any other response would require operator action ~ spray would occur. A such as a reset of RBCCW or an initiation of containment spray. request from the plant staf f to the Plant Ana,1ysis Section.to analyze the design bases behind RBCCW isolation and containment spray will be forwarded to the appropriate Engineering Section for l In the meantime, the plant has taken steps to insert a l response. time delay in the RBCCW isolation circuitry, which will hopefully l preempt any further unnecessary isolations. (3) Torus Level Readout Instrumentation Analysis of the functional operability requirements of the present l torus level measurement system should be completed, and an improved system proposed. Two possible schemes for consideration are. i t b s

~ ,i O z.:: gw _ a.: .s_ x. _ _. a3 _. Replacement of present vertical scale instrument with one-(1) a. having a digital display and/or recorder trace ' capability. and i ~ b. Possible control board location change. (2) a. Decrease in measurement frequency to allow better ~ vertical scale resolution, and b. Procedure to calculate straight-line, or curve fits to data points, and Any attendant technical specification changes required. c. (4) Computer Analysis Additional analysis of this transient is to be performed by the Safety Analysis and Plant Control Section of Systems Engineering. Part of this analysis will include the use of the RETRAN computer code. RETRAN will identify, more specifically, results of situations more serious than_ those that took place, but which might ,have occurred during the course of the transient. A summary. of this work can be forwarded to the 00RB, if desired. (5) Industry Distribution Despite the fact that it was a non-reportable event, the USAC (Nuclear Safety Analysis Center) people have expressed interest in this transient analysis. Consequently, we will make arrangements to ' forward a copy of this report to them, after obtaining the necessary approvals. Data Collection and Storage (6). I Gathering all the data for this event was dif ficult and time [ consuming. Although we recognize that there is limited capability l for automatic collection of data at Oyster Creek, we recommend that i for all significant events (e.g., reactor trips, etc.) the data base on the Prime Computer should be stored on magnetic tape and all . control room strip chart recorders should be stored in one place for ease of analysis. This will be helpful not only if the event turns out to be unusual, but also data from routine events are necessary-i to benchmark computer codes and training simulat6rs. I l l l i I ,.,,,.,.., _ _, - ~ _ _ _.. -... _ _. _. _ _ _ _ _ _ _. _,, _ _ _ _ _ __. _ _ _ _

u_.,. m.. __,,_u TOR NO. i 239 / WJService PAGE 1 OF 4 TITLE Reactor Water Level Transient Oyster Creek NGS: APPROVAL DATE REV

SUMMARY

OF CHANGE 1 Introduction (Abstract), Page 1 a) Added following to distribution: J. A. Camire 3-11-81 J. R. Thorpe, J. A. Camire, K. R. Goddard P. S. Walsh 3-19-81 Introduction (Abstract) Page 2 1 a) Change " triple-low level indication received" J. A. Camire 3-11-81 to " triple-low level signal received" P. S. Walsh 3-19-81 b) Changed "the separation of" to "the partial separation of" 1 Introduction ( Abstract) Page 3 a) Changed "is with response to the annulus" to "is with respect to the annulus" J. A. Camire 3-11-81 P. S. Walsh 3-19-81 b) Changed " area separation design is a factor" to " area separation is a f actor" Page 2 Main Abstract, 1 a) Changed " triple-low level indication received" J. A. Camire 3-11-81 to " triple-low level signal received" P. S. Walsh 3-19-81 b) Changed "from the separation of" to "from the partial separation" l Question No. I response, Page 4 a) Changed " triple low level actuation" to "triplo low level signal actuation" b) Changed "of a triple low was" to "of a triple-low signal condition was" c) Changed "that a triple-low condition was" to "that a triple low signal condition was" J. A. Camire 3 81 P. S. Walsh 3-19-81 d) Changed " triple low water / level signals" to " triple low water level signals" 1 1 A000 0017

3 TDR NO. i Service 239 EI PAGE 2 op 4 Oyster Creek NGS: Reactor Water Level Transient REV

SUMMARY

OF CHANGE APPROVAL DATE 1 Question No. 1 response, Page 5 a) Changed "of the triple low level in both" to "of the triple low level signal in both" b) Changed "the triple low indication of the" to "the triple low signal indication during the" c) Changed "At the point when the triple-low signal initiated" to "At the time when the bypass valves opened and a triple-low signal initiated" d) Changed "that the triple-low occurred" to "that the triple-low signal occurred" e) Changed "and a " level" was reached" to "and a pressure difference was reached" f) Changed "to the triple-low initiation" to "to the triple-low signal initiation" J. A. Camire 3-11-81 i g) Changed "by the computer, but that" to "by P. S. Walsh 3-19-81 the computer, and that" 1 Question No. I response, Page 6 a) Changed " seal destruction, and the pumps would" to " seal destruction, if the pumps J. A. Camire 3-11-81 we re no t" P. S. Walsh 3-19-81 1 ' Question No. I response, Page 7 ~ ~ a) Changed " lost the level would" to " lost the true water level would" b) Changed " loss, but would maintain the artifically ~~ high indicated level. This indication" to " loss, while maintaining an artifically high mixture level. The indication" c) Deleted " Compounding the indication descrepaner, l 1s the fact that no direct indication of the l core area level was operable" l d) Changed "It is assumed" to "It was evidently a 3-e) an " time a triple-low trip was" to [e p ,j 3_ g time a triple-low signal was" \\ A000 0017

TOR NO., 239 M ] Service Ir OF TITLE PAGE 3 4 Oyster Creek NGS: Reactor Water Level Transient APPROVAL DATE REV

SUMMARY

OF CHANGE 1 Questien No. 2 response, Page 9 a) Changed " lowering the water level continuously, until a triple-low water level was reached" to " lowering the true water level continuously until a triple-low level signal was received" b) Changed "13d" to and" c) Changed "for the triple-low level to be reached" to "for the triple-low level signal to be received" d) Changed "cause the triple-low level, but" to J. A. Camire 3-11-81 "cause the triple-low level signal, but" P. S. Walsh 3-19-81 e) Changed "for triple-low level initiation" to "for triple low level signal i'nitiation" 1 Question No. 2 response, Page 10 a) Changed "Although there is no data or indicator reading reflecting the vessel level inside the shroud" to "Although the indicator for measuring level inside the shroud area was not operational during the event" b) Changed "the triple-low level transmitters should" to "the triple-low level switches should" c) Changed "the annulus region, now momentarily s tagnant, and the leveling" to "the annulus region, and the leveling" d) Changed "why the triple-low level on the ~ event" to "why the triple-low level signal indication on the event" e) Changed " board annunciator for triple-low - to " board annunciator alarm for triple-low indica tion" J. A. Camire 3-11-81 f) Changed "from the separation". co_"frca the P. S. Wal sh 3-19-81 partial separation" 1 Question No. 3 response, Page 12 a) Changed "could then be completed from" to J. A. Camire 3-11-81 could then be computed from" P. S. Walsh 3-19-81 1 Corrective Action, Page 13 a) Changed " steps to insert a time delay in the RBCCW" to " steps to have a time delay inserted J. A. Camire 3-11-81 P. S. Walsh 3-19-81 in the RBCCW" 1 Appendix A, Page 1 and 2 J. A. Camire 3-11-81 a) Changed " Sequence of Events" to " Plant P. S. Walsh 3-19-81 Reported Sequence of Events" b) Added "p..m." to lis t of tbnes. A000 0017 '~

L. ..~,. <x.. __._w. _ :. w t-(rj g llSerVice 239 TITLE PAGE 4 OF 4 Oyster Creek NGS: Reactor Water Level Transient REV

SUMMARY

OF CHANGE APPROVAL DATE 1 Appendix B, all curves a) Added specific times for each tic mark on " TIME" axis on all transient curves. b) Added " ANNULUS" to those transient curves J. A. Camire 3-11-81 labeled only " Reactor Water Level" P. S. Walsh 3-19-81 1 Appendix C, Cover Page a) Changed " Recirculation flow = gallons / minute" J. A. Camire 3-11-81 to " Recirculation flow gallons / minutes x.01" P. S. Walsh 3-19-81 1 Appendix H, Cover Page, Test a) Changed "0yster Creek - LER 80-52/3L"to J. A. Camire 3-11-81 '" Oyster Creek - LER 38/3L" P. S. Walsh 3-19-81 b) Replaced LER 80-52 with LER 80-38 1 Table of Contents a) Changed "0yster Creek - LER 80-52/3L" to J. A. Camire 3-11-81 "0ys ter Creek - LER 80-38/3L" P. S. Walsh 3-19-81 O t l e A000 0017

p g . w i-a._ _.c__ -_,_-~ _- . s ' '~ t., '^ s I: t 3.. TABLE' 0F CONTENTS I. Abstract II. Response to Question No. 1 ,r III. Response to Question No. 2 IV. Response to Quest'lon No. 3 V. List.of Corrective Actions -VI; Appendix A: Original Plant Sequence of Events VII. Appendix B: Transient Curves

~"

VIII. Appendia C: P3MS Transient _ Digital Data 'IX. Appendix D: GE responses to various NRC and JCP&L questions X. Appendix E: "Bourding Loss 'o f Coolant Inventory Transient for the i Oyster Creek Plant" by the EXXON thselear Company. Inc. (*/a rt ia 1) \\ LER 79-022) X I. Appendix F: AIS003-09 (Big Rock Point XII. Appendix G: Re:spense from Oyster Creek ( A. H. Rone) to PA-193 XI;II. Appendix 3: Oy s t er Cre ek - LER 80 -38 / 3L ' Dri f t o f Tri pl e - Low Water Level Switche s' XIV. ' Appendix I: FRC Inspection Report for i July 9 - August 1, 1980 Report No. 50-219/80-25 (pa rt ial) e j,. a p. ,c' edw. -~h + ,s

o REPORT TO CORB Action Item No. 375 Oyster Creek N.C.S.: Reac t o r Wa ter Lcvel Transient of July 17, 1980 O M W M e

ABSTRACT At the Sist meeting of the General Of fice Review Board a presentation was given by Mr. J. L. Sullivan, Manager of Operations at the Oyster Creek 17, 1980. In an effort N. G. S. on a startup transient that occurred July 375 was to obtain more information regarding this matter, COR3 Action ItemThe contents of the issued to Mr. R. F. Wilson f or response. are as f ollows: including possible resulting 1. Investigate July 17 startup transient, (Note: If appro-events, to determine seriousness of situation. priate, results should be forwarded to NSAC and Owners Group.) Identify, if possible, why the triple-low level signal was received 2. during the July 17, 1980 startup transient. Investigate the leak race determination capability (accuracy, (If improvements are found to be necessary, they should be 3. added to the list of corrective items, taking their proper place in etc.). the priority.) The response to these items have been completed by the Plant Analysis Sec-All available information and data relating to tion of Systems Engineering. The following then, the subject transient have been reviewed and analyzed. are brief summaries of the results of investigation of each of the Action J Item questions: Several possible situations have been postulated which might have occurred during the transient with more serious consequences than 1. These situations are safety-bounded, however, by what took place. inventory transient - a complete the most limiting loss-of-coolant from the " Analysis of the Loss loss of feedwater flow. An excerpt of Feedwater Transient with Isolation Condenser Actuation" developed This docu-by EXXON Nuclear Co., Inc., is included in Appendix E. ment was generated in response to the triple-low level transient that occured on tiay 2,1979. Briefly, it calculated that if feed-water flow goes to zero 3.5 seconds after a scram from 100% reactor ~~ power, and a level of 12.34' TAF, approximately 19,300 lbe of cool-anc inventory is lost by the time the MSIVs (Main Steam Isolation ~ A minimum Valve) are signaled to close by a " low-low" indication. It is downcomer water level of 5.36' TAF occurs at 35 seconds. that about 4 seconds af ter a full power worth noting at this point This is the maximum power the scram the power level is close to 7%. As is discussed, the isolation condensers are capable of handling. plant was operating at close to 7%' reactor power during the worst and the water inventory loss was period of the July 17 transient,This occurred just prior to the 11,600'1bm. approximately onlyreactor scram on " low" level, when the bypass valves were Since feedwater was operable, the level was immediately open. raised to the normal point. 1_

1 2. We believe that the G.E. analysis concerning the effective coolant density is the major contributing effect in the cause of the triple-low level signal received during the event. However, the data illustrates a strong contributing effect from the partial separation of annulus and core areas and some possible effects from the presence of a pressure wave. 3. Before a final decision can be made in dealing with the question of torus level indication, further analysis must be made. Specifically, the functional operability requirements of the present measurement system should be defined from the original design criteria. As this is better understood, an effort to match the present arrangement with human factor engineering principles can be attempted. In the near future, a major review of the entire Oyster Creek control room instrumentation is to be undertaken in this manner. Two possible solutions which might be considered in this review are presented. As is evident from these summaries, several areas related to operational control of the plant are in need of upgrade. Each has safety related con-siderations which must be treated to preclude similar events from happening, possibly more severe than those that occurred during this transient. A series of curves of the various plant parameters important to an under-f standing of the transient have been generated to illustrate their behavior during this period. This information was obtained from that supplied from the plant as part of the PSMS digital trend record. Additionally, notations ( have been made on some of the curves showing points of interest. l O, I m. e 4 S I O g e 9 k n W M

r COR3 Action Item No. 375 Investigate July 17, 1980 startup transient, including Question No. 1: (Note: possible resulting events, to determine seriousness of situation. if appropri, ate, results should be forwarded to NSAC and Owners Group.) The following is a brief summary of the events that occurred At approximately 8:30 p.m. on the night

Response

17, 1980 transient. during the July operators were attempting to control reactor pressure dur-l 17, 1980 of July As pres-ing a normal ascent to power with the pressure regulator system. sure in the vessel continued to rise it was realized that the regulating the bypass valves were not open-system was not operating properly, in thatSome control rods were subsequently ing to adjust the increasing pressure.All their attempts to open the bypass valves at this time failed. When the vessel pressure reached 1050 psig, the "D" EMRV inserted by the operator. JElectromatic Relief Valve) automatically opened until the pressure dropped Coincident with the valve closure, 1000 psig, at which point it closed. an automatic RBCCI (Reactor Building Closed Cooling Water) isolation was to The operator pushed the reset and manually opened the isolation received. During the time the relief valve was open, a sharp rise in water This was caused by " swell", the voiding and flashing valves. level was indicated. Af ter of the reactor water to steam, pushing the observed level upwards. As the pressure control was sr*11 not } the EMRV closed, the level dropped. functioning after the relief valve closure, vessel pressure began to in-of control rod crease again, but at a slower rate reflecting the effect In order to avoid a second opening of the j insertion commenced earlier. CiRV, the operator decided to try checking the reset mechanism for the #2 vacuum trip. Despite the presence of a light indication which supposedly signaled a properly reset vacuum trip, when the reset was engaged the bypass to set vessel pressure at the demand valves immediately opened in an effort Realizing this, the operator value which by now had been set very low. instantly tripped the #2 vacuum trip causing the bypass valves to reclose. period of time when the valves were open, a second sharp During the short rise in water level was indicated, much larger than that produced by the Steam flow and feed flow EMRV, but caused by the same voiding mechanism. sharply increased, as they would be expected to behave during a rapid de-Since the bulk boiling and resulting voiding in the pressurization event. the flux dropped immediately also, as core is a negative reactivity effect, After the valves evidenced by the APRM (Average Power Range Monitor) curve. closed, pressure drop was arrested, the voids were collapsed, and'the rapid the "true, solid" water drop in indicated level momentarily terminated atThis "true' level, approximately 10'11" TAF level, before turning upward. and the (Top of Active Fuel), was below the " low" water level trip point,However, the reactor scrammed. The drop in level to the normal range of about 13'6" TAF (see Appendix B). the excessive loss of vessel water in-level was produced by two effects: ventory caused by a higher steam flow vs. feed flow during depressurization, and the surge effect, or ' splash', caused by the fall of water after void This surging mechanism would temporarily cause the indicated collapse. -annulus area to fall until the movement of water due to the pressure O e _____ _ _- - __- - __-- _- _

difference between the core and annulus reached equilibrium. This effect would serve to increase the magnitude of the level drop beyond what might be due to inventory loss. No obva7us reason can be determined for the spurious isolation of the RBCCW system upon closing of the CiRV. Normally, isolation shot)i only occur upon a triple-low level signal actuation, but at that point no indication of a triple-low signal condition was apparent. We understand that it is possible for a spurious trip to have occurred without any actuation due to the fact that this isolation circuitry is new and was just installed during the last spring outage. The possibility also exists that a triple-low signal con-I dition was in effect at some point during the EMRV evolution which, although it did not trip the event recorder or the annunciator alarm, might have caused the RBCCW isolation relay to trip out. RBCCW isolation is designed to occur after two of the four triple-low level switches are actuated. With the actuation of one switch, the event recorder should activate to note the actuation and to increase the speed of the chart paper. It was believed that the. spurious RBCCW isolation signal was received af ter the EMRV had closed. This was based on verbal reconstruction of the events by the three senior reactor operators present in the control room during the event, not on any recorded information, since none is available. A more plausible explanation would have placed the isolation with the actual triple-low received at the time of the bypass valves opening. However, repeated discussions with the operators present at the time confirm the sequence of f events correct as initially reported. So, it is difficult to determine why it is that the RBCCW isolation occurred at all during the EMRV closing. \\ Ruling out the assumption that it was a spurious affect, and therefore hap-pened at that point in time only coincidently with the EMRV closure, we have attempted to find another reasonable cause for the isolation. Af ter some review of the data, a plausible connection is that two triple-low switches were engaged at some point, and jumped "in" and "out" fast enough to avoid being detected by the triple-low annunciator and event recorder, probably because they are part of an old, slower acting circuitry, but not able to escape the notice of the newer RBCCW isolation electronics. The reverse was true during the byphss valve opening. The triple low water level signals l jumped 'in' and 'out', slower than before, and were picked up by the event recorded circuitry, but this time not by the RBCCW insolation circuitry. The explanation for this, we believe, lies in the combination of signal speed and electro-mechanical component construction. Short, pulsed signals can cause effects which are not normally expected, due to the characteristics of the mechanical portions of relays. When signals are pulsed, unusual effects on the operation of the device become apparent. This is due to the fact that the pulse tim,es.in question are comparable to, say, the decay times for magnetic fields, or the times needed to overcome inertia. Further investigation of the triple low level circuitry and its j components must be made to determine the extent to which these effects are l present, before a complete undertstanding of the cause of the physical events that took place can be afforded. I

i The events are connected by the presence of the triple-low level signal in both cases. However, it can be seen that the absolute water level peaks of The EMRV peak has been cut off by the scan each transient do not match. for each side of l frequency of the computer, but using a least squares fitAdditionally, the triple-low the peak, the peak point was reconstructed. i signal indication during the bypass valves opening transient occurs at a level much lower than its peak. Af ter closer examination it can be seen that the dif ference in points is not overly large. (No te that these are core area levels and only give us an indirect measures of annulus levels not is occurring in the core area.) At the time when the bypass valves opened and a triple-low signal initiated, the two water level indication of what instruments measuring annulus level read 6 64' and 6.30', respectively. j This is an average of 6.47' + 0 17'. When the E3RV level peak is reconstructed it is seen to have a top value of Assuming that a triple-low signal occured at that point, although it 6 18'. indicate as such, it should be allowed the same percentage range In this case, 6 18' did not deviation as was present during the bypass valve. peak. Comparing the median values of both peaks, 6.47' vs. 6.18', the + 0.16'. 3 5", is well within the 2-3% accuracy range of difference 0.29' or about Additionally, the upper range of the the triple-low ins truments (3"-4. 5"). EMRV peak at 6.34', overlaps the lower range of the bypass vaave peak at Although there is a similar basis for comparison of triple-low trip 6.30'. the signals from the annulus area levels of the two depressurization events, from the RCR (Re-triple-low level signal was not obtained until the flow Fortunately, the ~ actor Coolant Recirculation) pumps were reduced as well. computer scan occurred during the time when the recirculation flow was in~ dicating a low flow condition. From the event recorder trace it is known time, perhaps l the triple-low signal occurred only for a very short that And we have no way of knowing if the recirculation flow milliseconds. indicates the actual minimum or not. In any case, it is clear that more this time, than was being inventory was being lost from the vessel, at activated all of the four l added, and a cressure difference was reached that Since this low triple-low switches, despite any setpoint variations. l to the triple-low signal initiation recirculation flow is important mechanisms as will be discussed later, we must assume that some additional drop in recirculation flow, i.e. annulus / core separation, was present beyo'nd it occurred between scans during l what is indicated by the computer, and that Also, it may be said that if the RBCCW isolation occurred the EiRV opening.it may have done so from an initiation by only two triple-low - at this point, relays, those whose setpoints are in the most conservative direction in relation to the others (See Appendix H). appears, after generating the curves of the various nucit ar system parameter.s f rom the PStiS data supplied and comparing these c urves with each j It other, that the actual sequence of events as viewed by the camputer a for the most list of events, with the exception of some estimated times and absolute values of parameters. there are several possible situa-In examining the seriousness of this event We have analyzed four tions which could have occurred during the transient. events in which additional failures were assumed. l -S-

a 4i They are the following: (a) Bypass valves failing to open for any reaso0; (b) Bypass vaives failing to reclose for any reason; (c) RBCCW System fails to reset af ter trip isolation; and (d) Erroneous reactor water level indication due to rapid depressur-ization. A discussion of each is.given below.- ( A) Bypass valves failing to open for any reason. Should the bypass valves have failed to open when they did, the pressure in the reactor would have continued to increase as before, with the EMRV lif ting to relieve pressure. A controlled shutdown or manual scram could easily have been initiated and any excess pressure controlled with the isolation condensers. Feedwater flow, and all ECCS (Emergency Core Cooling Systems) were operable during this time. Similar action would, of course, be taken in the event the CMRV also failed to open, with other relief valves opening, and an automatic scram of the reactor oc-curring if the pressure reached the trip setpoint. (B) Bypass valves failing to reclose for any reason. Calculations of net inventory loss during a 25 second transient period when the steam flow jumped from a stable value due to the bypass valves opening, until valve closure, show that about 11,600 lba of coolant were lost. Had the bypass valves remained open, a " low-low" level (7'2" TAF) would have been reached approximately 46 seconds af ter the initial opening, based on extrapolation of the available data. This " low-low" level assumably would isolate the reactor vessel if it was detected. However, the feedwater system would be capable of maintaining inven-tory. Evidence from analysis of the parameters, notably water level, show that due to " swell" of the apparent level the transmitter was '" fooled", since indication on the Yarway read approximately 6.88' when in f2ct the " collapsed" level was only 3.76'. This was obvious when the bypass yalves finally did close. More will be said on this in (D). f (C) RBCCW System fails to reset af ter trip isolation. l In the event of a spurious RBCCW isolation similar to the one that oc-curred, but one where the RBCCW fails to reset, the cooling water for the RCR pumps' seals would become overheated causing seal destruction, l if the pumps were not eventually tripped manually. With the feedwater l pumps operable, enough injection water should be available to make up any loss of inventory adequately, despite feeding through the RCR pumps and lines, since after the DIRV reclosed the reactor would again be essentially isolated, except for leakage by the RCR pump seals. With i i , i i

-e F the loss of RBCCW, temperatures in the drywell would rise due to pump seal leakage, and the loss of cooling water to the drywell coolers. However, containment spray would have been available if temperatures or - pressures warranted its use. (D) Erroneous reactor water level indication due to rapid depressurization. When a rapid depressurization in the reactor vessel occurs for whatever reason, the reaction of the water level is to -be drawn upwards to some This is ca'used by the mass of fluid rapidly escaping from higher level. the vessel, and by the void formation in the water caused by the The difference between the level presence of bulk boiling in the core. indicated during this time, and the level that would be measured without voiding, the " collapsed" level, may be significant. During the trans-ient period when the bypass valves were opened, the maximum indicated 6.88' (14.05' TAF). However, after the valves closed, and reading was the " collapsed" level was again established, the reading was 3.76' (10.93' TAF). This is a difference of 312'. As net inventory is lost the true water level would gradually drop to reflect the loss, while maintaining an artifically high mixture level. The indication problem is a critical area of importance arising out of this transient as far as our analysis is concerned, and there seems to be no obvious solution. ~ It was evidently assumed during design that communication between the l the annulus area and the core area would always be maintained such that annulus area level would be strongly representative of'che core water level. However, the results of this transient, and at least one other, show that there are times when this communication can be lost, or severely reduced. During one five'second period of the transient, the RCR pumps dropped 53-88% in flow at which time the annulus level dipped the same slightly, showing a level on one meter of 13.47' TAF, when at time a triple-low signal was present corresponding to a collapsed water l level of 4.67' TAFt Further analysis of this phenomena is made in the response to Question No. 2. Summary These situations are safety-bounded by the most limiting loss-of-coolant inventory transient - a complete loss of feedwater flow. An excerpt from the " Analysis of the Loss of Feedwater Transient with Isolation Condenser . Actuation" developed by EXXON Nuclear Co., Inc., is included in Appendix E. This document was generated in response to the triple-low level transient that occured on May.2, 1979. Briefly, it calculated that if feedwater flow f goes to zero 3.5 seconds after a scram from 100% reactor power, and a level of 12.34' TAF, approximately 19,300 lba of coolant inventory is lost by the time the MSIVs (tiain Steam Isolation Valve) are signaled to close by a " low-low" indication. A minimum downcomer water level of 5 36' TAF occurs It is worth noting at this point that about 4 seconds af ter at 35 seconds. a full power scram the power level is close to 7%. This is the maximum power the isolation condensers are capable of handling. E

4 As was discussed, the plant was operating at close to 7% reactor power, during the worst period of the July 17 transient, and the water inventory loss was approximately 11,600 lbm. This occured just prior to the reactor scram on " low" level, when the bypass valves were full open. Since the feedwater system remained operable the level was icunediately raised to the normal point. N e e e g-O E W = e o O e e _.-

E f Identify, if possible, why the triple-low level signal was Question No. 2: received during t'he July 17, 1980 startup transient. There are several possible reasons for a triple-low level indi-

Response

cation occuring at Oyster Creek during the July 17 transient: (a) Annulus / Core Area Separation; 1 (b) Pressure Wave; and { (c) Effective Coolant Density. A discussion of each is presented as follows. The Annulus / Core area separation theory (A) Annulus / Core Area Separation. assumes that due to the rapid depressuriza l tion in flow from the recirculation pumps occurred making it impossible l for enough feedwater to enter the core area, thereby restricting its This scenario obviously has . access to the source of cooling water. With l important repercussions for a plant of Oyster Creek's design. pr continued heat in the vessel, the remaining water in the core area would be heated to t f l steam, thus lowering the true water level continuously, until a l triple-low level signal was received. The possibility of this separation'was presented to G.E. by the NRC. G.E. rejected the possibility of any long term pump cavitation and the Their remarks can be examined in resultant annulus / core separation. Basically, C.E. maintains that the prin-i Appendix D, Questions 2 and 4. ciple cause of the observed low pump flow condition was the reduction in L driving head produced by the density difference between the annulus and At RCR (Reactor Coolant Recirculation) pump l core area, not cavitation. minimum speed conditions this is the major component of flow, and during the transient, voiding and flashing in the vessel a) reduced the net j I For these positive suction head, and b) increased hydraulic losses.almost immediately after a: reasons then it was observed that pass valves were opened, all five of the BCR pumps showed low readings However, G.E. maintains that on their flow recorders (See Appendix 3). i even with a total loss of RCR flow it would require more than a minute l and a half of boil-off for the triple-low level signal to be received. i We agree with this portion of the G.E. analysis and have concluded that l l this mechanism alone did not cause the triple-low level signal, but was a major contributing factor. The pressure wave theory was initially believed to have l (3) Pressure Uave.been the most logical mechanism for triple-low level signal initiation. The wave fronts generated by the closure of the EMRV, and the bypass cap of the transmitter first, causing the the uppermost I valves arrive at to interpret the presence of a level lower than is actually this wave was believed to have been instrumentDuring the July transient, However, no present. strong enough to cause initiation of the triple-low signal. j mention of this effect is found in the G.E. explanation, and if this e O e -9_

effect is in fact present, it is certainly masked and/or outweighed by the sharp water level drop, seen on. the reactor level curve #2 as a valley between two peaks, caused by the reduction in recirculation pump flow. Although no visible effects of a pressure wave are evident, we cannot completely rule out the possibility of some impact on the trans-ient. (C) Effective Coolant Density. According to the C.E. analysis (see Appendix D) either a true reduced water level, or a reduced ' effective coolant density' will cause a triple-low level signal. The ' effective coolant density' is defined as the void fraction in the separator standpipes. During the transient event it is postulated that the core area was close to a void fraction of 34%. Although the indicator for measuring level inside the shroua area was not operational during the event, an idea of the level can be seen by examining the curves of annulus reactor water level. The sharp level increases observed are attributable to 'the " swell" in volume of coolant caused by the depressurization of the ves-sel. The formation of vapor voids displaces the, liquid coolant and 'causes the f ree surf ace as interpreted by the instrumentation to be moved upwards, when in fact the true " collapsed" level is less that the original level because of a nat loss of inventory. This is seen graph-ically when, after the bypass valves close, the voids collapse and the annulus level drops to approximately 3 80'. As calculated by C.E., during vessel depressurization the ' effective coolant density' was such that the triple-low level switches should not l have actuated, since the void fraction was still below that necessary for actuation. However, at the point when the vessel pressure was close to its minimum, the flows of the recirculation pumps were severely restricted due to the reduction in NPSH effect of the near saturation conditions. The net loss in water inventory was at its maximum and level in the core area dropped due to continued steam production. The drop in the annulus level is attributable to both the increased feed-unter flow sharply cooling the water in the annulus region, and the' l leveling off of pressure drop in the vessel. As these effects will serve to collapse the voids, the level indicates a drop. The recirculation pumps then recovered and reestablished flow to the vessel. These events occurred in a very short time frame, and this explains why the triple-low level signal indication on the event l Erecorder was seen to trip "in" and "out" in such a short time, and consequently neither activated the control board annunciator alarm for triple-low indication, nor caused RBCCW system isolation. Summary We believe that the G.E. analysis as outlined above in (C) is the best esti-mate for the cause of the triple-low level indication received during the ~ July 17,. 1980 Oyster Creek event. However, the data illustrates a strong contributing effect from the partial separation of annulus and core areas l and some 'possible effects from. the presence of a pressure wave. ,e _

l Investigate the leak rate determination capability (accur-3: Question No. (if improvements are found to be necessary, they should be acy, etc.). added to the list of corrective items, taking their proper place in the priority.) The leak rate determination capability refers to the identified into the primary containment.

Response

and unidentified leakage of reactor coolant The following excerpt from the technical specifications for Oyster Creek outlines the requirements f or leak rate determination: ( 3.3 Reactor Coolant D. Reactor Coolant System Leakage from uniden- " Reactor coolant leakage into the primary containmentthe total leak-In addition, tified sources shall not exceed 5 gpm. exceed age in the containment, identified and unidentified shall not the reactor will be If these conditions cannot be met, 25 gpm. placed in the cold shutdown condition." At the present time, the standard procedures for the operating personnel at Assuming the maxi-the plant calls for a torus level reading every 8 hours. mum technical specification unidentified leakage of 5 gpm occurs to the operator would need a device capable of torus during this time, the plant The maximum accurately estimating a change of 2400 gallons in the torus. and minimum water volumes in the torus are specified at 92,000 cu. ft. and 82,000 cu. ft., respectively. At a temperature of 60*F, this equates to a maximum of 688,211 gallons and a minimum of 613,405 gallons or a difference The two full scale torus level indicators at the plant of 74,705 gallons. are calibrated for a 40" water range,,or 1870 gallons per inch of torus The problem in reading these meters becomes apparent when water height. realizing that on a typical 6" meter scale graduated for 40" of height,Therefore, an op one inch is about 5/32" of the meter scale. a technical specification violation using the vertical =eters on 'the control board would be required to determine a dif f erence in pointer reading of about 3/16" over an eight hour period on a meter that is only knee high. It should be noted that although unidentified leakage finds its way to the torus, the drywell sump will also accumulate leakage from unidentified sour-Theref orey an even smaller The total will be the sum of the-two, of torus level increase may need to be determined to assess its con-ces. prior to comparison amount tribution to the entire unidentified leakage volume, with technical specification limits. There are two recorders which monitor the torus level in the control room in These recorders are located on a back addition to the-two level indicators. One is labeled as a " wide panel in a different area from the indicators. range" recorder and the othe The relatively narrow range scales when compared to the 40" indicators. " narrow range" recorder measures torus level from +3" to +7", a span of 4" t

w~. w.. ~ _. ~. 1 - u w _= _a (7,480 gallons). The " wide range" recorder reads level f rom -5" to +15", a range span of 10" (18,700 gallons). Tracking of the torus level by these recorders is automatic. A hard copy trace is conveniently made, which can be later used for trending or leakage calculations. However,there are times during plant operation evolutions when these recorders are not on scale and the torus level measurement for technical specifications must be made from the indicators. Summagt Before a final decision can be made in dealing with the question of torus level indication, further analysis must be made, specifically, the functional operability requirements of the present measurement system should be defined from the original design criteria. na this is better understood, an effort to match the present arrangement with human f actor engineering principles can be attempted. In the near future, a major review of the entire Oyster Creek control room instrumentation is to be undertaken in this manner. Two possible solutions which might be considered in this review are presented below: (1) A review of the technical specifications concerning torus level and identified / unidentified leakage may find that an increase in the time between torus level readings may be sufficient, with present instrumentation, to obtain on indicated rate of leakage. Given some constant leakage into the torus, for instance, the measurement of the level as followed by the operators would be a series of data points which could be fit with a straight line. The maximum leakage could then be computed f rom that line and compared with technical specification l limits. The possibility of a change in the technical specifications in regard to these limits should also be considered. (2) A changeout of the present vertical level indicators is the plant pref erred method for dealing with this problem. It is suggested that. L, the merits of a digital indicator be thoroughly examined. With a [ digital LED type readout little reading error should remain, precluding the necessity for allowing large periods of time to pass betve.en indicator readings to minimize error. Additionally, a wide range j recorder could be installed, compatible with the scale of the indicator. It would be of some value in trending or computing, leakage _, [ rates, when the situation warranted, as during those times when the " narrow" and " wide" range level recorders were unable to track. In the event that an instrument changeout is decided upon as the correct solution, a re-evaluation of the location of these instruments on the ( control board might also be in order. l i i 1 ~ r

4 CORB ACTION ITEM No. 375 Corrective Action (1) Reactor water level indication Although the sudden opening of the bypass valves ma'y not happen again as in the future which would it did, there might be a ' failure at some pointIn this case a quick and cause a.similar set of conditions to occur. We under-early automatic scram would prevent a more severe transient. stand that at present a wide range instrument has been installed which measures the core area level, but only when the RCR pumps are not run-the core area level is effected by a ning. This is due to the fact that number of hydraulic forces present during plant operation which will not Al-allow the actual reading and the real core level to be the same. though this indicator may not provide an absolute measurement of the obtained, with true level, it_might be possible to. correlate readings those of the annulus, thus developing a data base for determining what to the annulus. the normal range of level operation is with respect Tracking of this core level vs. the annulus level would additionallyThe assist in the measure of com=unicability between the two regions. primary purpose of this indicator, however, would be to transmit rapid I changes in core area level as.would be experienced during water level It is expected that any change in indicated level should be transients. accurate, to a high degree, of a similar level change in the core area. It is this data that would be of paramount importance during future transient analysis where annulus / core area separation is a factor. Investigation into the necessary engineering design information asso-ciated with this arrangement should be initiated to determine the feas- ~ in this appli-ibility of employing an indicator / recorder instrument cation for full time operation. (2) RBCCW System Isolation As a result of the spurious RBCCW isolation that occurred during the reexamination of the design initial phases of the July 17 transient, bases for the initiation should be reviewed. It appears that if the an actual loss of coolant, then the RBCCW system can be isolated without possibility exists for several temperature related problems to develop in the drywell before automatic initiation of containment spray would Any other response would require operator action such as a reset A request from the occur. of RBCCW or an initiation of containment spray. plant staff to the Plant Analysis Section to analyze the design baIses behind RBCCW isolation and containment spray will be forwarded to the In the meantime, the appropriate Engineering Section for response. l plant has taken steps to have a time' delay inserted in the RBCCW isolation circuitry, which will hopefully preempt any further unnecessary isolations.

t ~ (3) Torus Level Readout Instrumentation Analysis of the function'al operability requirements of the present torus level measurement system should be completed, and an improved system proposed. Two possible schemes for consideration are: 1. (a) Replacement of present vertical scale instrument with one having a digital display and/or recorder trace capability, and (b) Possible control board location change. 2. (a) Decrease in measurement frequency to allow better vertical scale resolution, and (b) Procedure to calculate straight-line, or curve fits to data points, and (c) Any attendant technical specification changes required. (4) Computer Analysis Additional analysis of this transient is to be performed by the Safety Analysis and Plant Centrol Section of Systems Engineering. Part of this analysis will include the use of the RETRAN computer code. RETRAN will identify, more specifically, results of situations more serious than those that took place, but which might have occurred during the course of the transient.- A summary of this work can he f orwarded to the GORB if desired. [ (5) Industry Distribution s Based on analysis of this transient; despite the fact that it was a non-reportable event, the NSAC (Nuclear Safety Analysis Center) people have expressed interest in this transient analysis. Consequently, we will make arrangements to forward a copy of this report to them, after obtaining the necessary approvals. ~ (6) Data Collection and Storage ~~ Cathering all the data for this event was difficult and time consuming._ Although we recognize that there is limited capability f or automatic collection of data at Oyster Creek, we recommend that for all signifi-cant events (e.g. reactor trips, etc.) the data base on the Prime Com-puter should be stored on magnetic tape and all control room strip-chart recorders should be stored in one place for ease of analysis. This will be helpful not only if the event turns out to be unusual, but also data from routine events are necessary to benchmark computer codes and train-ing simulators.

I L.

l Report to CORB ' Action Item No. 375 ~ ~ 3 Appendix A j Original Plant Sequence of Events s e e a O 9 9 4 e O e

_ __ _._ _...._ _ m,...... m....__.... ~ I PLANT REPORTED SEOUENCE OF EVENTS ^ l 8:30-8i35 pC Increasing power and attempting to control Reactor Pressure I with EPR/MPR and Bypass Valve opening. Reactor level (feed-water) controls were in manual mode. Bypass Valves not responding. Tried Bypass Valves opening jack with no success. 8:35-8:38 With pressure increasing to $1050 psig, "D" ERY cpened to reduce pressure. "0" stayed open for 14 seconds until pressure dropped to.S1000 psig at which point it closed. At this point operators received RSCCW isolation signal and the 3 RSCCW isolation Valves to the Drywell closed. Operator pushed the reset and manually opened the Isolation Valves. Other than the Isolation Alarms for RECCW no other related alarms or signals were present. 8:36-8:41 With EMRV "D" closed and the Turbine Pressure Control still not functioning, reactor pressure started increasing from 1000 psig to s1032 psig at wnich point, because of Control Rod insertion and resultant power decrease, the pressure started to turn around and slowly decrease. 8:41:30 Operator at the pressure control still unable to control F Rx pressure with normal controls (Bypass valves) and thinking pressure increasing to ERV set point, went to back panel and reset the 12 Vacuum Trip. This action caused the Bypass Valves to open on demand causing rapid depressurization of (' the Reactor from 4028 psig to 944 psig. 8:41:33 Operator realizing the effect of the Bypass Valve opening, tripped the 32 Vacuum Trip causing the Sypass Valves to go closed. Just prior to this action the rapid depressurization and the high coolant temperature caused voiding in all regions of the Reactor vessel with corresponding level increase of s2 feet and loss of suction on Recire. Pumps resulting in low Recire, flow. 8:41:47 When the Bypass Valves closed the pressure change and resulting. Hydraulic disturbance caused the Rx triple low sensors (4) to s actuate (Bounce in and out) and pick up the event recorder. At this time there was no heavy loss of inventory and the sensor actuation is believed to have been caused by the pressure transient sensed in the annulus and in the reference leg of the sensors. WithF the configuration of this instrument the reference leg senses from the annulus and the variable leg senses frcm the Top of the Core region. It is believed that the pressure transient was seen first in the reference leg and this created an imbalance in the DP to actuate the sensors. 8:42:11 With the Bypass valves closed, the pressure started increasing and Reactor Water level decreased due to collapse of the steam voids. Reactor Level decreased to the Law Level scram setpoint and the Reactor scrammed. Reactor level further decreased to approximately 37" on the Yarway ($10'3" above top of core). The operator then increased level manually and maintained approximately 75" on Yarway (13'5"TAF).

-. '. :. :., = .+ PLANT REPORTED l SEQUENCE OF EVENTS Page 2 8:43 pm Reactor pressure continued to decrease and experienced same l operation with pressure below' saturation temperature and resultant flashing to steam as evidenced by recire flow and Temperature. Reactor shutdown continued with mode switch in shutdown and the Reactor Coolant was brought to less than 212eF at 6:30 a.m. The NRC was notified of the scram at 9:40 and latei updated via the hot line at 4:15 a.m. on the condition and results of our investigation. An investigation was conducted of the events and parameters associated with the scram. Participating in the investigation were the following personnel: A. H. Rone, K. O. E. Fickeissen, H. Howey, J. L. Sullivan and John Thomas of the NRC. / e er- .4 eum 6 1 :.. e M enus i

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~ Report to CORB Action Item No. 37 5 Appendix C PSMS Transient Digital Data The units given below for the parameters indicated should be used Note: when working with the,PSMS data. Feedwater flow = Thousands of ib/hr Steam flow = 'Ihousands of lb/hr . Recirculation flow = gallons / minute x.01 O 4 O' e l )
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4 ~ ' ,eOITTPUT FROM PSMS DICITAL TRE11D-RSTFLO RPRVID RPRtfAR RF.ACWL REACWL CTHPOV M 3 11 M S 1 I I I 2 I K Hl.II/IIR PSIG PSIC FEtT FEET MVfH e l-7 17 23 29 231581.50 966.151. 978.56 5.95 5.97 122.61 5.93 5.96 117.03 17 20 29 291577.50 967.63 979.69 5.93 5.94 117.03 77 17'20 29 351577.50 968.63 9110.75 5.119 5.93 117.03 17 23 29 4115141.50 969.96 9111.08 5. 1111 5.90 117.03 7 57 2*> 29 4715110.00 970.56 9113.00 5. 11 7 5.90 117.03 7 7 17 20 29 5315fl0.00 971.G4 9114.06 5. 11 5 5.88 117.43 7 17 20 29 591579.00 972.75 9115.06 7 17 20 30 5151!3.50 974.13 9116.13 5. 11 3 5. 11 7 117.0 F 7 17 2D 30 1IIGita.50 975.31 9117.19 5. 11 2 5.84 117>.03 7172030 1715112.50 975.96 91111.I9 5. 11 0 5. 14 3 117.03 7 17 29 30 23I5116.00 977.31 9119. 3 8 5.79 5.112 117.03 5.76 5.1:0 117.51 7 17 20 30 2915115.00 978.23 990.38 5.75 5.78 117.51 17 20 30 351532.50 978.118 991.38 5.73 5.76 117.51 7 7 17 20 30 4115182.50 9110.13 992.44 5.72 5.75 117.51 i 7 17 20 30 4715117.50 9111.69 993.69 5.70 5.73 117.51 e, 7 17 20 30 5415115.50 9112.56 994.94 7 17 20 30 5915415.00 9153.50 996.19 4 68 5.78 117.51 7 17 20 31 6 5116.00 9116.75 997.38 5.67 5.69 117.51 7 17 20 31 1815I18.50 9fl6.31 9911.63 5.65 5.67 117.51 7 17 20 31 1715117.50 937.19 999.fil 5.63 5.65 117.51 7 17 20 31 23:587.50 9118.631601.00 5.61 5.64 117.51 1 7 17 20 31 2915118.50-9119.631002.13 5.60 p.63 117.64 7 17 20 31 361590.00 990.961003.19 5. 5 11 5.60 117.64 7 17 20 31 421590.00 991.561004.31 5.55 5.59 117.64 7 17 20 31 471591.00 992.941005.30 If.53. 5.57 117.64 7 17 20 31 531591.00 993.631606.44 5.51 5.55 117.64 t 7 17 20 31 591592.50 994.11111007.50 5.49 5.53 117.64 17 20 32 51591.00 996.061001).50 .f.47 5.50 117.64 5.46 5.49 117.64 7 17 20 32 111591.00 997.061009.50 17:594.50 99fl.311010.50 5.44 5.47 117.64 7 17 20 32 5.42 5.45 I17.64 7 17 20 32 231593.50 9911.75101I.50 5.41 5.43 117.77 7 17 20 32 291593.50I000.001012.44 5.38 5.41 117.77-7 17 20 32 351596.001001.001613.44 5.37 5.39 117.77 7 7 17 20 32 411597.001002.191014.44 5.35 5.38 117.77 7 17 20 32 481597.001002.941015.69 5.34 5.37 117.77 17 20 32 541597.001004.001016.31 5.34 5.35 117.77 7 7 17 20 32 591597.001004.751017.25 7 17 29 33 61598.501005.301018.19 5.31 5.34 11,7.77 5.28 5.32 117.77 7 17 20 33 111599.501006.691019.06 ' 5.26 5.29 117.77 g 7 17 20 33 1715911.50I007.691020.00 O 6 'g a i g .I s 8 e y i f i. e. 'l ? 0 o n
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~~+ + + + + + + t t + eS eS et e3 C eS e2 -s eS e3 = = = = = = = - = = c = O c c c = C C O N. CI C3 C3 08 0101 C3 C8. e.l el C. C. C C C C o C C O n n k2 n n c et t3 v3 t2 = = = = =.=.= A )C P. =~~
e. el el ef el et el et el en et e c e e e e e e c e @ @ @ @ @ @ @ @ @ e = c3 = e c c o m m a et f.l. CI ?! Cl 03 C.3 08 C1 c.i f.a C3 CI Cl Cl CI C.8 Cl Of C3 e.3. c3 C C C3 e.3 c3. C c.3. c.3. C C C.3. C c.3. C c.1. c.3. C Y
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$Y e4. C.l i.l C8 08 cl C4 0 8 i.l e.l C.l ?J. c1. C.I N. C.I C3 C. C. C. C. C. T3. N. O. =. n. C3.==. C. C. e. e. O. C. C. o. C. C.==. - U w gg gg W W3 W3 v3 r3 t3 t2 n n n t1 W3 W3 W3 t1 n W N3 W3 W3 W3 W3 W3 W3 W3 4 @ W3 n r3 t3 t319 W1 W3 p us 45 ut US @ + C C3 fl = 0 C C C C = = el C3 + @ P. C C el es t19 0 N C c3 C = N @ W3 @ C e =. C3 t1 N U P. CI Cl 01 f4 C.l f a el c4. CJ. C.I C.4 C3. CJ. f3. C.l C3. f3. CI C4 C3. C. C. C. t. O. o. W3. C.3. C. e. e. e. e. o. e. C. C. C. C. w A gal A ut n n L1 t2 W3 t3 W3 n n n US t3 W3 n k2 W3 n t3 n t3 r3 t3 O @ @ W3 t3 us ti t t t t t t W3 o kt et C C n v @n.n O f + 9 9 e C n. e C C C @ C = @ e O + n,I -- @ e. = n. C. C O_ n. e nnCCn-CCt= ~ Q C.. e.. = e. e. e. e... @ = n. C. . C. =. n. =..C m... . C.. e. g 2.,. rfs f E el c1 + t2 4 t c3 e C - el + + @ @ N O e C.= - - 04 + ? e el + n N c C.= C W.3 @ t e C.= A h 31 es c4 N cl f 3 is el el il cl 01 c3 ~3 c3 C1 c3 c3 c3 e2 -3 c1 + + v elor3 C o C C o = Q C. C..C C C C. C C C C C C C C C C C C C C o C C C o. e C C G C o. C. @. @. C C. C. C. C. C. l m. = C. r3. n. @. e. e. =. =. C. ?. O. -. C = + e = C W3 v + @ n - e C e3 + e M 4 @ c n e. c3 C c.3 + 4 l f.4 Q C C e + + C C'3 - + @ C 3 C. e. @. C. 3. P. Y. t. C. r. C. =. Q. 4. e.. O. O. k3. O. t. =.D.=.. v. v3 .L. (p) g -.= rf) c e C - c3 C t14 N C. e C - el et c3 + n @ N c" ? e cl es C e e3 + @ N o o el e2 r3 4 N c3 M A A r/J g C C - - = = = = = = - cl es el fl c3 21 !! 72 ts il el fl e=ceoeoeeoCCCCoC I. l 'A C C C C C C C C C. C C. C C C o O. C. C C C C. o. C C e e o e e e e e o o o C o C C C. c g t I C .J
~~= CCCCCCoccCCCCCcCCCooCCoCCCCCCCCCCCoCCCCC Ng W1. n. n. r3. W. C. C. Li.n. C. W. C. o. r3. n. W3. n. n. n. W3. n. G. C. o. C. o. t2 o. C. o. n. o. o. o. O. t1. E3. v3. C. C.~~
C N g g (^l ,.2 C e e e e3 fl ?! + W1 P. + t. N e e e e e o e o P. C @ + t 4 e e C - e + r c @ N @ C C (s. v g x ceeeCCCCCCCCCoCCCC=C=CCCNka@@hNrN h > N r. tC C y c e n. n @ @ @ @ @.@ @ @ @ @ @ @ @ @ @ @ @- @ @ Li. n @ n. n n n n n. n n n n n. n. e. n L . t3 = O U U ". = 03 C3 C3 9 9 US =.=.C4C C T T n = = Cl CI C t t t1 W3 N C8 C v t n n a = = CCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCC ~ M CCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCC N Cl ti N to C8 ti ci tI tt ::
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C.3. C C C C 3. C C O C C C C "3 C C C. 03 C3. C. O. C1 C.10103 O.J. 08 c.i c.a el C1. c f.3 C c.c. c c c c F.3 C3 t3 to C== c3 t3 @ C C - ?! t3 t3 P. N e C C O N N t* N to N O N O t* @ @ @ @ @ @ @ It3 n n .... e.n es en e.s e.s e3. e. e. n. e. e. n. n. e. v. n. e. n. e. c. o. n. n. e. e. n. e3. e. n. :3. n. n. es. e3. e3. e3 g3 gg W W n W3 W3 n v3 n t3 L1 e n t3 at3 us t3 t3 41 tr3 It3 t3 t3 t3 e n trJ n as n n n n n n n as us n t3 n w3 m3 M i& ..J
~~j C = + @ N.=.C4 f.l f3 C4. f.iil 03 C C M M C3 C 73 C O C C C 73 M C 73 C3 73 C* c3 C e3 c3 c3 C C es n N cll C e at + t3 @ @ @ @ v3 d C3 + n n + + + + + T3 e3 et CS f3 CS CS et *4 3~~
- ld .a -==== m y 3: W lt3 t3 n v3 = W3 e n W3 t3 t3 t's t3 n t3 e r3 v3 n e L1 t3 t3 e t3 t3 t3 t2 r3 t3 k2 v3 t3 e a WS n t3 ks n y CYe F-M.tl 0 e3 =. @ e 4 e e C n e e - e CS C O C + c C t O + C C.= 0 t1 + @ c +.c t3 C C.= C c3 9 .2 g'g @ f3. t2. C. @ n. f3. @. 73.. @. n. O. tt3. J.* C1== e. C. e. t3. C. O. c. *J. @. at3. C3 e. W3. C3. C. t3 c3 O @ f E. .= q !=. c t= et f3 e t3 @ N N C e C C - et el c f3 t + c n + 9 a'3 e3 e3 et et ?! = = = C C C e e e C c3 0 M c3 et et el 08 et el et 73 -3 03 -3 e3 -3 c3 -3 M r3 CS 73 e3 c3 03 S c3 73 f3 as e3 c3 M M e3 ~3 el ?? ci es es ;4 .O O O C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C.= C C g O O C3 O @ + @ its O E3 C - + e c: n t3 C @ * - e v t3 n C M = C = :::: = @ @ - + @ M ** e3 e.=3 O
== O. O. T. t3. t=. C. C.I C. C. C' @ O. tw N. C. C. @. C. =. c. N. *.l O. @. =. n. O. O. O. O. C. Q. e. O. @. e. =. 0 7, g=g ~ll c. J O C C3 !! O v t. @- @ to t.w= O.= e e C - el - ?! *l - - - O C C e e e C C O N P. P. 4 @ @ @ k2 /3 g = = 01 el el eI fI eI ?! 01 el fI ea fI - = = - - = = -. -===== = CCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCC g g C 3 O CCCCCDCCCCCCCCCCCCCCCC'3CCCCCCCCCCCCCCCCC 3
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.= C C atit=?!03t3 ale 3@C3 5witsr3ft@t3e3*fet73"I- - C - C C e C E C O O O C:: O O O O O = 0 C O O O O O O O O.** O O O O O O O O O O O O C O O t= c3 O y n a W3 n n L*J L1 L3 K3 W3 n t2 n n n n n t1 n a n_ n 40 t3 t3 O t'3 K3 n n n t3 C C n_ n_ n WS v_3 0
= 3 -_ = os e c @ - e., e n - = + e @ - = n e n e: G e c e - O n C n - t. n o n - P. + e @ es e. C3 ?! O t v t3 t1 - = = CJ 04 C t T C t2 - = f4Cl C 9 y n = = C1C 03 9 1 W3 o
== 3 M tw t P. r* N N tw C * * ** ** " " " *'" "* e e e e e e e e e C C C C C C C C C C C = = = C C2 e3 c3 73 C3 c e3 C 03 C O c3 03 O C c3 C3 M C3 C C c3 73 C C 7 e v v v v v v v v f 9 9 9 E CCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCCC,, CJ C8 C4 CA C4 C8 et C3 Cl ea C3 C4 CJ C4 C4 ca ta 04 :3 CI C3 c4 CJ CA Ca ti ci el C8 C4 c4 ci cl CI C4 f4 C8 ct C4 C4 4 ..,- 3
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O. P N t* N N t.*. t.w. t* t* t* b t* b N. P t.*. t.w. t.o t.w. t* P N P= N P N N P.*. P= b N N N t.w N.= N N P t.o to. - ~ - E ' tw t* t* t* r. t. t* t* N tw t* N t. to t* t. N N N N to t* t* to N t-N t N t N to t. N t-N t* b t* t*
f, ~ OITTPlfr FROM PSMS DICITAL TREN3 N D.H M S ILWFLO RPRWID RPRNAR REACH 1. REACWL CniPOW I I I I 2 I (NLit/IIR PSIC PSIC FELT FEET MhTil 5.36 131.51 @ $8 Vdt b.I #MM 7 17'20 41 231581.501016.001028.19 5.33 JY gefe f kWe /rV 82 7 17 jl3 41 291379.0010 8 5.3111027.88 5.33 5.36 132.71 .ppyra -/e r ~ /'e,g e,a) v4C"drh43 ~ 7 17/20 48 352632.50 975.44 986.81 6.22 6.18 132.714-gp ofm g fg jI 7 17 20 41 421934.00 9411.63' -1.00g 6.46 6.75 132.71 s ' ~7 17 20-41 471553.00 944.69 -1.00 n 6.64 6.30 132.71 4-. rgiptg g [f j agvf L [O) 7 17 20 48 541547.00 944.19 -1.od A 6.87 6.88 132.71 7 17 20 41 591547.00 944.69 -1.00f 6.05 6.03 132.71 7.17 20 42 61546.00 945.96 -l.00 y 4.61 4.63 132.7 7 17 20 42 181549.50 9414.13- -1.00 g 4.23 4.26 132.71 RtabC bC#4M 3. 11 8 3.92 432. 7 17 20 42 171546.00 9418.38 -1.00 4 32.7,1'p,5 oc h N t'r w 7 17 20 42 331548.50 9411.38 -1.00 3.76 3.80'1 7 17 20 42 291547.00 947.118 -1.00f3.90
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OITPPUT FRort PSMS DIGITAL 'IREND M D 11 M S APRM APRM APRM APRM AP8M APRM APRM APRM FhTEMP FWFl.0W 1 2 3 4 5 6 '7 8 I I / ElI AhT .'.f / 5 F PR X F PR K F PR X F PR X F PR X F PR % F PR X F PR DEC F 4H91 7 17 20 37 24 7.51 9.41 II.26 9.23 10.47 8.88 8.37 9.04 112.28 373.40 7 17 20 37 29 7.52 9.43 II.27 9.25 10.415 11. 9 0 8.38 9.05 182.22 373.40 i 7 17 20 37 36 7.53 9.43 11.30 9.26 10.51 8. 85 8 8.40 9.04 112.22 373.40 7 17 20 37 48 7.55 9.45 11.27 9.26 10.51 8.90 8.40 9.05 112.09 (173.40 7 17 20 37 47 7.54 9.43 11.26 9.24 10.49 11. 9 0 8. 3 11 9.02 112.22 373.40 7 17 20 37 54 7.51 9.:18 11.23 9.21 10.42 8. 11 5 11. 3 1 8.99 112.16 373.40 7 17 20 37 59 7.52 9.39 11.22 9.19 10.44 11. 11 4 8.32 11.98 1 2.06 373.40 7 17 20 Gil 5 7.44 9.31 11.20 9.17 40.36 8.77 8.31 11. 9 3 112.16 373.40 7 17 20 311 Il 7.41 9.31 11.89 9.14' 10.34 11. 7 7 8. 2 18 11. 9 3 111.97 373.40 7 17 20 311 la 7.09 9.28 11.37 9.13 10.33 11. 7 5 8.25 11. 9 1 111.97 373.40 7 17 20 Ull 24 7.34 9.23.11.14 9.09 10.29 11. 7 0 11. 2 3 11.118 111.88 373.40 7 17 20 38 29 7.34 9.24 11.13 9.09 10.29 8.71 11. 2 0 fl.llil I l l.Illi 363.13 7 17 20 38 36 7.29 9.23 11.09 9. 0 11 10.25 8. 6 11 11. 1 6 8.116 111.94 363.13 i 7 17 20 311 41 7.26 9. 1 81 11.09 9.03 10.20 fl. 63 11. 1 2 8. 11 1 111.84 363.13 7 17 20 311 48 7.16 9.09 11.01 8.97 10.11 11. 5 6 fl. 05 8.77 111.91 363.13 7 17 20 all 33 7.16 9.04 10.98 8.94 10.06 11. 5 3 8.03 11.72 111.111 363.13 .l.80 10.00 8.44 7.97 11. 6 6 111.81 363.13 7 17 20 all 59 7.30 8.96 10.92 7 17 20 39 3 7.02 8.113 10.84 8.79 9.92 8.38 7.89 11.58 111.72 363.13 7.17 20 39 12 6. 11 7 8.52 10.75 8.63 9.77 11. 0 9 7.77 8.42 111.01 363.13 7 17 20 39 18 6.71 8.23 10.67
11. 4 6 9.62 7. 11 4 7.66 8.27 Ill.66 363.13 7 17 20 39 23 6.59 8.02 10.59 11. 3 4 9.'50 7.61 7.55 11.13 111.72 363.13 7 17 20 39 29 6.39 7.75 10.47 8.16 9.30 7.30 7.41 7.94 Ill.66 343.61 7 17 20 39 35 6.25 7.57 10.40

11. 0 5 9.17 7.09 7.30 7.83 til.56 343.61 7 17 20 39 41 6. 0 11 7.39 10.30 7.93 9.00 6.(18 7.17 7.66 111.63 343.68 7 17 20'39 48 5.90 7.21 10.20 - 7.77 8.79 6.66 7.03 7.51 111.53 343.61 7 17 20 39 53 5.81 7.13 10.13 7.70 8.70 6.58 6.94 7.43 111.53 343.61 7 17 20 40 0 5.66 6.93 10.02 7.59 8.55 6.42 6.79 7. 2 11 111.50 343.61 7 17 20 40 5-5. 5 11 6.96 9. 9.1 7.55

8. 4 fl 6.35 6.73 7.23 III.63 343.61 7 17 20 40 11 5.52 6. 1111 9.91 7.48 8.41 6.29 6.68 7.16 Ill.53 343.61 7 17 20 40 17 5.44 6. 11 2

9. fl7 7.44

14. 3 3 6.22 6.60 7.09 111.50 343.68 7 17 20 40 23 5.44 6. 11 6 9. 85 3 7.45

11. 3 4 6.23 6.60 7.10 111.44 343.61 7 17 20 4C 30 5.47

6. fl5 9.98 7.4 fl (1.37 6.25 6.63 7.13 111.44 353.37 7 17 20 40 35 5.48 6. 11 7 9.93 7.50 ft.3fl 6.27 6.65 7.15 111.38 353.37 7 17 20 40 41 5.52 6.90 9.96 7.53 11. 4 1 6.30 6.68
~~7.18 181.41 353.37 j 7 17 20 40 47 5.53 6.91 9.98 7.54 8.43 6.31 6.70 7.20 111.38 353.37~~
./ 7 17 20 40 54 5.56 6.92 10.01 7.55 8.46 6.34 6.73 7.20 111.38 353.37 7 17 to 40 59 5.58 6.93 10.02 7.56 8.48 6.35 6.74 7.22111.34353.377.)h/ 7 17 20 41 6 5.61 6.95 10.03 7.57 8.51 6.38 6.77 7.23 111.31 353.37 7 17 20 41 12 5.63 6.95 10.03 7.59 8.51 6.36 6.76 7.27 III.31 353.37 7 17 20 41 17, 5.64 6.93 10.03 7.57 8.51,.6.36 6.77 7.23 111.25 353.37 i I i ,t.. a*" s .sf i I ). - i. g-a. *,
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y . ' OtrPPirI* FROM PSMS SICITAL TREN) z M D.D M.S APRM APRM APRM APRM APRM APRM APRM APRM FWrEMP FWFLOW ? 1 2 3 4 5 6 7 8 I 1 I N F PR N F PR N F PR N F PR N F PR N F PR N F PR E F PR DEC F CPM Y 17 20 45 24-4.34 3.18 5.52 2.85 2.56 2.69 1.26 1.95 107.72 30fl.69 7 17 23 45 29 0.34 3.16 5.52 2. 15 5 2.56 2.09 1.26 1.95 107.78 308.69 7 17 20 45 35 0.34 3.18 5.52 2. 11 5 2.56 2.09 1.26 1.95 107.72 308.69 7 17 20 45 41 0.34 3.16 5.40 2.85 2.56 2.09 1.26 1.95 107.56 3011.69 7 17 23 45 47-4.34 3.16 5.50 2.85 2.53 2.05 1.26 'I.94 107.66 3011.69 717234553 0.33 3.15 5.47 2. 11 4 ~ -2.51 2.05 1.26 - 1.94 107.44 3013.69 { 7 17 2) 46 'O 1).34 3.I6 5.49 2.85 2.52 2.06 1.26 1.94 107.53 308.69 1.26 1.94 107.31 308.69 7172046 6 e.34 3.16 5.50 2.84 2.52 2.06 717234613 0.34 3.16 5.50 2. 11 3 2.52 2.06-I.26 I.94 107.38 308.69 717 23 46 la 0.34 3.16 5.50 2. 11 5 2.52 2.06 1.26 8.94 107.25 3014.69 7 17 23 46 25-0.34 3.16 5.50 2.85 2.52 2.06 I.26 1.92 107.31 1103.69 7 17 23 46 32 6.34 3.16 5.50
2. f15 2.51 2.06 1.26 I.98 107.19 285.04 7 17 23 46 37 6.34 3.15 5,50 2.85 2.52 2.98 1.26 1.94 197.25 2115.06 7 17 23 46 44 0.34 3.16 5.50 2.85 2.52 2.06 1.26 1.94 107.13 2115.06 7 17 23 46 50 0.36 3.16 5.52 2. 11 7 2.54 2.08 1.27 f.97 107.89 2115.06 717234654 6.34
'3.15 5.50 2.85 2.52 2.06 1.26 1.95 107.13 215.06 7 17 23 47 0 0.33 3.16 5.52 2. 14 5 2.52 2.06 1.26 8.94 107.13 2H5.06 7 17 O 47 7 0.34 3.15 5.50 2.85 2.58 2.06 1.26 I.92 107.03 2115.06 737234714 0.33 3.16 5.50 2. 11 5 2.52 2.06 1.26 - 1.94 107.03 2115.06 - 7 17 2B 47 19 0.33 3.36 5.48 2.114 2.52 2.06 1.26 8.94 107.03 2145.06 i 7 17 23 47 26 O.33 3.16 5.50 2.85 2.52 2.06 1.26 1.94 106.98 2110.95 717234731 0.34 3.15 5.50 2.85 2.52 2.06 I.26 1.94 107.03 2110.95 7 17 23 47 37 0.33 3.15 5.50 2.85-2.52 2.06 .1.24 1.94 106.98 2110.95 , T.- 7172')4743 0.33 3.16 5.50 2. 11 3 2.51 2.06 1.24 I.94 106.91 2110.9 5 7 17 23 47 49 0.34 3.16 5.50 2.84 2.52 2.06 1.24 1.94 106.78 2110.95 7 17 28 47 56 9.34 3.16 5.50 2. 11 5 2.52 2.05 1.24 1.94 106.84 280.95 7 17 23 411 0-6.34 3.15 5.53 2.H4 2.52 2.05 1.24 1.94 106.78 2863.95 7 37 23 411 7 0.33 3.15 5.50 2.85 2.52 2.06 1.26 1.94 1 0 6.7 2 2410.9 5 7 17 23 411 14 8.34 3.15 5.50 2. 11 5 2.52 2.06 1.24 1.94 106.(14 2110.95 7 17 23 4fl 20 0.34 3.16 5.50 2. 18 4 2.52 2.06 1.24 1.94 106.72 280.95 7 17 23 411 26 0.34 3.20 5.48 2. 11 5 2.52 2.05 1.24 1.94 106.72 2110.95 7 17 20 411 32 0.34 3.20 5.50 2. 16 4 2.52 2.05 1.24 1.94 106.63 2110.95 7 17 23 48 37 0.34 3.16 5.50 2. 11 5 2.52 2.06 1.24 1.94 106.72 2110.95 717234844-6.34 3.15 5.50
~~2. Il4 2.52 2.05 1.26 1.94 106.63 2110.95 7 17 20 411 50 0.34 3.15 5.50 2.84 2.52 2.05 1.24 1.94 106.56 2110.95 7 17 20 411 55 0.34 3.15 5.50 2.84 2.51 2.05 I.24 1.94 106.50 2110.95 7 17 23 49 2-0.34.~~
3.15 5.50 2. 11 5 2.58 2.06 1.24 1.94 106.44 280.95 7 17 2D 49 7 4.33 3.16 5.47. 2. 11 4 2.52 2.06 1.24 1.94 106.44 280.95 7 17 23 49 14 G.34 3.15 5.50 2.85 2.52 2.05 1.24 1.94 106.44 288.95 7 17 O 49 19 g.34 3.15 5.48, 2.85, 2.52 :2.05 1.24 I.94 166.38 200.95 Q.4f e e i. l. ~ i. I Se ,1 a
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P' OUTPITT FROM PSMS D1CITAL TREftD s' 3O M S APRM APRM APRM APRM APRM APftM APRM APRM FVTEMP FWFLOW I 2 3 4 5 6 7 8 1 1 CPM N F PR K F PR 5 F PR E F PK K F tR M F ER X F PR % F PR DEC F -l e );- ie & 20 49 26 0.33 3.15 5.49 2.84 2.52 2.05 1.24 1.94 106.38 280.95 L 7 23 <9 31 0.34 3.36 5.50 2.84 2.52 2.05 1.24 1.94 106.311 290.71 7 2) 49 37 0.34 3.15 5.50 3. 11 5 2.51 2.05 1.24 1.94 106.*lt 290.71 7234944 0.33 3.15 5.4ft 2. 11 5 2.52 2.05 1.23 1.94 106.25 290.71 P 7234949 0.34 3.15 5.50
~~2. Il4 2.51 2.06 1.24 I.94 106.31 290.78 8~~
~~7 23 C9 56 0.34 ts.t3 5.50 2. 11 4 2.52 2.05 1.34 1.94 106.16 290.71 72353 1 0.33 3.16 3. 4 11~~
~~2. fl3 2.51 2.05 1.24 1.94 106.25 290.71 72353 8 0.33 3.16 5.19 2. 11 4 2.52 2.03 1.24 1.95 106.16 090.71 7 23 US 13 0.34 3.15 5. 4 11 2. 11 4 2.51 3.03 1.23 1.94 106.16 290.71~~
,7 23 53 19 0.33 3.15 5. 4 11 2. 11 4 2.52 2.05 1.24 1.94 106.09 290.71 7005325 0.34 3.15 5.4 fl 3. 11 4 2.52 2.05 1.24 1.94 t06.09 292.25 L' 7::35032 0.33 3.15 5. 4 11 2. 11 4 2.51 2.05 1.23 I.94 105.97 292.25 7 23 .~.3 Ull 0.33 3.12 5.4fl 3. 11 0 2.48 2.03 1.23 1.92 106.00 292.25 7 23 53 43 0.30 3.16 5.48 2.82 2.48 2.03 1.23 1.92 106.63 292.25 7235050 h.33 3.I9 5.4!! 2.11 5 2.51 2.05 I.24 1.94 105.9I 292.25 7 29 53 55 0.33 3.15 5.48 2. 11 4 2.51 2.05 1.23 1.94 105.97 292.25' 7 23 51 0 0.33 3.15 5. 4 11 2. 11 4 2.51 2.05 1.23 1.95 105.84 292.25-7 22 53 11 0.33 3.15 5.48 2. 11 4 2.51 2.05 1.23 1.94 105.111 292.25 7 23 SI 13 0.34 3.16 5. 4 11 2. 11 4 2.51 3.05 1.24 1.94 105.91 292.25 6 7 29 51 20 0.33 3.15' 5. 4 11 2.(14 2.51 2.05 1.23 1.94 105.f14 292.25 37 20 51 25 0.34 3.I5 5.46 2. 11 4 2.51 2.05 I.23 1.94 105.114 290.20 37 20 51 31
~~0.33 3.15 5.4fl 2. 11 4 2.51 2.05-1.23 1.94 105.fl4 290.20~~
~~]7 23 51 37 0.33 3.15~~
5. 4 fl 2.114 2.52 2.05 1.23 1.96 105.114 290.20 D7 20 51 43 0.34 3.15 5. 4 11

2. f14 2.51 2.05 1.24 1.94 105.114 290.20 07 23 51 49 0.33 3.15

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2. !!4 2.51 2.05 1.20 1.94 105.114 290.20 07 23 51 56
~~'O.33 3.t5 5.46 2. 11 4 2.51 2.05 1.21 1.92 105.75 290.20 17 20 52 2' O.33 3.15 5.46~~
~~2. !I5 2.51 2.05 1.23 1.94 105.114 290.20 l3~~
) 17 20 52 7 0.33 3.15 5.46 2. 11 4 2. 4 11 2.03 1.20 1.94 105.69 290.20 l17 20 32 14 0.33 3,15 5.46 2. 11 4 2.51 2.05 3.20 1.94 105.69 290.20 l 17 20 52 20 0.33 3.15 5.46 2.84 2.51 2.05 1.22 1.92 105.75 290.00 l i 17 23 52 25 0.34 3.12 5.48 2. 11 5 2.52 2.06 1.22 i.94 105.75 291.22 l j 17 20 52 30 0.33 3.11 5.46 2. 11 4 2.51 2.05 1.20 1.94 105.69 291.22 17 23 52 37 0.34 3.14
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'ic i I.* f 81 If * 'i*j .)' 2 OUTPUT FROM PSNS DICITAL TREND j Y.*- f,- - 2. >no RECFLO RECFl.O RECFLO RECFLO RECFLO RECTEM RECTEM HECTEM HECTEM RECTEM '.f':"2. 8: l ji I 2 3 4 5 1 2 3 4 5 1 M 1) 11 M S CPM GPM CPM GPM GPM DEC F DEC F DEC F DEC F DEC F d-s ? *; .1 l 7 17 20 29 23 110.111 117.67 114.69 104.77 109.09 535.47 541.fil 541.97 5311.69 540.63 ^ a ;, ', * ~ 109.13 535.59 541.91 541.97 5318.69 G40.7dl }' 7 17 20 29 29 !!O.3a 116.151 113.119 105.69 107.00 535.711 542.03 542.19 539.00 540.94 s +. 4 "i l 7 17 20 29 35 110.72 114.69 114.311 105.63 l 7 17 20 29 41 110.75 1111.41 114.911 104.fl9 107.64 536.09 542.41 542.47 539.31 541.22 ~ i 542.53 539.31 541.22 e .' t' ' *.$'i ' 104.77 107.511 536.16 542.41 7 17 20 29 47 110.47 115.23 114.50 1011.13 536.44 542.72 512.711 539.59 541.53 7 17 20 29 53 111.17 11'4.69 114.25 104.77 107.1111 536.47 542.72 542.711 539.50 541.53 ~~ ' 115.05 104.511 543.16 539.1111 541.111 j 7 17 20 29 59 111.27 116.14 114.02 104.77 1011.4(I 536.72 542.91 r 7 17 20 30 11 110.75 115.66 114.31 103.61 107.52 536.91 543.06 543.211 510.06 541.91 7172030 5 109.89 115.11 104.77 107.81 536.114 543.16 54:1 211 540.06 541.97 ? i, 7 17 20 30 17 110.59 114.92 113.16 542.211 105.25 loft.13 537.03 543.311 543.50 540.3 3 7 17 20 30 23 !!0.50 115.17 114.19 542.41 i 103.97 103.03 537.25 543.50 543.66 540.41 7 17 20 30 29 111.30 116.09 113.34 7 17 20 30 35 110.05 115.711 112.91 104.64 1011.31 537.22 543.50 543.66 540.41 542.41 1 7 17 20 30 41 110.11 115.66 113.211 105.50 1011.25 537.44 543.59 543.111 540.56 542.47 544.13 540.f14 542.711 / r 104.5fl 107.03 537.72 543.94 106.75 537.66 543.84 544.00 540.69 542.66 7 17 20 30 47 110.27 116.27 114.14 7 17 20 30 54 109.86 116.03 113.64 103.511 543.94 544.19 540.01 542.84 ~ e 105.25 107.45 537.111 106.66 537.97 544.13 544.25 541.00 543.00 7 17 20 30 59 110.66 II3.G3 113.47 104.41 543.28 7 17 20 31 6 110.66 114.69 114.31 103.36 107.14 5311.25 544.38 544.56 541.38 I l' 5311.25 544.31 544.50 541.31 543.22 7 17 20 31 11 110.27 115.411 312.94 113.53 105.02 106.11 103.4fl 105.38 5311.25 '544.50 544.69 541.38 543.3fl 7 17 20 31 17 110.211 I l3.fl9 g' 7 17 20 31 23 110.23 113.03 113.53 104.64 5311.50 544.63 544.75 541.53 543.50 103.35 T' 543.75 \\ .,3 ; b 7 17 20 31 29 110.17 113.95 113.09 103.06 105.73 5311.711 543.00 515.06 541.81 .i 7 17 20 31 36 1031.92 133.113 113.64 104.52 5311.116 544.94 545.13 541.91 543.111 7 17 20 31 42 109.30 115.30 113.00 102.69 105.56 539.06 545.211 545.311 542.09 514.13 7 17 20 31 47 109.77 !!4.14 112.113 103.30 103.114 539.00 515.13 545.2fl 542.09 544.13 j 7 17 20 31 53 109.00 113.70 113.38 103.711 7 17 20 38 59 109.50 113.95 112.23 102.94 104.34 539.06 545.311 515.50 542.211 544.19 7 17 20 32 5 109.13 113.211 113.95 103.73 104.34 539.16 545.41 515.59 542.41 544.31 e 7 17 20 32 la 109.56 115.36 113.119 103.23 103.73 539.31 545.59 545.72 542.53 541.50 546.03 512.711 544.69 102.63 104.16 539.59 545.81) 7 17 20 32 17 109.22 112.77 113.34 104.511 539.69 545.711 545.97 542.711 544.75 3,_ t".. 7 17 20 32 23 109.56 113.47 114.02 103.30 104.41 539.94 546.09 546.22 543.00 544.94 7 17 20 32 29 110.20 114.75 113.70 103.17 104.77 540.06 546.22 546.34 543.16 545.06 ,,1 7 17 20 32 35 1011.55 114.46 113.70 102.14 104.83 540.19 546.34 516.53 543.2a 545.22 .I 7 17 20 32 41 109.22 112.52 111.78 102.39 546.53 543.22 543.22 7 17 20 32 40 109.13 114.02 112.06 102.56 104.77 540.13 546.34 7 17 20 32 54 107.94 114.31 112.64 102.45 103.68 540.50 546.53 546.69 543.50 545.30 i 102.111 540.31 546.53 546.75 513.44 545.44 8.f ". 7 17 20 32 59 109.41 112.89 111.55 102.50 103.67 540.25 546.47 546.75 543.50 515.44 102.111 7 17 20 33 6 109.50 113.64 112.00 102.08 103.97 540.50 546.69 546.97 543.75 545.59 4 7 17 20 33 Il 109.22 111.30 112.119 545.73 104.52 540.50 546.69 546.97 543.81 7 17 20 33 17 109.03 113.64 113.28 101.53 .I i ? t 8 8 i
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~~RECFLO DECFLO RECFLO RECFLO RECFLO RECTEM REUTEM nECI'EM RECTEM RECTEM' 8-I-~~
~~'l*., 2 3 4 5 1 2 3 4 5 M D M M8 CPM CPM GPM CPM CPM DEC F DEC F DEC F DEC F DEC F ,5g t' .t i, e l...'1 j 7 17 20 33 24 109.98 !!4.25 112.91 102.14 103.91 540.31 546.69 546.91 543.66 545.59~~
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'e 7 17 20 33 29 108.411 113.89 112.30 102.02 103.97 540.31 546.75 546.97 543.75 545.66 I l* 7 17 20 33 35 109.38 113.16 112.06 101.113 104.511 540.4 8 546.C4 547.03 543.Ill 545.72 'd 546.114 547.13 543.118 545.72 '9 7 17 20 33 41 1011.36 113.77 112.67 101.34 103.48 543.41 547.19 543.811 545.72..* X ef,I'. 111.33 102.14 104.34 540.56 516.91 7 17 20 33 47 109.06 113.47 7 17 20 33 54 109.19 !!3.70 113.22 101.47 102.118 540.63 547.03 547.28 543.94 545.8 7,. 102.1811 104.09 540.69 547.03 547.34 544.13 545.811 I" 1 7 17 20 34 5 110.23 111.63 112.67 102.08 103.48 540.71) 547.19 547.50 544.19 546.03 i 7 17 20 34 11 110.41 114.14 111.81 546.03 7 17 20 34 18 108.116 114.19 113.28 102.33 104.89 540.04 547.34 547.56 544.31 I l 7 17 20 34 23 Ill.63 113.22 112.00 102.20 103.23 540.71) 547.211 547.50 544.25 545.97 7 17 20 34 29 110.08 112.16 112.39 101.53 103.42 540.04 547.41 G47.72 544.38 546.09 s 546.03 101.47 103.78 540.84 547.19 547.56 544.31 7'17 20 34 36 180.23 114.08 112.110 110.14 101.22 101.119 541.00 547.34 547.72 544.50 546.16 7 17 20 34 41 109.53 114.69 101.53 103.I14 540.711 547.19 547.63 544.311 546.03 i 7 17 20 34 48 109.80 114.02 112.61 102.33 103.36 540.114 547.19 547.63 544.311 546.09 g 7 17 20 34 53 109.00 115.30 112.00 104.03 540.94 547.28 547.63 544.50 546.03 7 17 20 35 0 109.110 113.53 111.118 102.14 104.22 540.94 547.211 547.72 544.50 546.09 s 102.20 7 17 20 35 5 10ll.I3 113.70 112.19 113.09 102.27 103.97 548.00 547.50 547.1111 544.75 546.34 f 7 17 20 35 11 109.47 113.34 100.50 102.111 541.00 547.41 587.711 544.63 546.22 s. 7 17 20 35 111-109.22 111.78 112.83113.09 112.16 101.66 103.48 541.16 547.56 547.94 584.75 546.4I 7 17 20 35 23 109.70 547.1l11 544.75 546.22 D f M O/Edl0
l,*, 9.P 7 17 20 35 30 1011.89 112.67 112.86 102.39 103.42 54I.06 547.41115.42 112.13 101.28 104.113 541.16 547.5 7 17 20 35.35 1011.119 7 17 20 35 41 110.31.114.31 115.11 103.98 105.116 541.31 547.711 5415.13 545.06 546.59 106.97 541.31 547.72 548.06.545.00 546.53 108.110 106.05 541.44 547.811 G43.25 545.13 54 6.69 *&- O A&2 V c 1.0)t)
- + 7 17 20 35 48 110.47 116.39 116.45 104.95 . i: 7 17 20 35 53 111.55 112.09 117.25 97.311 541.06 547.34 547.94.544.75 546.47 ~ i 7172036 0 107.811 104.16 1111.77 106.97 93.38 540.56 546.59 517.56 544.25 545.118 7 17 20 36 5 108.38 104.89 1811.77 911.6 6 98.13 539.94 545.711 547.03 543.59 545.28 I 7 17 20 36 II 106.59 105.02 120.30 1111. 1 4 flit.31 539.44 545.211 546.69 543.16 544.84 7.17 20 36 17 104.211 105.38 119.50 114. 011 1111.116 539.06 544.63 546.03 542.41.544.25 -7 17 20 36 23 103.55 104.22 120.05 112.22 717203629 103.30 103.13 122.67 112.33 116.45 538).69 544.69 545.711 542.211 543.94 7 17 20 36 35 103.30 99.39 128.119 82.70 117.52 5311.25 544.31 545.311 541.91 543.59 7 17 20 36 41 101.28 100.44 121.64 81.55 115.75 5311.25 544.31 545.13 541.81 543.38 t 7 17 20 36 47 100.61 100.67 118.66 82.33 84.59 53I1.19 544.31 545.00 541.66 543.211 t 7 17 20 36 54 101.41 100.00 108.73 84.23 115.33 5311.03 544.25 544.94 541.66 543.28 7 17 20 36 59 103.47 100.06 99.16 115.39 157. 4 8 538.13 544.38 544.94 541.75 5,43.28 7 17 20 37 6 100,73 101.28 97.63 115.39 87.77 538.13 544.50 545.00 541.81 543.38 117.31 5311.34 544.75 545.13 542.03 543.59 7 17 20 37 11 100.55 100.22 97.23 85.84 7 17 20 37 17 100.38 100,13 96.95 87.28 87.52 538.34 544.69 545.06 542.03 543,59 4, I e g , [e
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r .OITTPirr FROM PSrtS DIClTAL *IREND M RECTEM REC 7EN N S RECFLO RL,LO RECFLO RECFLO RECFLO RECTEM IlECTEM RECTE I a 3 4 5 1 2 3-4 5 CI'M CPN CPM CPM CI'M DEC F DEC F DEC F DEC F DEC F M 'S. H 86.58 53fl.56 544.94 545.28 542.19 543.81 117.50 538.56 545.06.345.311 542.2ft 543.1111 7 17 23 37 24 99.67 101.05 98.39 86.64 fl6.64 53fl.7fl 545.22 545.50 542.47 544.03 7 17 23 37 29 99.06 100.43 97.05 8 6. 1111 543.94 117.211 53fl.7fl 545. I3 545.44 542.41 7 17 20 37 36 911.45 100.03 97.91 86.77 87.55 539.00 545.44 545.66 542,63 544.25
7 17.20 37 41 97.20 10t.41 98.1I 87.41 87.55 539.22 545.59 545.711 542.711 544.311 7 17 23 37 47 97.91 101.72 97.17 117. 3 1 1111.02 539.86 345.44 545.72 542.72 544.*116 7 17 20 37 54 97.91 100.50 97.59 117.441111.011 539.38 545 72.545.97 543.00 544.63
7 17 23 37 59 97.20 100.55-98.30 117.16 117.77 539.22 545.66 545.97 543.00 544.63

7 17 20 3fl 5 97.53 9fl. 88 911.75 87.50 117.119 539.311 545.1111 546.09 543.16 544.75
,7 17 20,311 Il 97.311 100.19 911.3 0 117.58 init.38 539.50 54tn.03 546.22 543.2fl 544.94546.16 546.47 543.44 545.06 911.1 7 99.52 87.09 ' 7 17 23 311 18-96.22 I 7 37 20 315 24 97.53 99.73 98.45 87.61 111). 3 1 539.fil 1114.75 539.94 546.34 546.59 543.66 545.22 97.23 100.03 911.7 5 87.73 545.22 Ilfl.3 8 539.81) 546.34 546.59 543.59 7 17 20 Ull 29 9t!.75 117.50 7 17 23 all 36 96.92 100.41 545.50 1111.44 540.19 546.59 546.116 543.81546.75 546.97 543.94 545.59 7 17 29 3fl 41 97.63 99.94 99.22 87.41 9 9. 1111 99.09 87.09 7 17 23 314 53 95.211 99.42 99.23 117.44 11(1.02 540.31 7 17 20 311 4ft ' 97.23 547.13 544.13 545.72 87.38 540.50 546.91 547.13 544.19 545.78 7 17 20 311 59 9 5.113 99.55 9fl. 94 87.61 117.64 540.50 546.91 816.3f1 540.69 547.13 547.50 544.56 546.09. 7 17 20 39 5 95.61 100.311 911.3 6 87.77 544.56 546.09 7 37 20 39 12 96.06 99.67 99.13 815.311111.25 540.69 547.13 547.41 87.70 540.711 547.19 547.50 544.63 546.16 7 17 20 39 til 95.36 101.05 98.55 87.52 (17.47 540.94 547.34 547.72 544.75 546.22. 7 17 20 39 23 94.48 98.84 98.66 117. 511 547.711 544.94 546.47 7 17 20 39 29 94.7fl Ofl.91 99.16 {l7.41 117.77 541.06 547.41547.72 54fl.06 545.13 546.69 7 17 no 39 35 93.19 99.33 98.86 87.211
117. 6 1 541.31 (111.75 541.3ft 547.72 541).13 545.214 546.84 7'I7 20 39 46 9 5.(13 100.55 98.02, 811. 3 1 111).19 541.318 547.7fl 5418.13 545.214 516.114 7 17 20 39 415 95.09 100.16 97.91 117. 5 5 547.63 548.06 545.22 546.114 7 17 20 39 53 95.00 97.95 99.97 117.25;811.50 54I.31 545.3fl 546.97 7172040 0 95.48 99.39 99.09 1111. 3 1 (111.911 541.3fl 547.711 5411.31 1111.011 541.44 547.72 5411.25 545.3tl 546.91 7 17 20 40 5 95.64 911. 6 3 97.23 87.116 811.56 54I.I6 547.50 5411.00 545.I3 546.75 7 17 20 40 11 95.39 99.00 9 6.911 89.17 88.19 541.00 547.50 547.94 545.00.546.59547.94 545.00 546.59 7 17 20 40 17 96.03 100.00 93.38 119. I I 7 17 20 40 23 95.73 9 9.511 91.25 inn.f16 117.09 541.00 547.41 11(1.69 541.00 547.34 547.78 544.94 546.53 7 17 20 40 30 95.42 99.77 90.20 119.0 5 547.94 545.06 546.59 7 17 20 40 35 96.22 99.39 98.73 113.1 4 8(1.I9 541.16 547.41 87.5(1 540.94 547.19 547.711 544.75 G46.41147.03 540.69 54
~ 7 17 20 40 41 96.67 99.61 91.25 117. 7 3 7 17 20 40 47 95.116 100.92 92.28 011. 0 2 546.41 811.19 540.94 547.34 547.88 544.f14 96.77 100.98' 90.94 1111. 7 5 117.03 540.741 547.13 547.56 544.63 546.16 7 17 20 40 54 -95.113 100.92 90.64 118.0 8 87.16 540.69 547.13 547.56 544.63 546.16 7 17 20 40 59 7 17 20 41 6.96.64 100.60 91.06 811.3 187.52,540.63 547.03 547.56 544.63 546.09. 7 17 20 41 12 95.86 100.55 91.48 89.05 92.53 88.31 7 17 20 41 17,96.28 101.22 i { j g g _.I
= 8 OtTTPITT FROM PSMS CICITAL TREND RECFLO RECFLO RECFLO RECFLO RECFLO RECTEM RECTEM RECTEM RECEEM RECTEM M M C M S 1 2 3 4 5 1 2 3 4 5 CPM ~;l'M GPM CI'M CPM DEC F DEC F DEC F DEC F DEC F l 540.711 547.13 547.63 544.69 546.16 p gg v4/v,.r OKM 7 17 23 41 23 9 6.311 99.73 91.25 119.53 87.31 544.50 546.03 CgJC/ 7 17.23 41 29 95.97 100.98 92.16 89.05 117.47 540.56 546.91 547.41 544.50 545.97 9-6/* V"*/*ft 7172341 35 95.97 99.83' 94.66 90.03 87.64 540.63 546.97 547.41 79.911539.50 545.13 546.59 543.59 544.75e 7W##M # ## '### Iy 10.53 540.41 546.22 547.19 544.31 545.72 7172341 42 31.09 47.13 2 7.311 10.59 7 17 20 41 47 90.39 109.13 95.55 109.31 7 17 23 41 54 110.114 112.67 111.63 105.92 105.31 538.34 543.75 543.22 542.19 543.211 7172341 59 111.75 1018.95 107.20 104.89 103.78 537.111 543.00 544.63 541.66 542.72 !!9.42 536.7fl 542.19 543.66 510.63 541.44 536.22 543.03 543.44 540.19 541.16]869UM MM 7 17 20 42 6 102.94 98.11 97.69 911.7 2 7 17 20 42 11 99.48 911.27 96.31 9 8. 4fl 117. 6 1 7 17 23 42 17 94.94 100.31 92.09 89.98 fl3.44 535.44 541.31 542.711 539.3fl 540.31 g 93 7 fsg 7 17 23 42 29 97.31 99.45 9 6.211 98.25 86.00 533.211 53fl.34 541.59 537.81 5311. 69 +-T: SJTI.O y r,9 9.y,gg> g#4 r. 7 17 20 42 23 93.44 101.72 91.73 (19. 2 3 (14.81) 534.711 540.94 542.53 539.16 539.94 7 17 23 42 35 100.41 105.02 9 7.114 95.03 87.67 529.ltfl 533.59 539.06 534.7tl 535.91 527.47 534.41 529.56 530.75 7g g p;/ g 91.06 517.53 520.56 5211.69 522.22 524.53#-7:SM.f 3 j f : 9'lg.g (h J 85.63 524.41 7 17 20 42 41 102.20 108.55 98.55 97.81 8 ksq 7 17 20 42 411 101.119 105.25 101.53 96.19 7 17 20 42 53 101.05 10fl.110 100.16 95.52 92.70 514.41 517.114 525.28 519.38 521.25 /Q 7 17 23 42 59 91).05 106.59 99.64 94.05 88.92 511.22 515.22 521.06 516.06 517.3% 7 17 20 43 6 94.78 106.17 94.11 91.73 86.lill 509.75 514.215 5111.59 514.34 515.50 7 17 20 43 11 91.73 102.33 92.70 87.70 84.66 509.91 515.311 517.66 514.22 515.13 7 17 2D 43 Ill 89.23 101.28 87.77 811.2 5 82.22 514.63 521.97 520.16 5111.56 519.31
~~2. f.92 52 M il4~ T: M *7b p, g,yf f 717204323 (15.56 100.92 81.23 85.63 79.31 516.97 525.22 522.13 520.75 521,75 4" J p3 S I ' ' ^~~
-I.00 519.94 52fl.3fl 525.59 523.41 g_fgd f - -1.00 520.63 5211.16 527.16 523.84 !i35.59 4-Tu 52M,b ,6 *)I s y 7 17 23 43 30 52.39 96.53 62.02 49.09 7 17 20 43 35 -I.00 87.73 -1.00 -1.00 527.25 527.63 523.47 :l'12fl4-- T = S3I % 7 17 23 43 41 -1.00 (19.4 1 -1.00 -1.00 -1.00 520.50 7 17 20 43 48 72.86 1111. 3 1 61.41, 59.39 64.011 519.03 524.91 526.811 521.91 523.91 7 17 20 43 53 115. 5 9 91.55 112.73 111. 9 4 79.44 518.16 523.84 526.34 521.13 523.34 7 17 20 44 0 90.09 9 5.116 116.114 82.70 84.17 516.2ft 522.13 524.97 519.41 521.66 7 17 20.44 -6 flin.31 99.36 117.67 113. 3 4 (12.61 515. fill 521.94 524.22 519.06 521. 3 3 80.91 515.94 522.311 523.f14 519.22 521.19 72.36 516.22 522.711 52't.7fl 519.56 521.44 4-Ts h/-% N 8/b OJer~W '*? 7 17 23 44 12 114.50 97.78 83.09 76.50 7 17 20 44 17 84.95 92.22 77.73 74.411 7 17 20 44 23 73.61 77.55 72.02 66.34 63.59 516.114 523.09 524.09 519.1111 522.00 7 17 20 44 30 77.30 69.39 73.35 65.80 72.02 516.63 522.47 523.53 519.211 521.69 7 17 20 44 35 82.13 78.70 72.20 71.66 69.113 516.41 522.22 523.09 5115.114 528.31 717234442 116.411 93.89 83.44 83.50 79.02 516.00 522.19 522.66 Sill.56 520.111 7 17 20 44 47 811.911 95.67 83.61 84.18 84.53 515.69 522.19 322.66 Sill.41 520.69 717294453 89.42 97.53 86.58 84.31 113.47 515.72 522.16 522.69 518.38 520.75 7 17 23 44 59-89.72 95.31 84.47 110. 8 4 112.70 515.31 521.111 522.41 Sill.09 520.44 7172345 5 91.48 95.39 86.03 81.81 *B4.14 515.50 521.78 522.53 518.22 520.50 ) 7 17 23 45 17 bl.98 93.50 87.52 86.55 '84.95 515.22 521.66 522.41 518.03 520.53 9-Ta f.70.J.?j h 785',/3, 6,7 = F#o/'Is 7 17 23 45 12 93.11 96.73 117. 8 0 (12173 84.14 515.31 521.78 522.53 518.22 520.63 y f e'. 8 i i 8 l l 9 A ? l )., 1
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~~OITTI'UT F110M PSFDI DICITAL 'mEFID RECFLO RECFLO RECFLO RECFLO RECFLO RECTEM RECTEM RECTEM RECTEM RECTF4 I~~
2 3 4 5 1 2 3 4 5 M D 0 M S CPM CI'M CPM CPM CPM DEC F DEC F DEC F DEC F DEC F ?- V/ 17 23 45 24 93.75 98.11 89.91 87.41 83.56 585.06 521.31 522.13 517.72 520.31 7 17 20 45 29 95.09 97.91 88.44 117. 5 5 116.33 585.13 528.25 522.09 517.78 520.34 ^ 7 17 23 45 33 96.06 9 8. 811 90.58 (19.1 7 116. 51) 5 1 4. 81) 520.94 521.114 517.47 520.13 7 27 20 45 41 95.55 103.67 9 0.118 117. 2 2 116.06 5 t 4.63 520.75 521.59 517.19 519.114 e 7 17 2D 45 47 95.03 102.50 93.44 1111. 4 4 !!7.06 514.50 520.50 521.50 517.16 519.75 8 7 17 23 45 53 96.67 99.36 92.41 116.64 84.63 584.34 520.34 531.31 516.94 519.59 7172346 0 95.52 102.75 91.13 119.7 2 (15.55 514.09 520.19 521.19 516.78 519.44 I i 7 17 20 46 6 97.27 101.48 92.83 (19.72 116.00 533.111 519.711 520.811 516.44 519.16 7 17 20 46 13 911.9 1 101.28 91.92 ~ 9 0.111 90.75 513.66 519.69 520.69 516.44 518.94 7 17 ut 46 18 99.58 104.16 9 3. 1111 93.69 93.80 513.41 519.56 520.44 556.50 5111.711 90.75:513.215 519.44 520.34 516.56 5111.69 7 17 20 46 23 98.56 9 7.I14 95.70 93.02 91.311 513.03 519.16 520.06 516.44 5811.34 7 17 20 46 33 101.34 99.03 94.72 93.94 7 17 20 46 37 99.97 104.28 96.73 94.48 90.64 512.97 518.97 589.97 516.41 5 111. 21) 90.pa 512.59 5ffl.72 519.63 516.13 517.97 7 17 23 46 44 100.80 102.27 9 6.211 94.72 91.v5 512.41 5111.47 519.41 516.00 517.72 7 17 20 46 50 108.05 101.53 96.53 94.75 7 17 20 46 54 100.55 105.02 96.25 94.97 91.19 512.311 5111.47 519.311 516.00 517.711 ~ 7 17 20 47 0 101.53 102.56 96.73 94.84 92.53 512.38 5111.23 519.22 515.118 517.59 i 7 17 23 47 7 100.67 104.211 96.36 94.48 92.05 511.91 517.97 5111.711 585.63 517.28 7 17 23 47 14 100.(16 103.00 96.44 95.33 91.911 511.91 587.91 5111.84 515.63 517.16 7 17 23 47 19 100.110 101.36 95.70 93.118 92.89 511.75 517.78 5111.59 515.48 517.00 7 17 23 47 26 100.98 100.50 97.02 95.31 91.55 511.211 517.50 518.34 515.22 516.75 7 17 22 47 31 100.211 102.011 97.05 95.09 92.22 511 1111 517.47 5111.28 515.25 516.72 7 17 20 47 37 100.06 102.81 96.03 94.17 92.41 5to.91 517.10 517.97 515.06 5t6.311 7 17 20 47 43 101.17 103.30 96.70 94.17 92.22 510.711 516.97 587.711 514.114 516.22 7 17 20 47 49 100.44 105.44 96.95 95.36 93.20 510.69 516.75 517.59 584.69 516.00 I 7 17 20 47 56 101.59 101.66 96.03 95.25 91.61 530.53 516.59 517.41 514.47 515.111 91.19 510.211 516.44 517.19 514.311 515.69 7 17 20 4tl 0 102.02 106.34 9 6.119 94.18 7 17 20 411 7 100.67 103.73 96.77 94.55 91.110 510.06 516.22 517.03 514.16 515.44 4 92.16 510.06 516.25 517.00 514.09 515.41 7 17 2L 411 14 100.28 106.66 96.50 94.61 7 17 2D 411 20 99.39 103.97 96.53 95.25 92.41 509.69 515.111 516.66 513.72 515.13 7 17 20 411 26 100.86 101.47 96.09 9 4.711 92.70 509.53 515.69 516.50 5t3.66 514.91 7 17 20 411 32 100.113 101.95 96.64 94.42 92.34 509.31 515.50 516.08 513.41 514.75 7 17 20 4fl 37 100.16 102.94 96.56 94.55 91.48 509.31 515.41 516.25 513.34 514.72 93.38 509.00 515.22 516.06 5t3.22 514.47 7 17 2T 4fl 44 100.61 103.42 97.011 94.94 7 17 20 48 50 100.09 104.119 98.02 94.97 92.22 501).94 515.13 515.118 513.03 514.34 7 17 20 411 55 101.17 104.159 95.28 95.09 90.64 5011.69 514.811 515.72 512.111 514.19 7 17 20 49 2 100.19 104.119 96.03 94.00 98.55 501).38 314.53 5t5.44 512.50 513.91 7 17 23 49 7 101.05 104.45 96.67 9 5. 511 93.33 5011.47 514.56 515.41 512.59 513.91 7 17 20 49 14 100.38 105.31 96.41 94.61 92.41 5011.00 514.16 515.06 582.13 513.50 i 717234919 100.61 106.11 95,77 95.03 94.00 507.91 584.03 514.94 512.00 513.41 ; e t t ~ .l 9 g 0 4 s 3.' ! ? a a i y..s g 8 l e t,,g
5 u... OtrFPtrF FROM PSils DICITAL 'IRF,f0 I RECFLO RECFLO RECFLO IIECFLO RECFLO RECTEN RECI'EN RECTEN RECTEN RI:CTEN.
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2 ' 3 4 5 I 2 3 4 5 CI'M CI'N CI'M CI'M CPM DEG F DEC F DEC F DEC F DEC F 7 17 20 49 26 100.41 104.77 96.50 95.45 93.08 507.84 513.91 514.88 511.81 513.28 { 7 17 20 49 31 100.61 107.39 96.38 95.33 92.09 507.59 513.59 514.56 511.53 512.97 7 17 20 49 37 200.19 106.30 96.41 94.94 93.44 507.16 513.22 534.16 511.06 512.59 7 17 20 49 44 100.44 104.77 96.59 95.36 92.03 506.94 513.03 513.97 510.84 512.38 7 17 20 49 49'100.41 105.92 96.80 96.19 92.41 506.81 513.03 513.91 510.718 512.31 .7 37 20 49 56 100.67 106.169 96.36 95.25 92.09 506.56 512.69 513.50 510.2f! 511.91 7 17'20 50 l' 99.30 109.28 97.41 95.03 98.61 506.44 512.59 513.41 510.22 511.91 7 17 20 50 8 101.34 107.154 97.63 95.06 93.66 506.06 582.25 513.09 509.161 511.50 7 17 20 50 13 100.09 106.59 9 8.011 94.30 93.20 506'.06 512.31 513.03 509.f14 511.44 7172050 19 100.38 106.11 96.34 94.E8 91.98 505.69 511.91 512.75 509.44 Gil.16 7 17 20 50 25 99.511 107.u3 97.92 94.72 93.14 505.47 511.69 512.50 509.13 510.91 7 17.30 50 32 100.61 106.53 911.3 6 95.73 93.44 505.31 511.44 512.31'509.00 510.69 7 17 20 50 38 99.35 107.511 941.7 5 94.42 93.11 505.34 511.41 312.31 508.94 510.72 7 17 20 50 43 99.411 106.114 101.211 95.58 92.59 505.03 511.22 512.06 50ft.69 510.47 7.17 20 50 50 100.16 106.97 102.08 94.30 95.03 504.111 510.91 511.81 5011.47 510.22 7 17 20 50 55 99.36 107.45 102.02 94.61 93.88 504.66 510.78 511.59 5011.31 510.00 .7 37 20 51 0 99.22 306.110 100.67 94.78 95.36 504.25 510.311 511.22 507.78 509.63 7 17 20 51 8' 98.55 106.41 100.61 95.73 95.19 504.00 510.22 510.97 507.53 509.44 7 17 20 51 13 97.95 107.111 102.45 94.48 95.39 503.97 510.16 510.91 507.53 509.25 7 17 20 51 20 99.77 105.25 101.11 94.94 96.64 503.69 509.84 510.59 507.16 509.00 7.17 20 51 25 93.48 105.92 99.64 94.94 94.48 503.41 509.63 510.38 506.94 508.75 "I 7 17 20 51 31 97.41 106.72 100.86 94.91 94.97 503.2ft 509.53 510.311 506.111 50ll. 69 7 17 20 51 37 99.30'107.52 101.66 94.55-96.31 503.13 509.08 530.06 506.66 5011.38 7 17 20 51 43 97.53 107.45 101.05 9 4.114 95.89 502.91 509.00 509.98 506.44 5018.22-e 8 '7"17 20 51 49 100.09 106.66 101 97 94.97 95.16 502.66 50fl.9 6 509.75 506.19 5011.00 7 17 20 51 56 - 911. 54 107.33 103.23 93.69 97.00 502.31 501).53 509.31 505.84 507.59 7 17 20 52 2 96.25 105.50 105.311 94.48 96.06 502.28 5011.47 509.25 505.111 507.53 7 17 20 52 7 98.59 104.52 105.02 94.42 96.09 502.00 5011.16 5011.86 505.47 507.16 - 7 17 20 52 14 911.72 107.20 102.63 94.23 94.94 501.69 507.91 508.69 505.23 507.00 7 17 20 52 20 98.33 104.58 103.13 94.30 94.36 501.56 507.84 508.63 505.09 506.97 7 17 20 52 25 911.59 105.98 102.63 94.05 95.73 501.311 507.53 508.31 504.114 506.63 7 17 20 52 30- 911. 5 2 107.14 105.38 95.64 96.67 50s.19 507.41 508.16 504.66 506.50 7 17 20 52 37 98.11 106.23 103.97 94.48 95.25 500.91 507.00 507.114 504.31 506.16 j 717205243 911. 1 1 106.59 104.83 95.92 94.94 500.59 506.78 507.59 504.06 505.94. 7 17 20 52 50 99.52 106.36 104.16 9 4.118 95.97 500.47 506.63 547.41 503.94 505.63 7 17 20 52 55 98.17 107.17 105.59 95.48 95.97 500.16 506.25 507.13 503.66 505.34 7 17 20 53 I 99.00 108.06 106.30 96,13 95.39 500.19 506.19 507.13 503.50 505.41 + 7 17 20 53 7 99.00 107.08 105.19 95.33 96.67 499.94 506.13 506.88 503.34 505.22 7 17 20 53 13 98.45 106.17 103.67 95.22 94.72 499.72 505.81 506.66'503.13 504.97 ~ 7 17 20 53 20: 98.63 106.53 105.25 95.58 95.95 499.50 505.75 506.44 502.97 504.72 f-I O D {
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l-F Report to GORB Action Item No. 37 5 Appendix D G.E. Responses to Various NRC and JCP&L Questions - L m 9 ed as O O O M M g m l l- ~_m__
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~~__2 . :- w m.:....c 4. h' . Question 1: Why was there a 10*F temperature increase in the recirculation loop with a corresponding decrease in pressure? How does that relate to pump cavitation?~~
~~Response~~
Following the reactor scram, the feedwater flow rapdily increased by greater than 100'.. This, coupled with reduced flow from the separators, resulted in a short-term non-equilibrium temperature reduction in the annulus. Consequently, the temperature in the recirculation lines decreased faster than that which can be attributed to the pressure decrease and equilibrium feedwater mixing. As the flow and mixing conditions begin to stabilize, the enthalpy in the annulus increases resulting in the approximate 10*F temperature increase noted in the recirculation lines. Ultimately, the mixing reaches an equilibrium condition and the recirculation line temperatures begin to decrease in accordance with the pressure decrease. The increased feedwater flow and non-equilibrium mixing. resulted in increased subcooling at the pump suction. As the mixing approaches equilibrium, the subcooling decreases, which decreases the available net positive suction head {NPSH), below a certain value of NPSH (dependent on the particular pump), errou acion may occur. cas.h2 ou Question 2: Did the pumps actually cavitate or was the instrumentation fooled by two phase flow.
~~Response~~
Using the classical definition, cavitation is said to exist in a closed rapidly flowing stream when the fluid vapori:es, forming bubbles in the steam which disturb the flow and by their subsequent collapse produce vibration, noise, and rapid flow deterioration. Pump cavitation is principally of concern in low pressure systems since the volumetric change associated with vapori:ation of water is much more significant at low pressure. Pumps in high pressure systems are less sensitive to pu=p suction subcooling anddonothaveclassicalcavqq3tioncharacteristics. This is supported by General Electric test data which shows that with two phase flow at the pump suction, some loss in pump head is noted, but it does not rapidly go to :ero. This also accounts for normal flow being indicated with saturated conditions or near saturation conditions existing at the suction of the pumps. (1) NEDO-10329, " Loss-of-Coolant Accident and Emergency Core Cooling - Models for General Electric Boiling Water Reactors", April 1971. ._-______-________-___________________-__________0
? l -1 1 Thetransient[ data'indicatesthatsomepumpsexperiencedsaturatedconditions In this sense " cavitation" at the pump suction for a short period of time. may have occurred. HowcVer, the classical cavitation concern is not regarded as the principal cause for the indicated low flow. Because the pumps were operating at minimum speed, most of the driving head comes from the density difference (natural circulation). During the transient, flashing acted to decrease the available NPSH and increase hydraulic losses resulting in A net increase in the w+F4 fraction due momentary reduction in flow. to increased boiling in the core then occurred, whic$"rAestablished natural circulation flow conditions. Questions 3: Would the APRM's be expected to show a change due to changes in flow at the conditions which existed at the time of the transient? i
~~Response~~
Prior to the reactor scram, the APRM's would be expected to show effects of significant flow changes. However in the transient which occurred, the APRM's were reficcting effects of the rapid changes in pressure and flou variations. However the flow changes would be expected to appear as second order effects. After scram, the APRM's would not be expected to provide indications of core power level. Question 4: During pump cavitation is the annulus separated from the core region? During the Oyster Creek transient could this have resulted in a triple . low level?
~~Response~~
As noted in response to Question.2, in high pressure systems, pump -cavitae-ion does not result in large vapor fractions which would block the flow through the pump. Also, review of the transient data shows that at the low flow state, one recirculation loop was providing nearly normal flow. Thus, pump cavitation did not separate the annulus from the. core region during'the transient. The triple low level: signal was not due to loss of recirculation flow or separation of the core from the annulus. To boil-off the mass of liquid existing in the seprator stand pipes and region between the core spray sparger and the shroud head would require the recirculation flow to be zero for more than one and one-half minutes. Since a low flow indication prior to the triple low level signal existed only momentarily, (one indication-in a 12 second period), the triple low level signal was not .due to boil-off associated with annulus / core separation. 4
i 4 . Question 5: What is the significance of pump cavitation and how long can the pump be allowed to cavitate?
~~Response~~
Pump ca.itation can result in some performance degradation and, if allowed .to continue for a long period of time, can result in damage to pump components. Per the. tests' referenced in response to Question 2, high pressure pumping system can operate with saturated and two-phase conditions at the suction inlet with only minor degradation in the pump head capacity Curve. The period that a pump can operate in cavitation l's a function' of pump component design constraints. The pump manufacturer should be consulted to provide equipment life-time estimates. - Question 6: Under worst case conditions how long will pump cavitation occur? Will these conditions be self limiting or will they require operator action?
~~Response~~
Boiling water reactors normally operate with about' 25 BTU /LB subcooling in the recirculation loops. This subcooling is provided by feedwater flow and is sufficient to protect against pump cavitation during transient conditions. At the higher core flows, a core flow runback interlock exists at 20'4 feedwat_er flow to provide added. protection. Pump cavitation is most likely to occur when operating with low feedwater flows, i.e..minimun subecoling. This state occurs with low power, low recirculation flow operation associated with plant startup. If a rapid-depressuri:ation event should occur while in this mode, pump cavitation can occur at intermittent periods during the transient. Pump cavitation condit' ions would ultimately cease to exist once: a) The source of the rapid depressuri:ation was eliminated and the pfant reached an equilibrium operational mode, or b) Low low level is reached and automatically the vessel is isolated, the recirculation pumps are tripped. .The key point is that once the plant reaches a steady operationalmode with feedwater flow available, cavitation conditions cannot exist.
/YRC CIMRTlnU CM N W:' ?/h. t., i. ; + l: ; 6 -(. w,/ r) .fru.ric.nt-: 3, o l Questien 1: 4 II. l 4
~~I Provida an explanatics dy the. triple leu level sicaal occurred during the.~~
~~y t Oyster Creek transient. t . h.,. I~~
~~Response~~
~~1 3 and its relation to. t. Figure 1 deicts the triple low water level sensor systs: f,, g key factor pressure vessel corponets. - .i i. )~~
it 'can;be shown that the sensor reading (UP) is given by

-
i-t i d ^ )Y' N f 'h 1 }' i;- .+ aP + 6P = 9 ~ h gp Z S'i1E ~ h. ?. .i..:,l AP Separator dynamic pressure loss- [ .. -q SEP t Elevation head ebeve.the cane spray sparger (tw ph:se idxture)" ' ;;- - ', ' ? AP '7 indicative of the water. le rel- !. t y Elevatics head of sten batmtn theyate surface ara tf:e'c: par.I 'AI SMTZ i E prassure tap 1 Elevatien hesd betreta ths ccm spr y sparger. and penetraticn h ~ ? '(408-375)T
~~/, =.'. c, p 3 2rouch the reacter pressure vessel~~
,, g. _ 'f' Specific voluse of. saturated water in the vessel yf Elevatics head bebeen the upper and Tener tap reacter pressure. " (SM-4c0)Tg nssel clevaticns-Specif c volume of instrument' line wat r asstimd tD De. at 100 Fi j-0 Yg,
~~~ z Normally the steca dryer pmssure drop rould also appecr' fn this egiatist, but -~~
because of the lee care pcwr and cerres:s> ding low steam flow (=2% of' rated) during-the event this pressure droo was essentially Irro. Thereiere the water levels.inside and outside the dryer skirt were tne sara. The tripla.. low sensur-is se t ta trip when the static cuid water level is 121" to 123" beler the upper tap. - This. correspani to a DP in. the range of.-4.22 psi to -4.23 psi (ir.cluding a ccrrectica for stem:s den-i sity).. t is reproxfmately 10 psf. !? zever..E -. CuHng rated power and fleu c:nditions,' A?g 155.' during la:v power and flow conditions'such ar existed during the event, A .sufficiently reduced either by a true reduced water level cr by a red 1ced effCctive 4 caola:tt density (high vald fractica in separ:ter sts7dctges).Wara th.? carer 7he i < 1atter state armer.tarHy occarved d'e.rhg t'.ve trusiet, resAtty h tbc tH9 e Min.. 1 > jc ' ) level signal, ( f;y. r &st prior to tha enening of th'e bypass val res, the core e'tist quality was 1.64T. f and the void fraction was approxi=ately 22% With the rapid deprnsurization.: t".e:- [ care exit quality incredsr.d to 3.I'6% (24$. vofd f>sctica). Tha depressurizztien, } alone did not cause tha CP to ha less thaa -4.22 ps!. hVwerer, considering the -
1
~~'.) los recirculatics. fica prior,,to the triple loe siqaal", the cere exit c:;Itty was-approximately 9', (555 void fraction) and the CP =-4.25 psf.~~
~~f. this case the wat.ar.y~~
~~"{ level was approxi:sctal'y 141' incnes abcve the spuger, but tha. reduced effective : .u ..l m "s s . s.. _ c..; y, '.., b.* .N~~
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4-coolant density could not offset the other elevation heads in the instrucent loop. . This resulted in the wsentary triple icw signal. The triple icw indication *Rs momentary because the.lcer recirculation flou existed only for a very shcrt time j-during the transient. g: ~ 2 Qeestion 2: .e.4e,b........ 6t for tM RCtm 'isolatMf..... c 4 .t.#o.,i;.d.,eu.i. d.;n.:.,af..li .. ~.... .r. ce reas at.57 ie::ab.y.m m. r:*f 3.y Q Q,atr Po,..jo d e n 1 d. s a w p:f'%.',.c.::..~....gf ;'s3.;.]..:.!_.p'Q;l .c la,1.4 ").'U. :.m...a...., w. s.1(..'.,QK:.G yYf;C yi' ,f* w a: ~. ....:..,.n.. .m .r ...c V '. L Aalysis of the transient data shows that. a tripic ';w level signal w:s not present when the reactor building closed cucting cater (RS U3) holatica occurred. Prior ta the electrematic relief valve (BV) opening and subsequent RD CG isclation. .The the plant was operating at about M pcwcr, 341 flow with 5 BTWt.3 subcaoling. GV opening reduced pressure by about 42 psi. Saturstion ccnditions in the.recircu-lation Iceps were not reached du-ing this depressurizatico as dezenstrated by the . flow and temperatura data and suppcrting calculations.. A!ditienal voidin2 in the"., a core was calculated to.be sr.sil (~M quality increase), a.d the triple icw sensor-Since t.'c triple low level-pressure diffarential DP uns calculated to be -3.25 psi. sensor was net set to trip ur.til the op was less than 4.22 psi, a triplc Icz. level .c; , signal.was not pra:ent when the RS CC 1 solation occurred.-. c".* bestion 3:- .~ j.. Was' cacaunicatio, lost betzeen the annulus and the. cere?. 4
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~ ~;.,. ~ ... g _Tesponse:f n.. 'Aalysis of: tne transient data indicates that local voiding in'the recirculatiovr1- ~ iocas may have cor.ntarily reducmd the ficw in foer of the five loops. NGmIver.. ?z 5 the data indicata tfnt this reduced fics state e::isted only m:=cntzrily, and that'.. the voiding.in t*ta recirculatien. loc;:s did not tusult in a ex:=pleta vsper blockage'. in any pics or lecp.. Therefore, the cc:ctmicatica of coalant bet:mmen the acncl.se end the core wt.s not lost ducir.g the transient. Ciculated void tractioas within the recirculation (xmps at. the suction of t$e~ I. pumps wcre less than 33. The steam quality at the suctierr of the pass was calculated. to be 2.tHi'.' based on the atarage recircuhtien leap te:peratu.c anda, assu=ing a constant enthalpy da;r.essurization frcs steady c::erating conditions., I-i before the stess bypass valves opened. Because of the kinetic energy irparted to i 1! the tec-phase mixture at the in6eller pm-;phery, an editional 0.05t staan giality' Q is generated within the recirculatica pun:as. Since t!;e saferity of the "luid inside-the pegs was greater than.fE liquid on a volusa: basis, a cc=pletz vapar blockage - S, within the.recirculaticn pucps would not occur. Aso, the ' recirculation pept are located at the icrest point of the systes, below' the pressure.vesse.i, and the piping does not contain in'rerted "U's" conduci're.to-5 trapping vapor.- Taus.any vapor in er near the pumps sculd have a tendency to ' rise.. ,.4 ir.to the pressure vessel pr1rrfdtng added assurance that a coepiete vapcr blockage.. ~. L.p,7 3.$.,. ---. g. . neuld nats. occur.,..,l The retMctica inirecirculation flats is attributed: to facrected pressure'1'osses in. the recirculation 1 cop due to two-phase flow, no. net increase 'in the. buoyancy beboen t'-e core amt the asnulus, and a net d: crease in p=cptng head (325 laxar). Thisa-tini:ced ficw ' state existed only merientarily (as indicated by tha data) until the.* escass enorgy la the recirculation:1 cap was rescved by. vaporization and additional ?"- ..n
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L e. b voiding in the core (still an enerhy sourca) generates a buoysncy differenc.e. i 'tetween the core and the annulus. This nl. establishes tt::2 flee of coolant I through the recirculatton loop. ![ It is concluded that' at no time did a c_~:pleta vapor blockge, or state of pun cavitation, occur wMeh resultad in tM loss of c::s:::mic2 tion of ccolant .Batocen the annulus ar.d the care. t ~ .:\\ t t 4.. 3 .~. 5 ). s s ..e r. s a i s. 7. g o 1 s s. ~,. *. .t, l., f ,g . 4 5 .,V..;, s .r ,e a
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) Report to CORB Action Item No. 375 Appendix E " Bounding Ioss of Coolant Inventory Transient f or the Oyster Creek Plant" by the Exxon Nuclear Company, Inc. w e8 e em 9 l e 4 s e ap 4 O A 4 4 O ... __i
BOUNDING LOSS OF COOLAMT INVENTORY TRANSIENT FOR Tile OYSTER CREEK PLANT INTR 000CTi0tl The May 2 scram event, is bounded by the most limiting loss-of-coolant i In centrast to the May 2 tory transient, the complete loss of feedwater flow. scram, which showed a net coolant inventory increase (8070 lbm) following in .tiation (l'), the loss-of-feedwater-flow transient would involve a decrease coolant mass inventory of approximately 19,000 lbm during the period from initiation until system isolation, as discussed belcw. Previous loss-of-feedwater flow analyses (2a; 2c) were performed without In those-analyses, the reactor was the function of.the isolation condensors. allowed to pressurize with pressure equalization obtained with the relief valv The utilization of relief valvcs provided additional conservatism in the analyse In both since the resulting mass inventory loss would follow sys tem isolation. ld the previous and current analyscs, the net change in system mass inventory The previous analyses assumed be about the same to the time of MSIV closure. that, following the early stages of the event, the depressurization would b -limited by operator control of the isolation condensers. The loss-of-feedwater transient was re-analyzed, incorporating the fun l the plant sys tem .of the is'olation condensers, to more realistically simu ate The results responses as they would occur following initiation of the event. of. the re-analysis and the modeling of the isolation condenser is de the following sections. t
. A!1ALYSIS OF THE LOSS OF FEEOUATER TRAtlSIE!1T U!TH ISOLATI0il C0f;CEilSER ACTUATIOff The loss-of-feedwater transient analysis was perforced for a full core of E!1C 8 x 8 fuel using the El(C Plant Transient Analysis' Code PTSBh'R2 (2b.) The transient was initiated from a full power icvel of 1930 til. The initial water i3.?A ~ r 6. IV <*w 'acN
~~1. s 3 '.~~
level was assumed to be one foot below nonnal operating level to minimize the gg initial system coolant inventory. System responses for the first 125 seconds of the transient are shown in Figures 1.1 and 1.2. Feedwater floti d'ecreases to zero in 3.5 sec, reducing the downcomer inven- ~ tory. The downcomer water level falls rapidly, reaching the icw level reactor scram setpoint (11 f t. - 5 in. above top of active fuct) at 4.5 sec. The low-low icvel setpoint (7 ft. - 2 in above top' of active fuel) is reached at 15 sec. At this point, the following events occur: 1.. Main steam isolation valves (MSIV) begin closing (10 second closing tine). 2. Main recirculation pumps trip. 3. Isolation condenser return valves signaled to open. 4. Core Spray pumps are signaled to start. The reduced recirpulation flow after the pumps trip significantly reduces the rate of change of downcomer water level. A minimum downcomer water level of 5.36 ft. above the top of the active core occurs at about 35 seconds. After MSIV closure, the isolation condenser heat removal system initiates system depressuri za tion. The change in pressura vessel coolant inventory up to the time of MSIV closure is given by: t MSIV t MSIV c c r f W Wfw t )= o OMsys tem " ms st=0
. and was calcul'ated to be - 19,300 lb. That is, more mass was lost in steam m The minimum cri tical flow to the turbine than was made up in feedaater flow. power ratio does not decrease below its initial steady-state value, and the maximum do'me pressure during the incident was 1047 psia, below the setpoint of the relief valves (1065 psia). Beyond the first 125 seconds of the transient analyzed above, the sequence The limited amount of inventory makeup available from the is straightforward. control rod drive flow is not expected to raise downcomer level at a sufficient rate to clear the low-low level indication. Since the various safety systems have-been actuated at this level, no credit is taken for operator intervention. The system will continue to depressuri::e until core spray flow is intreduced' to the vessel at approximately. 285 psig. The water. level in the core at this point has been calculated to exceed the. Iow-low low level setpoint (4' 8" abave the active. fuel). This level estimation is based upon a fully collapsed level of the fluid mass at saturation conditions. Following initiation of core spray the level will recover and the event tenninated. ISOLATION CONDENSER MODEL The Isolation Condenser is a model addition to the base PTSBWR2 code As,part of the overall model, it receives steam flow from the first steamline .The return node and returns condensed steam to the bottom of the downcomer. valve characteristics are shown in the Figure 1.3. A delay of 10 sec. from time of low-low level setpoint to start of valve opening and a 20 sec. opening time are assumed. The rated condenser flow was takcn to be 330,000 lbs/hr per condenser.. ~ 'N_
4 EFFECT OF MODEL A001_ TION TO PREVIOUS ANALYSIS The Isolation Condenser Model is external to the basic calculations of the code, and would only be activated during a transient wherein the low-low level setpoint is reached. During no event previously analyzed (2a, 2c, 3),' with the exception of the loss of feedwater transient, did the water level reach this setpoint. Thus, the model addition would not change the results of transients other than loss of feedwater. e e S e e e e 4 9 e M e 0 k
' ~ ' ' ' ~ ~ ~ . t 1 REFERENCES l R. E. Collingham, J. D. Kahn, C. E. Leach, K. P. Galbraith, " Evaluation of 1. the Oyster Creek Reactor Core Liquid Level Following the Inadvertent React High' Pres'sure Scram on May 2,1979, XH-NF-79-49, May 11,1979". Amendnent 76 to the Oyster Creek Nuclear Generating Station Facility D
2.
tion and Safonty Analysis Report, Jan. 31, 1975. This document included the following reports: JD Kahn and MS Foster, "P'lant Transient Analysis of the Oyster Creek - with Exxon fluclear 8x3 002 Fuel Assemblies, XII-74-43, Revision 2, 2a. January 1975. JD Kahn and MS Foster, "PTSBWR2 - Plant Transient Simulation Code fo Boiling Water Reactors, XN-74-6, Revision 3, January 1975. 2b. JD Kahn tnd MS Foster, " Plant Transient Analysis of the Oyster Cree Fuel Assemblies, XN-74-41, Revision 2,- 2c. wi di Exxon Nuclear 7,t7 002 January 1975; JD Kahn, " Additional Plant Transient Analyses of the Oyster Creek GWR 3. Exxon Nuclear Fuel Assemblies, XH-75-51". September 1975. i W em em t m e E 7 G
_s .._..__.r_ t t 1 1 i I l l i Re po rt to CORB Action Item No. 37 5 ~ i Appendix F AI 8003-09 (Big Rock. Point - LER 79-022) t + I I r l i r [ t t i i 1 i i i i i f l l l t
7 l j .. ~ AI No. 8003-09 OPERATING EXPERIENCE REVIE*4 AC'" ION ITDi ASSIG' MENT FORM. I. EVENT DESCRIPTION Unic/ Docket No. Big Rock Point (50-155) Source' irw 7a_no9/n1y_1 8/22/79 Event Date Event /Cause Description Design review of primary system water level sensors, LE RE09 (4 units) and LE REOS (2 units) revealed that during postulated loss of coolant accident conditions the automatic initiation of ' reactor scram, containment ~ isolation, core spray, and reactor depressurization; that are initiated by these systems might not function due to flashing that could occur in the reference line during rapid depressurization of the primary system. No hazard to the public occurred. Reportable per tech spec 6.9.2.A(9). This item represents a generic design shortcoming that was identified by the NSSS vendor. The temperature compensation features of. all six sensors were modified and calibration and set-point specifications were approved by technical specification change (amendment 31) dated 11/2/79. II. APPLICA31LITY Preli:ninary review of this event indicates possible applicability in the following area (s): Procedure,s Systems pres Components Type p v-nr r. 3,.. 1,v,,1 sensors Mfg. y; m y Model No. Operations ] Maintenance O Training Design RadiologicalO Protection Other O Yes No Supplementary Information Enclosed - O G
W~,- . u. m_ .. k PA So 7 / S67U C6 TR DATE TASK RECORO ^ SECTION CEPT. SECTION MAN AGER
~~OPERATING EXPERIENCE REVIE'J PROJECT ME AI HUMBER NU 8ER 94 N M8Ea t~~
6 7 9 9 20 21 25 26 3 32 35 f%Co '7, llal^ D,0.3,o9.,.... /,3 oial /ol/Sild A#4 ~ c$c k ENT PLANT ANALYSIS RECCMMENDATION 79'8 Cl 68 69 36 ek /'8IMlNkvl kdiMiTINII lbdV!6id l$ MSidkiilIlll h0!d1 if G ll IR!s!,dFolahse loluxl/W61 K'o'OAl I i i l i I I I I I I i LJE/d :1 l'MI-ioiM t Pdco 'nMeivioi4l7Mo Will i I I I I I I l i I I I Ii!I IililiiliI zi /l,l C N T" h 4!M / /[ Cl ITit/! d IA!a9 81bil!(!MIOl/I(l/ T d/l I ! I IIIII!lI!I h I I IdF! l'T!M/ I51 Dd Mll6]M !Bld d / \\6d l 7toi 101.lG,i l ilIIiilili 41 i 1 !4Wlol IslF4icl/ l@l l*M Wcts b'tl7 e MW I I I ! I IIIIIIIIII si d l Icloigk!s'c.Prlik/Id 14:.l7ttletaf I ! I I I I I I I I I I i i I I I I I I I I I !st 11 ll11111111 111111 I11111111111 1llI111111171 sl ill!1111111 11 l'ill 111ll1I1111'1111111111 ai illlilllilli I 'l11111 IIllllllli 1 I 1 1 ! ! 1 1.1 1 \\/ COMPT.ET10N REQUIRED BY ATTACHMENTS ? ORIGINATOR DATE ADDITIONAL DESCRIPTION DATE RESPONSIBLE ENGINEER. DATE SECTION MANAGER DATE PLANT NfALYSIS MANAGER CC: ACCO C024
i . I c -- -.= - ->_..a 1 L l i t-i 1 1 J 9 }. 4 + 1 .) 1 i t Report to CORB Action Item No. 37 5 A Appendix-G Response from Oyster Creek ( A. H. Rone) to PA-193 4 J i i i 4 f 4 I I, .l e 4 ^ I f t .j .h' = t + a e t I 1 i
~.. L { V d. He t..<d 's : Inter-Office Memorandum Decerber 19, 1980 m mm ). _ y. r-r -w= Resp'onse to Rcquest for Assistance [j-- 7} = contained in letter fran P. S. Walsh 1.1,.2.. W. _.% 1,wu lect to K. O. E. Fickeissen, Jr. Subject Cyster Creek Transient of July 17, 1980 P. S. R11sh L caHom ~ Oyster Creek _= This mero is intended to respond to the question posed in the subject letter relative to the Oyster Creek July 17, 1980 event. For clarity the questions are restated herein follcwed by the response. Caestien 1. Khat caused the Peactor Building Closed Ccoling Water isolation? Is the depressurization by the Dav opemng and closing related to the Bard isolation? Can we expect similar isolations to 1 occur in the future? Can we or should we prevent similar future isolations? Please verify by list those carbinations of sigrals that will cause isolation of the Peactor Building Closed Cc61in, Water (RBCCW) systen as occurred during the transient. I ~ Resp 3nse During the 1980 Fefueling Outage the RECCW Syston isolation valves were nedified such that they isolate on triple icw level 4 or on coincident signals of high drywell pressure (2 psig) and low low Rx water level (7'2" ATAF). Prior to this nodificaticn RECCW isolaticn was a manual operator action. At the time RBdCW isolated during this event ncne of the autccatic' isolation signals were known to be in. This determinaticn is based upon inspection of the event recorder traces and cperator rccollection of the plant status urmdiately preceding, and during, the isolary a event. It should be notcd that the drywell high pressure ar: Je r Irv water icvel signals actuate several = other auta W cc.ns, none of which occurred at this time. Had the isc.5..g fn due to a triple lcw level event the event = recorder shc 4d rA a mitiated at the time the isolation occurred. The sama rel'ays are involved in initiating the event recorder as are used iri the isolation icgic for RDTW. It is possible, = bcwever, for a slight difference in contactor arm tensions to = 3 allcw one contact to renain closed while the other ocens if the ~ relay armature travels only part way. This a:nditicn is highly unlikely since it would have to occur #or nore than one relay,. but is possible particularly for trip signals which are " pulse-like" in nature. It should be stressed that the coupling of the RECCW isolation event to the same general time period as the cperation of the D-E Dmv is not suppartable by any recorded evidence. The tm e frame was established based upon the recollection of the Shift Super-visor (SFO), Group Operating Supervisor (SPO) and the undersigned (SFO), all present in the control roca at the tre of the event. L e
( ; 4.. The undersigned interviewod,the aforerentioned operating personnel separately following the event and found no major disagrca ents between their recounting of the events and that of the undersigned. However, it is recognized that severe time frame distortion can occur under stressful situations, and it is possible, though not highly likely, that the RBCCd isolation actually occurred at the r time of the kncwn triple low level event. This possibility was incdiately recognized but was discounted based upon subsequent discussions by the aforcrentioned licensed personnel present in the control rocm and their comen recounting of the event. In strmurv, the isolation of the RDCCd raniins an enigma. Ecw w ts it determined that the triple low level indication had Question 2. Event recorder? Cent. col board alarm? Alarm been activated? printer? If triple icw level initiation is an isolation signal for RPCCW, why was there no isolation? Was the isolation relay unable to trip during the time the triple Icw signal was "in"? Is it a time delay relay / system? Sicw acting? Diref contacts? Ecw was it determined that the triple low level indication Is it possible to determine the subsequently " bounced out"? period of time that elapsed between the "in" and "out" indications?
~~If not, can an estimate be made? Please forward a copy of the event recorder output for the transient period (7/17/80, 20:25:23 to 20:49:19) showing reactor water level during this period.~~
~~We triple lcw level indication was detezrtined based upon the~~
~~Response~~
event recorder. The event reccrder ronitors the status of four relays which correspond to the four triple low 1cw level sensors. 'Ihe triple icw level event was characterized by the r.urentary sinultnncous operation of all four relays. The p21se duration was indeterminately short and appeared as a line retraced upon itself. No alarms or other indications were observed to be present indicative of a triple low level condition. It should be noted that the determination that a triple lou level event had occurred was made subsequent to the reactor scram and restablication of plant param2ters. The timing beem the scram signal and triple low level event was superimposed upon data extracted frcm the plant ecmputer in a best estirrate fashion. It cannot be explained why an isolation of the RBrd did not occur at the time of the triple low level event other than to assum2 that it did occur and was not correctly placed in time.- This possibility is addressed above in the response to question one. Khat would have happened if the bypass valves had rcrained open l Question 3. l longer? Would the event have,beccue more severe? i 't e ,i r-g ~
2
~~Response~~
The response to this question can only be answered by detailed mulytical nodeling of this event and/or other similar depressur-ization scenarios. It is conjectured that a longer depressurization interval would not have produced any results significantly nere severe than those experienced during this event. This conclusion is based upon the fact that the depressurization did not threaten We ability to maintain adequate core cooling, even though the , reactor was not scranned and operating under high voided conditions. Cperation at less than 354 Mdt ( 18% pwer) is permissable at any pressure and is used as part of the licensing basis for the plant. Question 4. Our docket currently analyzes the spurious cpening of one bypass valve or 14RV. Should we analyze the open.tng of all bypass or 2 relief valves at both high and lcw pcwer?
~~Response~~
Whereas it is expected that the results of such analysis will yield few surprises, it may prove interesting to perform the suggested analysis factoring in oparator errors and cquijrnent malfunctions which could result in operation at interrediate pwer under extren21y high voided conditions. Consider the follcwing scenario: 1) Plant is operating at approximately 50% power when a failure occurs in the pressure regulator. ~ 2) All nine bypass valves open as a result of the failure and a depressuripti<'n of the reactor begins. 3) The operator transfers the node switch out of Run to Startup. This acticn will not cause a scram and will block the closure at the MSIV's and~ the scram attendant thereto which is designed to prevent operation at, p::wer when less than 825 Psi - 9 Although the above scenario is unlikely, it requires only one cquipicnt malfunction and one operator e_mr to yield a condition of high power and high mids. Question 5.. Khat hunun errors and mechanical failures contributed to the event? Could better procedures or human-engineered controls help prevent future errors during startup?
~~Response~~
The event was initiated by failure of the No. 2 vacuum trip to reset although the available irdications indicated that a reset had occurred. The event was aggravatcd by the operator reseting the 12 vacuum trip while the pressure regulator was set to a value lower than Rx pressure. The available instrumentation and controls used to nonitor the plant during the event were satisfactory. Dunng the post event analysis conducted by the Plant Staff arai the NBC operator actions were discussed and were considered to be gcod overall. Furtherrore, no ptocedural deficiencies were identified as contributing factors to the event. Howver, a procedural change was made to test that #2 vacuum trip does indeed reset while the Rx is still low in pressure.
Were is an a'dditional safety cencern which was not addressed in any of the discussions related to this event. Prior to axiifying the F1CCN icolation i valves to'autcrratically icolate on the aforcmantioned signals, there existed i nat only one ucchanism which would cause a loss of the dt,fwell coolers. mechanism was a Contaiment Spray Systen auto start signal. Bis assured that i i sone mechanism for cooling the contai:wnt would be in effect upn loss of the drywell coolers. It has now been denenstrated that another mechanism exists for icosing drywell ccoling via isolation of the Rard drywell, and that this condition can occur under conditions which do not cause an auto start of contairnent spray. We consequences of a loss of drywell ecoling event could yield centainrrent tarperatures kull in excess of those for which the contai=ent is designed. ne drywell liner desi'gn targerature is 280 F and under these conditions assuming an adiabatic heat up of the drywell, the liner te@erature would tend toward the te:peratura of the reactor vessel (i.e. 549 F at 1020 psig). H is could lead to 0 a loss of contaircent integrity, distortien and or loss of Rx water level indication, as well as loss of function of other cquiprent inside contairment critical to safety'. It should be stressed that the operator would have little or no indication that the contaiment was heating up other than drywull pressure l increases and sane tengerature instrurrentation in the centrol roon. It is requested that Plant Analysis investigate the concern discusscd above to determine the realistic maxirum tog 2rature that a loss of dt,jwell ecoling could yield, and the tirre required to reach the maxinum allcuuble contair:nent temerature. Sasad upon this evaluation, recortrendation shculd be made to the Project Engineering Group within Tcchnical Functions to redify the PSCCW Syston isolaticn Icgic to eliminate this as a possible mechanisa for a less of drywell cooling event. An Engineering Roquest is being forkurded to D. Grace, together with a copy of this nuro, asking for an improved contaiment ta:perature ncnitoring systen. Emergency procedures are being revised, consistent with the G. E. a:ergency Procedure Guidelines and the time table related thereto, to provide operator actions for high dry-sell tamperature synpts. ]f ~ ~& A. H. Pcne Engineerire Y2 nager AHR:dh cc: R. Barrett. mwr wh ~ ,J. T. Carroll, Jr. 'J. DeBlasio ~ K. O. E. Firkeissen, Jr. I. R. Finfrock, Jr. D. Grace E. J. Growney R. W. Keating J. Knubel i M. Laggart J. P. Maloney i J. L. Sullivan, Jr. N. G. Trikouros l
- -_-. ~ - -. _ - = - 1, e l 1 8 s Report to CORB Action Item No. 375 Appendix H Oyster Creek - LER 80-38/3L '*Drif t of Triple - Low Water Level Switches" ~ s M* e p 6 9 e .a.' 1 s e 0 e . s 4 9 e G T
act a : -: _. ~. m ,_ww -tc.:- . _ = _ __ ~:: ~ s QYSTER CREEK NUCLEAR GENERATING STATION Forked River. New Jersey C8731 License Event Report Reportable Occurrence No. 50-219/80-38/31. Report Date September 25, 1980 Occurrence Date August 26, 1980 Identification of Occurrence Exceeding a limiting condition for operation as per Technical Specifications, Section 3.1, Table 3.1.1, Function G.2, when reactor triple low water level sensor RE188 exceeded its required setpoint during surveillance testing. This event is considered to be a reportable occurrence as defined in the Technical Specifications, paragraph 6.9.2.b.l. Conditions Prior to Occurrence Steady State Power Power: Reactor 1563 MWt Generator 491 MWe 12.1 x 10j gpm Flow: Recirculation Feedwater 5.76 x 10 lb/hr Description of Occurrence: On Tuesday, August 25, 1980, at approximately 1130 hours, while performing routine surveillance testing of the reactor triple low water level sensors, RE188 tripped at a level which was less conservative than that specified in the Technical specifications. Tests on all level sensors yielded the following data: Pressure Switch Desired Manom'eter As Found As left Designation Readino at Tric Point ("Ho0) ("H90)_ ("H2O)_ System I RE18A <126 123 123 RE18C <126 123 123 System II RE188 1126 128 123 RE180 <126 126 123 The "As Found" value of 128" H O corresponds to a water level 54" above the 2 active fuel. The Technical Specification limit is 56". - - - -. - - - - _ _ ~ _ _ _. - _. _ _ _.
y .= m...x _ ~ _ +- Reportable Occurrence Page 2 Report No. 50-219/80-38/3L Accarent Cause of Occurrence Sensor Repeatability Analysis of Occurrence Failure of pressure switch RE188 to actuate at its prescribed setpoint would have delayed initiation of reactor triple low water level indications. However, due to the existing logic configuration, the redundant switch, RElSD, would have actuated to initiate the required functions at the required Technical Specification limit. The safety significance of this event is considered to be minimal since sensor RE18B's non-conservatism resulted only in a temporary loss of redundancy in the system. Corrective Action -Reactor triple low level sensor RE188 was reset to trip with its prescribed limits. An engineering study is in progress regarding the feasibility of replacing the existing sensors with a solid state system. Failure Data ITT Barton Differential Pressure Indicatin.g Switch / Switch Model #288A ( Adjustable Range 0-150 inches H O 2 e. e- = a gm O I ) l O 9* 9 e
r h ~ '.s. .i..~,n.. ,<:.1y4.. --,.4,: e : -.dc... i e:v.44 n. a. f ..l 1 I t Report to CORB Action Item No. 375 Aopendix I NRC Inspection Report for July 9 - Augu s t 1, 1980 Repo rt No. 50-219/80-25 (partial) L 4 e9 mum O S 1 I s e
mam ~ .. ~ '. :.-[. ..q U.S. NUCLEAR REGULATORY COMMIS'5IW" OFFICE OF INSPECTION AND ENFORCEMENT ' P .=. e. 9 ~ . Region.I Report No. 50-219/80-25 Cocket No. 50-219 Li. cense No. OPR-16 Priority Category C Licensee: Jersev Central power and Licht Comoany Madison Avenue at punch Bowl Road ~ Morristown New Jersey 07960 Facility Name: Oyster Creek Nuclear Generating Station Inspection at: Forked River, New Jersey Inspection conducted: July 9 - August 1,198 Inspectors-h h~ s?d
~~origgs, nior n~~
~~. neace.or Inspector care signea d)-. ~, ehsleo J incmas, Resicent Reactor inspector icat'e signec. cate signed _ Approved bf: Af~~
~~h. Jh - [0 jf. R. Keimig( Chief, Rgtor Projects cate signed~~
/ Section M.1, RO&NMranch Insoection Summary: Insoection on July 9 - Auoust 1,1980 (Recort No. 50-219/80-25) Areas Insoected: Routine inspection by the resident inspectors (139 hours) of: ' followup of operational events that occurred during the inspection; review of plant operations (startup); tours of the facility; log and record reviews; and followup of IE Bulletins and Circulars. Resu'l ts: No items of noncompliance were identified. Region I Form 12 (Rev. April 77) o
,e e p- , =, 'Q DETAILS 1. Persons Contacted _ J. Carroll, Station Manager K. Fickeissen, Support Superintendent W. Garvey, Director, Station Administration E. Growney, Engineering Supervisor, Acting T. Johnson, Supervisor, Station I&E Maintenance J. Maloney, Operations Supervisor J. Sullivan, Plant Superintendent The inspectors also interviewed other licensee personnel during the course of the inspection including management, clerical, maintenance, and operations personnel. 2. Ooerational Events 17,1980 at 8:35 p.m. during power ascension following plant startup, operators attempted to control reactor pressure by operation On July The operators reduced the of the steam bypass valves (SPV's). pressure setpoints on both the mechanical oressure reg Control red insertion was begun, but.before the heatup and pressure increase could be teminated, one electromatic relief valve (ERV) open. In about 20 seconds opened at a reactor pressure of 1050 psig. When the ERY pressure decreased to 1000 psig. and the ERV reseated. shut, The RECCW isolation was cooling water (RSCCW) system occurred. As reactor system pressure i immediately reset and flow restored. again increased, it.was detemined that the low condens This The trip reset was actuated by the operator. was energized. As reactor pressure action caused all nine SPV's to open fully. decreased rapidly due to opening of the BPV's the operator tripped The resultant pressure transient caused a momentary _. shut the BPV's. triple low water level. indication (Technical Specification value of i 4 feet 8 inches or greater above the top of the active fuel) on the I event recorder followed 24 seconds later by a reaccer water low level l inches j - scram (Technical Specification value of greater than 11 feet 5Following above the top of the active fuel). review was conducted by the inspe f the instrument recorder charts, the event recorder chart, and theIn ad process computer print-out. The investigation with the on-shift reactor operators and supervisors. led to the following conclusions: The recirculation loop isolation valves remained open on all five loops allowing sufficient water flow between the core and
y annulus regions to provide " accurate water level indication on both the Yarway and GEPAC -instruments. Indicated water level on these in'truments, which.have both reference and variable s taps in the annulus region, did not go lower than 6 inches below the icw level scram-setpoint. No alarm actuations, automatic containment isolations, or emergency core cooling system actuations that would have indicated a double low water level condition (7 feet 2 inches above the top of the active fuel) occurred. No isolation of the RBCCW system occurred at the time of the event recorder indication of triple low water level. The triple low water level indicated by the event recorder was apparently caused by a hydraulic disturbance in the annulu's ~~ area of the reactor vessel resulting frem rapid closure of the BPV's. The triple low level sensor is a differential pressure switch with a high pressure tap on the core spray system sparger in the core area and a low pressure tap in the annulus area. When the BPV's closed, the rapid pressure increase in the steam headers and the annulus regicn in conjunction with the time lag associated with equalizaticn of pressure acrcss the steam dryers, caused the pressere in the annulus to be momentarily higher than the pressure in the core regien. This caused a mcmentary erroneous indication of triple low water level that activated the event recceder. The indication was of short enough duration that the electrical relays in the RBCCW isolation circuit were not activated. Subsecuent surveillance testing was performed on the RBCCW isolation ~ ~ system to determine the cause of the mcmentary RSCCW isolaticn upon closure of the electrematic relief valve. All isolation actuation signals (triple icw water level by itself or double low water level with high drywell pressure) were tested and no malfunctions were found. The cause of the RBCCW isolation was determined to be a spurious mcmentary actuation of the isolaticn circuitry. Investigation of the malfunction of the low condenser vacuum trip revealed that misadjustment of a 71mit switch on the trip mechanism ' caused the indicatir.g lights in the centrol roan to indicate the trip as reset.when in fact it was not. The limit switch was adjusted and the low condenser vacuum trip was tested satisfactorily. The licensee has agreed to change Precedure 201.2, " Plant Heatup to Hot Standby", to require verification of bypass valve operability prior to reaching full operating pressure to preclude recurrence of pressure control problems on startup.
+ ,, w Question 1. Do the Recirculation Pumps have a vibration trip or alarm, which may indicate ~ -pump cavitation? What do the procedures tell the operator to do? How long does he have'to do required action?
~~Response~~
-Each' recirculation' pump does have a vibration alarm. The alarm is set at.4g's and annunciates In the control room. During the events of July 17, 1980 none o' these alarms annunciated. The action required by procedure for a pump high vibration situation requires the operator to: a) check. pump speeds to ensure all pumps are'in unison, b) adjust out of sync pump' If necessary and reset alarm, c) If alarm will not reset, change pump speeds and try to reset alarm, d) If alarm will still not reset remove the pump from service. No time limit is specified for carrying out these actions. However, discussions with the pump manufacturer indicate that it would be hours before mechanical ~ damage would occur. Question 2. Did the Recirculation Pumps show decreased amperage when recirculation flow decreased? . Response LUnusual or of f normal. pump motor amperages were not noticed by the operator during this event. Question 3.. Could there have been a core-annular region disconnect when recirculation flow went..to zero or cavitated? If so, how long before a manual-action is necessary? How'Is this addressed in the emergency procedures? Response. As long as there is no physical blockage in all~the recirculation lines then there can be no " disconnect". The worse situation that can occur is that the reactor shroud and-the-core region will act as a manometer. The zero recirculation flow Indicated durIng the transient-does not necessarIly mean .there was no f low, but only that the f low was ai.the lower end of. Instrument sensitivity. This~ low recirculation flow is not unexpected during transients of this nature. Question-4. What do procedures tell operator. for appropriate action when recirculation flow
~~goes.to zero?~~
-a
~~Response~~
While the symptom of no recirculation flow is not specifically addressed by emergency procedures, there is an emergency procedure which addresses the closure of~the. recirculation loop valves. This is part of the scram procedure ~(Stetion Procedure 532) and warns against closing off all five (5) loops. Question 5. Whatz does the operator do if he gets a triple low alarm? Is this addressed in -the emergency procedures?
~~Response~~
~~The-procedure for_.a triple-low level alarm requires that the operator do the following:~~
~~a.. confirm with other level indicators, b.~~
check that at least two (2) recirculation loops have both the suction and discharge valves are open, c. verify core spray pump operating, d. verify ADS actuation if other required signals are present. Other plant emergency procedures that address pipe breaks also include specific actions for the operator,if a triple low level alarm occurs. Question 6. What effect does the operation of the recirculation pumps have on level measurement?
~~Response~~
The Oyster Creek Station has several diverse water level measurement devices. All the reactor vessel water level measurement devices utilized at the Oyster Creek f acility are delta-P devices and some take measurements in the vessel annular region while others measure inside the core shroud. Since these are delta-P devices, the operation of recirculation pumps will af fect the water level measurement especially those measurments inside the shroud. Our letter dated October 30, 1980 partially describes the differences between the inside the shroud indications versus the annulus' indications. It should be noted that the mixture level inside the shroud during normal power operations (power > 10%) is always up_to the top of the moisture separators. It is the separators which actually separate the water from the steam mixture. Question 7. Indi_cate whether your procedures instruct the operator to keep RBCCW available when it is beneficial for transient mitigation, if it automatically isolates during an event.
~~Response~~
Yes,the plant emergency procedures do address the steps to be taken to keep the RBCCW available for those transients for which RBCCW~Is beneficial for transient
mitigation. This is identified in Station Procedure 507.1, (Reactor Building Closed Cooling Water System failures) Sections 4.4.1 and 4.4.2. I 1 .}}

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