ML20083R424
| ML20083R424 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/12/1979 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| References | |
| NUDOCS 8304040007 | |
| Download: ML20083R424 (87) | |
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i REPORT ON THE MAY 2, 1979 TRANSIENT
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4 AT THE OYSTER CREEK NUCLEAR GENERATING 4
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JERSEY CENTRAL POWER & LIGHT COMPANY l
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.s 3-4 DATED: MAY 12, 1979
.o 8304040007 790512 PDR ADOCK 05000219
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3 REPORT ON Ti!E MAY 2, 1979 TRANSIENT AT THE
.O OvSTER CREEx NuCtEAR ceNERAT1Nc ST^T1oN TABLE OF CONTENTS
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Page No.
1.
Description of the Transient and Sequence of Events 1-1 2.
Analysis 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 Reoccurrence 3-1 A.
Procedural Changes 3.A-1 B.
Operator Training 3.B-1 C.
Surveillance Changes 3.C-1 D.
Physical Changes 3.D-1 4.
Supplemental Information 4-1 v
O 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-1 E.
Level Instrumentation 4.E-1 F.
Future Actions Under Consideration
- 4.F-1 Appendix 1 - General Electric Analysis
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Page 1-1 DESCRIPTION OF TRANSIENT AND SEQUENCE OF EVENTS RELATED TO SCRAM 0F MAY 2,1979 AT j Q OYSTER CREEK NUCLEAR GENERATING STATION INITIATING EVENT:
On May 2,1979, at 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br />, an inadvertent 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 4
condenser high pressure initiation switches being tested.
The technician performing the test was in the process of verifying that the sensing 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-03B due to a hydraulic disturbance associated with valve manipulations O
required by procedure to verify the position of the excess flow check valve. The hydraulic disturbance also caused a momentary trip of the isolation condenser initiation switches (RE15A and RE15B).
These sensors were not closed long enough to initiate an automatic initiation of the isolation condensers, since a time delay is involved in the initiation logic. However, these sensors also are used in the automatic 1
i recirculation pump trip logic which did operate in tripping the four operating-recirculating pumps.
No automatic time delay is involved in this logic.
l INITIAL CONDITIONS:
Plant Parameters at the Time of the Scram:
Reactor Power 1895 MWt-Reactor Water Level 79"Yarway(13'-4"Abovethetopoftheactivefuel)
(See Figure 2 for water level reference tabulation) 6.4' GEMAC 4
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Page 1-2 Reactor Pressure 1020 psig Feedwater Flow 7.1 x 106 lbm/hr Recirculation Flow 14.8 x 104 gpm Equipment Out Of Service:
Relevent to Event Sequence:
A.
Oneofthetwo(2)startuptransformers,SB(Bank 6),wasoutofserviceas permitted by Technical Specifications, to perform an inspection of its associated 4160 Volt cabling.
SB supplies offsite power to one half of the station electrical distribution system (see Figure 3) when power is not available through the station auxiliary transformer. The 4160 Volt buses which receive power from SB are 1B and 1D.
Bus 1D supplies power 4
l to certain redundant safety systems. Bus 1D is designed to be powered from #2 Diesel Generator in the event power is not available from either the auxiliary transformer or startup transformer.
Bus 1B supplies 4160
~O Volt power to non-safety related systems and hence, does not have a diesel backup power source.
B.
One of the five (5) recirculation loops (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.
EVENT SEQUENCE: (To=1350)
TIME OF EVENT (Sec)
EVENT DESCRIPTION 0
A reactor scram occurred for the reason previously
!.O described coupled with a simuitaneous automatic tr4P of the four operating Recirculation Pumps. The Control 4
Page 1-3 TIMEOFEVENT(cont)
EVENT DESCRIPTION (cont)
Room operator verified that all control rods inserted and proceeded to drive-in the IRM and SRM Nuclear Instrumentation. At this time 4160 Volt power was being supplied from the auxiliary transfomer during the coastdown of the Turbine Generating System and the f
Feedwater System was in operation.
Recirculation flow 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 decreased to the pressure regulator setpoint as steam flow decreased. Reactor 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 lA and 1C successfully transferred from the auxiliary trans-former to the SA (Bank 5) s'tartup transfomer.
Since SB (Bank 6) was out of service at this time, power was lost to Buses 1B and 10. As designed Buses 1B and ID separated through operation of breaker ID and a fast start of Diesel Generator No. 2. occurred to power emergency loads on Bus 1D.
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TIME OF EVENT (cont)
EVENTDESCRIPTION(cont)
Loss of power to Bus 1B resulted in loss of Feedwater Pumps B and C and Condensate Pumps B and C.
Although power was available to the A condensate and feedwater
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pumps, via Bus 1A, the A Feedwater Pump tripped on low suction pressure.
Since water inventory was leaving the Reactor Vessel through the Steam Bypass Valves to the Main Condensers and a high capacity source of high pressure makeup water was not available, reactor water level and. pressure decreased.
In addition, the loss of power to Bus 1B caused the B Cleanup System Recirculation Pump to trip which, in turn, caused an isolation of the Cleanup System due to O'
low flow through the cleanup filter.
Furthermore, one condensate transfer pump and the operating fuel pool cooling pump tripped. An unsuccessful attempt was made to restart the A feedwater pump.
(The reasons for the restart failure are described later.)
(EventRecorder) 13.6 Reactor water level decreased to the Low level scram setpoint which is 11'5" above.the top of the active d
fuel region.
f (EventRecorder) 16.8 The output breaker on the No. 2 Reactor Protection i
System M.G. Set tripped due to loss of power to the
- O drive motor. The output voltage from the M.G. Set had
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Page 1-5 TIMEOFEVENT(cont)
EVENT DESCRIPTION (cont) been maintained by flywheel action since the time of the turbine trip. Power to the M.G. Set drive motor is fed indirectly through Bus 1D which was deenergized at this time.
31 The No. 2 Diesel Generator Breaker closed and supplied power to the 1D Bus.
A second control rod drive pump started.
43 Reactor water inventory continued to decrease due to steam flow to the main condenser.
In anticipation of a Low Low Reactor Water Level automatic isolation of the reactor (which occurs at 7'2" above the_ top of the O
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active fuel region), a manuai reactor isolation was initiated to conserve inventory by closing the Main Steam Isolation Valves.
This action was taken at an indicated water level of approximately 30" on the Yarway 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 ampli.fied by g-the effects of introducing cold feedwater into the vessel during the 13 second period prior to the Turbine Generator Trip. The cold feedwater reduced the steam
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voiding inside the vessel thereby causing a shrink in water level.
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Paga 1-6 TIMEOFEVENT(cont)'
EVENTDESCRIPTION(cont) 49 The Main Steam Isolation Valves fully closed, thus j
stopping the loss of water inventory from the vessel thereby causing an increase in reactor steam pressure.
Indicated reactor water level started to increase shortly after isolation, when reactor decay heat re-established a steam void distribution.
(EventRecorder) 59.6 The reactor mode switch was transferred from RUN to REFUEL.
76(1 min.16sec.)
To establish a sink for the removal of decay heat from the reactor, the B isolation condenser was placed into service. At this time, the Control Room operator O'
ciosed tne x and E rec 4rcuiation loop discharge v.,ves (these valves take approximately two (2) minutes to close).
It is postulated that at this time, both B and C loop discharge valves were also closed. The conclusion that the five recirculation pump discharge valves were closed is based upon loop temperature response later in the event and is further supported by the Low Low Low level at 172 seconds. The D loop was isolated previously.
(See the equipment out of servicesection).
(EventRecorder) 90(1 min.30sec.)
The reactor Low water level alarm cleared due to the Q
water added from the isolation condenser to the
. Primary System.
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P' age 1-7 TIMEOFEVENT(cont)
EVENT DESCRIPTION (cont) 96 (1 min. 36 sec.)
The B isolation condenser initiation valve fully
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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 recirculation loop temperature did not change appreciably.
A, B, and C recirculation loop temperatures 4
increased slightly.
The heat-up is attributed to natural circulation through the partially open discharge i
valves carrying hot water (536*F) waming the lines previously cooled by the effects of cold feedwater, The reduced flow area between the lower downcomer and lower plenum area, due to the slow closure of the i
jh discharge valves, started to cause a shift in water inventory from the core area to the upper and lower downcomer region. The shift was due to the isolation condenser returning condensed-steam from the core area to the downcomers.
The water inventory shift continued
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as the discharge valves moved to the full closed position.
(EventRecorder) 172 (2 min. 52 sec.)
The reactor Low Low 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
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the Low Low'Lew water level instrument. This does not i
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EVENTDESCRIPTION(cont) necessarily indicate an inventory loss from the core Q
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but rather a redistribution of water and steam voids above the core.
186sec.(3 min 6sec)
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 was due to heat transfer between the hot recirculation loop piping and the water in the piping.
A-U Reactor pressure continued to decrease as a result of isolation condenser operation.
4 250(4 min 10sec)
B isolation condenser was removed from service to reduce the rate of cooldown of the Primary System.
Removal of the condenser caused indicated water level i
to decrease.
The decrease in indicated water level was due to a return of water to the core region from the downcomer region through the five (5), two-inch (2")
bypass valves around the recirculation loop discharge valves.
During this period, the net water inventory 1
effect was a storage of water in the recently secured isolation condenser.
The recirculation loop discharge temperatures reached equilibrium and followed a slow cooldown trend.
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Page 1-9 TIMEOFEVENT(cont)
EVENTDESCRIpTION(cont) 270(4 min 30sec)
The reactor pressure increased due to the effects of removing B isolation condenser. The rate of decrease in water level shifted from a ramp of approximately 37 in/ min to 2 in/ min. The reason for this change is the isolation condenser tube assembly was completely filled. The flow through the five (5) 2" bypass valves continued, accounting for the change in slope.
4'50(7 min 30sec)
Both isolation condensers were placed in service.
This caused an increase 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 h
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 48sec)
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 approximately 14.4 feet above the top of the active' fuel (88" on 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 active fuel where it remained until approximately 1212 seconds when A e
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Page 1-10 TIMEOFEVENT(cont)
EVENTDESCRIPTION(cont) isolation condenser was removed from service.
The O
reactor pressure continued to decrease and all recir-culation loop temperatures continued to trend downward.
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.
s 540(approx).(9 min)
The four (4) Low Low Low water level indicators were verified locally to be below their alarm setpoint which is 10".
The reading appeared to be at or below O ~
the instrument's lower level of detection.
810 (approx) (13 min A recheck of the triple Low water'1evel indicators
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30 sec) showed that the pointers were active (moving) although
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they continued to read below their alam point.
The instrument was at or slight'ly above its lower level of detection.
1212 (20 min 12 sec)
A isolation condenser was removed from 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.
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P.aga 1-11 TIMEOFEVENT(cont)
EVENT DESCRIPTION (cont) 1488(24 min 48sec)
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.
1914 (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 pump.
The pump failed to start due to a tripped overload on the auxiliary oil pump which is interlocked in the pump starting sequence.
The indicated water level started to increase due to the' action of the operating isolation condenser transferring water to the downcomer region. When the C recircula-tion loop.was stirted the loop temperature in-creased from approximately 400 F to.470*F. The other recirculation loop temperatures continued to trend down. At this time Low Low Low alarm may have cleared.
Page 1-12 TIME OF EVENT (cont)'
EVENT DESCRIPTION (cont) 2208(36 min 48sec)
The A Feedwater pump was successfully started by
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locally starting the auxiliary oil pump which satisfied the required starting interlocks.
Indicated water level increased to a level corresponding to 13'8"
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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.
2340(39 min O sec)
The 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.
O' The tow tow tow water levei alarms were unown to be cleared 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 initiated at this time to bring the plant a
to " cold shutdown condition".
2700 (45 min 0 sec)
Reactor Protection System #2 restored and scram reset.
3600(1hr.)
The SB transformer was returned to service and Buss 1B was energized.
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Page 1-13
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l REACTOR PARAMETERS:
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i Figures 1-4a and 1-4b are a trace of reactor pressure, saturation i
temperature, annulus water level, recirculation flow, and recirculation loop temperatures from the time of the trip to 45 minutes later, when the tran-sient was over.
They are annotated with significant events during the period.
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Elev. T Active Fuel (G+4)
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Flange 660 Stea:n Line 591 Top of Steam Separations 84'8" 539 186" 100 8'
130" 3
Top of Indication Turbine Trip 529 90 REOS Hi Alam 4
528 175" 89 ID14 Runout Peset Normal 519 8
6.4' In Alarm 500 61 ID14 In Ievel 80'7" 490 11'5" 51 RE0S Bottczn of Steam Separation 485 Bottczn of Dryer Skirt 477 Feed Inst. "0" 443 0
Io Io Ievel 439 7'2" 0 RE02 Feed Idne Penetration 422 Io Io In Ievel 409 4'8",
O RE18 Core Spray Nozzle 408 Ibp of Active Fuel 69'2"
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Vessel "0" 39'9"-
0 Call other Ievels rounded off to nearest inch REF'c2 GE prints 104R858, 148F712, Burns & Poe prints 2063 FIGURE 2 a
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l DESCRIPTION ~ 0F 'Ill5 OYSTER CREEK PLANT The Oyster Creek Plant is a General Electric, 5 loop, forced recirculation, 1930 MWt, Boiling Water Reactor (BWR) with a Mark I con-tainment system. The steam supply system consists of main steam piping, feedwater piping, ahd recirulcation pumps and piping. The system is also l
l equipped with a cooling system consisting of circulation piping and con-l l
densers to provide for heat removal via natural circulation thrcugh 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, f
The containment system 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
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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 operated either automatically or manually to relieve excess pressure or depressurize the system.
Each of the 2 steam lines is also equipped with 2 isolation valves.(1 inside and 1 outside of the con-tainment vessel) to isolate the pressure vessel either automatically or manually as needed. The S 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 l.
reactor low-low-low water level, and core spray booster pump discharge. pres --
sure exists simultaneously for a period of 120 seconds or less. This is to ~
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o depressurize the system to allow for core spray flow into the pressure vessel. The main steam isolation valves are closed automatically on de-1 tection of any one of the following signals: (1) main steam line high radiation, (2) high steam flow in the main steam lines, (3) high tempera-i ture along the main steam lines, (4) main steam line low pressure, or (5) low-low reactor water level. These valves may also be closed manually a
by the operator.
4 The feedwater piping delivers feedwater through 2 check valves (1 inside and 1 outside of the containment vessel) and a locally operated I
stop valve, inside containment, to the feedwater sparger within the an-nular region (downcomer) of the reactor. This water mixes with the re-1 circulation water in this region and is then delivered to the core through t the recirculation loops.
The recirculation pumps take a suction from the annular region of the pressure vessel,between the vessel wall and the core shroud,through a normally open suction valve and discharge water threugh a discharge valve equipped with a 2" bypass valve into the bottom of the pressure i
vessel. The rated flow capacity of the. combined recirculation loops is 6
61 x 10 lb/hr. Each recirculation loop is 26" diameter piping contain-ing motor operated suction and discharge valves (equipped with 2" motor i
operated bypass valve) and a variable speed recirculation pump. There are j
5 such recirculation loops and all suction, discharge and bypass valves are normally open during operation. Recirculation loops A and E have a
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10" connection on the suction side of the recirculation pump upstream of the isolation. valve. - These' connections are the return lines from the.
2 isolation condensers.
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The isolation condensers are connected to the reactor vessel i =
l) steam region and the suction side of r'ecirculation loops A and E providing a loop for natural circulation through the reactor core. The isolation 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.
All valves are motor operated and normally open with the exception of the I
outside containment valve (DC motor operated valve) on the outlet piping j
which is normally closed. This system receives steam from the reactor i
vessel which is condensed within the tubes by surrounding water on the shell
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l side and returns the condensed water to the recirculation loop. The heat
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i transferred to the water on the shell side causes it to boil. The re-i sulting steam is vented to atmosphere. The driving force for this system i
j 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-
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sure or low-low water level after a maximum of 15 seconds time delay. The i
system may also be actuated manually by the operator.
Steam from the reactor drives the main turbine / generator, is then condensed and returned to the reactor via three 1/3 capacity condensate pumps, three 1/3 capacity feedipumps, and'the feedw' ter piping.
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a The 3 condensate pumps discharge to a common header feeding various heaters j
and coolers. Discharge water from the 3 intermediate pressure heaters feed the suction side of the 3 feedwater pumps which discharge through the 3 high pressure heaters into a common header feeding two 18" lines which run to a tee inside containment. Each of these lines thenL feed two 10" l
(cedwater lines to the. reactor..
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l The condensate, feedwater, and recirculation pumps are powered from the station non-vital 4160 volt buses IA and IB which during operation receive power from the auxiliary transformers. Startup transformers SA and SB provide power to buses IA and IB during plant shutdown. Condensate pump 1A; recirculation pumps A, C, and E; feed pump 1A; and cicanup re-circulation pump A receive power from bus IA while condensate pumps 1B and IC; recirculation pumps B and D; feed pumps 1B and IC; and cicanup recirculation pump B receive power from bus IR, Startup transformer SB provides power to bus 1A and startup transformer SB supplies power to bus IB.
Figure 3 presents a schematic diagram of this distribution system as well as the emergency power distribution.
In order to monitor system performance, instrumentation is pro-vided to monitor reactor water level, reactor pressure, valve position, recirculation flow rate, and other system parameters. Reactor water level is monitored by three icvel measuring devices; "GE/ MACS", "Yarways, and "Bartons". Two reference legs outside_the. vessel are provided for-
. level indication and protective functions. Eight "Yarway" differential pressure cells, four "Barton" differential pressure. cells, and three "GE/MAC" differential pressure cicctronic transmitters provide for IcVel indication and protective functions. Reactor level is indicated both locally at instrument racks and remotely in the control room.
Level indication in the control room is provided by both the "Yarway" and "GE/MAC" instruments. While the "Bartorf' instrument provides 1cvel indication at the instrument rack, it provides only an alarm function in the control room. All level indicator variable legs sense Icyc1 in the annular region of the pressure vessel except for the 4 "Barton" Low-Low-Low OO Icyc1 indicating switches. These switches indicate 1cvel inside the shroud 1-17
above the core. Local indication of this level is provided on an instrument
- O rack with a remote aiarm function in the contro1 room. rioure e pre-sents a diagram of the icvel 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 the time of the Oyster Creek event all systems were in the normal line up with the exception of the startup transformer SB and re-4 circulation loop D.
Startup transformer SB was relloved 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 valve open, and a plate was installed over the loop opening.
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2.A REACTOR WATER AND STACK 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 tims "B" recire. pump was off. The con-ductivity had risen from 0.10 pmho/cm in the morning sample to 0.37 pmho/cm I
which is normal after a scram. The gamma spectrum analysis showed the fission product activity concentrations to be in the normal expected ranges for the I
condition.
Iodine 131 was up approximately a factor of 2 which is expected l
due to depressurization.
A reactor water sample was taken at 1640 from the cleanup inlet. "B" O
reeire r# r a deea ter ea ea orier t this se ete-ce r rise er the re-a sults with those of the previous sample showed good agreement. Conductivity I
had dropped to 0.30 pmho/cm. Gamma spectrum analysis showed normal isotopic decay and the effect of the cleanup demineralizer removal.
i Reactor water samples were taken at 2105 from both the cleanup irlet and "A" recire. loop. The samples showed good agreement on the para-meters-checked. The conductivity in the C.U. inlet sample was down to 0.18 pmhos/cm. Radioactivity IcVels 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 reactor water sampics were taken on 5/3/79 at 0641, 0840,1547, and 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.
Isotopic analysis showed stable conditions in the water. When the sample taken on 5/4/79 0 0800 continued the stabic trend,'the reactor water s
]
analysis was returned to the normal once per day frequency.
2.A-1 n.,
O
l-(]J Additional backup analysis of the reactor water radioactivity concentration is being maintained with a continuous on-line gamma spectrum analysis. This multichannel analyzer 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 hours0.0186 days <br />0.448 hours <br />0.00266 weeks <br />6.129855e-4 months <br /> and integrated over an eight hour period. The results of this analysis com-pare favorably with that of our grab sampics taken during the period, and do not indicate any abnormal fission product Icvels 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 data taken during subsequent startups so that expected levels during those periods are seen.
Stack Analysis The stack particulate and charcoal filters were removed and analyzed on 5/3/79. The results were compared to the filters removed on S/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 comparison is made also with previous scrams during this cycle.
Data for previous scrams also includes stack releases during subsequent startups.
a M
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Rx Scram at 1350
/
May2,]9' O
TABLE 1 0820
[5gS-2-7 5-2-79 5-2-19 2105 bib 5 79 5-3-79 On Iane 5-2-79
,p.1640gM B
A A
tbte*5-2-70 Recirc P m 0641 Analysis sm B Off 4
. nre-1617 a
pCi/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 FDL det.ect.'d
~
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 1.08 E-2 7.97 E-3 1.09 E-3 2.27 E-2 I-134 6.50 E-1 4.08 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 E-1 Xe-133
<l.34 E-3 5.19 E-2 1.2_9 E-2 1.19 E-2 8.40 E-3_ 2.19 E-2 l
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. 80 E-2 2.40 E-1 7.01 E-2 1.67 E-2 1.19 E-2 1.58 E-3 3.10 E-1 Sr-92 1.00 E-1 4.05 E-1 1.02 E-1 8.79 E-3 5.34 E-3
<3.0 E-4 4.68 E-1 Mo-99 1.69 E-2 4.20 E-2 1.97 E-2
<l.00 E-2 1.56 E-2
<3.8 E-3 7.83 E-3 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 8.30 E-3 Total Iodine 1.02 E O 8.97 E-1 4.59 E-l 1.84 E-1 1.69 E-1 1.13 E-1 4.68 E-1 Y
Np-239 1.35 E-2 1.47 E-1 9.71 E-2 1.48 E-1 1.45 E-1 6.50 E-2 1736E-2 w
Gross S 8.16 E-1 1.87 E-1 Gross a 1.51 E-6 2.39 E-5 i
pH 6.13 5.50 5.97 6.14 Conductivity 0.10 0.37 0.30 0.18 0.25 0.15 Sus. Solids 220 ppb C1-
<20 ppb
- Contintous on line spectrum analys -s integratml over 8 ly >ur collectj on period r tarting at 1611 on 5-2--79 e
b h
4
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O
=2 RE7CIOR VATER ANALYSES 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 0825 I 131 1.05 E-3 1.69 E-3 1.23 E-3 1.08 E-3 4.72 E-4 4.75 E-4 Not Det.
132 1.12 E-1 1.00 E-1 9.71 E-2 8.51 E-2 7.33 E-2 6.01 E-2 l 4.53 E-2 133 ND ND 134 l
MD j
ND l
135 ND ND Xe 133.
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 2.52 E-3 3.47 E-d
<3.3 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 Ba 140 6.38 E-3
<2.57 E-3
<2.56 E-3 3.72 E-2 a.ll E-3 2.36 E-3
<2.3 E-3 La 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 k2.0 E-4 Np 239 5.58 E-2 3.53 E-2 3.16 E-2 1.97 E-2 9.25 E-3 5.06 E-3 3.42 E-3 P
. Gross Beta 1.59 E-1 1.43 E-1 1.05 E-1 6.84 E-2 l6.3d E-2 4.26 E-2 Gross Alpha
<l.0 E-6 1.27 E-5 8.11 E-6
<l.0 E-6 kl.0 E-6
<l. 0 E-6
. L-pII 5.9 6.4 6.05 6.1 5.8 Conductivity 0.13 0.42 0.41 0.37 0.37 e
8 h
a 4
g.p,,,.
g, III O
(,,)
REncion hWrER MhL.
, - DEC. SCIW1 12-13-78 9 1851 V
SIARIUP 12-18-78 @ 1521 8:20 0:30 8:30 8:30 8:20 8:30 8:30 PARN4rfrER 12--13-78 12-14-78 12-14-78 12-15-78 12-16-78 12-17-78 12-18-78
_uci/ml I-131 6.16 E-3 2.15 E-3 5.2 E-3 9.12 E-4 6.42 E-4 2.7 E-j nrr 2.68 E-3 T-117 7.92 E-2 4_11 E-2__ 1.66-E-1 4.07 E-2 4.79 E-2 3.5 E-2 4.61 E-2 I-133 7.39 E-2 1.90 E-2 2.64 E-2
<4.77 E-4
- 1. 35 E-2rn m<3.18 E-4
<3.58 E-4 I-134 7.07 E-1
.7_99 E-1
<1.91 E-1
<1_na E-1 1.86 E-2ni m 1.51 E-2
<3. 59 E gn I-135 1.68 E-1 1_54 E-2 1_1s E-?
<n_n7 E-4
<5.91 E-4 4.58 E-4
<4.61 E-4 Xe-133
<6.49 E-1 6.57 E-1
<l.39 E-3 4.78 E-3 2.73 E-3 1.28 E-3
<3.58 E-4 Xe-135 2.40 E-2 1_40 E-1 1_ns E-7 5.16 E-3 4.20 E-4 1.92 E-4
<1.95 P-4 Sr-91 5.11 E-2 1.70 E-2 2.8 E-2
<1.61 E-3 c7.61 E-4 6.93 E-4
<6.88 E-4 Sr-92 1_53 E-1 1_2s E-2__
7_s9 E-7
<2.51 E-4
<1.95 E-4 1.67 E-4
<1.46 E-4 Fb-99 1.84 E-2 3.27 E-1 1_01 E-? m m<2.85 E-3 <l.59 E-3 1.50 E03
<l.56 E-3 Tc-99m 7.44 E-2_
1.94 E-2 1.33 E-1 4.30 E-2 2.93 E-2
?_n9 E-7
<?_q6 E-7 Total Iodine 1_03 E-0 2.09 E-1 4.16 E-2 8.06 E-2 5.28 E-2 4.88 E-2 No-219' 4,61 E-2 3.79 E-2
_2.9 E-1 3.83 E-2 9.91 E-3 6.24 E-3 1.O E-2 w
Y*
Gross B 8.63 E-1 4.65 E-1 7_34 E-2 3.77 E-2 2.81 E-2 3.12 E-2 Gmss n ND 1.88 E-5 3.87 E-4 5.11 E-5 6.35 E-5 2.79 E-5 pH 6.9 6?
6.3 6.25 6.55 5.94 annrine t-ivi t-v n2
_ _0.15 0_35 O.27 0_4 0_Sn Sus. Solids 230
<20
<?n 760 lan
<?n Cl-
<?n 34
<20
<20
<20
<20
O '~
RE7CIOR WATERWYSES - DEC. SCRAM CONT.
~
O TABLE 4 5:12 8:30 8:45 Pe N,
12--19-78 12-20-78 12-21-78 pCi/ml I-131
'4.92 E-3 5.16 E-3 9.09 E-3 I-132 5.19 E-2 6.28 E-2 9.46 E-2 I-133 6.35 E-2 7.13 E-2 1.19 E-1 i
I-134 4.75 E-1 6.31 E-1 8.1 E-1 l
I-135 1.48 E-1 1.53 E-1
<3.6 E-3 4.
<8.34 E-4 1.06 E-3
<l.18 E-3 Xe-135 1.84 E-2 3.03 E-2 2.65 E-2 Sr-91 3.19 E-2 3.91 E-2 7.77 E-2 opn Sr-92 1.22 E-1 1.39 E-1 1.02 E-1 opn Mo-99 1.28 E-2 7.71 E-3 5.41 E-2 s
Tmm 2.96 E-2 1.24 E-2 9.17 E-2
'Ibtal Iodine 7.43 E-1 9.23 E-1 1.03 E =0 Np-239*
2.22 E-2 2.24 E-2 1.78 E-2 w
kE Cmss a 3.79 E-1 6.96 E-1 8.93 E-1 Gross n 7.02 E-5 1.71 E-5 8.02 E-5 pil 5.98 6.1 6.03 Conductivity 0.33 0.15 0.15 Sus. Sol W 670
<20 110 cl-
<20' 31 30
=
d a
5 a.. mm
REPCIOR MTER ANA
- JAN.,SCImM l-15-79 0 3552 SIARIUP.- 1-18-79 9 1848 TAPTE 5 8:26 1:50 9:00 9:00 5:56 8:00 8:15 PAPNETER l-15-79 1-16-79 L-17-79 l-18-79 l-19-79 1-20-79 1-21-79 uCi/ml -
'l.31 E-2 5.99 E-3 2.69 E-3 2.28 E-3 1.34 E-3 6.71 E-3 6.65 E-3 I-132 2.34 E-1 1.73 E-1 1.69 E-1
<l.4 E-1 9.75 E-2 9.17 E-2 1.04 E-1 I-133 1.48 E-1 4.01 E-2 9.54 E-4 1.57 E-2 n1 m9.48 E-3 8.38 E-2 8.45 E-2 I-134 9.66 E-1 1.01 E-Im m8.43 Ef%" 1.37 E-3 9.68 E-2 6.90 E-1 7.27 E-1 l
I-135 3.05 E-1 3.14 E-2
<l.09 E-3
<1.04 E-3 2.95 E-2 1.74 E-1 1.79 E-1 Xe-133
<2.11 E-3 4.76 E-3
<l.15 E-3
<9.84 E-4
<6.44 E-4 5.95 E-3
<l.36 E-3 l
Xe-135 3.51 E-2 9.98 E-3
<4.4 E-4
'4.24 E-4 4.32 E-3 1.89 E-2 2.61 E-1 i
.9 & 91 6.78 E-2 5.19 E-2
<1.24 E-3
<1.3 E-3 5.8 E-3 4.36 E-2 5.4 E-2 Sr-92 1.78 E-1 1.28 E-2 3.37 E-2 m m 6 E-Inun 2.13 E-2 1.67 E-1 1.71 E-l' Mo-99 1.70 E-2 1.82 E-3 3.56 E-3 8.82 E "
8.71 E-3 1.71 E-2 1.06 E-2 l
Tc-99m 1.79 E-1 1.41 E-1 1.06 E-1
<l.25 E-3 3.24 E-2 4.82 E-2 7.43 E-2 Tnt n1 Tnaim 1 67 rco 7_5 E-1 1_71 E-1 1_58 E-1 2.35 E-1 1.05 E-0 1.10 E-0
@ 239' a nn v-?
3.73 E-1 1_89 E-1
<2.73 E-3 2.95 E-2 1.95 E 2.33 E-2
.m
~?
l Gross 8 1.07 B-0 7.03 E-1 2.98 E-1 1.95 E-1 1.59 E-1 7.09 E-1 7.33 E-1 gross a ND un 2.69 E-5 3.95 E-6 2.72 E-5 2.65 E-5 ID I
pII 6.25 6_1s 6_n 6.1 6_91 6.2 6.25 cnnaiv.H vi,-v n ian 0.175 0.2 0.14 0.16 0.12 0.112 t
Sus. Solids
<20 400 500
<20
'330
.<20 980 Cl-
<20 36
<20
<20 43
<20 24 1
a i
i i
i l
t
- o.
N
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~
REACIOR MTER MMLY
. -- SCP5M 2-6-79 e 1110 STARIUP 2-6-79 0 2300
~
8:46 8:23 4:55 9:00 i
PARNCIER 2-5-79 2-6-79 2-7-79 2-8-79 i
uci/ml -
i I-131
'6.5 E-3 6.8 E-3 3.5 E-3 6.85 E-3 1-112 1.3 E-1 1.45 E-1 1.8 E-1 1.11 E-1 L
I-133 8.0 E-2 8.17 c-2 2.6 E-2 8.19 E-2 I-134 7.1 E-1 7.24 E-1 2.1 E-1 6.5 E-1 I-135
-1.8 E-1 1.78 E-1 6.9 E-2 1.63 E-1
~
<l.33 E-3 2.7 E-2 2.11 E-3 i
Xe-135 1.81 E-2 8.7 E-3 2.03 E-2 I
Sr-91 3.69 E-2 1.3 E-2 3.3 E-2 l
Sr-92 1.2 E-1 5.2 E-Z 1.2 E-1 Pb-99 2.0 E-2 1.6 E-2 1.0 E-2 T&99m 1.0 E-1 6.9 E-2 7.8 E-2 1
Total Iodine 1.11 E-0 1.14 E-0 4.92 E-1 1.01 E-0 Np-239*
2.45 E-2 1.35 E-1 3.4 E-2
. g b
do Gross 8 8.35 E-1 9.09 E-1 4.04 E-1 8.03 E-1 Gross a ND 1.51 E-6 1.17 E-5 1.48 E-6 nir 6_Os g_s 6.1 5.8 nnnrine tivity
- n. _w; D.122 0.2 0.14 Sus. Solids 40 280 170
<20 Cl-50
<20 40
<20 8
e O
y*
m m
uw.
4 STTsCK DTdA - DECDER SCRAM:
Scram 12-13-78 0 1G51 Startup 12-18-78 0 1521 Start Stop Release Rates Q
Stack Sample (S-101-78 9:30 - 12/11/78 I-131 0.41 pCi/sec I-133 1.25 I-135 17.5 (.46 pCi/sec) ttal 'htch. Spec % 11.5%
(S-102-78 9:30 - 12/11/78 8:30 - 12/15/78 I-131 0.34 I-133 1.59 (.48 pCi/sec)
I-135
.22 h tal Tech Spec % 11.94%
(S-103-78 8:30 - 12/15/78 8:17 - 12/19/78 I-131 0.096 I-133 0.102 (.100 yei/sec I-135 0.489 Total Tech Spec % 2.5%
45-104-78 8:17 - 12/19/78 8:10 - 12/21/78 I-131 0.43 I-133 0.94 (0.59 pCi/sec)
I-135 2.43 e
2 tal Tech Spec % 14.8%
O e
4 4
4 e
s
\\
e O
TABl.E 7 2.A-9 s
.-w.
Scram Jan. 15 0 1552 Startup Jan. 18 @ 1848 O
. 8,1979 - Jan 12 -
n1 Jan I
.23 pCi/sec I
~;*)$ pCi/sec (0.34 pCi/sec)
S-5-79 I
.72 pCi/sec 2btal Tech Spec 8.58%
f33 Jan. 12, 1979 - Jan. 16
.39 y see 5
S-6-79 I
(0.51 pCi/sec)
Scram 1-15-79 Ubtal 7bch Spec 12.75%
131 Jan.16,1979 - Jan.19 1
.19 pCi/sec If35
.07 pCi/sec (0.19 pCi/sec)
S-7-79 I
.01 pCi/sec on line 1-19-79 Ybtal Tech Spec %
4.7%
l31 Jan. 19, 1979 - Jan. 22.
I 1.04 1.33 (1.09 pCi/sec)
S-8-79 1.41 Ubtal Tech Spec %
27.35%
O s
I TABLE 8 W
'2.A-10 I
e --
-n.
m e
February 1979 Scram Stack Data Scram 2-6-79 9 1110 Startup 3-6-79 6 42300 12-79 g31 0.45 pCi/sec 1-30-79 to 2-2-79 3 33 0.98 pCi/sec
(.46 pCi/sec) i 3 35 0.78 pCi/sec
% Ibch Spec 'Ibtal 11.5%
S-13-79 3 31 0.24 pCi/sec 133 2-2-79 to 2-6-79 I
0.78 pCi/sec (0.25 pCi/sec) l g5 0.83 pCi/sec
% 'Ibch Spec 6.3%
l S-14-79 l
2-6-79 to 2-9-79 3 31 0.32 g33 1.12 (0.33 pCi/sec)
Scram 2-6-79 3 35 y,y7 on line 2-7-79
% h ch Spec 8.26%
,S-15-79 p31 0.48 2-9-79 to 2-11-79 3 33 0.81 (0.43 pCi/sec) h 0.91 4
% h ch Spec 10.76%
O 4
e 4
h 1
e TABLE 9 2.A-11
~
.96
i e
4 May Scram - Stack Data Scram 5-2-79 9 1350 i
O s-40-79 F
0.241 to 5-1-79 d33 1.59 (0.26 pCi/sec) 3 35 2.05 4 Tech Spec 6.5%
i i
S-41-79 dh 4
5-1-79 to 5-3-79 0.32 pCi/sec
[g35 scram 5-2-79 0.85 pCi/sec 0.47 pCi/sec
% Tech Spec 8.3% -
t l
S-42-79 d
0.0858 3
3 5-3-79 to 5-7-79 I
0.0218 (0.087) l g35 Not detected 1
% Tech Spec 2.2%
)
l
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t
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4 iO~
4 I
i i
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t
+
9 w
TABLE 10 4
. O f
i
. i 2.A-12
.y.,,,,.,
,v.w.
=
w r
2.B.
CORE WATER LEVEL ANALYSES O
The previous section described the extensive activity surveys which were conducted to determine that no fuci damage had occurred as a result of the event. No activity indications above normal levels were observed. This section outlines the analyses 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 Icycis. Even with the simultaneous
, recirculation pump trip, the pump coastdown retains an increasing flow to power ratio until natural circulation flow is established. Thus, the critical power ratio stays above the operating level prior to the scram.
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 maintained above the core. Therefore, the objective of the analyses is to
~( }
determine the minimum mixture Icvel above the active fuel region.
There are two basic evaluations which may be performed to determine the core mixture Icycl. The first method is to conduct a mass and energy balance on the core accounting for the bolloff 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 allocation process to distribute the available-mass through the system depending upon known volumes along with known-levels or known thermodynamic conditions.
~
Both of the above approaches make use of 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 1cyc1 indication (172 seconds) is utilized as an initialization point..At that time, water levels in both the downcomer 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.
9
- 2.B-1 r
Analyses of the minimum water Icyc1 were calculated by both Exxon
(
).
Nucicar Company and General Electric Company. The Exxon evaluation included both the bolloff approach and the mass allocation technique. The General Electric analysis determined the water Icyc1 by the boiloff process. Details of the GE analysis are shown in Appendix 1.
A summary of the results of the various analyses are shown in Table 4
2.B.I.
As can be seen, each evaluation concluded that the core remained adequately covered by a two-phase mixture Icvel throughout the course of the event. Therefore, no loss of fuel integrity would have been expected as a consequence of this event and indeed no indications of fuel failure were observed.
e O
O 4
5 e
g 2.B-2' 1
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O TABLE 2.B.1 RESULTS OF WATER LEVEL ANALYSES OYSTER CREEK SCRAM OF MAY 2,1979 Exxon Analyses
- GE Analyses
' Initial Mass Inventory Above Core 7.12 ft.
(two-phase misture)
.31,600 lbs.
Type of Analysis Inventory Boiloff Boiloff'
(
N Minimum Mixture Level Above l
?
Top of Active Fuel 1.62 ft.
2 ft.
2.38 ft.
k Time of Minimum Mixture Level 29 min.
20 min.
12 min.
L t
- Preliminary results, final report in process.
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. Operator action specified by Procedure 501 in response to the l
VESSEL LEVEL TRIPLE LOW annunciator has been revised to include lessons i
]
learned from the Oyster creek scram of May 2,1979.
j 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-j fication and if instructions in the procedures could have untributed to o
the incident under investigation.
(
Standing Order #23 related to the operation of the isolation con-i j
densers has been deleted. The required actions necessary for manual ini-tiation of the isolation condensers have been incorporated into plant.
O Procedure 307.
n j
All Standing Orders have been reviewed to ensure that they do no't j
impede or preclude the automatic operation of an engineered safeguard system.
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The procedures reviewed as a result of IE Bulletin 79-08 and the Oyster Creek scram of May 2,1979, are listed below. The considerations that were reviewed against are noted.
Procedures for which a change request has been submitted are marked by an 3
asterisk.
NOTES PROCEDURE NOS. AND TITLES
- 204.1 - Scram Recevery.
+
e*301 - Nuclear Steam Supply System
]
b,c
- 307 - Isolation Condenser System 4
b,c
- 317 - Feedwater System i
b,c
-*318 - Main Steam System and Reheat System a,b,c, e*501 - Annunciators and Alarms
)
a
- 502.1 - Loss of 230 KV Lines
,r. L a
502.3 - Loss of 4160 Volt Bus 1A (IB, 1C, 1D)
./
a 502.4 - Loss of the Reactor Protection System Power q
a 502.5 - Loss of 125 V D.C. Power a
- 502.6 - Complete Loss of AC Power a
- 503 - Instrument Air Failure la 504 - Service Water Failure a,
c 505.1 - Recirculation Flow Increase a,b.c
- 505.2 - Recirculation Flow Decrease a
506.1 - Rod Drop 1
a 506.2 - Loss of CRD Hydraulic System a
506.3 - Abnormal Control Rod Motion 506.4 - Rod-to-Drive Coupling Failure i
a a,b,c, e*506.5'- Scram System Failure s
a, c
507.1 - Reactor Building Closed Cooling WateriSystem Failure
- 507.2 - Turbine' Building Closed Cooling Water System Failure a,
c T
s a
508 - Loss of Vacuum s
g O-so9 - taadvertent oPenns of Turbia 8xPass valven) y 3.A-2
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- 510 - Turbine Trip a,b.c.d *511.1 - Feedwater Pump Failure a
- 511.2 - Condensate /Feedwater System Rupture i
i a
511.3 - Feedwater Flow Control Failure t
a 511.4 - Loss of Feedwater Heaters a
512.1 - Loss of Generator Excitation i
512.2-GeneratorExcitationEquipmentMalfuncIion a
a 512.3 - Loss of Generator Stator Cooling i
a 51't -
Generator Trip i
a,b.c.d.e*5% -
Reactor Isolation Scram a
515.1 - Small Piping Leaks in the Turbine Building 1
]
a 515.2 - Small Piping Leaks in Reactor Building a
- 515.3 - Small Piping Leaks in Drywell i
i a
- 516.1 - Main Steam Line Rupture Outside Drywell Qa l
- 516.2 - Piping Rupture Inside Drywell, Offsite Power Available j
a
- 516.3 - Piping Rupture Inside Drywell with Loss of Offsite Power 1
l
.a
- 516.4 - Isolation Condenser Line Break Outside Drywell j
a
- 516.5 - Piping Rupture Inside Drywell with Loss of Offsite Power and j
One Diesel Generator Inoperable a
- 517
- Significant Increase in Off Gas Release Rate i
a 518
- Inadvertent Liquid Release to Discharge Canal j
a
- 519
- Loss of Containment Integrity i
a 520
- Hurricane.
~
a 521
- Hazardous Condition on the Refueling Floor
]
a 522
- Inadvertent Poison Injection a
523
- Condenser Tube Leakage a
524
- Cleanup Filter Cake or Demineralizer Resin Breakthrough
,O a
- 525
- Loss of Drywell Cooling' 3.A-3
- 1 i
a 526.1 - Fire in Plant Areas Other Than Control Rcom a
- 526.2 - Fire in the Control Room Oa
- 527.1 inadvertent Reisef vaive Actuation Whii. at Power a
- 527.2 - Failure of Relief Valve to Reseat - Reactor Scramed
- 527.3 - Loss of Faedwater - Electromatic Relief Valve Failure a
a 528 - Tornado a
- 529 - Emergency Containment Purge 530 - Loss of the Reactor Shutdown Cooling System a
a 531.1 - Loss of SRM Instrumentation a
531.2 - Loss of IRM Instrumentation a
531.3 - Loss of APRM Instrumentation a,b,c d.e*532 - Automatic and Manual Reactor Scram a,b.c
- 534 - Loss of Reactor Cooling Mechanisms During Reactor Shutdown
'535 - Inadvertent Reactor Criticality a
i a
536.1 - A0G System a
536.2 - Radwaste Service Water 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 537.1 - Radwaste Building Closed Cooling Water Failure a
a 538 - Off-Gas Explosion 539 - Response to Malfunction of Meteorological Instrumentation a
e*603.3.001 - Recirculation 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 lo Lo Water Level Functional Test e*636.2.001 - Diesel Generator Automatic Actuation Test J
3.A-4 h
'l
NOTES:
Oa-aeview considerations:
1.
Are the operators directed to override automatic action of safety systems except when continued operation would result in unsafe plant conditions?
2.
Is primary containment isolation on automatic initiation of core spray prevented or hindered by this procedure?
j 3.
Is inadvertant, undesired pumping, venting, or release of radioactive liquids 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 determine plant status?
b.
Review Considerations:
1.
Does this procedure contain instructions which could be interpreted to direct closure of recirculation loop isolation valves in all loops?
c.
Review Considerations:
O i.
Does this procedure require aeditionai caution statements to adequateiy 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-recire pump trip?
e.
Review Considerations:
1.
Does this procedure require revision to adequately reflect the recirculation pump trip modification?
2.
Did the instructions in this procedure contribute to the May 2,1979, Oyster Creek incident?
3.
These procedures have been reviewed to ensure that adequate guidance to operators on mode switch changes following a transient have been provided.
l O
3.A-5
_+
m
3.B OPERATOR TRAINING lO The training planned as a result of the Oyster Creek incident of May 2.1979, includes both short-term and long-term training. Short-term training has two parts. 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 anticir*ted as a result of the incident. The second part is a detailed
, review of the specific operating and emergency procedure changes resulting from the incident. Long-term 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 include nroblem solving, simulated drills, and abnormality response sessions. To address conflicting indications, the operators will O
receive training to help them better analyze plant conditions, verify the most conservative 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 following the approval of the procedure
~
changes.
The long-term training is planned to include a continuing emphasis on abnormality response.
O
~
e 3.B-1
J 3.C SURVEILLANCE CilANGES The following steps and actions have been taken to minimize the likelihood of an instrument surveillance test initiating a similar event.
4 a.
An investigation has been initiated into the possibility of replacing t,he existing excess flow check valves with more' cuitable device that would not require valving and will allow for continuous monitoring.
b.
The instrument surveillance valving techniques will be j
reviewed with the instrument technicians prior to q
i startup in order to ensure an understanding of proper I
valving, and the notification of surpervisory per-O
- i.
sonnel of procedure deficiencies.
l c.
The per.formance of surveillance tests will be evalu-1 4
ated on a case by case basis when either startup transformer is out of service.
7 d.
The possibility of procuring an analog instrument i
system that will not require any valving in i
order to do surveillance testing is under inves-tigation.
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Platic covers have been placed over the control switches for i
the suction and discharge motor operated valve s.
This will require
).
the operator to lift the cover prior to actuating the closure of the l
l valve and minimize inadvertent closure of all five suction or discharge valves in the recirculation lines.
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4.A Test Program for Startup and Power Ascension A.
General 4
1 1.
This section provides a brief description of the startup test
,I program. The testing will cover prestartup, startup to hotstandby, and power range testing. Table 4.A-1 is the j.
startup checkoff sheet that will be used during the testing.
.,i B.
Prior to Startup l-1.
Prior to reactor startup, an interference. check will be performed j-on all control rods. This check will ensure that all' rods will i
stroke fully using " Normal" drive pressure, and ensure that all rods will drive in on a scram signal. This testing will ensure j
that no gross core distortion exists.
-C.
Normal Startup
~
i, 4
1.
During the normal startup, the following comparisons and analysis j
will be performed.
I a.
A comparison of the estimated Critical Position with actual j'
critical configuration - ( > 173*F).
~
b.
An analysis of the reactor coolant after heatup to
,- (
250*F.
c.
An analysis of the reactor coolant after heatup to 500 psig.
l d.
When primary pressure is approximately 1000psig, a performance check will be made of scram times on Group 1 and Group 2 rods.
l 2.
After reactor startup, with reactor power level raised to approxi-p mately 20% rated power, analysis will be' performed on the reactor coolant and off gas.
3.
As reactor power is increased from 20% to > 98%-the following analysis will be performed:
}
Coolant I'odine Offgas
- j.
- p 30%
- X 4~
40%
X
-X l
-50%
X I
60% -
< X X
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1X 80%
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Criteria During Startup to Full Power 1.
O to 50% of rated power level a.
If the reactor coolant I-134 or I-135 concentrations or stack release rate exceeds those measured at full power prior to shutdown, the reactor will be held at the existing level at which samples were taken until subsequent 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 > 985. of rated power level a.
If reactor 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 1
subsequent 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 I
condition until the problem is resolved.
E.
Reactor Startup Program Checkoff Sheet
.O 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.-
Stack Release Rate <1.25 times that prior to shutdown a.
Follow Technical Specification frequency of sampling analysis.*
b.
Stack Release Rate 1.25 to <1.5 times that prior to -
shutdown Augmented sampling and analysis program to detect "
any continuing failures.
. c.
Stack Release Rate >1.5 times that iprior to shutdown
~
Reduce power level to maintain < 1.5.
- d.. Stackgas. Release Rate > 2.0 tines 1
Isolate the reactor and resolve.
~
- A second sample shall be analyzed to donfirm the high values.
a o.
'4.A-2 A
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j The normal frequency of sampling and analysis of off gas to satisfy Technical Specification requirements is as follows:
1.
Samples are taken.on Monday, Wednesday, and Friday. They are analyzed for gross gamma activity after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay to determine a ratio of long lived to short lived ac-tivity.
2.
The sample taken on Wednesday is analyzed for nobic gas isotopic concentrations.
t 3.
If the ratio obtained in 1 differs from the previous analysis by more than 20%, a new isotopic analysis will be performed.
1-At full power operation the normal daily reactor water iodine analysis will be performed.
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PRE-STARTUP Supv.
PORC Interference checks on all control rods
/
O Reactor coolant analysis - total Iodine, Cl, conductivity /
'v NORMAL STARTUP Supv.
PORC Compare ECP with actual critical configuration
/
Heat up to 250 F. and analyze coolant iodine
/
Heat up to 500f and analyze coolant iodine
/
1000 osio - scram tima test Grnon 1 and Gennn 2 rnds 20% rated power - analyze coolant and off gas
/
30% rated power - analyze coolant iodine
/
40% rated power - analyze coolant iodine
/
40% rated oower - analyze off oas 50% rated power - analyze ceolant iodine
/
60% rated power - analyze coolant iodine
/
_60% rated power - analyze off gas
/
70% rated power - analyze coolant iodine
[
(()
80% rated power - analyze coolant iodine 80% rated power - analyze off aas
[
90% rated pcwer - analyze coolant iodine
[
>98% rated power - analyze coolant iodine
>98% rated power - analyze off gas
' /
/
REMARKS:
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4.B SUPPLEMENT TO BULLETIN 79-08 Please refer to the response to question 4 to IE-79-08 which g
was submitted by JCP&L and add the following supplemental information:
The reactor vessel instrunentation 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.
t The liquid phase of the annulus reactor water communicates with the core area under the shroud via the reactor recirculation lines while the steam phase in the annulus connects to the shroud through the steam -
4 separatcrs which connect to the top of the shroud 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 level (mass equivalent) within the shroud over the core level is accurately indicated because liquid equalization occurs 4
through the recirculation piping. Additionally, the General Electric re-4 qponse (Attachment 1), verifies that the recirculation flow rate with only one loop open is sufficient to prevent bciloff from re-ducing water level within the shroud and the reactor will function under normal natural recirculation flow conditions.
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4.D ISOLATION CONDENSER OPERATION
{
The isolation condenser line break sensing system is-intended to initiate l
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 i
sensed, high condensate flow rate conditions exist upon isolation condenser initia-tion. This is due to a surge of cold water stored in the condenser condensate return piping and tube bundle.
Since density correction is not perfonned on the dp signal, the cold higher density water further amplifies the sensed dp. A time
, delay is incorporated in the 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 would occur within one minute of a true lirie break event.
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 perfonned in November 1972 (report attached) indicates that the normal operating dp across the condensate line flow sensors to be 6-10 inches of water.
However, the initial transient dp can exceed 60 inches of water and remain above the isolation setpoint of 27.5 inches of water for 30 seconds from the time of initiation. The testing perfonned in 1972 was with 5 recirculation pumps in service at minimum flow (- 4.8 x 104 gpm).
It should 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.
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4.D-1 J
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4.C EVENT RECORDER OPERATION DURING THE MAY 2,1979, SCRAM Description of Event Recorder Operation Any of 60 signals switch the' event recorder from slow speed operation 1
(3/4 inch / hour) into high speed operation (6 inches / minute). Simultaneously, two (redundant) three minute timers are initiated.
The event recorder remains in high speed until both of the timers have timed out, at which time slow I
speed operation resumes. This sequence is independent of signal status.
This sequence of events may be modified by operator action as follows:
a)
Moving the on/off switch ~from on to off to on resets the initiation logic.
If all 60 signals have reset, the event recorder will return to slow speed 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.
b)
Opening the front door of the recorder stops the forward movement of the chart, regardless of the chart speed.
The logic and drive power to the event recorder is-from continuous in-strument Panel #3. The power to the recording styluses is from the 125 volt DC system.
Recorder Operation on May 2,1979 All aspects of the event recorder operation were normal following in-itiation on May 2,1979. The initiation of the triple low level sensors at 172 seconds is approximately the same time as the completion of the three minute timing sequence.
It is evident that the on/off switch was placed in the off position by persons unknown.
The event recorder was. discovered to be in the O
off-position at approximately 2000. hours of the same day.
O w.d 4.C-1 M
A review of available information pertaining to reactor scrams since 1972 O demonstrates that when the recirculation pumps are tripped and the isolation condensers are placed in service, inadvertant isolation does not occur. This information is presented in tabular form as follows:
3 a
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'O Isolation Condensers Rx Thermal Recirculation Placed in Service Recirculation Isolation Power Flow in GPM Pumps Tripped Condenser IsolatG
-Scram No./Date (MWt)
(Rated = 16 x 104 gpm)
Yes No Yes No
- 53 April 13, 1972 1857 15.9 x 104 A and B were initiated.
/
/
gpm (LoLo Level)
"C" failed to trip.
4
- 72 September 25, 1974 1920 15.8 x 10 gpm A and B initiated.
/
/
'f 78 May 4, 1976 1765 14.9 x 104 A and B initiated.
/
/
gpm (LoLo Level) 4
- 83 September 9, 1978 1310 16.0 x 10 gpm A and B initiated.
/
/
- 86 December 13,1978 1917 16.0 x 104 A and B initiated.
/
/
gpm 141 Press
- 89 February 2, 1979 1920' 15.4 x 104 A and B initiated.
/
/
gpm (Lolo Level) 4
- 90 May.2,-1979 1895 14.8 x 10 B initiated.
/
/
(Hi Press) i 4.D-3 L
}
In reviewing the past scram data, it was never found that when the recircu-()1ationsystempumpstrippedthattheisolationcondensersisolated. As presented in the tabulation above, five times since 1972, the isolation condensers were initiated without isolation when accompanied by a trip of the operating recirculation pumps.
It is therefore concluded that the isolation condensers will function as intended when accompanied by a trip of the operating recireulation loops.
It is noted that the isolation condenser automatic initiation signals are the same as those for automatic recirculation pump trip.
O m
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4.D-4
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ISOLEMON [ONUSER IEST s
nQ Date:
November 14, 1972'
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Purpose:
,1.
Detemine the nagnitude of the steam and condensate leg d/p spikes 2.
Detemin~ the duration of the above d/p spikes e
3.
Calculate condenser heat renoval capacity, if possible Initial Conditions:
'lhe reactor tas in the "RUN ICDE" at a power level of approximately 400I&t.
Reactor level control was in the automatic node and reactor 7
pressure was maintained at 1010 psig by use of the bypass valves under
~
~
the control of the EPR.
Recirculation flow tas at the minimum value (20 Hz).
The APRMs were set such that 100% was equivalent to 1200 IMt.
[ Results:
Each isolation condenser was put into service twice and the resulting transients were recorded.
Figures 1-A and 1-B are the d/p vs. time plots
. for the. condensers. Each figure contains the d/p's for the steam leg elbow and tir condensate return elbow for both Run #1 and Run #2. The pertinent test results are found in Table 1 and co. nents follow:
For Run #1, the condensate return leg d/ pts for both condensers exceeded the range of the recorder (0-40 inches of H 01 and also exceeded 2
the range of the local d/p gages (0-60 inches of tater). This magnitude for the spike is normal when an isolation condenser has been in the
' standby condition for a period of time prior to its being put into service.
For this condition, the condensate return leg tater tenperature is cool and the driving' head is relatively large even though the condensate return isolation valve may leak slightly. With this valve not leaking, the d/p spikes would probably reach g:' eater magnitudes than observed.
The steam leg d/p for Isolation Condenser "B" vas not recorded due to a loose electrical connection at the reccrder, which accounts for the missing plot on Figure 1-B.
Each condenser tas in operation for about 6 3/2 minutes for. this run.
..*.pJ For Run #2 all recorded parameters remained on scale. Since Run #2 was performed a relatively short time after Run El, the condensate return legs had not cooled ccepletely to their normal "in-starxiby" te:rperatures.
(Tnere is no tenperature indication on the return legs) hence, the lower tragnitude
.' d/p spikes.
i
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An atte:rpt was made to deternine the heat capacity of the
- Iso.htion Condensers. The mthod available to perfom the
.:'O:
. level when-the -Isolation Condenser is in operation. At the ca1cu>atioa re11e= ca the ^ra4 siste-to a' ee=1=e cae 9o er e
low power level and non-nonral core power distribution cxperienced during the tests, the APRM System could not be
~
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relied upon to provide enough accuracy to make a good determination.
j Results of the attempts gave indication that the removal capacities
.g were anywhere from 2.7% of rated power to 5 9% of rated power. 'Ibe
. condensers are designed for a heat removal capacity of approximately 3% or rated power.
' It is recoamended that this pcrtion of the test be performed at higher power levels (75%) so that the APRFa and other plant parameters will provide more reliable infomation. If this is done, lowever, the heat removal capacity under designed conditions
'will still not be detemined since the recirculation pumps (not
-- 7: -- running for the design conditions) will be at near rated capacity w.
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//[/f-/72.
ISOLATION CONDENSER Test PATE:
ynsy A Un 7A Llu si.6 Urs:TS Ru.n L R09 2.
Run i Run 2 70 m'n.
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4.E LEVEL INSTRUMENTATION
())
Table 4.E-1 (attached) details the indication and recording caphbilities as well as auto functions for all reactor vessel level instrumentation installed at Oyster Creek.
In Addition, the GE/MAC narrow range level signals A and B input to plant performance computer.
The narrow range GE/MAC indicators (2) and recorder are the only level j
monitors that have unique units of level. Their range is zero (0) to eight (8) feet which matches quantatively with the span of "T0TAL FEEDWATER FLOW" signal (zero to 8(x 106) lb/hr) which shares a common recorder chart and scale.
~
Consideration had been given to change narrow range level units to inches of H 0 to coincide with units of all other level instrumentation, and to that 2,
l extent, parts to modify indicator, recorder, and feedwater controller level
[)
setpoint scale had been ordered.
4 e
/~)
LJ 4
4.E-1
e e
G TABLE 4.E'-l
SUMMARY
OF REACTOR VESSEL WATER LEVEL INSTRUMENTS h
ACTUATION OR CONTROL DENSITY SENSOR INDICATION ANALOG QTY.
DESCRIPTION ID N0.
LOCAL REMOTE RECORDER RECORDER FUNCTIONS COMPENSATION 1 - Core Spray Init.
" Low-Low" RE-02A Yes Yes 2 - Cont. Spray Init.
Level 3 - Reactor Isolation Compensated Indicating RE-02B Yes Yes 4 - Containment Isolation during calibration 4
Switches No N
for conditions 5 - Recirc Pump Trip (Yarway)
RE-02C Yes Yes 6 - Isol. Cond. Init.
of operating 7 - SGTS Init.
temperature and RE-02D Yes Yes 8 - Annunciators pressure.
RE-05A Yes No Yes*
,, Low,,
1 - Low Level Scram
- Level 2 - High Level Turbine 4
Indicating RE-05/19/
Yes Yes Yes*
Trip & Target Relay No Switches 3 - Annunciators (Low Level )
(Yarway)
RE-058 Yes No Yes*
?
RE-05/19E Yes Yes Yes*
N RE-18A Yes Yes
,, E "-lU"' lU",
Setpoint is 1 - Auto Depressurization compensated for RE-18B Yes Yes System Initiation weight of steam 4
Ind cat ng No No Switches 2 - Annunciators above variable leg (Barton)
RE-18C Yes Yes and temperature of reference 109 RE-18D Yes Yes Narrow 1 - Feedwater Control 2
Range 2-Annunciator (Hi/LowLeve !) Auto density comp.
ID-13A Yes N
throughout range Level 3 - Feed Pump Runout of level and pressure, (GE/MAC)10-13B Yes Protection Reset j
Wide Range (GE/MAC)
ID-12 No Yes No None NOTES (1):
Variable legs of sensors listed 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 operator action, the follow-ing modifications are being considered:
1.
Develop sequence of events recording capability which would provide event recording for a period of hours.
This might be done by use of a panel alarm automati: log-ing recorder along with a plant computer capable of moni-toring more plant parameters.
Evaluate a wiring modification that would require the
.2.
C.R. operator to hold the control switch for-both re-(N circulation pump suction and discharge valves for the O
two minutes required to fully shut the valve; or electrically interlock the recirculation pump suction and discharge valves to prevent closure of all valves simultaneously.
3.
Review what can be done to minimize instrument surveillance testing which might cause a reactor scram while a startup bank is out of service.
4.
Continue the engineering review of the desirability of g
i installing a modification that provides solid state sensing devices to replace existing mechanical devices for trip actuation.
()
4.F-1 i
I-
-r g
i 1 mad
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 surveillance test.
6.
Change the scale readouts on the two existing level inst. readouts in the control room (Yarways and GE/MAC recorder) so they are equivalent, and in-dicate water level above the core.
~ 7.
Investigate ways to improve the reliability of 4
)
the feedwater system.
i i
8.
Provide control room indication of low-low-low i
j water icvel.
i 9.
Provide overload bypass switches for the feedwater
(}
j pumps and oil pumps in the control room.
s 1
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4 4.F-2
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APPENDIX 1 W
(
NATURAL CIRCULATION FLOW R
~
ui, H1; A.
CALCULATIONS OF IN! ITAL CONDITIONS, 9
A l Following scram and pump trips. natural circulation is established. The h
natural circulation core flow rate is of the order of 107 lb/hr and un-j evaporated water spills over into the downcomer. As the discharge valves are closed, this flow will decay to approximately 200.000-230,000 lb/hr.
j :
In addition. control rod drive cooling water is available. The ninimum 2
flow at which drainage of water starts occurring in the separators cor-I responds to the situation where the inlet flow to the core cannot make n
up the boil-off of steam.
j At 3 minutes into the trar.sient, 11
).
1895 m + 0.0333 (May-Witt)
Core power
=
63.1 N
=
325,000 lb/hr Evaporation rate
=
Flashing in core 38,000 lb/hr r
p Total W9 35 4,000 lb/hr
[
Leakage flow from bypass region (1 psid)
[
= 840,000 lb/hr (thru channel-tieplate leakage path).
Core inlet flow 840.000 + 250,000 = 1,100.000 lbs/hr -
The variatim of the flow rates and void fractions 'vs. elevation are y
sketched below. The vapor flow rate increases in the core corresponding to power input vs height. The inlet flow is boosted by the leakage from g
9 the bypass. At-the top of the core the leakage flow is returned to the 1
y bypass. This forms a natural circulation loop between the core and the bypass. The net liquid flow in_ the upper plenum and separators is thus the' difference between the core exit flow and the leakage flow.
- 3. C/
3-f a
1 3W j
y 1.
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at pu 4
=y y?
S The void fraction increases with height in the core. 4he large area and
'(
low vapor velocity 11 the upper plenum lead to a lower void fraction in j
the upper plenum. The vapor accelerates in the separator standpipes pro-d ducing a high void fraction.
The void fraction in the three regions are calculated belo.<:
a
,il J
5F# tac
-l W(
SurentoRS y
- 4
, - n w
j; 2
2
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UPPER A
COPI PLENUM SEPARATORS _
.' ~
(exit) 4 t
?
W9 (1b/l,ti) (includes
'364,000 443.360 452,000 flashing) 5, m.
W lb/hr) 736,000
-104.000
-114,000 f
?)
3
- Drift Flux C
1.3 1.0 1.3 0
0.82 ft/sec 1.0 0.82 Parameters 7gj 5l Cross sectional p
I area v 56 ft 186 30.0 5
W g_
N 2900 ft/hr 1173-7100 L
j
=-
'9 hP /
^
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4 g
l0
~-
W
-\\
jf f
278 ft/hr
-11.8,
-8.0
=
i AP
' e' d
9 i
jaj g+j 3178 1112
^ - -
7020 9
u i
3 i -
a=
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q 0.41 0.24 O.59
- 3600-(
i JC + p9J e
w
- i o
i I
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'~
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,3'
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j s \\ ~'
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\\ A the i; ore'.inle?,(lyx has(tietrensed below the vaporizal'longte, the level the sepayator[wi jj fall. t he swinkff pressure, p Icularlydurigs, T
4
. ypressurizativD,~Taaf accc) crate this process.
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T 4
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, Sefore the level falls, the total static heads inside the shroud are:
t Above BAF: 13.2 * ( 1-0.20) + 5*(1-0.25) + 9' (1-0.59) = 14.31 ft, f
Above TAF: 1.2 (1-0.4) + 5* (1-0.25) + 9* (1-0.59) = 7.44 ft.
hg Water inventory inside shroud before separators drain:.
Volume Msss (_l,b_),,
Bypsss:
614 26.955 Core (including unhested region 835 28,490 aboveTAF):
Upper plenum:
783 29,780 h
Separator standpipes 186 3,561 2
O f
In order to get the Triple Low Level Alarm, collansed level cust drop to 5.b' f t above TAF. Using the region void fractions this implies that two phase level F
must be at 2.6 ft into the separator Standpipes. This corresponds to approxi-gI mately 31,600 of fluid above the active core.
F g.'
i
[
Note: BAT - Bottom of Active Fuel TAF = Top of Active Fuel i
e e
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[
J
z u
h B.
Cfd.CjlLATIO!! 0F NATURAL CISCULATION FLOW 1
s O The #aturai circuieti#a rio
<ro= the do "ce er to the core d ne"es aa
- 1 the static heads inside and outside the shroud:
5
(
3.
4 N
W f
i!l L
e j
3 f
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A g,
7
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6
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The static head inside the shroud will be history dependent based on the net Water flow into the core region.
However, based on the results (to follow) r that inflow is of the same order as the boiloff, the level is expected to stay near the middle of the upper plenum. The void fraction will gradually
.)
decrease inside the shroud as the power falls.
However, the two values of recirculation flow chosen should bound this effect.
- . gj-i,
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+
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y g
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- ]
Pressure drop in the external loop ACB is controlled by the resistance j
in the five 2" bypass ifnes. This pressure drop must be balanced by the pressure gain along BDA.
fu
\\l E
O ACB "
l
- bypass, (feet of water) f where A = flow area of 2" line (schedule 80) (assteed) 2.953 'in
=
!b u
120 ft(for4 elbows) j K
=
bypss Crane
+
8fg(forgatevalve)
(P Handbook 4
j
+
72 fg (pipe friction)
F f
= 0.019 g
1 J
Exit loss
=
4.8 K
=
Substituting these values.
A Atg W" (lb/hr)2 n
(1) 10 1
(ft) 0.407
- 10
[
t Inside the vessel.
d v
[
~AP
~ b II-"l h (2)
BDA
- Ndwnc e r i
g i = tore, upper plenum, separators.
Assuming the void fractiors calculated earlier and an average level at the midddle of the upper plenum, H
U.0 (5)
OEBDA =
downtomer i
v s
9 6
wa. - -
,,m-~
~ new,v
- e-
O A
Level in the downcomer ranges from 136" to 170" above the top active fuel (i.e.
280" to 310" above BAF).
- Using. AP
=
w,2 20
,,,o7 3o gg, 33,o) g,)
t O For the low downcomer level of 280" (23.3 ft).
205.000 lb/hr W,
t r
For the high downenmer level of 310" (25.85 ft).
W 229.0D0 lb/hr
=
r Tlie uncertainty in this calculation is primarily the static head inside the shroud.
flowever. Since a smaller static head produces a larger recirculation flow. compensating effects are introduced.
In any event. the recirculation flow (together with CRD flow) should be sufficient to make,up the bo11of f after the first few minutes.
]
- Mnre accurate interpretation of the level accounting for density changes and flashing in the downcor'ar beltne the oressure tap location leads to a variat.1on of 27D" to 310" above LAF.
4 em 4
.l
?
-x
()
I 1,nyentory Inside the Shroud i
l The inventory calculation has been improved by consideration of subcooling of the lower plenum, flashing and stored energy effects.
thss balance inside shroud:
i 1
M-Win ~ N90ut Win - Natural recirculation flow + CRD flow 1
J Wgout = Steam outflow due to flashing and evaporation t
I t-l(em INin-W9aut)dt (2)
Mt'Hinitial
- ti
)
3 1
h M - Total mass inside shroud at time t t
I Minitial : Total mass inside shroud at 3 minutes g
The region inside the shroud is composed of four regions-j
'1.
Lower picnum and control rod guide tubes. This region will j
remain single phase liquid, but change in temperature.
I 2.
Core region 3.
Bypass region 4
Upper plenum (and separators).
la j g j o a
lj f
tr 1
8
3'jl 41
- 14.
l
'\\
s L,'
3
!)
]
lower Plentra and Guide Tubes j
+ 0
(
- ((
- N h
~
in in out out
)i Total internal energy in lower plenum E
a enthalpy leaving lower plenum h
out 1[
0-heat from vessel walls
'7 Various assusptions can be made about h If the plenun is perfectly out.
.J mired, h cortc5 ponds to the average niene tenperature. The worst out 1
case for the level calculation in the upper plenum (i.e. the one that i
yields the lowest level) is to assume perfectly stratified flow. The j (,)
outflow enthalpy in this case remains the initial enthalpy until the f;,
initial rass in the plenum is entirely replaced.
1 Y
(4) 5 kp LP P
e
if
Yh f
(5) d M
=
W,-Wcut 9
on Approximating internal energy by enthalpy, equation 3 becomes
- 4 jn in-4'out o d40 (6) 9
( 7)
- W h
h Substituting (5) into (6)
= Win (hj n--hout)+0 II) '
IM e (h I
out for H 27 raindes O
tg = hinitial
$n
=
7 y (,
in 1!
p k
- f 0
q.
!3 4
4 61
.g
^
a3 R.1
].
Table 3.1 shows the calculation of lower plenum temperature and mass using 9
this pro::edure. The heat from the vessel wall was esticated to be fairly lj small and neglected. This is conservative for the level calculation.
?
j Tw3-Phase Regions _
]
Fro 6n the energy equation, vapor generation rate in a particular region
.~
A pf (1-a) dhf P I^
3 04 V(1 p a dh9-V h
de (8) m fg J
g dp N
n.
N In the core both negative and positive P ar.4 used; in the bypass and upper f
plenum only depressurization is considered.
1 b
The steam flow leaving the core and bypass regions (and generated in the ui O
- 9eer 9'ea ") is c 'c=1 *ed s==i"9 auest-staticcoaditicas (5 e d>>
small).
dt
$I 1
k
( r yg + apg d_vo, p } Ag (9)
W g
g g
dp i
The ste3n flow leaving the shroud
,i l
- 90ut " 9
- N9 bypass
- N9 upper plenum
(
}
Core
}
l The mass in each region is calculated as below:
j w gr y +nPg T dvc. P L 9
A
'(11) j j
a
=
exit 9
(12)
.1C + V~gj o
/
E- 0.5 (ixit + " inlet)'
M=
Ep +. (1-li) P -
'V -
(14) g f
-' M 7s W is.nr h *ek 4 ti
^r=
.a..
b E
I O t
t Table 3.2 lists the system parameters used in the calculations. The pres-sures and pressure rates were obtained frora the data from the site. The decay heat factors correspond to the Itay-Witt curve.
It is estirated that stored energy release from the core is less than 5% of the decay power after the first three minutes. Since the May-Witt values are conservative by 15-23; compared to the new best estimate ANS Stan6;. decay heat values more than compensate for stored energy effects.
Using the data in Table 3.2 the steam flow rates and void fractions were calculated as a function of time for the various regions. as given by equations 8-14.
These are tabulated in Table 3.3.
Tabics 3.1 and 3.3 provide sufficient information for a mass balance using Equation 1.
t O
1 hen. 8,,,,,,,,,,,
M, - n,,, -- s,,, s s c
t.
-M plenum (15) 1oer Table 3.4 shows the calculation of total mass and upper plenum mass, using equations (1). (2)
~nd (5). Two values of recirculation flow at 200.000 3
3 and 230,000 lbs/hr were used. At the higher flow, a minimum mss of 20,000 1bs (33" collapsed level) was reached at between 9 and 1S minutes. Beyond e
this time, the inflow is able to overcome the effects of vapor generation and increased density in the lower plenum.
For the lower flow, a minimtra value of approximately 16. BOO lbs (?a" collapsed level) was reached at 11-23 minutes. An increase occurs there-
- after, e
The variation in mass and collapsed level is shown in Figure 1.
s.
~
i 1
- g. -- ~. -. :
-, ~
, -..,. ~......
s.
W ue o
.A kl
,4 ;-
.O w
Table 3.1
.a 4
1 power Plentra Mass Calcul_ation
'li h = 4583 (hin-hout)40 (BTlj/lb/ min) 1.521x10b ~
)
(using II - 1.43x105 lbs. dP =-0.04188 between h = 512 to 356 BTU /lb 6
dii
?
l h = 77 3TU/lb.
h hout in calculated by weighting enthalpics of recirc and 1
CRD flow )
a i
Time h
h U -
h Ti V
kss in out t
4 BTU / min 1
3 3
410 51 2
-3.07 512 0.02091 1.341x105 f
4 410 512
-3.07 509 0.02085 1.345x105 I) 5 41 0 512
-3.07 506 0.0208 1.348x105
)
6 410 512
-3.07 503 0.0207 1.355x105 1
7 410 512
-3.07 500 0.02057 1.357x105 A
(
8 410 51 2
-3.07 496.5 0.0206 1.362x105 l
9 410 512
-3.07 493'.5 0.0205 1.362x105 f
10 410 512
-3.07 490.5 0.02048
- 1. 369x105 I
12 382 512 S
-3.92 484.5 0.0?039 1.375x105 I
l 14 362 512
-4.52 476.7 0.02025 1.385x105 1
16 340 512
-5.18 467.6 0.02005 1.398x105 4
z
]
18 330 512
-5.48 457.2 0.01988 1.411x105
)
20 320 51 2
-5.78 446.2 0.01957 1.426x105 a
il 25 313 512
-5.00 417.3 0.01924 1.458x105 r
[:
30 305 410
-3.28 387.3 0.0188 1.492x105 5
33 30) 410 377.5 0.01867 1.502x10 1
I
+
x__
i k
a-e..-
.m, r
Ef"Y.32ESSWrWp&l;;M;ikW**M.6*N"?" ^ 'A *"*U"3"Wi'W~yOI55$fETESQif@5555
, %'.id.194ki@lbi,L%d;MMMM NU8WEhE
- P
%G.:54 M Lah
\\
Ir.1 cem a
mU TABLE 3.2 Based on 1895 WT Wg = 10000d/1.054hfg i-f Tine P
i' Qd Qd h rg Wg 7
(min)
(psig)
(psi /sec)
(pu)
(nM )
(Btu /lb)
(1b/sec)
T p
l
?
3 920
-1.3 0.0346 65.57 660.9 94.13 4
850
-1.3 0.0326 61.78 674.6 86.89 I
5 850 1.3 0.0310 58.75 674.6 82.63 f
6 920 1.3 0.0294 55.71 661.0 79.96 7
1000 1.3 0.0280 53.06 647.5 77.75 l
[
8 950
-1.65 0.0272 51.54 657.0 74.43 h
9 830 1.65 0.0263 49.84 680.5 69.49 I
30 780
-1 0.0255 48.32 690.6 65.38
[
12 730
-0.23 0.0244 46.24 700.8 62.60 14 710
-0.23 0.0234 44.34 704.9 59.68 l
16 i 680
-0.23 0.0226 42.83 711.1 57.14 I
18 650
-0.23 0.0220 41.69 717.7 55.11 1
g 20 620
-0.23 0.0215 40.74 724.2 53.37 25 760 0
0.0204 38.69 694.6 52.85 L
30 620 0
0.0192 36.38 724.2 47.66 f
33.33 510 0
0.0184 34,89 749.3 44.18 O-
r :,..M w;.;;;.M... ~..,..<
,...-...R. ;:; p.n_M_ _.,....,~.>,,.,_.,. :"?..lg,Ri,,",,!,,t*"*M,.%,
b p
,,,,,,a, l'1
.e e in 1". 4 kd pjv C
TABLE 3.3 t!:h El flashing Rate f
Upper a
a a
H M
M r$
Tim Cove Byor.ss Plenun core Bypass up core By
'uD
[j (ain)
(lb/sec)
(1b/sec)
(lb/see)
(Ib)
(lb)
(lb)
- n
~
O
~
D[3 (3) 10.7 9.9 12
.21
.065
.25 26,600 26.955 31,590 y
4 10.7
- 9. 9 12
.21
.065
.25 26.600 26,955 12[j 5
- 10.7 0
0
.20 0
.24 26,900 29,570
- f; 6
- 10.7 0
0
.19 0
.23 27,236 29,570 8
L; 7
- 10.7 0
0
.19 0
.22 27,236 29.570 y
8 10.7 0
12
.20
. 055
.23 26,600 28.430 9
10.7 0
0
.20 0
.21 26,609 29,300
- s M /~T (10) 6.4 4.7 5.8
.18
.04
.19 28,290 28,430 t,j V 12
- 1. 5
.97 1.1
.17
. 01
.17 28,620 30,000
}
{}
14 1.5
.97 1.1
.17
. 01 16 28,620 30.000
~
(16) 1.5
.97 1.1
.16
.01
.15 29,590 30,000 s
10 1.5
.97 1.1
.16
. 01 15 29.590 30,000 y
20 1.5
.97 1.1
.15
.01
.14 29,913 30,000 ig 25 0
0 0
.15 0
.13 29,913 30,296 4
T (30) 0 0
0
.14 0 -
.13 30,236 30.296 i
1 33.33 0
0 0
.14 0
.13 30.236 30.295 i
a]
e k
,n Ik i
a v
4 u
18 l
llt I4 l
L
N
(([
Nj g
?f0 b
TABl.E 3.4(a) 1/
- 74 Oyster Creek Inventory Calculation - 1__
.IIf Recirc Flow = 230,000 lb/hr oM f[
Time out in M
M "up i
(min)
_(ib/sec)
_(1b/sec)
(1b/sec)
(lb)
(1b) lh S
3 126.7 79.2
-47.5 219,245 31,590 p
4 119.5 79.2
-40.3 216,6111 28,256 s
- 1l'
5 71.9 79.2 7.3 215,620 23.650 6
69.3 79.2 9.9 216,130 23,824
- s
~
7 67.1 79.2 12.1 216.790 24.284 i
8 97.1 79.2
-17.9 216,442 25,512 O
j 9
80,2 79.2
- 3.0
-215,876 21,156 8.
]
10 83.3 79.2
. 4.1 215,726
'22.106 12 66.2 79.2 13.0 216,222 20,140 14 63.3 79.2 15.9
_217,956
'20,874~
l
- )
16 60.7 79.2 18.5 220.020 20,658 18 58.7 79.2 20.5 222,398 21.708 f
20 56.9 79.2 22.3 214,958 22,455 25 52.9 79.2 26.3 232,218-26,249 30 47.7 79.2 31.5 240,928 31,194 33.33 44.2 79.2 35.0 247.561 36.830 k
~
a t-IS m.
7.__., _ -...
. - :. : =.
.. m.
~w E
i;d h;
al o 1ABLE3.4(b) y Oyster Creek Inventory calculation - 2 Recirc Flow - 200,000 lb/hr
(
Tir.c out in k
M up N
N (min)_
(1b/secl (1b/sec)_
,(lb/sec)
(1b)
(1b)_
h.
d 3
126.7 70.9
-55.8 219.245 31,590 pf 4
119.5 70.9
-48.6 216,113 27,758 E..
5 71.9 70.9
- 1.0 214,613 22,643 f
6 69.3 70.9 1.6 214,631 22,325 e
7 67.1 70.9~
3.8 214.793
,22.287 y-8 97.1 70.9
-26.2 214,121 22,891
@ O 5'. 5 7.9 1.4 213,377 18,657 10 83.3 70.9
-12.4 213.047 19,426 9
12 66.2 70.9 4.7 212,585 16.465
]?.
14 63.3 70.9 7.6 216,792 19.672 u,
36 60.7 70.9 10.2 217.860 18,470
!)
IB 58.7 70.9 12.2 219,204 18.514 20 55.9 70.9 14.0 220,776 18,263 w
d 25 52.9 70.9 18.0 224,076 18.067 S) a 30 47.7 70.9 23.2 230,256 20,524 9
33.33 44.2 70.9 26.7 235.241 24,503 s
g d
1 50
~
~
w y
aft 4
"2
M l E.V E L AGWC C.O R E (IN C+1E.S { ~"
e, a
.r a
no Ln n
en m
j y
.3 g
j>e d
..l..
.j!1 d._
I
.d.,I j
y
- q..
g
... ) e.
3 z,
l
..! w:
..)
.o I
j
...I.,i,
,s
.i
.m 3
n.
1 l
D l
I; w...
i q
l O,-
n
--l
_i i
a t
- o. j t
j
..._.I t
4 4
i
.o
. (*' __.
i 1
o, t
ra...
3 o.
{
e i
l
!. - ---- ',l.
l
.g' r.-
q i
i j
IM
^
'I l
st j I
i j
+
j i
a
+
l
)c i3.s i
e t
i b M.
i i
l i
i, -
i N
i "a,
i i
i 2_
i s
v
- 4 C) i 1
J3 l
I
.c a
I I
~
9.
,, q
. 7.. _._..
i
__....._4 1.i 1, i
i i
i i -._
l i
P..
i.
I f
i i..
t.
4
_a.
i 1
\\.
a i
t i
l' i
1 g.
4 i
3 c.._. 4 b
- n
.4 j
i 6
}6 N
l a
p i
i i
?\\
t i
i 1
i s l
\\
l-I
'M (,,
t 1 q.. -.
3 i
i i
h.
l I
i g
i i
i s
i N
. i
~!
e g
t.u I
I
.g.-
- p...
j j
1 s
-[.
I
'tl.
i l
l l
1 j
i i
r, "F'I F
.o i
q_
1 J.
s, c
i.
i l.
4 3 3
.j
.l
,i
- u..
.l 3
.i...
j ay_
i
.a, i
i
.A
....,i.
=
q.
4.
1.s 3
-1 l...
y l
t a
qs-
.. N_. j.
s.
/. u)..__.
Q I
4 l.
l
..l.....
y
_e i
.a
-,1 g7__
C t. _ -
a 4,
i,_.
._...}
( }..........
.-r-
_1 l
_..f. j t.. y,
_. __..l.
9 i...
=.
g_ "'
i 1
. 1....
.g l
....g
.wc p.4 j
m
.g O
.M.
e n v
.~
,a)
N
_oo 1
I t
, a'
)
Ovster Creek s
~
~ Addendun
- )
Cafculation of Natural Circulation Flow with 1 (one) External L)op Open C,
4 Utilizing the saurs rathods of sections A + B, the natural circulation flow M
rete can be calculated by:
a I
= (W /(S A)
(Eloop+ K pump /2E)
ACB r
l33 vbern:
s AP
=-AP
=H
- 19.0 (ft.)
ACB LiDA dcr.mcomer d
j ua Flow through open loop,1bs/hr 4
r
{ = density of water in loop - 47 lbs/ft d r~N b (>
d A = 3.33 ft (26" schedule 80 pipe) y 1.25 vessel entrance + crit losses d
Y
=
b op a
!5
+1 20 5 c3 bows 40.50 2 gate valves
-10.801 floc ele.snt s@
40.25 straight pipe.
L 4.0 total loss coeff. (Crane WRr values) lg:
=
,l
= 21.0 fired rotor (vorst caso vn free rotor)
Y K
- f recirc pua (from Byron-Jacimon Test Data)
'\\j;e r
= 25.0
' Total Yb s
6 9
19.0 ft = atatic head inside shroud'st core flow of % 2 x 10 lbs/hr Q.
Solving For U :
1,,- -
r
? C' 1/2
- H.0)1/2 W *fA II EIE IU p
cotal d
r H
o.a5 x 10" (n - 19.0)1/2 II-u-
d 7
6 j4 m
h
1 W
W O
'For tbs low downcomer level of 280" (23.3')
g.
l 6
W
- 1.76 x 10 lba/nr r
b L
For the high downcomer level of 310" (25.8')
t i
W 6
2.22 x 10 lba/hr
}'
r =
g 1
I Since the other four loops would be unpplying fluid through the 2" bypass linea..
L the ratal fics to the core venid be:
r 0
6 r - 1.76 x 106 + (0.8)o.205~x 10 W_ - 1.93 x 10 1ba/hr N
g L
9 6
6 T 2.22 x 106 + (0.8) 0.229 x 10 V
= 2.40 x 10 lbs/hr
[
3 N
Thus the rceirculatico fim rate (which is about 5 to 6 times bolloff rate)
(j O
-its 11 e 1aor r - 1
i=1*=t ta > veat
- 11-orr 'r =
d=c$"a -^ter h
level within the shroud and the reactor will function under normal natural
?
cireviation flow conditims. With additions) loops open, this flow rate wonid I
'd be such greater (spproxinstely 2 x for 2 loops. 3 x for 3 loops, etc) sud thus
{
g provide even greater margin to boil off.
l E
- (
r 6
t y
i U
k
. 1 i,,
J i
g 5
r s
I C, v
? :.
g}
,- u g-(1 !.
- - _ _ _ _ _ _ _ - - _ _ _