ML20085B429

Rept on 790502 Transient at Oyster Creek Nuclear Generating Station

ML20085B429
Person / Time
Site:	Oyster Creek, Crane
Issue date:	05/12/1979
From:	JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML20085B408	List: ML20085B429
References
TASK-*, TASK-GB NUDOCS 8307080135
Download: ML20085B429 (171)

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{{#Wiki_filter:- ,' 9 /- - s 4 I l I REFOR CN THE hay 2,1979 TRANSIENT AT THE OYSTER CREEK NUCLEAR GENERATING ST.2' ION l i gy JERSET CENTRAL PC~7ER & LIGHT COMPANT i DATED: MAY 12, 1979 83070BP135 790512 PDR ADOCK 05000289 S PDR a .n-.-- mm r r, ._. - -., -,~-- -.-

s y REPORT ON THE MAY 2,1979 TRANSIENT AT THE OYSTER CREEK NUCLEAR GENERATING STATION TABLE OF CONTENTS Page No. 1. Description of the Transient and Sequence of Events 1-1 2. Analys a of Core Water Level and Core Integrity 2-1 A. Reactor Coolant and off Gas Analysis 2.A-1 B. Water Level Analysis 2.B-1 3 Actions to Prevent Rwecurrence 3-1 l A. Procedural Changes 3.A-1 l B. Operator Training 3.B-1 C. Surveillance Changes 3.C-1 D. Physical Changes 3.D-1 4. Supplemental Information 4-1 A. Test Program for Startup and Power Ascension 4.A-1 B. Supplement to Response to Bulletin No. 79-08 4.B-1 C. Event Recorder Operation 4.C-1 D. Isolation Condenser Operation - ~ - - - - -- - - ~4.D - E. Level Instrumentation 4.E-1 F. Future Actions Under Consideration 4.F-1 Appendix 1 - General Electric Analysis e e e 1 i l l ,er + ~+% ,_,.,,.,._r ___,.m.,, _,-w,.. ,,--w e--,-...--%.,,r.,y e -r-..%,. _- r ,.-~w.-.-

i 't J DESCRIPTION OF TRANSIENT AND SEQUENCE OF EVENTS RELATED TO SCRAM OF MAY 2,1979. AT OYSTER CREEK NUCLEAR GEE 3ATING STATION INITIATING EVENT: On May 2,1979, at 1350 hours, an inadvertant reactor high pressure scram occurred during required surveillance testing on the isolation condenser high pressure initiation switches. Two (2) sensors (RE-03A System I and RE-03B System II) (see Figure 1) of the four reactor high pressure scram sensors share a common sensing line with the isolation condenser high pressure initiation switches being tested. The technician pe'rforming the test was in the process of verifying that the sensinh line excess flow check valve V-130-1 was open when the scram occurred. The scram has been attributed to a momentary simultaneous operation of switches . RE-03A and RE-038 due to a~ hydraulic disturbance associated witn valve manipulations required by procedure. to verify the position of the excess flow check valve.. The hydraulic dishrbance also caused a me:::antary trip of the isolation condenser initiation switches (RE15A and RE158). These sensors were not closed long enough to ir'itiate an automatic initiation of the isolation condensers, since a time delay is involved in the initiation logic. Hawever, these sensors also are used in che automatic l recirculation pump trip logic which did operate in tripping the four operating recirculatine pumps. No automatic time delay is involved in this logic. INITIAL CONDITIONS: Plant Paramet3rs at the Time of the Scram: Reactor Power-1895 MWt Reactor Water Level 79" Yarway (13'-4" Abeve the top of the active fuel) l (See Figure 2 for water levelreferencetabulation) 6.4' GEMAC e e e - G Te ,e v,-

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,-t 4 Page 1-2 ~ P.eactor Pressure 1020 psig Feedwater F"4cw 7.1 x 106 lbm/hr Reciretr1ation Flow 14.8 x 104 gpm Ecuiument Out Of Service: Relevent to Event Secuence: A. One of the two (2) startup transformers, SB(Bank 6), was out of service as permitted by Technical Specifications, to perform a'i inspection of its f associated 4160 Volt cabling. 58 supplies offsite power to one half of the station electrical distribution system (see Rgure 3) when power is not available through the' station auxiliary transformer. The 4160 Volt buses which receive power from SB are 18 and 1D. Bus 10 supplies power , to certain redundant saf3ty s'ystems. Bus 1D is designed to be powered from #2 Diesel Generator in the event power is ret available frem either - the avtiliary transformer or startup transformer. Bus 18 supplies 4160 Volt power to non-safety related systems and henes, does not have a diesel backup power-source. ~ B. One of the five (5) recirculation loo p (D) was not in service due to a faulty seal cooler cooling coil. The pump suction valve was open, the discharge valve was closed, and the discharge valve bypass valve was open. No other systems and/or components important to the event sequence were out of service. b EVENT SE0VENCE: (To=1350). TIME OF EVENT (Sec) EVENT DESCRIPTION + 0 A reactor scram occurred for the reason previously l described coupled with a simultaneous aut:xr.atic trip l of the four cperating Recirculation Pumps. The Control l L.

l g 1 o j. TIME OF EVENT (cont) DENT DESCRIPTION (cont) Room operator verified that all cm:ol rods inserted and proceeded to drive-in the IRM and SRM Huclear Instrumentation. At this time, 4160 Volt power was being supplied from the auxiliary transfomer during the coastdown of tM Turbine Generating System and the Feedwater System was in operation. Recirculatim fkw started decreasing due to pump coastdown. Steam flow started decreasing due to loss of heat production (scram) but feed flow remained at the full power flow ~ rate. Reactor vessel pressure decreued to tne pressure i regulator setpoint as stesa flow decreased. Reacter water level began decreasing due to steam void collapse in the core. 13 The Turbine Generator tripped at the no load trip point which initiates an automatic transfer of power to the startup transformers. Power to Bus 1A and 1C successfully transferred from the auxiliary trans-fonner to the SA (Bank 5) s'tartup transformer. Since SB (Bank 6) was out of service at this time, pcwer was lost to Buses 18 and 1D. As, designed, Buses 13 and 10 separated through operation of breaker 1D and a l fast start of Diesel Generator No. 2 occurred to power emergency loads on Bus 1 D. O e e b w, - -.*-,.u ,,..v.e.-,__.--.,--,-_.-,,--+,.m-,m,__--,-,. -...,,-..v-, ,.----m.rr,, ,,,s,,

ef 1 l m ~ IIME OF EVENT (cont) EVENT DESCRIPTION (cent) Loss of pav e to Bus 13 resulted in loss of Feedwater Pumps B and C and Condensate Pumps B and C. Althugh power was available to the A condensate and feedwater pumps, via Bus 1A, the A Feedwater pump tripped en 1cw suction pressure. Since water inventory was leaving the' Rea; tor Vessel thromah the Stoam Bypass Valves to the Main Condensers and a high capacity source of high pressure makeup water was not available, reactor water level.andpressuredecreasef. In addition, the loss of power'to Bus 1B caused the ~ B Cleanup System Recirculation Pump to trip which, in tur.n, caused an isciation of the Cleanup System due to 4 .g low flow through the cleanup filter. Furthermore, one ' condensate transfer pump and-the operating fuel-pool cooling pgap tripped. An unsuccessful attempt was made to restart the A feedwater pump. (Thereasons for the restart failure are described later.) (bent Recorder) 13.6 Reactor water level decreased to the Low level scram setpoint which is 11'5" above. the top of the active s fuel region. (EventRecorder) s 16.8 The output breaker on the No. 2 Reactar Protection i System M.G. Set tripped due to loss of power to the, i drive motor. The output voltage from the M.G. Set had i e e a O w. -..-. f --.=.-...-.-.--. -.-...i- -

'A 4.-b t s. l l ~- TIME OF EVENT (cont) EVENT DESCRIpT!ON (cont) hen maintained by flywhesl action since the time of the turbing trip. Power to the M.G. Set drive motor is fed indirectly through Sus.1D which was deenergized at this time. e 31 The No. 2 Diesel Generator Brerhar closed and supplied power to the 1D Bus. A sicond control rod drive pump started. ~ 43 Atactor water inventory centinued to cecrease due to steam flow to the main condenser. In anticipation of a low Low Reactor Wat' r Leve1' automatic isolation of e the reactor (which occurs at 7'2" above the top of the active fuel region), a manual reactor isolation was initiated to conserve inventory by closing the Main Steam Isolation Valves. 's i This action was taken at an indicated water level of approximately 30" on the Ya'rway instrument which corresponds to 9'8" above the top of the active fuel region. It should be noted that the decrease in indicated water level and pressure was amplified by the effects of introducing cold feedwater into the vessel during the 13 second peric( prior to the Turbine Generator Trip. 'The cold feedwater reduced the steam voiding inside the vessel thereby causing a shrink in water level. r --~ es---, .y--w.-, ,,~,.-,----,..,,,-m.--_,,,-__,..,--..e.,-.--- .-m -..m--, ,,--,y -.-,.,,,..-,..---,_-.,-w ,r

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[ i l l. l TIMEOFEVENT_(cont) EVENT DESCRIPTION (cont) 49 The Main Steac Isolatier. Valves fully closed, thus stopping the lors of water inventory from the vessel thereby causing in increase in reactor steam presrure. ~ Indicated reactor water level started to increase shutly after isolation, when reac' tor decay heat re-established a steam void distribution. (EventRecorder) 59.6 The reactor mode switch was transferred from RUN to ~ Rerua. j s 76(1 min.16sec.) To establish a sink for the removal of decay heat frem 1 the reactor, the B isolation condenser was placed into service. At this time, the Control Room operator closed the A and E recirculation loop discharge valves i (these valves take approximately two (2) minutes to close). It is postulated that at this time, both B ( and C locp discharge valves were also closed. The conclusion that the five recirculation pump discharge valves were closed is based upon loop temperature l respenst later'in the event and is further supported 't Y s Lcw Low Low level at 172 seconds. The D loop & raolated previously. (See the equipment out of servicesection). (EventRecorder) 90(1 min.30 sac.) The reactor Low water level alarm cleared due to the water added from the isolation condenser to the e Primary Systam. + a h ---,-,nv----e a v----m--w -- --- w

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o? Page I-7 TIME OF EVENT (cont) EVENT DESCRIFTION (cont) 96 (1 min. 36 sec.) The B isolation condenser initiation valve fully opened after 20 seconds. The temperature of the E recirculation loop, which serves as the B isolation condenser water return path, decreased due to the effects of cold water from the isolation condenser. The D recirculatbn loop temperature did not change 1 appreciably. A, B, and C recirculation loop temperatures increased slightly. The heat-up is attributed to s natural circulation through the partially open discharge. valves carrying hot water,(536'F) warming the lines previously. cooled by the effects of cold feedwater. The reduced flow area between the lower downcomer and j 1ower plenum area, due to the slow closurs of the discharge valves, started to cause a shift in water inventory from ths. core area _to the upper and lower downcomer region. The shift was.due to the isolation (; condenserreturningcondensedsteamfromt5ecorearea to the downctmars. The water inventory shift c:ntinued as the discharge valves moved to the full closed position. (Event Recorder). l 172(2 min.52sec.) The reactor Low Lcw Low water level instrument trip Last recorded point on the event recorder. point was reached. This was probably caused by the voided mixture in the separators having drained to the . upper plenum, causing a reduction of static head above the Low Low Low water level instrument. This does not e 5 1. , - -.-~ -. ~. - -, - _.,.. _ - -. -,. - -,.,, _ - _ -, -. _ _. - - -, - - -. ~. - - -., - - -,..

j TIME OF EVENT (cont)- EVENT DESCRIPTION (cont) necessarily indicate an inventory loss from the core but rather a redistribution of water and steam voids above the core. 186sec.(3 min 6see) All recirculation loop discharge valves fully closed. ' At this time, based upon closure initiation, the cooldown of the E recirculation loop stopped and a heat-up began. The indicated reactor water level increased due to the shift in water inventory. Recirculation loops A, B, and C continued to heat up. The mechanism of the heat up wet due to heat transfer between the hot recirculation loop piping and the water in the piping. Reactor pressure continued to decrease as a result of

i. solation condenser operation.

R 250(4 min 10sec) B isolation condenser was removed from service to reduce the rate of co 1down of the Primary System. Removal of the condenser ciused indicated water level to decrease. The decrease in indicated water ievel was due to a return of water to the' core region from the downcomer region through the five (5), two, inch (2") bypa'ss valves around the recirculation loop discharge valves. During this period, the net water inventory effect was a storage of water in the recently secured i isolation condenser. The recirculation loop discharge \\ temperatures reached equilibrium and followed a slow cooldown trend. l i l -w- ..,-.g, ,m. ,,,o _,,,,__,_.-,-w-mm_, __mmy,--.,-.v ,~.-,,v--r, .--.e--

t s TIME OF EVENT (cont) EVENT DESCRIPTION (cont) 270(4 min 30sec) The reactor pressure increased due,to the effects of removing B isolation condenser. 'Ihe rate of dictef,se in water level shifted from a ra::tp of approxi' mat'ly 37 e in/ min to 2 in/ min. The reaso for this~ change is the isolation condenser tube assembly wass:ompletely filled. The flow through the five (5) 2" bypass valves contir.ued, accounting for the change in slope. 450(7 min 30sec) Both isolation condensers were placed in service. This caused an incre.se in indicated water. level and a decrease in pressure. The A recirculation loop tempera-ture decreased because cold water from the A isolation condenser entered the A recirculation loop which is its return path to the reactor. A portion of the water passed through the loop via its 2" bypass valve, thus causing the cooldown. ( 528 (8 min 48 sec) To slow the rate of cooldown, the B isolation condenser was removed from service. At this time, the indicated water level reached a maximum of apptry.imately 14.4 feet above the top of the active fuel (88" en Yarway). This is considered to be above normal water level for full power operation. When the B isolation condenser was removed from service, indicated water level decreased to 13'8" above the top of the activa fuel where it remained until approximately 1212 seconds when A -esy v ,--w e,-,,---w .-v4--. - e we-e,------%---*ry-, -..w---- ,_--we -.e-e-- .---+-

i f., n. .-.o TIME OF EVENT (cont) EVENT DESCRIPTION (cont) isolation condenser was removed frem service. The reactor pressure continued to cecrease and all recir-culation loop tempert'ures continued to trend d:unward. Indicated water level was stable at this time because the. head of water in the downcomer' region was sufficient to establish equilibrium between the water entering the core region via the 5 two inch bypass valves and the condensed steam returning to the downcomer from the isolation condensers. 540 (approx) (g min) The four (4) Low Low Low water level indicators were verified locally to be below their alarm setpoint which is 10". The reading apeared to be at or below che instrument's lower level of detection. 810(approx)(13 min A recheck of the triple how water level indicators 30sec) showed that the pointers were active (movin' g) although they continued to read below their alarm point. The instrument was at or slightly above its lower level of detection. 1212 (20 min 12 sec) A isolation condenser was removed fra service, thus stopping the removal of inventory from the core region. Indicated water level decreased as the water in the downcomer region flowed into the core region. Reactor - pressure started to increase due to the decay heat steam production. l l 1 I -~--,,,e ,--n--,- --ne,n..w..,ea. ,.-.+__..- - n-ac_n-,,,,n,. - - - -. -, - - - -,. - -,,,.. a.

a t. w Page 1-11 TIMEOFEVENT(cont) EVENT DESCRIPTION (cont) 1488 (24 min 48,sec) The isolation condensers were used several more times to control the reactor cooldown with pre-dictable increases in indicated water level and reduction in pressure. This mode of operation continued until 1914 seconds. 1314 (31 min 54 sec) In order to more correctly determine the plant cooldown rate C recirculation pump was started and the discharge valve was opened. It was noted that the indicated water level dropped approximately 3 feet in less than 2 minutes. ( The C recirculation pump was shutdown and isolated to investigate the reason for the drop in level. In response to the indicated water level drop, additional attempt was made to start the A feedwater pt.ap. The pump failed to i start due to a tripped overload on the auxiliary I k oil pump which is interlocked in the pump { starting sequence. The indicated water level started to increase due to the' action of the operating iso 1' tion condenser transferring water a to the downcomer region. When the C recircula-tion loop eas started the loop temperature in-I k creased from approximately 400*F to 470'F. The I 1 other recirculation loop temperatures continued to trend down. At this time Low Low Low alarm may have cleared..

9 m 'TIMEOFEVENT(cont) EVENT DE,5CRIPTION (cent) 2208 (36 min 48 sec) The A Feedwater pump was successfully started by locally starting'the auxiliary oil pump which satisfied the required starting interlocks. Indicated watar levei increased to a le' val corresponding tc 13'8" above the top of the active fuel region. Realization occurred that the indicated water level and core water . level may not have been the same when it was recognized that the five recirculation loop discharge valves were closed. 23G (39 min 0 see) Thn A recirculation pump was placed in service at a flow rate of approximately 1.9 x 104 gpm, thus removing the disparity between water level measuring systems. .( A The Low Low Low water level alarms were known to be c.leared at this. time. Indicated water level dropped approximately three feet to 11'4" above the top of the (, active fuel. The A recirculation loop temperature rose from 375'F to 465'F when it was placed in service. . Steps were in!tiated at this time to bring the plant to " cold shutdown condition". 2700(45 min 0see) Reactor Protection System #2 restored and scram reset. 3600 (1 hr.) .The SB transformer was returned to service and Buss 1B was energized. I ._,_____,m_

i Page 1-13 REACTOR PARAMETERS: Figures 1-4a and 1-4b are a trace of reactor pressure, saturation terr +erature, annulus water leyal, recirculation flow, and recirculation loop temperatures from the time of the trip to 45 minutes later, when the tran-sient was over. (hey are annotated with significant events dur,ing the period. 1 t 0 m I 1

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~ MSCRIPTION OF THE OYSTER CREEK PLANT The Oyster Creek Plant is a General Electric, 5 loop, forced recirculation,1930 MWr, Boiling Water Reactor (BWR) with a Mark I con-tainment system. The steam supply system consists of main steam piping, feedwater piping, and recirulcation pumps and piping. The system is also etuipped with a cooling systas consisting of circulation piping and con-densers to provide for heat removal via natural circulattan through the reactor. Various instrumentation and control systems are provided to moni-tor system performance and control operations. Figure 5 presents a sim-plified diagram of the above piping systems and their interconnections. The containment sy:; tem consists of a containment vessel (drywell) around the reactor vessel and recirculating system attached to a suppres-sion chamber (torus). Steam released to the drywell is vented to the torus where the steam is condensed by the torus water which can be cooled by heat exchangers. The main steam piping inside the drywell is equipped with 5 relief valves which can be cperated either automatically or manually to relieve excess. pressure or depressurize the system. 2ach of the 2 steam lines is also equipped with 2 isolation valves (1 inside and.1 outside of the con-tairment vessel) to isolate the pressure vessel either automatically or manually as needed. The 5 relief valves operate automatically on hi'gh pressure to blowdown to the torus where the steam is condensed. These valves also actuate automatically when high drywell pressure reactor low-low-low water level, and core spray boc>:er pump discharge pres-sure exists simultaneously for a period of 120 seconds or less. This is to 1-14 O e 9 ,,,o-,

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6 depressurize the system to allow for core spray flow into the pressure t vessel. The main steam isoletion valves are closed automatically on de-taction af any one of the following signals: (1) main steam line high radi'ation, (2) high steam flow in the main steam lines, (2) high tempera-ture along the main steam liner, (4) main steam ifne low pressure, or (5) low-low reacto; water level. Rose valves may also be closed manually by the operator. The feedwater piping delivers feedwater through 2 check valves (1 inside and 1 outside of the containment vessel) and a locally operated ( stop valve, inside containment, to the feedwater sparger within the an-nular region (downcomer) af the reactor. Als water mixes with the re-circulation water in this region and is then delivered to the core through the recirculation loops. ( The recirculation pumps take a suction from the annular region of the pT. essure vessel,between the vessel wall and the core shroud,through a normally open suction valve and discharge water through a discharge valve equipped with a 2" bypass valve into the bottom of the pressure i vessel. The rr.ted flow capacity of the combined recirculation loops is 61 x 106 lb/hr. Each recirculation loop is 26" diameter piping contain-ing motor operated suction and discht.rge valves (equipped with 2" ascor operated bypass valve) and a variable speed recirculation pump. D ere are 5 such recirculation loops and all suction, discharge 'And bypass valves are normally open during operation. RecirculationloopskandEhavea 10" connection on the suction. side of the recirculation pump upstream of the isolation valve. These connections are the return lines from the 2 isolatun condensers. 1-15 e%%.-- -w-s y-es, ,---g--, ,, y m m w %_y,,,,,_,_,_-.4gy.ygww,m, .__w-,w,w ww mw-,--,--us-mm---__,- - - -_-a ---+-a-- w-

i l l The isolation condensers are connected to the reactor vessel steam region and the suction side of reciteulation loops A and E providing l a loop for natural circulation through the reactor core. The isolati.en con-denser piping is 10" diameter piping with 2 isolation valves in the con-denser inlet piping and 2 isolation valves in the condenser outlet piping. l All valves are motor operated and normally open with the execption of the 1 outside containment valve (DC motor operated valve) on the outlet piping which is normally closed. This system receives steam from the reactor vessel which is condensed withiin the tubes by surrounding water on the shell side and returns the condensed water to the recirculation loop. The heat transferred to the water on the shell side causes it to boil. ' Die re - sulting steam is vented to atmosphere. The driving force for this system i is natural circulation due to the heating of water in the core region. This system is actuated automatically on detection of a reactor high pres-l t. I sure or low-low water level after a maximum of 15 seconds _ time.. delay._The system may also be actuated manually by the operator. Steam from the reactor drivss the main turbine / generator, is then condensed and returned to the reactor via three 1/3. capacity condenstte - pumps, three 1/3 capacity feed' pumps, and the feedwater piping. The 3 condensate pumps discharge to a common headez feeding various heaters and coolers. Discharge water from the 3 intermediate pressure heaters feed the cuction side of the 3 feedwater pumps which discharge t}. rough the 3 high pressure heaters into a common header feeding two 18" lines which run to a tee inside containment. Each of these lines then feed two 10" feedwater lines to the nactor. 4 1-16 i t -ee - wr w2,,-,- - i,, . s e-m

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4 The condensate, feedwater, and recircuistion pumps are powered l from the station non-vital 4160 volt: buses 1A and 13 which' during operation l l receive power from the auxiliary transformers. Startup transformers SA and SB pcVide power to buses lA and IB during plant shutdown. Condensate pump 1A; recirculation pumps. A, C, and E; feed pump 1A; and' cleanup re-circulation pump A teceive power from bus 1A while ccadensate pumps 1B and IC; recirculation pumps B and D; food pumps 1B and IC; and cleanup recirculation pump B receive power from bus 13. Startup transformer SB provides power to bus 1A and startup transformer SB supplies power to bus 1B. Figure 3 po>ents a scliematic diagram of this distribution system as well as the emergency power dist-ibution. In order to' monitor system perfor=ance, instrumentation is pro-vided to monitor reactor water level, reactor pressure, valve position, recirculation fic,w rate, and other system paiameters. Reactor water level _is_Donitored' by three level measuring devices; "GEA!ACS", "Yarways, and "Bartons". Two reference legs outside_1;he, vessel are provided for level indication and protective functions. Eight "Yarway" differential pressure cells, four "Barton" differential pressure. cells, and three "GEhtM:" differential pressure electronic transmitters provide for level indication and protective functions. Reactor level is indicat'ed both locally at instrument racks and remotely 'in the control room. Level indication in the control room is provided by both the "Yarway" and "GEAIAC" i.struments. While the "Bartort'instnaent provides level indication at the instrument rack, it provides only an alarm function in the control roca. All level indicator variable legs sense level in the annular region of the pressure vessel except for the 4 "Barton" Low-Low-Low -s' level indicating switches. These switches indicate level inside the shroud 1-17 .f ww-w.a-- - - -. - -,. y*.r9 i,.-.- --.my. y yw-

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l e i l l l [ above the core. Local indication of this level is provided on an instru=ent rack with a remote slann function in t.ae contrel room. Figure 6 pre-sents a diagram of the level indicators and associated alarm setpoints. Valve position indication is provided in the control room for the motor operated valves mentioned in the above discussion of systems. At tihe time of the Oyster Creek event all systems were in the normal line up with the exception cf the startup transformer SB and re-circulation loop D. Startup transformer SB was removed from service for maintenance. Recirculation pump D had been removed from the system due to a seal leak; therefore, the discharge valve was closed, suction valve open, discharge bypass v.ive open, and a plate was installed over the loop opening. 0 n- .~ ~. e e e l 1-18 e e- -", w .,,.,-n,__ - - ~ -,---,e.,-,, ,,,---,,-w.,--

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l: cuTSsDE I INSIDE I FIGURE 5 9tSIDE OUTSIDE DAYWELL l DRYWELL i garwett l oRywgLL l f. ,I VENT TO I 405/REuEr l ATMOSPHERE VALVES (S) i l I l ~ TO l l l l TORUS L woSoM i iw s l J y,,y MSIV I '?Jfofg e g L r J y<,c D< A h SSE Q ?) ^ i k l ISOLATloH CONDENSER (TWO) l CORE 1 I g I l l. \\ fr

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-a,. m 2.A REAC1DR WATER AND STACI GAS ANALYSES A reactor water sample taken at 0820 on 5/2/79 showed all parameters to be within normal ranges. The reactor scramed at 1350 on 5/2/79. A reactor water sample was taken at 1520 from the cleanup system inlet line after flow had been re-established through B loop. At that time "B" recire. pump was off. The con-ductivity had risen from 0.10 umho/cm in the morning sample to 0.J7 unho/cm which is normal after a scram. The gamma spectrum analysis showed the fission product activity concentratio.'. to be in the normal expected ranges for the 6 condition. Iodine 131 was up approximately a facter of 2 which is expected due to depressurization. A reactor water susple was taken at 1640 from the cleanup inlet. "B" recire, pump had been turned on prior to this sample. Comparison of the re-- suits with those of the previous sample showed good agreement. Conductivity had dropped to 0.30 umho/ca. Gamma spectrum analysis showed normal isotopic decay and the effect of the cleanup domineralizar remova'1. Reactor water samples were taken at 2105 frem both the cleanup inlet and "A" recire. loop. The samples showed good agreement on the ytra-meters checked. The conductivity in the C.U. inlet sample was down to 0.18 unhos/ca. Radioactivity levels continued to drop'at normal rates. A reactor water sample taken at 0641, on 5/3/79 showed all parameters within normal expected ranges for shutdown. - Four re ctor water samples were taken on 5/3/79 at 0641, 0840,1547, and 2000 hours. Isotopic analysis showed stable conditions in the water. When the sample'taken on 5/4/79 0 0800 continued the stable trend, the reactor water, analysis was returned to the normal once per day frequency. 2.A-1 lI We~e e si-+te e --ew e.w e,,.--+-v+ g-- - -w-+.e.- eic+-ee----ew--w-9w e -e e 5e%-**w----e- - ' -**-Nm---ww=rvm- - = ---h-=e= -w-- s--

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0 l s t t l l Additional backup analysis of the reactor water radioactivity concentration is bcing maintained with a continuous on-line gamma spectrum analysis. This multichannel analyter has been in experimental operation as a part of an EPRI project at the Oyster Creek Site, since December 1978. After the plant shutdown, a continuous on-line analysis was started at 1611 hours and integrated,over an eight hour period. The results of this analysis com-pare favorably with that of our grab samples taken during the period, and do not indicate any abnormal fission product levels in the water. Data on a number of. parameters measured in the primary coolant before and after shutdown are given in Tables 1 and 2. 'Similar data for three other scrams during the current cycle are presented in Tables 3-6 for comparison purposes. These tables also include dat: :aken during subsequent startups so that expected levels during those periods are seen. Stack Analysis ,The stack particulate and charcoal filters were removed and analyted on 5/3/79. The results were compa:ed to the filters removed on 5/1/79 and showed no unusual or abnormal releases as the result of the plant shutdown. Data is presented in the attached Tables 7-10, where compari:en is made also with previous scraas during this cycle. Data for previous scrams also includes stack reise.ses during subsequent startups. e e e 2,A-2 l l l l l t , ~, - --,,- - -- -,.-- __._ _.______.-....- - -- - - -, - - ----- ~ _,..,-,.~.. ~. -. _, - -... _ -,..,,.. -.... -., ~. - -, -

a. i me,stxt nnar.at newuadrs / ~ Rx Scram at;,,1350 my 2, l' 5 T.SIM.E 1 ' ~ (- 0020 5-2-7 5-2-79 5-2-7!8

b579 5-3-7.9 On I.ine 5-2-79 l

0641 Analysis': c% 1640 2105 1 Irc Ihmt D A A mte*S-2

• n Off I

on niairt 161; pCl/ml I-131 7.47 E-3 1.53 E-2 1.34 E-2 2.61 E-3 4.07 E-3 1.36 E-3 h t. cletcc I-132 1.33 E-1 1.74 E-1 1.79 E-1 1.43 E-1 1.37 E-1 1.11 E-1 1.34 E-1 I-133 7.35 E-2 5.18 E-2 5.63 E-2 1.00 E-2 7.97 E-3 1.09 E-3 2.27 E-2 ~ 1-134 6.50 E-1 4.00 E-1 1.41 E-1 6.69 E-3 1.02 E-2 <l.0 E-4 I-135 1.64 E-1 1.10 E-1 6.95 E-2_ 2.10 E-2 9'.27 E-3 W1. 0 E-3 3.11 P-1 Xo-133. <l.34 E-3 . 5.19 E-2 1.29 E-2 1.15 E-2 H.40 E-3 2.19 E-2 Xe-135 2.35 E-2 ' 3.79 E-2 1.33 E-2 1.49 E-2 4.75 E-3 2.49 E-2 5.22 E-2 Sr-91 2.00 E-2 2.40 E-1 7.0L E-2 1.67 E-2 1.,19 E-2 1.50 E-3 3.10 E-1 Sr-92 1.00 E-1 4.05 E-1 1.02 E-1 H.79 E-3 5.34 E-3 <3.0 E-4 4.60 E-1 i Ib-99 1.69 E-2 4.20 E-2 1.97 E-2 <1.00 E-2 1.56 E-2 <3.0 E-3 7.83 E-3 I Tc-99m 8.91 E-2 7.00 E-2 5.56 E-2 4.11 E-2 4.64 E-2 3.66 E-2 __ H.30 E-3 F 'Ibtal Iodine 1.02 E O 8.97 E-1 4.59 E-1 1.04 E-1 1.69 E-1 1.13 E-1 4.60 E-1 l l Y Np-239 1.35 E-2 1.47 E-1 9.71 E-2 1.40 E-1 1.45 E-1 6.50 E-2 1.36 E-2 l l Gross 3 8.16 E-1 1.07 E-1 ~ 1 Gross a 1.51 E-6 2.39 E-5 pli 6.13 5.50 5.97 6.14 Conductivity 0.10 0.37 0.30 0.18 0.25 0.15 Sus. Solids 220 Mb l C1- <20 gyb

• Continuous on line spectnsa analys.s intergral d over 8 Iv.ur collecti on period : tarting at 1611 on 5-2 79 5

D a AN.

i ~~ LEE 2 / nEwan m m mnases 4 5-3-79 5-3-79 5-3-79 5-4-79 5-5-79 5-6-79 5-7-79 Parameter 0840 1547 2000 0845 0812 0752 0025 I 131 1.05 E-3 .l.69 E-3 1.23 E-3 1.08 E-3 4.72 E-4 A.75 E-4 Not Det. 132 1.12 E-1 1,.'[0E-1 9.71 E-2 8.51 E-2 7.33 E-2 6.01 E-2 4.53 E-2 l 133 ND m l ~ 134 ND ND 1 135 m ~ Xa'133. ~ m 2.04 E-2 1.69 E-2 1.45 E-2 8.99 E-3 5.68 E-3 4.84 E-3 3.07 E-3 I Xe 135 2.33 E-2 1.43 E-2 1.10 E-2 2.Si E-3 3.47 E-4 c3.3 E-4 <3 E-4 Tc 99m 4.53 E-2 4.70 E-2 4.49 E-2 3.87 E-2 2.98 E-2 2.24 E-2 1,68 E-2 Da 140 6.38 E-3 <2.57 E-3 <2.56 E-3 3.72 E-2 4.11 E-3 2.36 E-3 <2.3 E-3 IA 140 2.34 E-4 <2.14 E-4 (2.06 E-4_ 2.19 E-3 5.42 E-d 4.22 E-4 c2.0 E-4 i l Np 239 5.58 E-2 3.53 E-2 3.16 E-2 1.97 E-2 9.25 E ~4 5.06 E-) 3.42 E-3 j i Gross Beta 1.59 E-1 1.43 E-1 1.05 E-1 6.84 E-2 6.34 E-2 4.2fi, E-2 j F Gross Alpha <l.0 E-6 .l.27 E-5 8.11 E-6 <l.0 E-6

1.0 E-6

<l.0 E-6 i l L pil 5.9 t 6.4 6.05

6.1 5.8 i

l Conductivity 0.13 h 0.42 _0.41 2 0.37 0.37 l I i I f j g I ! l h m

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1 2

9 5 6 1 3 0 1 5 9 1 6 1 7 1 5 7 0 -2 4 7 3 0 9 7 2 6 9 5 9 0 9 7 1 2 2_ ns 0 1 1 6 4 1 1 5 2 4 7 1 l 2_ 0 9 3 5 6 0? 2 5 ( 9 1 4 3 8 4 2 1 4 3 3 1 4 E-2 2 2 2 4 e 7 E _ E-E E E-E E eR E E E E E 0 05 7 4_ 5 7 31 2 7 7 7 0 6 1 1 0 0 6 1 4 7 5 1 0 n_ n,7 1 6 5 31 D. 1_ R. 3 1 n 0 3 82 4 2< L. i_ 3 ? 2 u 4 1 9 4 1 n 4 5 l 2 a, 7 3 f_ 0 e. L1 g i 2 0 1 3 1 2 1 ? 1 1 1 5 t 5 7 1 2 7 E-E-E-E-E- E E E E-M1 3 1 E E-E-E E R 04 Ce 31 E 1 9 9 1 1 9 5 0 5 ,S c 4 9 5 3 n_ 0 5 n_ 1 0 9 6 0 2 n 4 1 0 02 2 n_ 6 1 ? 3 7 1 i l i 2 31 1 2 2 4 1_6_0 E-5 1 2 <1 c 0 E I D1 [ .I 0 2 2 2 1 2 1 1 2 2 1 2 2 I - 2 7 E-E-E E_ R E-E_E_ E-E- E 9. '1 E E 04 9L7 n' 31 5 1 0 n0 7 4 9 1 q ? PAhm 02 1 9 5 5_ 4_ 7 2_ 9 7 Tt l 1 2 4 1 7 1 6 i 1, 1 1 3 , NS 1 _1 har ?_ 2_ 2 l 0 3 2 2 i 1 1 1 2 2 0 1 7 E-E-E-E-E E-E E-E E-E-E- 1:- i E E 11 03 n _ 9 21 /6 2 7_ 8. . 1 4 4 l 1 3 i 2 9 0 1 11 9 3 5 4 i6 6 82 C n 6_ 4 J1 e. l D 9_ 2 0 n L_ L_ _ 1 ? 1 u_ 6 7 7 7 1 6 2_ 5 1 1 7 0 H G_ D_ 2_ 1 1 5 1 NER O e e v n F s L H d v i d i i I o 1 l E I A a

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2 c

7 2 L_5 51 p 4 5 6 4 1 0 1 3 1 1 2 7 2 3 0 6 I e y n t s d i i R lx v l E c i i I I a t n _a T m E 3 5 9 cs 1 2 3 4 5 3 3 1 2 9 l 3 gs u W 3 3 3 3 3 1 1 9 9 9 a 2 as l o y 3 - 1 1 1 1 1 mr w t N e e r r s b 4 l o ul r c G ip C SC P I I I 1 I X X S S M I' I n Y. e t l 1l ll l{lll

i i ,t d1 2 1 1 3 1 2 1 2 2 0 2. 1 9 E E EE E E E E E E E E I 7 f 2 '51 5 4 5 7 9 6 1 1 6 3 0 1 3 5 1 s 4 7 0 4 1 3 3 12 6 0 4 2 7 3 6 0 4 J1 ) n 2 M_ 6_. 0 1 1 7 'l 2 7 2_ 5 01 6 1 0 7 1 1 9 3 2 2 1 1 3 2 2 1 2 2 0 2 1 5 E-E E-E E-E- E E E E E E_E E E S 7 2 1 7 0 0 4 5 9 6 7 1 2 S 5 9 S 2 1 0 0 00 02 7 1 3 9 7 9 0 3 6 70 Q9 0 Ji0 2 2 i Ll f. 7 2 _ 0L 6 9 0 6 1 5 1 4 1 1 4 s 3 2 3 2 2 4 3 3 2 3 2 1 2 5 1 7 E E-E- E E-E- E E E E E E_E E E 9 69 4 5 0 0 5 4 2 3 1 4 5 5 9 2 L6 51 3 7 4 6 9 4 1, 7 2 3 9 5 7 S1 0 3 3 0 3 4 2 5l 1 9 9 9 2 G 4 5 2 8 3 2 2 1 2 4_ 0 3 n E 5 i ,I 5 g j_ 3 i" _ 3 1 6 i 1 m i M 3 4 3 1 s e 9 3 1 2 3 n E E E P. E 7 E E E E I I E E E E 9 00 4 4 4 E 2 5 A 3 5 5 1 7 . 01 0 4 7 7 0 0 2

3. 6 0

2 5 7 9 9 1 1 0 0 5 2 5 3 2 2 9l l 1 9 4 1 0 l 1 2 1 1 f. 0 2 1 1 1n 1 e 1 0 g 4 n T 3 3 4 3 4 0 W1 l-2 3 1 1 1 1 5 3 1 4 9 F E E E E _ E-E-E E E IC e 7 E E E E E S, 3 9 5 4 7 6_ 6 3 9 0 9 9 07 9 9 4 4 0 1 4 2 6 5 3 50 7 f. 9 6 n 2 0 0 . 7 5 01 6 0 i A 0 f 9l 2 1 9 i l l 4 1 11 N 0 2 22 40 5 5 2 s 3 1 1 e J 1 g I - R a - 1 A 3 1 2 l 2 3 3 2 2_ 3 1 1 l T 9 1 7 E E E-E- E E E E E E E-S E E E 5 .r 06 7 3 1 L 4 6 G 9 0 21 0 3 5 7 0 6 'r U 7 0 0 1 2 5 7 0 51 9 7 J1 J4 J4 01 1 m I l 2_ 5 0_ 1l 5 1 4 1 3 4 t 1 j 1 ? 3 7 6 R l 4 A N T S 2 1 1 1 1 3 22 1 2. t 1 n ? 0 3 E-E-E E E E 3 E-E-E-E- r-u 9 t 9 T 7 l t r 1 a s m 65 A 1 4 0 6 5 1 1 l 0 0 9 l a 7 5 a f 3 4 6 0 1 3 5 77 7 7 E n 0

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2 ? 9 6 1 0 6 1 2 2 3 1 1 7 1 3 0 1 5 0 ~ . 3 2 - 0 3 1 2 1 2 2 2 2 1 1 1 5 2 6 0 E-E E-E E-E E E E E E E-E E-E - 3 9 2 2 7 5 - 2 5 4 7 .M 9 57 5 0 6 1 9 7 73 2 6 9 9 3 0 1 1 2 0 A 7 0 t 9 42 3 1 2 2 6 2 0 1 5 1 6 4 1 4 1 6 0 1 4 t . h 7 E ( G 2, 3 1 2 1 1 0 1 6 6 E-E-E-E E-E E-1 t 2' P 9 e a t 7 2 u. n 3 - 5 7 4 0 4 9 1 2 s1 0 0 26 0 4 1 2 7 1 0 5 l .S A l A 0 2 E r 02 6 1 0 7 1 l 9 1 s 0 2 '?

S j

f Y ! :l L = 0_2 1 ~ A 3 1 2 1 1 3 2 2 1 2 1 N E-E- E-E E E-E E-E- E-E E E-E- A 9 t 7 n m U 6 - A 3 1 9 1 5 5 s t n_ . 0 O 45 i 5 3 0 1 0 3 0 6 2 t 0 0 1 4 3 D ^ 02 u 6 1 0 7

l. 1

} 3 1 2 1 1 2 D H s_p4 J l' l l0 1 C hE tI a y i n t s f i d i R d v i E o i l 1 T p a t o E 3 5 m 9 v S 1 2 3 4 5 3 3 1 2 9 l 3 s s N 3 1 3 3 3 1 1 G 9 9 a 2 s s h f 1 l' 1 1 1 r t M t o o n s r r-b 2 a%_ A e i t r r n u L i M G G n_ n S C M S M P I I I 1 1 X 1 f r. >J 's ,1 I ,\\ \\* (' , '4 l l! i llt q' !~

s. e n N M - mq SGAM: Sc:am U 78 9 1951 Sta: :p, 12-13-78 9 1521 Stan St=:p Pa'a=== Rates Stack Sample,, 45-101-78

• 9:30 - 12/M/78 I-D1 0.41 vCf/sec I-D3 1.25 I-U5 17.5 (.46 uCi/se,

TM Tech. Spec % 11.5% e

1. S-102-78 9:30 - 12/H /78 8:30 - 124 5/78 I-131 0.34 I-133 1.59 (.48 vCi/se I-D5

.22 htal TW:fi Spec % 11.94%

1. 5-103-78 8:30 - 12/15/78 8:17 - 12/19/78 I-131 0.096 I-U3 0.102 (.100 uci/

I-135 0.489 2tal Tes. Spc % 2.5% 95-104-78 5317 - 12/lS/78 8:10 - 12/21/78 I-131 0.43 I-133 0.94 (0.59 vCi/s I-135 2.43 htal 2h::h Spec 414.8% e e e mee e 1 e o a e e e e e TABLE 7 e 2.A-9 h-c s i i b a v

l l Scr e Jan. 15 9 1552 Sta=g Jan. 18 9 1848 11*1 Jan. 8, 1979 - Jan 12 - .23 vCi/see y 33 i If d$ vCi/sec (0.34 pCi/sce) l Sr-5 -79 I~ .72 vCi/sec j 2 *al Tech spec 8.58% i Jan. 12, 1979 - Jan, 16 .39 pCi/sec 3 .80 5-6-79 .50 (0.51 vCi/sec) ~ scrum 1-15-79 mtal 'Doch Spec 12.75% Jan. 16, 1979 - Jan. 19 .19 vCi/.sec .07 vCi/sec (0.19 vC1/se::) s-7-79 A .01 vCi/sec f on line 1-19-79 2*4 Tw-h Spec % 4.7% Jan. 19, 1979 - Jan. 22. El 1.04 1.33 (1.09 vCi/sec) S-8-79 1.41 = m'al Tech Spe:: 4 27.35% e O 4 0 G g TASt.E 8 2.A-10 e e ew-- o r ,w~n,,.-..,,.,, ,,_,.,,-n-p., g -- -~--

---

s n

-.e---,-,. --n+-

..? i Fel=:c.a.y 1979 Sc::3mStackI$ta Sc:== 2-6-79 9 1110 Sta.:p 3-6-79 ? s2200 l uCihsec 5-12-79 0.45 1-30-79 to 2-2-79 0.95 uCi/sec (.46 vCi/sec) 0.78 uCi/sec % Tech Spec W.a1 11.5% 5-13-79 b 0.24 uci/sec 2-2-79 to 2-6-79 0.78 uCi/sec (0.25 uCi/sec) 0.83 uCi/sec % TW:h Spec 6.3% 5-14-79 2-6-79 to 2-9-79 E3 0.32 N3 1.12 (0.33 uci/sec) j c=a:n' 2-6-79 p5 y,y7 cn lina 2-7-79 % TW::h Spec 8.26% l 5-15-79 [ 0.48 [35 0.81 (0.43 uCi/sec) 2-9-79 to 2-11-79 Y 0.91 4 Tech Spec 10.76% ( t ~ l TABLE 9 2.A-11 4 1 qi i g .m upgi n,m u

1 .i l 1 l May Sc=a:2 - Stack rai:a Sc=am F2-79 91350 S-40-79 0.241 m 5-1-79 1.59 (0.26 vCi/sec) 5 2.a 4 Tech Spec 6.51 S-41-79 5-1-79 to 5-3-79 0.32 vCi/sec 3 sc=am 5-2-79 0.85 vCi/sec 0.47 vCf/sec 4 Tech Spec 8.3% 1 7 42-79 0.0858 3 5-3-79 m 5-7-79 0.0218 (0.087) 3 r mm t Tech spe= 2.2% e e 1 O e B e O e e D TABLE 10 l l 2.A-12 \\ t

/ 2.B. CORE WATER LEVEL ANALYSES The previous section describek the extensive activity surveys which were conducted to determine that no fuel damage had occ.:rred as a result of the event. No activity indications above normal levels were observed. This section outlines the ans,1yses and results which were com-pleted to demonstrate that no' loss in fuel integrity would have been expected as a consequence of the scram event. Because of the initiating scram, the power drops rapidly through the { fission power decay to normal decay heat levels. Even with the simultaneous recirculation purp trip, the pump coastdown retains an increasing flow to

• power ratio until natural circulation flow is established. Thus, the critical power ratio sta,ys above the operating level prior to the scram.

I The decay heat drops to sufficiently low levels after the first few ' seconds, that following recirculation pump coastdown the fuel may be cooled sufficiently in a pool" boiling mode, provided a two-phase mixture level is i maintained above the core. Therefore, the objective of the analyses is to determine the minimum mixture level above the active fuel region. There are two basic evaluations which may be performed to detentine the core mixture level. The first method is. to conduct a mass and energy balance on the core accounting for the boiloff rate as a function of decay heat and saturation conditions, balanced against makeup from natural circulation flow and control rod drive flow into the lower plenum. The second approach is to utilize a mass inventory ailocation process to distribute the available mass through the system depending upon known volumes along with known levels or known thermodynamic conditions. l Both of the above approaches make use e ? the fact that the system is isolated by MSIV closure and the mass at that time is retained and further augmented by Control Rod Drive flow. In order to determine the total amount of mass in the system, the time of low-low-low water level indication (172 seconds) is utilized as an initialization point. At that time, water levcis in both the downconer and core are known, one isolation condenser is in service, loop temperatures are known and the total system mass may be estimated. Subsequent changes in the mass allocation is utilized to determine the core mixture level. t 2.5-1

.s l l l l l l l l Analyses of the minimum water level were calculated by both Exxon l l Nuclear Company and General Electric Company. The Exxon evaluation included l both the boiloff approach and the mass allocation technique. The General Electric analysis determined the water level by the boiloff process. Details of the GE analysis are shown in Appendix 1. ~ A sg of the results of the various analyses are shown in Table

• 2.B.1.

As can be seen, each evaluation concluded that the core remained adequately covered by a two-phase mixture level throughout the course of the event. Therefore, no loss of fuel integrity would have been expected as a consequence of.his event and indeed no indications of fuel failure were observed. ( s 6 e a t 2.5-2 1 I' I l_

h (l i i I i TABLE 2.5.1 RESULTS OF WA111R LEVEL ANALYSES OYSTER CREEK SCRAN OF MAY 2,1979 a Exxon Analyses

• CE Analyses Initial Mass Inventory Above Core 7.12 ft.

(two-phase misture) 31,600 lbs. Type of Analysis Inventory . Solloff sailoff l = w Minimes Nixture f.evel Above ,y Top of Active Fuel 1.62 ft. 2 ft. 2.38 ft, i.e l Ti.w of Minimum Mixture Level 29 min. 20 min. 12 min.

• Preliminary results, final report in process.

I G

_ _ -~ 1 3.A PROCEDURAL CHANGES Operator action specified by Procedure 501 in response to the VESSEL LEVEL TTtIPLE LOW annunciator has been revised to include lessons learned from the Oyster Creek scram of May 2,1979. Plant procedures affecting reactor recirculation pumps have been reviewed. The procedure review included appraising whether the procedure had been adequately revised to reflect the recirculation pump trip modi-I fication and if instructions in the procedures could have contributed to the incident under investigation. Standing Order #23 related to the operatica of the isolation con-densers has been deleted. The required actions necessary for manual ini-tiatien of the isolation condensers have been incorporated into plant Procedure 307. ~~ All Standing Orders have been reviewed to ensure thaithey do noi ~ ' ' impede or preclude the automatic operation of an engineered safeguard ( system. 4 0 3.A-1 1 i \\' l r. ___...-._.__._.~,.._.--.__.__m.

y The procedures reviewed 'as a result of-IE Bulletin 79-0C and the Oyster Creek scram i of May 2, 1979, are listed below. The considerations that were reviewed against are t noted. Procedures for which a change regest has been submitted are marked by ai asterisk. i ' NOTES PROCEDURE NOS. AND TITLES i

• 204.1 - Scram Recovery 1

l e*301 - Nuclear Steam Supply System i b,c

• 307 - Isolation Condenser System i

b,c

• 317 - Feedwater System 1

I j b,c

• 318 - Main Steam System and Reheat Sys*m I

a;b,c, e*501 - Annunciators and Alanns

f j

a'

• 502.1 - Loss of 230 KV Lines i

a 502.3 - Loss of 4160 Volt Bus 1A (1B,1C,10) a-502.4 - Loss of the Reactor Protection System Pcwer { a 502.5 - Loss ol' 125 Y D.C. Power a

• 502.6 - Complete Loss of AC Power a
• 503 - Instrument Air Failure a

E04 - Service Water Failure I a, c 505.1 - Recirculation Flow Increase j a,b,e

• 505.2 - Recirculation Flow Decreasa a

506.1 - Rod Drop a 506.2 - Loss of CRD Hydraulic System i a 506.3 - Abnormal Control Rod Motion a 506.4 - Rod-to-Orive Ccupling Failure a,b.c. e*506.5 - Scram System Failure i a, c 507.1 - Reactor Building Closed Cooling Water System Failure j a, c

• 507.2-- Turbine Building Closed CooMag Water System Failure l

a 508 - Loss of Vacuum {k- ' 509 - Inadvertent Opening of Turbine Bypass Valve (s) a P 3.A-2 e

~ l a

• S10 - Turbine Trip a,b c d *511.1 - Feedwater Pump Failure a
• 511.2 - Condensate /Feedwater Systen Rup+wre a

511.3 - Feedwater Flow control Failure a 511.4 - Loss of Feedwater Heaters a 512.1 - Loss of Generator Excitation a 512.2 - Generator Excitation Equipment Malfunction a 512.3 - Loss of Generator Stator Cooling a 513 - Generator Trip a,b c d.e*514 - Reactor Isolation' Scram a 515.1 - Small Piping Leaks in the Turbine Building a 515.2 - Small Piping Leaks in Reactor Building ~ a

• S15.3 - Small Piping Leaks in Drywell e

a

• S16.1 - Main Steam Line Rupture Outside Drywe'.1 a
• 516.2 - Piping Rupture Inside Drywell, Offsite Iawer Available a
• 516.3 - Piping Rupture Inside Drywell with Loss of Offsite Power a
• S16.4 - Isolation Condenser Line Break Outside Drywell a
• 516.5 - Piping Rupture Inside Drywell with Loss of Offsite Power and One Diesel Generator Inoperable a
• 517

- Significant Increase in Off Gas Release Rate a 518 - Inadvertent Liquid Release to Discharge Canal a.

• 519

- Loss of Containment Integrity a 520 - Hurricane a, 521 ,- !!arardous Condition on the Refueling Floor a 522 - Inadvertent Poison Injection a 523 - Condenser Tube Leakage a 524 - Cleanup Filter Cake or Deminerali:er Resin Breakthrough a

• S25

- Loss,of Drywell Cooling l 3.A-3 l 1 - - -,.....,. ~, _ - -., _ _.. _...,.. _ _. _. _ -.,. _,. -. _ _ _., - _ _ _ - - - - _ - _. - _. - - -... _., _ _,.., ~. _

l 3 a 525.1 - Fire in Plant Areas Other Than Control Room 'a

• 525.2 - Fire in the Cc.1 trol Room a
• 527.1 - Inadvertent Relief Valve Actuation While at Power a
• 527.2 - Failure of Relfsf Valve to Reseat - Reactor Scramed a
• S27.3 - Loss of Feedwater - Electromatic Relief Valve Failure a

528 - Torn, ado a

• 529 - Emergency Containment Purge 530 - Loss of the Reactor Shutdown Cooling System a

531.1 - Loss of SRM Instrumentation a I a. 531.2 - Loss of IRM Instrumentation a 531.3 - Loss of APRM Inst 7 mentation a,b.c d.e*532 - Automatic and Manual Reactor Scram a,b.c

• S34 - Loss of Reactor Cooling Mechanisms During Reactor Shutdown a

535 - Inadvertent Reac+4r Criticality ( a 535.1 - A0G System a 536.2 - Radwaste Service Wa'ter Failure a 536.3 - A0G Closed Cooling Water Failure a 536.4 - Off Gas Building Loss of Power a 536.5 - Fire in A0G Charcoal Bed a 537.1 - Radwasta Building Closed Cooling Water Failurs a 538 - Off-Gas Explosion 539 - Response to ' Malfunction of Meteorolog'ical Instrumentation a e*603.3.001 - F4 circulation Pumps Trip. Circuitry Test

• 604.4.013 - Pressure Suppression Chamber (Torus) External Inspection e*609.3.003 - Isolation Condenser Automatic Actuation Sensor Calibration & Test e*619.3.004 - Reactor to Lo Water Level Functional Test e*636.2.001 - Diesel Gen?rator Automatic Actuation Test 3.A-4 l

( l l l l 1

e i l NOTES: j l a. Review Considerations: l l 1. Are the operators directed to override aut:xr.atic action of safety systems l except when centinued operation would result in unsafe plant conditions? 2. Is primary containment isolation on automatic initiation of core spray prevented or hindered by this procedure? 3. Is inadvertant, undesired. pumping, venting, or release of radioactive lisaids or gasses from primary containment possible per this procedure? 4. Is unrestricted resetting of isolation signals allowed by this procedure? 5. Is the operator given parameters other than reactor level to detemine plant status? b. Review Considerations: 1. Does this procedure contain instructions which could be interpreted to direct closure of rec 1rculation loop isolation valves in all loops? c. Review Consideraticns: l 1. Does this procedure require addit.ional caution statements to adequately warn the operator against closure of recirculation loop isolation valves in all loops? d. Review Considerations: 1. Does this procedure require additional guidance to adequately direct the operator in the event of a total loss of feed and 5-recirc pump trip? a. Review Considerations: 1. Does this procedure require revision to adequately. reflect the recirculation pump trip modification? 2. Did the instructions in this procedure contributt to the May 2.1979, Oyster Creek incident? 3. These precedures have been reviewed to ensure that adequate guidance to operators on mode switch changes following a transient have been provided. 3.A-5 4 i m,. ~. - - - ~., __ _.,. __,_-,. _ _, _. _ _.. _.,,._ m_.,

~ 3.8 OPERATOR TRAINING The training planned as a redult of the Oyster Creek incident of i May 2,1979, includes both short-term and long-term training. Short-term training has two ;: arts. The first part is a review session with all licensed Control Room Operators and their Group Supervisors to cover the sequence of events, lessons learned, and changes to procedures, equipment, and policy anticipated as a result of the incident. The second part is a detailed review of the specific operating and emergency procedure changes resulting from e .the incident. Lcng-tenn training will place additional emphasis on active participation by the Control Room Operators, and especially their Group Supervisors, in efforts to better prepare them for unexpected events. The training methods includ.e,nroblem solving, simulated drills, and abnormality response sessions. To address conflicting indications, the operators will receive training to help them better analyze plant conditions, verify the mostconservative-indication, and act accordingly. ( The incident will be reviewed with each licensed operator prior to shift relief. The procedure review will be completed with each shift during the regularly scheduled training week followir.2 the approval of the procedure changes. The long-term training is planned to include a continuing emphasis on abnonnality response. l l l 3.8-1 ' i l l t -.. -.. ~.. ...-..,__,___._m._ ---w---- - - - - - - -, - - ~ ~ + - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

s j 6 -- t ( r l 3.C SURVEILLANCE CHANGES l The following steps and actions have been taken to minimize the i likelihood of an instrument surveillance test initiating a similar event. a. An investigation has been initiated into the possibility of replacing the existing excess flow check valves with more suitable device that would not require a valving and will allow for continuous monitoring. b. The instrument surveillance valving techniques will be reviewed with the' instrument technicians prior to startup in order to ensure an understanding of proper valving, and the notification of surpervisory per-sonnel of procedure deficiencies. c. The performance of surveillance tests will be evalu-ated on a case by case basis when either startup e transformer is out of service. d. The possibility of procur$r.g an analog instrument system that will not require any valving in order to (a surveillance testing is under inves-tigatisr.. s i 1 3.C-1 M h ' r u. n.= 'N--wgy- ,93- + - = --~ gr 4, +nemi.

• ,,-,t-ewe

+tw+*t+-+-ter m --u**v--*--'*e v v* -N rv t-t*#--W-vWe*rw-owwe--*ww---e**----- a

---

- - -- - - -==- - - -

4 + ~. .... r i. g: 2 e 3.D PHYSICAL CHANGES l l Pirtic cov'rs have been placed over the control switches for e the stx;'aion and discharge motor operated valves. This will require the operator to lift the cover prior to actuating the closure of the valve and minimize inadvertent closure of all five suction or discharge valves in the recirculation lines. 9 e e O e 4 e t 9 g 9 e S 9 D e 9 e

• +9-FP-?

^ y1.- ge ws. -+-man

• -=3w-1f,w-.-m--y v w, w'

.m 9-Wm'e w --

• P---*g P.We*

we-9-r r y*w*--w -w'ew ww,eew--wy--4 w- -w e

m o 4.A Te'st Program for Startup and Power Ascension ~ A. General 1. This section provides a brief description of the startup test program. The testing will cover prestartup, startup to hotstandby, and power range testing. Table 4.A-1 is the startup checkoff sheet thct will be used during the testing. 3. Prior to Startup 1. Prior to. reactor startup, an interference check will be performed on all control rods. This check will ensure that all rods will stroke fully using " Normal" drive pressure, and ensure that all rods will drive in on a scraa signal. This testing will ensure j that no gross core distartion exists. ~o' C. Normal Startup i ( 1. During the normal startup, the following comparisons and analysis will be performed. A comparison of the Estimated Critical Position with actual a. critical configuration (> 173*F).. b. An analysis of the reactor coolant after heatup to i 250*F. An analysis of the reactor coolant after heatup to c. 500 psig. d. When primary pressure is approximately 1000psig, a performance check will be made of scraa times on Group 1 and Group 2 rods. 2. After reactor startup, with reactor power level raised to approxi-untely 20% rated power, analysis will be performed on the reactor coolant and off gas. 3. As reactor power is' increased from 2Gi to > 98%,the following l analysis will be performed: Coolant Iodine Offgas 3M X 4M X X 50% X 60% X X 70% X 80% X X 904 X ,y >98% X X e

4. A-1.

e +

~. l 1 D. Criteria During Startup to Full Power 1,. O to 50% of rated power level I l a. If the reactor coolant I-134 or I-135 concentrations or stack ) release rate exceeds those measured at full power prior to j shutdown, the reactor will be held at de existing level at which samples were taken until subaquent samples show the criteria to be met. If the sample results exceed a 2 time

• increase from the levels measured at full power prior to shutdown, the reactor will be shutdown and placed in the isolated condition until the problem is resolved.

2. 50% to > 98% of rated power level a. If retttor coolant I-134 6 I-135 concentrations or stack release rate exceed 1.5 times those measured at full power prior to shutdown, the reactor will be held at the existing level at which samples were taken until subsegaent samples show the criteria to be met. If the sample results exceed a 2 time' increase from the l 1evels measured at full power prior to shu down, the reactor will be shutdown and placed in the isolated conditi'on until the problem is resolved. E. Reactor Startup Program Checkoff Sheet 1. During startup to full power, each item on the Reactor Startup Program Checkoff Sheet, shown on Table 4.A-1, will be signed off by a cognizant supervisor and a PORC member as it is completed. F. Full Power Operation 1. The following stack release rate specifications will apply for 14 days after achieving full power operation. a. Stack Release Rate <,1.25 times that prior to shutdown Follow Technical Specification frequency of sampling analysis.* b. Stack' Release Rate 1.25 to <1.5 times thati prior to shutdown Augmented sampling an'd analysis program to detect any continuing failures. c. Stack Release Rate >1.5 times that prior to shutdown Reduce power level to maintain < 1.5. d. Stackgas Release Rate > 2.0 times Isolate the reactor and r : solve.

• A second sample shall be analyzed to donfirm the high values.

4.A-2 e 9 e

I 's t ine normal frequency of sampling and analysis of off gas to satisfy ] Technical Specification requirements is as follows: 1 1. Samples are tak m on Monday, Wednesday, and Friday. They are analyzed for gress gamma activity after 2 hours and 24 hours decay to determine a ratio of long lived to short lived ac-tivity. 2. The sample taken on' Wednasday is analy:ed for noble gas isotopic concentrations. 3. If the ratio obtained in 1 differs from the previous analysis by more than 20%, a new isotopic analysis will be performed. At full power operation the normal daily reactor water iodine analysis will be performed. e l I g e es eme e e-- e e ( O e 9 4 4.A-3 f r --we. e e-b--,~-r_ ,-._m,,.,, ,_g _y.,_7.- ,-.%_.m-,.%.,76.-, ---%.,,--,_-.m- ,-,...y-_,..y -,wi.,,_m,.,.,.,.. _.,,.,-.. _ ,,..,_vb_.,. = y

= s m REACTOR STARTUP FROGPAM CHECXOFF SHEET Refer to startup program schedule for details. All completed items must be signed off by a ecgnizant supervisor and a PORC member as beine acceptable. PRE-STARTUP M Suov. PORC Interference checks on all control rods / Reactor coolant analysis - total Iodine. Cl, conductivity l/ / NORMAL STARTUP fuov. PORC ( Comcare ECP with actual critical configuration Heat up to 250*F. and analyze coolant iodine / Heat up to 500f and analyze coolant iodine /! 1000 ado - ercam time teet f:rnnn 1 a nti craon7 eMe 20% rated power - analyze. coolant and off gas / l 30% rated power - analyze coolant iodine f 40% rated power - analyze coolant iodine / 40". ra ted newav-analv7e off one ~ 50% rated power - analyze ceolant fodine / 60%. rated power - analyze coclant iodine f 60% rated oower - analyze off oas / 70% rated power - analyze coolant iodine / 80% ratdocwer - analyze coolant iodine [ 80% rated power - analyze pff oas / 90% rated power - analyze coolant iodine / >981 rated cower - analyze coolant iodine / >98% rated power - analyze off gas / / REMARKS: J l 1 4 e e r w sm,- -,wr wm--,w,,,e, -_w,__,- -,..-w--,-,n--m -n---~,-~,a-,-,.n,,-,,-,,,,, m m-e --a.---- - - - -,, - -,, - -

f~~

c. -

.s I l 4.8 SUPPLD4ENT TO BULLETIN 79-08 Please refer to the response to question 4 to IE-79-08 which was. submitted by JCP&L and' add the following supplemental information: The reactor vessel instrumentation described in the above re-sponse,'except for the low-low-low level instrumentation measures the level in the vessel annulus external to the core shroud. The liquid phase of the annulus reactor water comunicates with the core area under the shroud via the reacter recirculation lines while the' steam phase in the annulus connects to the shroud through the steam sepaistors which connect to the top of the shrcud thus venting the core area to the annulus steam area above the annulus liquid phase. With reactor recirculation pumps stopped and the recirculation line valves open, the mixture leal (mass' equivalent) within the shYoud over the ~ ' ~ ~ ~ core level is accurately indicated because liquid equalization occurs I. through the recirculation piping. Additionally, the General Electric re-(ponse (Attachment 1), verifies that the recirculation flow rate with enly one loop open is sufficient to prevent boiloff from re- { ducing water level within the shroud and the. reactor will functier, under normal natural recirculation flow conditions. l l l I i \\ 4,B-1 1 1

r s. i 4.0 ISOLATION CONDENSER OPETsATION The isolation condenser line break sensing system is intended to initiate and achieve an isolation of the condenser within one minute after receiving a line flow signal of 300% or greater. This flow corresponds to a differential pressure (dp) of 27.5 inches of water across the condensate return piping flow sensors and 20 psid across the steam supply piping flow sensors. It is known that transient sensed, high condensate flow rate conditions exist upon isolation, condenser initia- - tion. This is due to a s' rge of cold watar stored in' the condenser condensate I u return piping and tube bundle. Since density correction is not perfomed on the dp signal, the cold higher ' density water further amplifies the sensed dp. A time I . delay is incorporated in tne system isolation logic to allow for transient dp's in excess of the isolation setpoint for 35 seconds. Based upon this time delay and a valve closure time of 20 seconds, full isolation wecid occur vithin ene minute of a true line break event. f - The isolation condensers return condensate to the suction of A and.E recirculation loops. The effects of recirculation flow cause the transient dp to .= be larger in magnitude and longer in duration. The higher the flow rate, the greater the effect. Testing perfomed in November 1972 (report attached) indicates that the \\ normal operating dp across the condensate line flow sensors to be 6-10 inches of I water. However, the initial transient dp can exceed 60 inches of h& tar and remain abcve the isolation setpoint of 27.5 inches of water for 30 seconds from the tima of initiation. The testing perfonned in 1972 was with 5 recirculation pumps in service

• at minimum flow (- 4.8 x 104 gpm). It shoul'd be noted that subsequent initiations do not yield similar results since the condensate stored in the condenser is hot and the density of the water is less.

l S 4.0. I D.

r ( [ ( l 4.C EVENT RECORDER OPERATION DURING THE MAY E,1979. SCRAM

cription of Event Recorder Ooeration Any of 60 signals switchthe' event recorder frem slow speed operation 4 inch / hour) into high speed operation (6 inches / minute).

Simultaneously, m;.; : (redundant) three minute timers are initiated.The event recorder remains high speed until both of the timers have timed out, at which time slow ed op: ration resumes. This sequence is independent of signal status. .. [ ~- This sequence of events may be modified by operator action as follows:

j. ?

Moving the on/off switch 'from on to off to on resets the ir.f tiation logic. If oil 60 signals have reset, the event recorder will return S to slow spead operation regardless of the time after initiation. If any of the 60 signals have not reset, the three minute timers will reset and time for an additional three minute period. . Op ning the front door of the recorcer stops the forward movement o'f the ~ chart, regardless of the chart speed. i The logic and drive power to the event recorder is from continuous in-w.5 i ment Panel #3. The power to the recording styluses is from the 125 volt L;, ystem. SMr Ooeration on May 2.1979 All aspects of the event recorder operation were normal following in-

ion on May 2,1979.

The initiation of the triple low level sensors at 172 h :- .ds is appmximately the same time as the completion of the three minute s:- ig sequence. It is evident that the on/off switch tas placed in the off ion by persons unknown. The event recorder was discovered to be in the Z criti:n at approximately 2000. hours of the same day. v-l 4.c-1 I T l

q

1. *s 4

Isolation C W ensers Rx Thermal Recirculation Placed in Service Recirculation isolation, Power Flow in GPM Pumps Tripped Condenser Isola 4 Scram No./Date (Wt) (Rated =16x10 gpm) Yes No Yes No

1. 53 April 13, 1972 1857 15.9 x 104 A and 8 were initiated.

/ / gpa (LaLa Level) "C" failed to trip. ) 4 4

1. 72 September 25, 1974 1920 15.8 x 10 gym A and B initiated.

/ / t 1 4

1. 78 May 4, 1976 1765' 14.9 x 10 gps A and B inittsted.

/ / (LoLo Level) i I 4

1. 83 September 9, 1978 1310 16.0 x 10 gpa A and B initiated.

/ / l } 4

1. 86 December 13, 15 M 1917 16.0 x 10 gpm A and B initiated.

/ / lli Pgss \\

1. 89 February 2, 1979 l'920 '

15.4 x 104 A and B Inttlated. / / gpm (Lolo l. Level) 4 l

1. 90 May 2 '1979 1895 14.8 x 10 8 initiated.

/ / (111 Press) I l l 4.0-3

~ In reviewing the past scram data, it was never found that when the recircu-lation system pumps tripped that the isolation condensers isolated. As presented in ' the tabulation above, five times since 1972, the isolation condansers were initiated without isolation when accompanied by'a trip of the operating recirculatien pumps. It is therefore concluded that the isolation condensers will function as intended when accompanied by a trip of the operating recirculation loops. It is noted that the isolation condenser tutsuatic initiation. signals are the same as those l for autoaatic recirculation pump trip. ) \\ ~ G 2 G 6 ~ m eo O 9 9 P e G G d o 4.0-4 a-s ow, - _ -. -., -,.. ~, ,,.... ~, --_..,..,. _,. .-em

rn ISCLATiett COMDasa '[jEI. Date: Nov h 14, 1972 ~ ~ gg. 4 1. Detemine the nagnitude of the stea= and condensate leg d/p spikes ~ 2. Date:::nine the du::stien of the above d/p spikes 3 Mr"hte condenser heat re:n:nral T M ty, if possible . ~., .t Iriitial C::nditions: l . - f,. Ube reactor tas in the "RUN FiZE" at a pcwer level cf apprcxi=ately 400 MWu. Reactor level mal was in the aute::atic nede and react = I.

c.. pressure tas raintained at 1010 psig by use of the bypass valves u. der the control of the EPR.

Recirculation flew tas at the udnf.:::m value (20 Ec).- The APRMs were set such that 100% was equivalent to

b '.,'

• 1200 rft a

.7 . '. Results: l!FchTsolatlen condenser was put into service t-dce and the resulting -(,. transients were r h ded. Fisces 1-A and 1-3 are the d/p vs. tine plots . for the. condensers. Each figure centains the d/p's for the steam les elbr and the 6.ondensate ret

• n elbow fo~ both Run il and Run 12. The pe.d ne.:

.( test results are fouM in Table 1 aM cc::=ents follow: For Run 51, the condensate return leg d/ pts for beth, cendensers exceeded the range of the ree:c-der (0-40 inches of H 01 and also exceeded 2 . the rar:ge of the local d/p gages (0-60 inches of tater). This cag.itude for the spike is ncecal when an isolation condenser has been in the standby condition fer a period of time prier to its beirg put into se-rite Per this coMition, the condensate return leg.:ater ta=perature is cool and the driving' head is relatively large even though the cen:iensate rett ~n isolation valve nay leak slightly. With this valve not leaking, the d/p ,. g spikes would probably rwn greater magnitudes than obse.9ted. The steam leg d/p fe Isolation Condenser "B*' was ret reccrded due to a loose electrical connectica at the rec = der, which acccunts fcr the udssing plot on Figure 1-B. Each condenser tas in operstien for about 6 3/2 ndnutes for. this run. For Run f2 all rw "ed para::sters remained on scale. Since Run f2 :.- perfor:ed a relatively short time after Run (1, the ecmdensate recen legs !ad not cooled cer::pletely to their normal "in-standby" te::peracures. (Tne: is no ten;perature $ndiestion on the return legs) he:ce, the '.cwer magnitud: - d/p spikes. .4.0 5. [ i t l l I. nua -h-.e*vgg--7*-e i->e-+3. 9 we-+,,..,,e.- -,...--,--.,+---ee-in+w--ve-wm'-emier---e-"a-+rM-wwa-----er-eesever*-p------ee- --r--e-- -a-

' +. e.., s ~. ..e. z. ..a. = An atte.pt ws made to dete: '.r.e the heat capacity of the ~ .. IscIation Ccrxiensers. The methcd available to perfe:u the , d *7m7= tion relies on the APF.'( Syste:s to dete:r.ine the pcwer ~,:.., ..., level when. the. Isolation Corxienser is in coe ation. At the ^ ~. low power level and non-no=al core power 'distributien ~ . experienced durirg the tests, the is?.9! System could not he relied upon to provide e:w.:gh ace:.acy to nake a good dete=iraticn. ~

• Results of the attempts gave 1:xiication that the re:moval capacities

. were ann:ere f9sa 2.7% of rated power to 5 9% of rated power. n:e .ccrdensers are desig.ad for a beat removal capacity of approximately 3% of rated power. 3 'I.t is recon =weded that this pcrtion of the test he performed

• at higher power levels 05%) so that the AP*xs an:1 other plant
• parameters w'1' ;rovide more reliable infc=ation. If this is dene, }xr ever, the heat re=cval capacity under desig:.ed corriiticas "will still not be dete: mined since the recirculatica pu=ps (not

'. (

- y:-- zu:ning for the design conditions) v* be at mar rated. capacity..

.. g.. .. ~,.. .. r..- .s ~ .s n.. s..,. t e t = .( ..a .... =... ~ ., g.e-. .v s ...a.n - 1 i s.. ..s . ~. e I 4.0-6 a e w

l e e s -e -.- e e en e

- ~ ~

e,,. .e e j w-

4

= .-.p e ~_.- _1 l l -6 A s=_q h. y" - -

6..

4 O., ="'*-- "e r, ' _nm e ee .e , 6... s

9. -

. - = - +* _ }^ 4 0 """*b ee d ~ M* 3 a .T. -4 2g g =d 2 M o

s. M

.~- g i 3 .h e

4.,-

e. 3 t uL =_

t

,7 '4 (. .s w. v'" ,_.)* .i

m..

- s.. v-= - 9.:.; M --e ,.e.. e==.. e.3.e e C 4 -. ~ ,r g W i=- L Ja r= * ( -~C g .,. I I I '. * *" t. \\ ~ . - ~ -. (" T d_'r" O-C y, ,s* 2y-y- 3 - - F.. 9 ,p.. - J. _. -m 9 g - ;r..... _ _ u. o. [ >0 . = = * *.... _1 ~ W ft w ..e,* l ,'m c.9

• W a sg s%

m ~. t =,=*% = e 3 s g es M bd W. a. mp w ~ a a'- d {. W3b h' ~ % pa-g .w_ ' _d -- ~---=c, g 4 ~$ ~ ...-.3.,_~ .te [ 1.

• Y

1. '8

( !IW 8 = e en 1 o ..i,. 4 9 Ii g ^9 = U s. O 9 e0.. _e e... 4i - _.ge- _ J, 4-M. 4 em ne 8

• 8*""*
• '"" A m%e Hw a.

-{,,e m _ o e -o e .=_s._. ? m yc..[9,- ~ _'t A g~n _._ l...... m. =~-: O-g.... .,. w,(, --e s. jg g f. 4 . ~ - - 4 e- _ .m. e em . es-a.m m ..m.. ems.p m m.

m..m.m.--. g m e M

m..es==. em.

e mm m .m_ e e -mmm 1 4 4.0-7 - - - = 'w ,e m

i y, M s- 'i, ;. l' f ~. *.- JE. .T CENTRAL 1.AR & I.100t? CCHP, c. orsna cRizx NucurJLR OUERATING STATION. l ~ II//i'[7z. ISOLATION CONDENSER Test m7ss vuur A UnstA. U"* '.B Unor s l l pan t gian g RanL R09 E- \\ s~ to aan. s.z mM. 4.s nun. s.s min. I \\ Tgsi Dur,a4*.ori %ive Openinq Tso'une 14 se<. 14 see. 14.s see. 14.s* sec. l l Yn%l Shell ' Temp. 214 *F 214 *F 2/6 *F /90 *F \\ Fahnt Sine.Il Tem p. 2 % *F 233*F 230*F 22.o*f yf to //i,jli flow Se(po'enf s 1 Reken Les /2.0 sec. //. Ssec /4.8 sec. R 2 Steam f.eq 'A 4 U<ab Above Scht in+ I 2ggjm Q 15 5 sec.

1. I sec.

/7 0 sec. W. b t Sisen, len l4 kgb EP<cscure s Refarn le9 M h.14 0 32.2 k. Ila o 730in./ho lo kr. Mao f 3.7S Psid 4.spaid t Sicam Les worintconoen 4.9 psig. I erah' g_ A Pecssuec p o i Rela <n Leq to h. Hao G.tSin.Ifo 4.t h. H, o 6.zs,;,. y, o -l 2 Stca m Leg 2.7Ses*rd

3. o p s',J xarsecoeoeo 3.o ps'sa p D8D Nr>T FEAcil SETibsNT e

.s

I l .w_ -a .e 6 A w

== . ig =..x, 2 m. u- = j *- @"r.. 3 c -V W .1* g,-

• .. * =

7 o-.c I u ) u._ ~ m y; - l w = . n.w g 'g-n Mn 1 4r \\ ) w 2- +- d L. m 2 J u .q gg i ..d = e- .m

{

.l.. ) =8 w 1 ,, =. Ze. .u. .g ,g. 2 l -e.. m ) a.- I

• m _

9 \\ -+- g-- - m \\ 1 g t \\ i = m .s 3 e.g Ig .l w w u s =. o x _ f I e .u u =

ex

_w-g Q es. '" =- = = ~- - , e. s = A .mo - ~ _h --7 4%% i hq.q Z-M &,_ .. _. f N g- _ ...L.- .T3 0.. _ ---.s N... __ @ ' t '* ~-~ _ ~ ;;c w_.g'll2.V.L5:".I.I.W - 4 1-t. t 3 i.. -} ) [*'..** a r- ,. w - ,s i . t. _ ).- _ =.. s e, ,s i .v n n._ _ i i l l. E l b i 1

4.E LEVEL INSTRUMENTATION Table 4.E-1 (attached) details the indication and recording capabilities l' as well as auto functions for all reactor vessel level instrumentation installed at Oyster Creek. ' / In Addf* fon, the GE/MAC narrN range level signals A and B input to j plant performance computer. The narrow range GE/RAC indicators (2) and recorder are the only level monitors that have unique units of level. Their range is zero (0) to eight (8) feet which matches quantatively with the span of " TOTAL FEEDWATER FLOW" signal (zero to 8(x.10s) lb/hr) which shares a coninon recorder chart and scale. Consideration had been given tio change narrow range level units to inches of H O to coincide with units of a?1 other level instrumentation, and to that 2 extent, parts to modify indicator, recorder, and feedwater controller level setpoint scal.e had been ordered. l L . i 4.E-1 t -a ___________.___________m_______m.__

( s e I i 8 TABLE 4.E. SUMARY OF REACTOR VESSEL WATER LREL INSTRUMENTS ACTUATION OR CONTROL DENSITY SENSOR INDICATION ANALOG E E DESCRIPTION 10 NO. LOCAL REMOTE RECORDER ftECORDER FUNCTIONS CONPENSATION 1 - Core Spray Init. RE-02A Yes Yes 2 - Cont. Spray Init. = tow. tow. Level 3 - Reactor Isolation Co*Pensated during calibration Indicating RE-028 Yes Yes 4 - Centainment Isolation No No for conditions 4 Switches 5 - Recirc Pump Trip (Yarway) RE-02C Yes Yes 6 - Isol. Cond. Init. Of Operating 7 - 5GTS Init. temperature and RE-020 Yes Yes 8 - Annunciators pressure. RE-05A Yes No Yes* .g. 1 - W Level Scrm* Level 2 - High Level Turbine 4 Indicating RE-05/19/ Yes Yes Yes* Trip & Target Relay 88 3 - Annunciators (Low Level l Switches (Yamay) RE-058 Yes No Yes* T RE-05/191 Yes Yes Yes* m RE-18A Yes Yes atow. tow _ Low, Setpoint is 1 - Auto Depressurization compensated for Level RE-188 Yes Yes System Initiatten weight of steam 4 Indicating No No Swltches 2 - Annunciators above var.lable leg (Barton) RE-18C Yes Yes and temperature of reference log. RE-180 Yes Yes Harrow 1 - Feedwater Control -l A I5 2 - Annunciator (Ill/ Low Leve l) Auto density comp. Range throughout range 2 tevel 3 - Feed Pump Runout of level and pressure. (GE/NAC) 10-138 Yes Protection Reset Wide Range y (GE/MAC) 1D-12 No Yes No None 1101ES (1): Variable legs of sensors Ilsted above sense level in downcomer region (annulus) except the triple low sensors which sense above core region at core spray sparger. (2): Reference leg condensate pots tap off of upper downcomer region for all sensors above except wide range level which taps into top of upper head.

4.F FUTURE ACTIONS UNDER CONSIDERATION As a result of this event and our review of the factors which contributed to the plant trip and subsequent cperator &ction, the follow-ing modifications are being considered: 1. Develop sequence of events recording capability which would provide event recording for a 1 eriod of hours. This might be done by use of a panel alarm automatic log-ing recorder along with a plant computer capable of moni-toring more plant parameters. 2. Evaluate a wiring modification that would require the C.R. operator to hold the control switch for both re-circulation pump suction and discharge valves for the two minutes required to fully shut the valve;. or electrically interlock the recirculation ymp suction and discharge valves to prevent closure of all valves simultaneously. 3. Review what can be done to minimize instraent surveillance testing which might cause a reactor scram while a startup bank is out of service. I 4. Continue the engineering review of the desirability of installing a modification that provides solid state ( sensing da'tices to replace existing mechanical devices for trip actuation. t 4.F-1 i......

sl. I }' i ~ ~ 5. Investigate procedural and/or excess flow check valve modifications which would eliminate the delicate evolution presently required to insure the. check valves are open following a survdllance test. 6. Change the s ute readouts on the two existing level inst. readouts in the centrol room (Yarways and Cf./NAC recorder) so they are equivalent, and in-dicate water level above the core. 7. Investigate ways to improve the reliability of the feedwater system. 8. Provide control room indication of low-low-low water level. 9. Provide overload bypass' switches for the feedwater pays and oil pumps in the control roon.. e e S e 4.F-2 4 0 h e ?

a l l ? i i 1 9 APPENLI.i ' i'l NATURAt. CIRCULATIC'i FLOW J 4 f'.j .j A. CALCOLatiON5 CF piTIAL C001TIO!G. The n'? Following scram and pr.o trips. natural circulation is established. 7 lb/hr and un- ,f natural circulation core ficw rate is of the order of 10 As the discharge valves 'j evaporated water spills over into the downcomer. 5 are closed,.this flow will decay to approximately 200.000-230,000 lb/hr. g In addition. control rod drive cooling water is availsbie. The rtininc j flow at which drainage of water starts occurring in the separators cor-j N responds to the sit.uetion where the inlet ficw to the core cannot make up the boil-off of steam. g, A dI At 3 minutes into the transient, M= 1895 M4

• 0.0333 (P.sy. Witt)

- 3I tore power = 63.1 754 3 = 325,000 lb/hr Evaporation rate ~ = 38,000 lb/hr a Flashing in csra = 35 4,000 lb/wr J Total W9 e ,f Leakage flow fror-bypass region (1 psid) I- = 840,000 lb/hr (thru channel-tieclate leakage p:th) Corn inlet f1cw 840.000 + 250,000 = 1,100.000 lbs/hr 3 The variation of the ficw rates and void fractions vs. elevatica ara The va;cr' ftw rate increases in the core correspending stetched below. to pcwer input vs height. The inlet flow is boosted by the leakage from At the top of the core the leakags flow is returned to the [ the bypass. This foms a natural circulation loop between the core and the a bypass. .- g The net licuid fim in the u;per plenin and separators i: thus bytess. .~4 A difference seween r.he t.orn exit ficw and the leakage flow. .2 - L. 2 e

r f . = n j 7 1' - 1. The void fraction increases with height in the core. The large area and i low vaper velocity in the u;:str plenr lead to a lower void fraction in -j the upper plenu=. The vapor accelerates in the separator standpipes pro-l i], ducing a high void fraction. The void fraction in the three regicos are calculated b410w: .g i 3, 3 'Y t 5FAf'RK W4 gygags 2 I 1. 2! o. 3 t 'tw I ,fg VPP J=

ogpsg, ia PLwurt J.

w hi 3 ,to g 'I NO Ns. 3 w 'e I, 1 ht .et J l} [kb@khN ,4' 4 s- ? 1 1 1

• e

'4 5 -i 7 Wh. 1 n W -w --,,,w ,. -. -mai--- m mw e--+--* -.----g----s - -we--- a-m

l S-f T 4 'li UPPER 4.: CORE P!.ENUM SEPARATURS -t H (exit) x i m ( V g (1b/hr) (includes 364,000 443,360 452.000 flashing) l im t W Th/hr) 736.000 -104.000 -114000 0 f 4 ? Drift Flux C' 1.3 1.0 1.3 -} - Para:natars 0.82 ft/sec 1.0 0.82 Yj - M 9 Cnss sectional 2 area 56 ft 186 30.0 4 W a o 2900 ft/hr 1123 7100 4-f .4 Q y f-9 s. N ? jf = f 278 ft/hr. -11.A -8.0 AP -4 9 4 j

• jf + j 3178 1112 7020 g

i_t - 1. e= j '.h o O.41 0.24 0.59 jC,t9j'2600 [j 9 Average a for enre = 0.21 s .O a] .m E As the core inlet flow has decrac>wd below the vaporization rate, the lavel 1 .,3 in the separator will fall. The swings in pressure, particularly during

• 7 pressurization, may accelerate this process.

.:H N. dn. 1 3 1 i l a 9 I = o ,__-..--.u. ,.-w,. - - - -. -, - -.. - - -,,,. _ - ~,

c' t .-....... ; -~n ~. . j 1 ? l .1 ' l l sefore the level falls, the total static heses inside the shroud are: 1 j Move BAF: 13.2 * ( 1-0.20) + 5*(1-0.25) i 9' (1-0.59) = 14.31 ft. ) Above TAF: 1.2 (1-0.4) + 5* (1-0.25) + 9' (1-0.59) = 7.44.ft. 14 1 Water inventory inside shroud before separators drain: 4 !.a} l j Yolume Mass (1.b_). ' ?. .25.955 Byp:ss-614 !.i s Core (including unhented region 835 28,490 above TAF): ] upper planum 752 29.750 1' Separatur standpipes 186 3,551 ? 4 1 In crter to gerthe-Triple Low Le9el Alam, collacsed level rrst drop to 3.5 ft i abeva TAF. Using the region void fract13.is this it:glies that two phase level 3 rzst bu at 2,6 ft into the snoarator standpipes. This correspor.ds to appr:xi-3 .j., mately 31,600 cf fluid above the active core. w 5s El 7-Note: BAT - Bottom of Active Fuel ~ TAF = Top of Active Fuel t 4 g 4

a

.I. vl x r 3n f._ i .s h,1 4-s ~ .--..-..,__.,-.m., o.- .___.m_.,y,,m..,,_,....._-..-,..~...,4-~.-.--- e. ,.m .... - - -.... ~., - - -..... - -

, 2 -

s 4 .. t i 1: '] 8. CALCgAT!C'! 0F NATURAL CIRCUI.ATIC!! Fl.C'4 ) l 1,- The natural circulation flow from the downtomer to the core depends en d the static heads inside and outside the shroud: h t 'b ,I [ 7 i ~ W b 1 2 _j 1 .i A \\ s C. g ( A 0 ' Fig F Nsee: NT C&b L y I i 1 p -.i-q The static head inside the shroud will be history dependent based on the net f water flow into the core region. However. based on the'results (to follow) I,. that inflow is of the same order as the boiloff, the level is expected to

s stay neer the middle of the upper planum. The void fraction will gradually w

f decrease insida the shroud as the power falls. Hewever. the two values of I -j recirculation flow chosen should bound this effect. f t i k __.________________.___..m_.

p l l l \\ 5 3 4 Pressure drop in the extarnal loop AC3 is c ntrolled by the resistan::e f in the.five 2" bypass lines. This praisure' drop rast be tralanced by the j pressure gain along BUA. \\2 W

1

1. ACB be,ars (fact of water)

(,) 29 .a a where A = flow area of 2" lina (schedule 80) (assim:cJ) y$ 2 l ,74 = 2.953 in

i f,)

120 ft (for 4 elbows) = Crane j. + 8fg(forgatevalve) F.andbed + 72 fg(ptpafriction) j f a 0.015 g 1 J Q. Exit loss 3:-3 4,8 K = p substituting these values. W"2 (lb/hr)2 W = (I) ]' 4 AC5 '.407

• 1010 (ft) 0 k

3 Inside the vesial. y .n -d. -AP II-*l h (2) goA. gdowncomer i 3 1

.]

i = cura, upper plenum separators. h,'.:( Assts:ing the void fractions calculated earlier and an average level at the 6 midd'dle of the upper planum. 'b -3 se,,, - w, 11.0 (s) I 1 1 3~ M .i 3 s a e C________m__

N ] I l [ l. l e 4 / l 3i. Level in the downesseer ranges frasa 134* to 170* above the top active fuel

i

,N (i.e., 280* to 33D" above SAF). *

J usins. Ap AP y

ac, soA. = J @k ( w,8 o.4c7 - 1020 - ( M,- 12.o) (4) f For the low downcaster level of 200" (22.3 ft). -- 6 y' W,

• 205.000 lb/hr

.m e , 5 For the high downcomer level of 320" (25.55 ft). .l W '= 229.000 1D/hr m V i .1 The uncertainty in thf a calculation is primarily the static head inside the -[9 shroud. However. since a sum 11er statle head produces a larger recirculation jl flow. campensating effects are introduced. In any event. the recirculation 3 flow (toget.her with CRD flow) should t>e sufficient to srake up the bo11of f aftee the first few seinut e. 3.y 4 .' i J .4.j Y il i 1 I'

• 94ere accurate intersretation of the f ewel accountiac for sens Ity changes and f

flashing in the sowncomcc below the p:*r*.sure tan location leeds to a variation l 1 of 270" to 310" alaove RAF. 1 I' ,[ S

(.

l 1 e G 9 . mw,. ..ee,- .,_.___.._,m ,..y.---,,.~.+.--_,,___-..-..._e....-_,.,,-,.-m. - -,.... _,, ~,. -,

,'t-7. 4 - .4 I -s f... .?. 4 1 3 3 hvent.ory Inside the $hroud A M i The inventory calculation has been Improved by consideration of subcooling i ? of the lower plena, flashing and stored energy tfacts: o $f Mass balance insida shroud: 8 f( M - V,- Wg,e (1) g o 3 .f

4 v,. Natural recirculation flow + cRD flow i

't a Staaza outflow due to flashing and evaporation N = out te Mt*Einitial U N -"I idt (2) in out t1 0 M - Total mass inside shroud at time t g h M,ggq,3 = Total mass inside shroud at 3 minutes [ g 'kj The region inside the throud is capssed of four regions: .I 1.. Lower plenu:n and control rod r;ide tubes. This region will reafn single phase liovid, but change in temperature. y a .] 2. Core region i q j 3. Bypass region 1 eg 4. Upper plena (and senarators). 5 .4 4 ~.)s.. t -1 .4 5 b__________________.._.______.._

) .S d.*.,. N4 9* .1 4 Lower Plenum and Guide Tubes i g - W,hg -W,,g,3. + Q (3) h i et Total internal energy in lower plenum E .4 f h,g enthalpy leaving lower plenm -y (,- heat from vessel walls j Q E ! Various assirations can be made about h If the pienue. is perfectly og. 4 - mixed, h corresponds to the average niane temperature. The worst g -G - case for the level calculation in the upper Diens (i.e. the one that ~ 1

3 yields the lowest level) is to assume perfectly stratified flow. The d(

outflow enth&lpy in this caza remains the initial enthalpy until the 1 initial mass in the plantaa is entirely replaced. ~

4 LPbP I4)

Y .{ kP 1 ( dp P Yf = Yh (5) W,-Wout M = j an -)...

ty Approxisating internal energy by enthalpy; equation 3 becomes e

.i (m)

• Win i 4,t out'0 (8) h h

y .?. substituting (5) intu (s)

d

$ = u,,(h,-h,)+q (7) - 3 o I' HSE 3houtM dh 4 h for M 27 minutes = out

• initf a'l g

9 r-in 1

1 I

e t l l i e -w o,--y..w,..-.w,, y..-.mm.-,,,_m,,-.w,--.- ..,.w.--+-%-,_,-,. . --.c--.-r--. - ~ - -. -. -, -.....,,,,. - - - - _. -,

---

e---

3 rl. : i 4* -l< .n .77 if 1 .1 .:{ Table 3.1 stows the calculation of lower plenum taperature and mass using ,S this procedure. The heat frara the vessel wall was estimated.to be fairly 4 small and neglected. This is conservative for the level calculaticn. J Gj -Two-Phase Reefe n .g j From the energy equation, vapor generation rate in a particular regica l \\ .ti, h

3 -

1 Pg (M) bg 0*YIh Pad 9* h g dP (S) fg 1 'g ] In the core heth negative and positive P are used; in the bypass and upper -] plenum only depressurization is considered. 3-The steam flow leaving the core and bypass re; ions (and generate.1 in the .S2 e .) upper planum) is calcularad asstaning quasi-staticcanditicas (i.e.. B ,-f sua11). }I W - (IgYg + apg eva, p } #g (9) g op j The steam flow leaving the shroud 2 a l W9 = Wg + W9oypass + Wgupper plenum l (10) l t out core 5 lli d ,4 The msr in math region is ralculated as below: 49 ] j = @ +opg dvc. P 1. g g A dp (11) I Ni a v j exit 9 (12) 9 .1C + v$ 0 s._ 1 E- 0.5 (a,9; + ainlet) (13) (1-B E (14) N* En + i f 9 'S y O Gb w______________ _.. _ _ _ -. _ _ _ _ _ _ _. _.

l* ~ 3 -j., 3 .1 y a .c .d 3 Table 3.2 lists the system parameters used in the calculations. The pres-a l sures and pressure rates were chtained from the data from the site. The decay heat factors corresocnd to the May-Witt curve. It is estimated that stored energy release from the c=re is less than 5% of the decy power after the first three minutes. Since the May-Witt values are conservative by 16-235 compared to the new best esticate AMS Standard, decay heat values more than co":pensate for stored energy effects. g Using the dats in Table 3.2. the sten:t flow rates and void fractions were ]- calculated as a function of time for the various regions. as given by 3 equations 8-14. These are tabulated in Table 3.3. m .~t Tables'3.1 and 3.3 provide sufficient infcrmation for a mass balance using 4'- Equation 1. c.? M M' Mupper plenum

• M

~M ~ bypass t core - ~- M g plenum -- (15) j Table 3.4 shws the calculation of t:ts1 mass and upper plenum mass. using il ' equations (1), (2), and (5). Two values :* recirculation flow at 200,000 .A

.3 and 230,000 lbs/hr were used. At the higher flow, a minimum mass of 20.000 lbs (13" collaps--i level) ws rn.

hed at between 9 and 15 :sinutes. Beyond d

this time. the inflow is able te overcome the effects of vapcr generation j and increased density in the lower plemmt. 4 ~ For the lamer flow, a mininsa value of approximately 16.000 lbs (23= N collapsed level) was reached at 11-23 minutes. An increase occurs there- .e Y after. 9 -[ .g The variatien in mass and collapsed level is shown in Figure 1. . *] a( 1y i i I'. u___-______.__ _ _ _ _

y m [-y,g-l l ~ z. a l f .j l 1 ] Table 3.1 e ~i j ger Plenun Mass _ Calculation m if }l h a 4583 (h d I

1. ")

in eut i 1.521xlG2 .,T (using W 1.43x105 lbs. dP. *-0.04188 between h = 512 to 356 57U/lb f dh II h,,g-h = 77 3TU/lb. h [ calculated by weighting enthalpies of rectre and g h ',. CAD flow ) 'y tY h 5 Y Mass h h, h U. g Time g out STU/ min e cd. 3 410 51 2 ' -3.07 512 0.02091 1.341x105 4 410 512 -3.07 509 0.020E5 1.345x105 3 b 5 41 0 $12 -3.07 506 0.0203 1.348x105 .k. 6 410 512 -3.07 503 0.0207 1.355x105 -3.07 500 0.02067 1.357x105 g 7._ . 41 0 51 2 8 410 51 2 -3.07 496.5 0.0205 1.362x105 ,1 9 41 0 51 2 -3.07 493.5 0.0205 1.353x105 10 410 51 2 -3.07 450.5 0.02048 1.363x105 5 12 382 51 2 g -3.92 484.5 0.02039 1.375x105 wT 14 362-512 0 -4.52 476.7 0.02025 1.385x105 h' -5.18 467.6 0.02005 1.398x105 ') 16 340 512 E 18 330 51 2 -5.48 457,2 0.01988 1.411x105 I 20 320 51 2 .5.78 446.2 0.01967 1.425x105 1 25 313 512 -6.00 417.3 0.01524 1.458x105 5 30 305 410 -3.25 387.3 0.0188 1.492x105 d 5 410 377.5 0.01867 1.502x10 pg 33 301

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,e g 3 j-M 1 -Q P .N TABLE 3.2 s.4 Q 3 Based en 895 D Wg = 10000d/1.054hfg } Time P P Qd Qd h {f W fg g J (min) (psig) (psi /sec) (pu) (%) (Stu/lb) (1b/sec) .i M 3 920 -1.3 0.0346 65.57 660.9 94.13 4 850 -1.3 0.0325 61.78 674.6 86.89 I 5 850 1.3 0.0310 58.75 674.5 82.53 ,j 6 920 1.3 0.0294 55.71 561.0 79.95 7 1000 1.3 0.0280 53.06 547.5 77.75 i i8 8 950 -1.65 ' O.0272 51.54 657.0 74.43 l 9 830 1.65 0.0253 49.84 680.5 69.49 -3 ,y 10, 780 -1 0.0255 48.32 630.6 65.38 [j 12 730 -0.23 0.0214 46.24 700.8 62.50 ~ i 14 710 -0.23 0.0234

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i* .r i 1 r i ; .c L t e,meer c-e+c [ - Addendum 3 3 C. Ca.fculation of Mat. ural Circularian Flow vi-b 1 (me) Ex =rnal bop Open j \\ 'J. 11t1113:ing the same swthods of saccians A + 3, the natural ciret:1= tion flew

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) rata esa be calculated by: 'j 1 = (v /(4 A)2 cr nop+ x pump /2g) 1 APAC3 z A 3 f3 1-i AP = - AP =R - 19.0 (ft.) 3 .ac3 RDA downcamar 'f 'V a Flow through.open loop, ibs/hr T m. 3 J_ p = density of water in loop. - 47 lbs/ft }l = 3.13 ft' (26" schedulu 80 pip) s A ,4 125 wesel entranca + axit Inszus 1 K = j g. 4- +120 5 albows O f, 40.50 2 gata valves .5 1 +0.50 1 flow element 30.21 straisht pipe s (crana EdEE vaines) -g Q = 4.0 total less coeft. 3' / = 21.0 firsd rotor (worst case vs free rotor) f E ** N (fras Byran-Jackman Tass Dats) k.1 h3 %,a = 25.o 6 lS.O ft = static head insida shroud a= care flow of % 2 x 10 lbs/hr r, (

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,g'N r,c' A r- .-p: g ...,7 ~f / t.- 4 a i for the Inv doun=omer Invul of 280" (23.3') 6 W - 1.76 x 10 2be/ar i e '.i For tha histi dowacomer 2avel of 31C" (21.8') 8 i. w 6 (

• E 2.22 r.10 lbs/hr a

9t Since the other four loops would be supplying f2uid through the 2" bypass lineen. i the total 11w to the core won 2d be: H.- 0 g = 1.76 x 10* + (0.E)o.205 x 10 W_ = 1.93 x 10 2.ba/hr I L. 'I I 6 r 0 v = 2.40 x 10 lbs/hr r 2.22 x 100 + (0,5) 0.229 x 10 3 o-Thus the racirculation flow Tata (which is ahe.g: 5 tm 6 ti ms bo11cff Tata) J with malv ar.a loop open is sufficidat to prevent boil-off fram reducin5 water i Inval within the shrmad and the reactor will functics u; der normal na.tural ciresla:lon f2mv condi.imms. With additional 1 mops apes, this flow rate m:Id i be such truster (apprezinstaly 2 x for 2 loops. 3 x for 3 loops, etc) and chus j ti previda even smat.ar margin to boil off. 3 1 .4.. a \\ 4 t k s I 4 .t E-3, ,k t 9 . k.. N .y. b l s

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