ML19305A847

Transient Assessment Rept,Reactor Trip on 800226, Preliminary Rept

ML19305A847
Person / Time
Site:	Crystal River
Issue date:	03/09/1980
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19305A844	List: ML19305A842 ML19305A847
References
07-80-02, 07-80-02-R2, 7-80-2, 7-80-2-R2, NUDOCS 8003180495
Download: ML19305A847 (126)

Text

{{#Wiki_filter:---- Transient Assessment Report Reactor Trip at Crystal River 3 Nuclear Station on February 26,.1980 Prepared for Florida Power Corporation I Babcock &Wilcox &T5"TITONE

TRANSIENT ASSESSMENT REPORT REACTOR TRIP AT CRYSTAL RIVER - 3 NUCLEAR STATION ON FEBRUARY 26, 1980 (PRELIMINARY) CAUTION: THIS IS A PRELIMINARY REPORT. IT IS BASED ON INCOMPLETE DATA AS AVAILABLE TO US THROUGH MARCH 9, 1980. THE INFORMATION AND CONCLUSIONS HEREIN ARE PROVIDED SUBJECT TO FURTHER VERIFICATION. Prepared for Florida Power Corporation By Nuclear Power Generation Division The Babcock and Wilcox Company Lynchburg, Virginia March 9, 1980 Report No. 07-80-02, Rev 02 l! Prepared By: _i s / 1, o E~ Reviewed By- ' V eviewed By: Reviewed By: [ ~ '2'i' r //.{ Reviewed By: / es J

1. 8 y

y Babcock & Wilcox Company

TABLE OF CONTENTS Page I. EVENT SYNOPSIS 1 II. PERFORMANCE EVALUATION AND RECOMMENDATIONS 5 A. Expected Plant Performance and Deviations 1. Initiating Cause Assessment 5 2. Loss of Instrumentation Function 10 3. Opening of PORY and Spray Valve 12 4. Control System Actions Before Reactor Trip; 15 Reduction of Main Feedwater 5. Reactor Shutdown 17 6. Initiation of HPI 17 7. RC Pump Trip 17 8. Closure of PORY Isolation Valve 17 9. Control of HPI 18

10. Safety Valves Relief 19

11. Steam Generator Cooling 19

12. Core Cooling 21

13. Restoration of Pressurizer Pressure Control 23

14. Restart of Reactor Coolant System Pumps 24

15. Release of Radiation to the Reactor Building 25

16. High Sodium in Primary System Water 26 B.

Safety Implications 30 C. Conclusions / Recommendations 32 III. EVENT DETAILS AND INPUT DATA 34 A. Initial Plant Conditions 34 B. Plots of Major Parameters 34 IV. COMPONENT TRANSIENT ASSESSMENT 82 A. Reactor Vessel 82 B. Reactor Vessel Internals 82 C. Steam Generators 82 D. Pressurizer 83 E. Reactor Coolant Pioing 84 F. Reactor Coolant Pumps 84 G. Control Rod Drive Mechanisms 84 H. Safety Valves 84

1.. Fuel 85 V.

REFERENCES 86 l

TABLE OF CONTENTS (Cont'd) Page APPENDIX A - INSTRUMENT STATUS A-1 THRU A-5 TABLE II SIGNAL VALIDITY TO COMPUTER, CONTROL ROOM INDICATOR, HOT SHUTDOWN PANEL, AND ICS TABLE II VALID OPERABLE ALARMS TABLE II VALIDITY OF NNI, DIGITAL INTERLOCX AND CONTROL FUNCTIONS FOR RC MAXEUP AND PURIFICATION SYSTEMS APPENDIX B - ANALYSIS OF CORE SIGNALS B-1 THRU B-7 APPENDIX C - SEQUENCE OF EVENTS REPORTED BY FLORIDA POWER C-1 CORPORATION - REVISION 5 THRU C-9 l

GLOSSARY OF ABBREVIATIONS dp Differential pressure del ta p del ta T Differential Temperature AT AFW Auxiliary Feedwater EFW Emergency Feedwater ESFAS Engineered Safeguards Features Actuation System FSAR Final Safety Analysis Report HPI High Pressure Injection ICS Integrated Control System Loop A, B Reactor Coolant Loop A; side with pressurizer connected to the reactor outlet piping. Loop B is the opposite side. LPI Low Pressure Injection WW Main Feedwater MT Main Turbine NNI Non-Nuclear Instrumentation OTSG Once Through Steam Generator PORY Power Operated Relief Valve RB Reactor Building RC Reactor Coolant RCS Reactor Coolant System RCV Reactor Coolant Valve RPS Reactor Protection System SEM Sequence of Events Monitor Tave Average Temperature of the Reactor's Inlet and Outlet Temperatures Tc Reactor Inlet Temperature Tcold l l .=

GLOSSARY OF ABBREVIATIONS (Cont'd) Th Reactor Outlet Temperat.e Thot X Power bus which supplies 120 VAC and plus and minus 24 VDC to NNI instrtmentation and control systems Y Power bus which supplies 120 VAC and plus and minus 24 VDC to NNI instrtaentati )n t i f

i I. EVENT SYNOPSIS On February 26, 1980 Crystal River-3 Nuclear Station experienced an automatic reactor shutdown. This synopsis of key events and parameters was obtained from the plant coguter's post-trip review and plant alarm sumary, the sequence of events monitor, control room strip charts, and the shift supervisor's log, as well as the sequence of events prepared by the Florida Power Corporation. Prior to the incident, the reactor was operating at approximately 100*. FP with Integrated Control System (ICS) in automatic. No tests were in progress and minor maintenance was being performed in the Non-Nuclear Instrumentation (NNI) cabinet "Y." 14:23:20* The upset began with interruption of one of the two 24 Volt DC (-25 sec)*** power supply units. This supply (called the "X" supply) powers much of the plant control instrumentation, transmitters and recorders. As a result of the failure, a number of erroneous alarms and indications were received in the Control Room. Erroneous signals were also supplied to the Integrated Control System (ICS) which controls the Reactor Power level, Main Turbine Steam flow, and Main Feedwater flow. Alarus recorded at this time indicated that the reactor inlet temperature was erroneously indicating approximately 5700F (instead of a normal 5550F) and that Reactor Coolant System (RCS) flow indication in one loop was erroneously indicating I one-hal f normal flow. These erroneous signals caused the ICS to start reducing feedwater flow. Simultaneously, an erroneous indication of average RCS temperature caused the ICS to begin increasing reactor power. 14:23:21* The NNI "X" 24 volt power supply monitor tripped the 120 volt (-24 sec) AC breakers supplying power to the 24 volt power supply. The Power Operated Relief Valve (PORV) on the pressurizer opened and was held in the open position by the power supply failure. The pressurizer spray valve was also opened slightly. Erroneous signals supplied to the ICS caused the feedwater to bcth Once Through Steam Generators (OTSG's) to be rapidly reduced, Steam Flow to the Main Turbine to be increased, and the reactor power to be increased. The combined effect of the feedwater to the OTSG being reduced and the reactor power being increased raised RCS pressure. 14:23:30** Main Feedwater (WW) flow rate was decreasing and RCS pressure (-15 sec) was increasing rapidly.

• Eastern Standard Time as Recorded on the Sequence of Events Monitor.
  • Times are from the 855 Computer alarm printer. Time difference between the Sequence of Events Monitor and 855 Computer is + 5 seconds.
• '2 Times related to reactor trip l

l..

Event Synopsis 14:23:45** The reactor tripped on high RCS pressure at 2300 risig. No (0 sec) alarms were available on the annunciator printer because it was,not functional at this time. The Main Turbine (MT) was tripped at this same time by automatic circuit response. The RCS pressure peaked at approximately 2320 psig and started to decrease due to post trip cooling and the open PORY. 14:25:50* The Reactor Coolant Drain Tank level alarmed high due to the (65 sec) accumulation of RCS coolant being released through the open F "V. 14:26:41* RCS pressure decreased to 1500 psig and the Emergency (176 sec) Safeguards System activated the High Pressure Injection (HPI) System which started two additional HPI/ makeup pumps. The full flow of three pumps was now being injected to the RCS. 14:27:04** The four Reactor Coolant Pumps (RCP's) were turned off in 14:27:07 accordance with procedures. Some reactor building service (199-202 sec) penetrations were isolated manually by the operator in accordance with procedures. HPI flows were balanced. 14:28-14:32 The PORY and pressurizer spray line isolation valve were (25E-495 sec) closed by the operator in accordance with procedures. This (Note 1) permitted RC pressure to rise as high pressure injection water flowed into the system (time approximate). 14:31:32* Reactor building pressure reached 2 psig. Reactor coolant was (467 sec) being released to the reactor building through the rupture disk on the Reactor Coolant Drain Tank. 14:31:49 OTSG A Rupture Matrix activated and tripped main feedwater (484 sec) pump 1A. OTSG A was dry at this time and OTSG B was nearly dry. 14:32:35 Steam Driven Emergency Feedwater (EFW) pump was manually (530 sec) started. 14:33 Motor-drive Emergency Feedwater (EFW) pump was manually (555 sec) started (time approximate). 14:33:11** RCS pressure 2361 psig. (566 see) 14:33:15 RCS code safety relief valve opened (time approximate). 14:33:30 (570-585 sec) 14:34:33* Reactor building dome high radiation alarm. (648 sec) (Note 1) At end of this section I

Event Synopsis 14:44:12* NNI "X" 24 VDC power supply re-energized. This provided (1227 sec) reliable indication of the primary and secondary parameters. 14:44:31* Reactor building pressure at 4 psig and the reactor building (1248 sec) isolated on automatic signal. Sodium hydroxide tank valve opened admitting NaOH to Decay Heat Removal System. 14:45 B&W notified of reactor building pressure and radiation (approx.) alarms. B&W established emergency connunications and support to CR-3 control room from Lynchburg. 14:46:10 The reactor building isolation was bypassed to allow injection (1345 sec) flow to the RC pump seals, balance HPI flow, and restore essential component cooling water. 14:48:24 Seal water and cooling water re-established to the Reactor (1479 sec) Coolant Pump seals. 14:51:57 The Steam Rupture Matrix on Once-Through Steam Generator (1692 sec) (OTSG) B activated because of low steam pressure and tripped MFW Pump 1B. The B OTSG water level was about 70% on the operate range. 14:52 HPI was throttled to approximately 250 gpm. (approx.) 14:53 Re-established letdown flow to reduce RCS pressure to aid in (approx. ) reseating RCS code safety valve RCV-8. 14:56 Bypassed OTSG A rupture matrix and feedwater to OTSG A was (approx.) re-established. 14:57:09 Bypass OTSG-B rupture matrix to regain FW control (at approximately 65% operating range). 14:57:15 Re-established RC pump seal return. 15:00:09 Water level was re-established in OTSG A. (2184 sec) i 15:15 Verified that natural circulation cooldown had been i (approx. ) established on both OTSG's with approximately 230F reactor l differential temperature. The Technical Support Center was manned. l l I - -

Event Synopsis 15:17 A Class "B" accident was declared and evacuation of ( app rox. ) non-essential personnel from the site was iniciated. 15:19 Commenced feeding OTSG B. (approx. ) 15:26 Received a low level alarm from the Sodium Hydroxide tank. (approx.) 15:49 OTSG A at high level in accordance with procedure. ( app rox. ) 15:50 Terminated HPI flow and established Makeup and Letdown control (app rox. ) of RCS pressure. At this time the RCS and pressurizer were completely filled with liouid (" solid" system). 16:00 Commenced pressurizer heatup to establish a steam space in (app rox. ) the pressurizer. 18:05 Established a steam space in the Pressurizer by increasing ( app rox. ) letdown flow. 21:07 Started Reactor Coolant Pumps (RCP) 1B and 10. RCS pressure (approx. ) was approximately 2000 psig, Tave was was approximately 4200F and pressurizer level was 235". This returned the plant to a norraal shutdown condition and terminated the abnormal transient. NOTE 1: The time of the PORY block valve closure is different from that of FPC and is based on B&W's interpretation of the supplied data. Calculations of the pressure increase on the RC drain tank show that with a 100 + 10 PSI rupture disc, the presure increase would require steam flow indo the RC drain tank for c. least 4 minutes after the PORY opened. The exact block valve closure time does not, however, significantly influence the course of the transient and has little effect on event interpretation. i

I II. PERFORMANCE EVALUATION AND RECOMMENDATIONS A. Expected Plant Performance and Deviations 1. Initiating Cause Assessment The distribution of essential AC electrical power at the Florida Power Corporation's Crystal River 3 nuclear plant is shown schematically in Figure II-1, adapted from the Final Safety Analysis Report. The principal control instrumentation, the "Non-Nuclear Instrumentation" is supplied by two sources, Wil-X, and NNI-Y. Instrumentation, transmitters, recorders, indicators, etc., is assigned to one of these two sources as discussed in Section II.A.2. The Non-Nuclear Instrumentation (NNI) power distribution, system for NNI-X channel is shown on Figure II-2. A similar power distribution system exists for NNI-Y and is shown on Figure II-3. The power distribution system provides + 24VDC as required for i signal processing modules, output modules and relay logics. 118 VAC power is provided for + 24 VDC power supplies, sensor power supplies, indicators, and electric to pneumatic converters. The initiating event for this transient appears to have been a short circuit on the +24YDC bus in the NNI-X channel. The power supply monitor sensed this bus short and within 0.5 seconds tripped both Si and S2. This automated function removed all DC power to the equipment served by the NNI-X supply. Two sources for AC power may be brought through shunt trip switches S1 and 52 on the auctioneer panel to the AC terminals of the + 24VDC power supplies. At the Crystal River 3 plant, NNI-X AC~ source 1 and 2 is supplied 120VAC by vital bus 3C, NNI-Y is supplied 120VAC by vital buss 30. The 24VDC outputs of power supplies are brought through auctioneering diodes to form buses on the power supply auctioneer panel. The auctioneering diodes block the lowest of the two supply voltages, preventing a low or zero supply output from loading down a bus. This arrangement results in an auctioneered +24VDC positive bus and an auctioneered -24VDC negative bus. The buses serve the + 24VDC loads in NNI-X. The 24VDC system is protected from high voltage supply output by overvoltage protection networks within the individual power supplies. If a power supply voltage exceeds an adjustable limit (approximately 27VDC), the overvoltage network forces the power supply output to zero volts. This will not affect the output to i the NNI since the diode auctioneering circuit will allow the other power supply to continue to function. __

The 24VDC power distribution system is~ monitored by a power supply monitor module. The power supply monitor receives inputs from the redundant positive 24VDC power supplies (pins 1 and 2) and the redundant negative 24VDC power supplies (pins 19 and 20). These inputs are used to initiate an alarm relay contact to annunciate to the operator either a loss of one of the AC ll8VAC power sources or failure (low or no voltage) of either or both positive and/or either or both negative 24VDC power supply. The alarms originate in comparator circuits that compare the power supply voltage to an adjustable setpoint voltage. The power supply monitor also monitors the +24VDC bus (pin 15) and the -24VDC bus (pin 18, pins 16 and 17 are comon). If either the positive or negative bus voltage falls below a setpoint (approximately 22 volts), the power supply monitor will initiate the shunt trip circuits on both switches S1 and S2. Switches S1 and 52 have an opening dropout time delay of approximately 0.5 seconds. Thus, if the bus voltage falls and does not recover in 0.5 seconds, the AC power will be removed from all four 24VDC power supplies. The opening dropout time delay allows the buss voltages to come up to normal when the AC power is restored. Switches S1 and S2 must be manually reset to restore power. Based on information received from the site within five hours of the transient and verified in discussions with FPC C&I technicians and B&W personnel at the site, the FPC technicians appear to have taken proper and timely corrective actions. They verified that the power supply overvoltage breakers had not tripped and attempted to reset 51 and 52. When they could not reset 51 and S2, they began troubleshooting the circuit. This effort led the technicians to believe that the power supply monitor had failed so they removed the monitor from the circuit. They were then able to reset Si and 52. The elapsed time from loss of 24VDC power to restoring 24VDC power to the NNI-X was approximately 21 minutes. Subsequent trouble-shooting and investigation by FPC and B&W personnel revealed that the power supply monitor was not defective. The source of the shorted +24 VDC-power supply appears to have been located in a Coop B wide range pressure voltage buffer amplifier. When the power supply monitor was removed and the power supply re-energnized; the short was burned out. Each power supply is capable of supplying approximately 15.5 amps and without the power supply monitor this current could burn through a shorted component. In summary, the NNI-X power supplies and power supply monitor all functioned as expected during a short circuit event. l 1

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"Y" POWER FIGURE 11-3 NNI-Y POWER 128V AC l l IERN. BLK FUSE RC ^ PANEL SP )Si FUSE MU Y U PANEL DH 4 +24V DC -24V DC CF 83 POWER SUPPLY POWER SUPPLY FUSE RC PANEL 2 a POWER SUPPLY SP a MONiiOR (IVP OF AB0VE) s i l I l l +24V DC BUS 24V DC BUS

2. Loss of Instrumentation Function This section identifies which displays, controls, annunciators. interlocks, and signals were valid and/or aiailable to the operator during the time that the NNI "X" channel DC power supplies were tripped. Appendix A provides the detail results of an evaluation of these signals and contains the following tables: Table 1I-1 identifies the valid signals to the computer, ICS, control room and hot shutdown panel. Table II-2 identifies the valid operable alarm inputs to the control room annunciator for the Reactor Coolant, Secondary Plant and Makeup and Purification Systems. Table II-3 addresses the digital interlock and control functions performed by the NNI for the Reactor Coolant and Makeup and Purfication Systems. 2.1 NNI Safety System Interactions The NNI System does not provide any inputs to the safety systems. The safety system process sensors are powered from their respective safety system channels. 2.2 Valid Reactor Coolant System Indications Valid RC narrow range (1700-2500 psig) pressure indication from Loop A and Loop B was available in the. control room. The pressure information is displayed on recorders and is derived from the Reactor Protection System. Valid RC wide range (0-2500 psig) pressure indication for Loop A and Loop B was available in the Control Room. The pressure information is displayed on Indicators (Loop A is also recorded) and is derived from the Engineered Safety . Features Actuation System. Valid RC wide range (0-2500 psig) pressure indication for Loop A was available at the Hot Shutdown Panel. The information is displayed on an indicator and is derived from the Engineered Safety Features Actuation System. l

Valid RC outlet temperature (520-6200 ) indication for Loops F A and B was available in the Control Room and at the Hot Shutdown Panel if the sensors powered by the "Y" cabinet were selected. The information is displayed on indicators. Valid RC inlet temperature (520-6200F) indication for Loops A and B was available in the Control Room if the sensors powered by the "Y" cabinet were selected. The information is displayed on indicators. Valid RC inlet temperature (50-6500F) indication for Loops A and B was available in the Control Room if the sensors powered by the "Y" cabinet were selected. The information is displayed on indicators. Valid uncompensated pressurizer level was available in the Control Room from RC1-LT3 via the computer. Valid uncompensated pressurizer level was' available at the hot shutdo;.n panel from RCl-LT1 and RCl-LT3 and is displayed on indicators. Valid core outlet temperature (70* - 2300*F) were available from the SPND outlet thermocouples. There are 52 outlet thermocouples. The readings are collected by a microprocessor, averaged, and presented as highest, average, and lowest reading. 2.3 Integrated Control System Effects The affected ICS input signals, and the resulting ICS response, are discussed in Section II.A.4. I l.

3. Opening of the PORY and pressurizer Soray Line Valve As described in Section II.A.1 above, power was lost to the NNI + 24VDC power bus. Approximately 0.5 seconds later, the -24VDC power was remcved by the power monitor. This sequence of losing power has the potential of causing signal monitors which are supplied by the +24 VDC NNI-X Supply to energize all "High" signal relays (See Figure II-4(a)). The consequences of this action at Crystal River Unit 3 (CR-3) would be an open connand to the PORY (see Figure II-4) and, until power is restored, the circuit not be able to send a close connand. The NNI would also send an open command to the pressurizer spray valve (see Figure II-5) for the period of time from the loss of the +24VDC to the loss of -24VDC (approximately 0.5 seconds). A close connand would not be sent until power is restored. In order to verify that the signal monitors used in the NNI would behave as described, two tests were conducted, one by CR-3 and one by Bailey Control Company (BCCO). The model signal converter tested by CR-3, (identical to the one in the CR-3 NNI) was an earlier version than the one tested by BCC0 but the results were similar. In both tests, the signal monitor was. powered by + and - 24VDC and the output state was monitored. The +24VDC power was momentarily removed and restored. The -24VDC was then removed and restored. The results of 'these tests were: a. CR-3 (using earlier version of signal monitor) - when +24VDC was removed, "H" relay energized. b. BCCO - when the -24VDC was removed, the "H" relay energized. The results of these tests reveal that it is possible for the NNI to issue a false command or alarm signal if the + and - 24VDC power supplies are lost in a set sequence ("+" first on older models of signal converters and " " first on newer models). i ) ).

+24 VDC Power from fl!!I-X PORY ACTIVATic". CIRCUIT 4 SIGNALkOF]R 1 SIGNAL /

• .0T USED During normal conditions when the RC Narrow Range Pressure exceeds RC MARR0w the high setpoint, the "H" con-H/L waCt is closed. When the pres-L-

RANGE PRES 3URE sure is reduced, the "H" Contact 2 opens. 27 g, When the pressure decreases be- -24(DCPower i 1 w the low setpoint, the "L" a from NflI-X 'l contact opens. When pressure n0T USEg returns to above th' setpoint, j 'T the contact closes storing the control circuit. (a) PORY Activation Circuit Signal tionitor P,efore the loss of the f24 VDC pdL - f 20 Dower supply, the "H" Contact was H ""* open and the "L" contact was h 20 / closed. The K20 relay and PORV y solenoid were deenergized. / / ] PORV $0LEF0lD RELAY j (NOT ENERGlZED) (NOT ENERGl2ED) Simplified PORY Control Circuit (b) n fore Loss of WI "X" Russ e At the time the 424 VDC power n ;d pdL

d20 supply was shorted, the PORV signal monitor clused the "H" M20

/ relay. The K20 relay ener-V gized closing contacts to the / PORV solenoid and a " seal-in" / contact in the K20 circuits i E AY E ERG D { (EMERGlZED) l AT THE Til4E OF LOSS OF NNI X +24 VOLT SU33 (c) Simolified PORV Contro! L1rcUit was trippe(d, the"H" opened, but After-the 24 VDC power supply 7 [_ 20 N -" ! L j / the K20 relay and the PORV sole- )d.20 / noid remained energized because Y the "L" contact could not open. / / l K2o PORY 30LEN0lO l RELAY (ENERGlZED) (ENERGIZED) l i FIGURE 11-4 l AFTER LOS5 0F NNI X BU33 _ _ Page 13 .(d) Simplified PORV Control Circuit i

PRES 3URIZER SPRAY VALVE During normal operation when the RC Marrow Range Dressure exceeds the +24 VDC Power 1 high setpoint, the "H" contact closes

1. rom Mt!I-X f

NOT USE which causes the pressurizer spray ~ ?f motor valve to energize in the open direction. As pressure decreases, signat n T the "H" contact opens and the spray RC MARROW RAMG,1 MONITOR valve remains at that position. PRESSURE nft When pressure goes below the low _js setpoint, the "L" contact closes y NOT USED and the spray valve's motor is en-f ergized to close the valve. As the -24 VDC Power b-pressure raises above the setpoint, L the "L" contact opens and the sys-from Mill-X "T-tem is restored. SIMPLIFIED PRE 55URIZER $ PRAY MOTOR OPERATED VALVE CONTROL --N -L Before the loss of /24 VDC, the spray val.ve was shut and the motor 00"II0' $[n CLOSE op,erated valve control was de-(DE-ENERGlZED) (DE-ENERGlZED) energized. BEFORE LOSS OF NNI X BUS At the time of the loss of NNI's "X" t24VDC Bus, the "H" contact f L yH closed which energized the Motor Operated Valve (!10V) open control. The valve started to open. NOV CONTROL CLOSE / RGlZTD) /( *" AT THE TIME OF LO35 0F NNI *X" 24V BUS Afterthet24VDCpowerwasremoved, the "H" contact opened. This ac- -n --t tion deenergized the MOV open con-trol and stopped the pressurizer i spray valve in a partial open posi-tion. h[,CONTR3L CLOSE (DE-ENERGlZED (DE-ENERGlZED) AFTER L0s5 CF NNI X 2tiV BUS PAGE 14 FIGURE 11-5 l

4. Control System Actions Before Reactor Trio 1. ICS Signal Status The interruption of the NNI "X" power supply caused several NSS process signals to fail which are inputs to the ICS. Among the signals important to the ICS performance are: RCS Hot Leg Temperature (T I h Steam Generator Startup Levels RCS Average Temperature (Tave) calculated from RCS hot and cold leg temperatures Steam Generator Operate Range Levels Feedwater Control Valve Differential Pressure Steam Generator Outlet Pressure (AaB) Total RC Flow - calculated from individual loop measurements The failure mode of all the signals listed above is to mid range. The failed signal reading and its relation to setpoint is as follows: Setpoint or Failed Signal Normal Value Value Th 600 570 T 556 570 c Tave 578 570 Total RC Flow 100% 50% Loop A RC Flow 100% 50% Loop B RC Flow 100% 50% Turbine Header Pressure 885 psig 900 psig SG Startup Level 25 inches

• 125 inche SG Operate Range Level 50%**

50% IFW Control Valve Differential Pressure 50 psi 50 psi Steam Generator Outlet Pressure 1000 psi 600 psig

• Low load or post trip level setpoint, RC pumps on
  • Post-trip level setpoint with RC pumps off ', }

l l

In the period before and after the reactor trip (up until NNI "X" power was restored), some, or all of the NNI signals listed above, reached the failed value. The preliminary data assessment indicated the following: Signal Loop A Loco B Th Failed OK T Failed Failed c T

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Failed----------

ave Total RC Flow Failed OK Loop RC Flow Failed OK Turbine Header Pressure

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Probable Failure-----

Steam Generator Outlet Pressure Failed OK SG Startup Level Failed Failed SG Operate Level Failed OK MFW Control Valve Differential Pressure Failed OK Based on the invalid input signals in the 10 to 25 second period following the power supply disturbance, the integrated control system (ICS) initiated the following actions: (a) A rapid reduction in feedwater flow to the A and B generators, responding to the indicated drop in That and the indicated RC flow reduction. (b) A control rod withdrawal to bring T back toward its setpoint. ave Power increase was limited within the ICS to 103%. (c) Opening of the turbine control valves in response to the indicated rise in turbine header pressure. This response was consistent with expected ICS performance, given the invalid input signals. i l l 5. Reactor Shutdown l The control system's actions caused an increase in the actual RCS T , and the RC system pressure, al though the PORY was open, wnf!h led to a high pressure reactor trip at about 2300 psig some 10 a to 25 seconds after the NNI "X" electrical upset. Pressure response is indicated on the wide range control room recorder chart, Figure III-3, which is constructed from the Plant Computer Data, and the Signal Event Monitor (SEM). This response was as designed. 6. High Pressure Injection Initiation i Following reactor trip, the relief of steam through the open PORY caused RC Pressure to fall rapidly as shown in Figure III-3. The Emergency Safeguards Features Actuation System (ESFAS) initiated high pressure injection automatically as designed, when the RC pressure fell to approximately 1500 psig. 7. RC Pump Trip B&W instructions specify that the reactor coolant pumps (RC?'s) shall be tripped inmeidately after initiation of ESFAS by low reactor coolant (RCS) pressure. Low RCS pressure alarms were receivet at 14:24:43, alerting the operators that RCS pressure was decreasing. The low pressure ESFAS initiation alarm was annunciated at 1.:26:41. Two RCP's were tripped manually at 14:27:04, and the final two were tripped at 14:27:07. 8. Closure of PORY Isolation Valve B&W Small Break Operating Guidelines specify that, when the symptoms of a small break occur, the following valves shall be closed. e Letdown Isolation Valve e PORY Block Valve e Spray Block and Control Valve l These actions are specified to isolate a small break, if the break is downstream of the PORY block valve cr the letdown isolation valves or between the spray block valve and the spray control valves. The operators had received a low RCS pressure alarm at 14:24:43, following the high pressure trip. A low pressure ESFAS actuation signal was received at 14:26:41, indicating RCS pressure of 1500 psi. In addition, a high drain tank pressure alarm had also been received. This pressure decrease indicated the possibility of a small break. Letdown had been manually isolated by the reactor operator following l reactor trip in accordance with appifcable procedures. The operator then manually closed the PORY block valve and the spray block valve in accordance with the Small Break instructions..

9. Control of High Pressure Injection The High Pressure Injection (HPI) system was actuated automatically on the icw pressure ESFAS signal. Three HPI pumps functioned, providing a flow rate in excess of 1000 gallons per minute at the minimum RC pressure of 1325 psig. Pump suction was from the Barated Water Storage Tank (BWST). The HPI water was injected through the normal safety injection nozzles, in the four cold legs between the reactor coolant pump discharge and the reactor vessel inlet. Initially, the operator chose not to throttle HPI flow. The operator apparently could not insnediately determine the validity of reactor coolant signals available to him. The operator therefore opted for HPI cooling to assure the core was properly cooled. After regaining NNI power (about 20 minutes after initial power loss) th'e operator could assess plant conditions. They were as follows: HPI cooling in effect, and RCS filled with water (pressurizer e level off-scale high, RCS pressure oscillating at about 2400 psi, subcooling margin greater than 1000F). e Natural circulation in B OTSG (differential temperature of about 70*F across this steam generator, steam pressure consistent with outlet cold leg temperature, steam generator secondary water level about 50". on the operate range). HPI injection was then throttled to about 250 gpm. This is consistent with instructions allowing throttling when pressurizer level is high and subcooling margin is greater than 500F. Natural circula: ion was then established in the A steam generator by raising the secondary water level and observing the proper differential temperature ( 30*F) across the gererator. The control room pressure trace shows RCS pressure oscillating between 2250 and 2400 psi. This indicated that the pressurizer code safety valve was cycling as HPI was adding water. HPI was throttled to determine if the safety valve would reseat. By balancing HPI flow with letdown flow, it was determined that the valve had reset. Some point HPI flow rates were read by the operators during HPI injection, but no continuous recording of HPI flow vs. time is available. Based on BWST levels before and after the incident, about 47,000 gallons of HPI wacer was injected. About 43,000 gallons were discharged through the PORY and code safety valve to the RC drain tank and thence to the containment building basement. The remaining 4,000 gallons filled the pressurizer steam space, and was discharged to a reactor coolant bleed tank when the pressurizer steam space was established. f l _.

10. PRESSURIZER RELIEF VALVES Only one of the two safety valves (RCV-8) was actuated based on evaluation of discharge pipe temperature readings. The actuation pressure was approximately 2400 psig. The normal set pressure is 2500 psig. The downward drift in set pressure was probably due to the seat leakage experienced by this valve prior to lifting. Seat leakage causes the thermal expansion of the valve internals, and the thermal expansion increases the areas exposed to the escaping steam as the valve nears its set pressure. The increased areas require lower steam pressures to generate the forces necessary to overcome the closing force of the spring.

During the two hours that the safety valve appeared to be open, it passed saturated steam, two-phase flow and water. The pressure and temperatures seen by the valve are recorded in Section III. The valve appears to have reseated at approximately 2300 psi or less than a 5% blowdown below the initial opening. Except for the low opening pressure of 2400 psi vs. the 2500 psi setpoint, the safety valve appeared to perform as expected. The PORY was opened at the beginning. of the incident as described in Section II. A.3. After approximately 5 minutes, the PORV isolation valve was manually closed. During the time the PORY was open, it is likely that the PORY passed only saturated steam, based on event simulation. The pressure range experienced was approximately 2320 to 1350 psig as shown in Figure II I-4. The PORY appears to have performed as expected.

11. Steam Generator Cooling (Reactor Trio until RC Flow Restored)

Upon reactor trip, the ICS performs the following normal functions: (a) Reduce feedwater flow to that value required to maintain 25" level on steam generator startup range instrumentation (50% on operate range if no RC pumps are operating). (b) Transfer turbine bypass setpoint from 885 psig to about 1000 psig for steam generator pressure control. (c) Transfer steam generator pressure control signal from turbine header pressure to steam generator outlet pressure. __

9 As a result of the input signal failures, a preliminary evaluation indicates: (1) The selected startup range level indications for both steam generators failed to mid-position, indicating a level of about 125 inches in both generators. Since the setpoint is 25 inches, feedwater was not added to the A generator and may not have been added to the B generator. The B steam generator pressure indicates that some feedwater entered the B OTSG after reactor trip and before RC pump trip. When the reactor coolant pumps wcre tripped at about 3-1/2 minutes into the event, the ICS transferred the steam generator level setpoint from 25 inches on the startup range to 50% on the operate range. This transfer provided a valid level signal on the B steam generater, which also had a valid main feedwater control valve delta P signal. The B steam generator startup feedwater valve opened, feed pump speed increased, and feedwater was introduced at about 1400 gpm to the B generator through the main feedwater nozzles. This feedvater originated from the Deaerating Feed Tank (DFT) at about 400F. This action is believed (from incore temperature data) to have initiated natural circulation cooling in the "B" loop. Auxiliary feedwater flow to the B steam generator was manually l initiated and controlled by the operator at about 10 minutes into the event. The B steam generator level was increased to about 92% on the operate range, and then manually controlled at 50% on the operate range indication. The A steam generator most likely had an invalid operate range level signal (50% setpoint, 50% indicated), so that no feedwater demand was created. The A-0TSG rupture matrix was actuated at about 14:31:49. There is no indication of feedwater to the A generator until 14:56:43 (33 minutes into the event), when the steam line break rupture matrix was bypassed, and auxiliary feed flow was establisned to bring the A steam generator level to 50% on the operate range. 1 (2) Short-term post trip steam generator cooling control appears to have been valid in the B steam generator (e.g., control at about 1000 psig setpoint--indicatina a valid pressure measurement), and invalid in the A steam generator (e.g., loss of pressure such that the ruptare matrix was activated leading to isolation from main feedwater). l While there is no available feedwater flow indication, the main and auxiliary feedwater flow to each generator may be inferred from steam generator operate range level response as shown in Figure III-34. l 4 Automatic actuation of the auxiliary feedwater flow may have been prevented by the mid-range failure of the steam generator startup range level signals. A valid low level signal is required from both steam generators to actuate the auxiliary feedwater system, or loss of control oil pressure on each main feed pump. Since these were invalid startup level signals and at least one main feed pump continued operating, no AFW initiation signals were actuated. The operator actuated AFW to the B steam generator at about 14:32:35 (8 minutes 30 seconds after trip), although the main fee &ater was supplying feedwater through the startup valve at this time. After AFW initiation steam generator flow control and pressure control were manual. After AFW initiation steam pressure began to decrease in the B steam generator, and reached the rupture matrix setpoint at about 14:52:00, which isolated the B steam generator and initiated trip of the 8 main feed pump. Feedwater flow to the A steam generator was re-established by AFW at about 15:00:00, and natural circulation cooling was obtained in both loops. At about 15:19, AFW flow to both steam generators was manually increased to achieve 957, on the operate range level. This level [ was maintained until RC Pump restart at 21:07. I

12. Core Cooling Before and during the beginning of the transient, the core was cooled by forced convection provided by the four reactor coolant pumps. This mode of cooling continued for. about the first two hundred seconds after reactor trip. At this time, high pressure injection was initiated and all four reactor coolant pumps were tripped. The heat transfer mode from the fuel was thermal convection to the RCS inventory. The heat was being removed from RCS by venting of steam and water through the pressurizer relief system, with the HPI system replacing lost water. A limited amount of OTSG cooling was also available, as described in Section II.A.11.

Although core outlet temperature data during the first ten minutes of the transient are incomplete, the highest core outlet temperature recorded in the data is 600'F. The minimum subcooling appears to have been reached at approximately six minutes after reactor trip. From this point on, subcooling margin increased as shown in Figure III-8. Heat transfer to the RCS fluid from the core, primarily in the thermal convection mode, appears to have continued throughout the transient. In these circumstances no fuel cladding failures would be expected. For approximately seven hours, decay heat was removed from the core by natural circulation convection cooling in the subcooled mode. At the end of this period, forced convection cooling was re:umed by the restart of two reactor coolant pumps. The evidence from measured radiation levels in the containment building and in the reactor coolant letdown system and from measured radionuclide concentrations in the reactor coolant and in the containment atmosphere is consistent with the conclusion that no new fuel cladding failures occurred in this transient. This evidence is i as follows: 1. The radiation montior on the letdown line, RM-L1 (which is recorded on RM-RS) showed a continuously decreasing activity level following reactor trip. Figure III-41 shows the decreasing activity levels recorded on RM-RS. The letdown monitor has a three-minute delay line to allow for the decay of nitrogen-16 so that the monitor will be more sensitive to the activity of other nuclides. If any significant number of fuel rods (approximately 10 to 20) had their clad breached by this transient, RM-L1 would not have shown this steady decrease in the coolant activity level. 2. The radionuclide concentrations in the containment atmosphere can be accounted for based on the activity expected to be released from those fuel rods that were leaking fission product activity into the reactor coolant prior to the reactor transient. This point is discussed in greater detail in Section II. A.15. 3. The dose rates measured by the radiation monitors in containment are consistent with what would be expected based on the release of activity from fuel rods that were leaking prior to the reactor transient if allowance is made for lack of homogenity of the containment atmosphere. If any significant amount of fuel cladding failed due to the reactor transient, the dose rates in the containment would have been much higher. Section II.A.15 presents the technical basis for this conclusion. These data do not preclude that a small number of fuel rods (10 to

20) may have developed cladding leaks during this reactor transient since a small activity release would have been indistinguishable from other activity release mechanisms which follow a nomal shutdown transient or could have been obscured by calculational and measurement uncertainties.

22 -

To further demonstrate adequate core cooling, the responses of various neutron detectors at Crystal River-3 (CR-3) have been analyzed to determine if they are indicative of any localized boiling or voiding as was seen at TMI-2. The sources of these data were from the alarm printout following the reactor trip (from 1421 to 1437, 1511 to 1528, and 1626 to 1637 hours on February 26, 1980), the strip charts from back-up recorders for selected self-powered neutron detectors (SPND's) and the source-range detectors. No localized boiling can be inferred from the core instrumentation responses. A description of the core instrumentation indications and their meaning is provided in Appendix B. Based on these analyses, we conclude that adequate core cooling continued at all times throughout the transient.

13. Restoration of Pressurizer Pressure Control After high pressure injection was terminated, pressure control was achieved by balancing letdown and makeup. RCS temperature was controlled using natural circulation and steam generator pressure control. Natural circulation was verified by measuring a delta T of about 290F across the core, cold leg temperatures within 50F of steam generator saturation temperatures, and constant core outlet l

temperatures. The plant was stable, with about 1400F subcooling. A review of makeup and letdown flow rates indicated that the pressurizer code safeties had reseated. A RCS coolant sample was taken and analyzed. for radioisotopes. Fission product isotopes indicated no fuel pins had been ruptured' during the transient. This was further verified by the letdown line radiation monitor readings. Furtner, the condenser air ejector radiation monitor verified that there was no abnormal primary-to-secondary leakage. Since there was no unusual coolant activity, the obvious decision to take the plant to cold shutdown on the decay heat system was made. Several methods of achieving cold shutdown were available, as follows: e Solid plant cooldown using natural circulation or forced flow e Reestablish normal pressurizer control, then cooldown by either natural circulation or forced flow. A joint decision was made between the B&W Emergency Response Team and Florida Power Corporation to reestablish normal pressurizer pressure control, establish forced flow, and place the plant in cold shutdown on the decay heat removal system. l 1

Pressurizer water temperature was about 5350F when this decision was made. A steam space could be established by heating the pressurizer water to about 6100F, corresponding to a saturation pressure of 2000 psi. The pressurizer heaters were activated to accomplish this. Since the Florida Power Company operators had performed a similar sequence during simulator training at B&W, no difficulties were anticipated, and none were encountered. Pressurizer water temperature increased about 400F per hour with all 0 heaters activated. When water temperature reached about 620 F, letdown flow was increased (to a bleed storage tank, to maintain proper makeup tank level control) and a steam space was formed. There were no unexpected pressure or temperature swings during this evolution.

14. Restart of Reactor Coolant System Pumps RC Pump s'eal data had been retrieved on an hourly basis and reviewed by B&W. All pumps were capable of restart. Normal procedures specify one pump per steam generator loop for cooldown. The B pump was selected in loop 1, to maximize pressurizer spray flow capability, if required, and, therefore, pressure control capability. The D pump was chosen for startup in loop 2.

There was a differential temperature of about 25'F across the core, before RC pump restart. After pump start, this differential temperature went to about 1-2*F, so bulk average temperature decreased upon pump start. A run was made on the B&W simulator to mockup pump start, which indicated a sharp upward spike of about 50 psi on RCS pressure indication due to a sudden flow increase at the RCS pressure taps. The actual system pressure would not spike. A pressure decrease of about 150 psi, and a presurizer level decrease of about 20 inches, was shown due to bulk average temperature decrease. Neither condition was detrimental to the plant, but the expected response was discussed thoroughly with the Crystal River Unit 3 Operations Technical Advisor so that the operator would recognize such changes as normal. Imediately before planned pump start, low oil level alarms were received on the upper oil pots on RC pump B&D motors. B&W evaluated these alarms as thermal contractions of the oil volume, and not detrimental to pump restart. This was discussed with Crystal River Three personnel and the alarms were bypassed. The operators were instructed to closely monitor the motor upper thrust bearing temperatures, and trip the pump if this temperature exceeded 185'F. i

RC pump B was started first, without incident. The pressurizer level only decreased five inches, and little pressure change occurred. The core differential temperature decreased to about 2*F, as expected. Cold leg temperatures, core outlet temperatures, system pressure, and pressurizer level were all normal. The D pump was started without incident. Upper thrust bearing temperatures stayed below 130*F, and oil level alarms cleared. The plant was in normal cooldown configuration. High letdown was continued until the pressurizer level was about 210 inches.

15. Release of Radiation into the Reactor Building Prior to the transient, the reactor was operating with reactor coolant activities corresponding to between 3.7 and 5.8 percent of the design coolant activities, as shown in Table II.A.15-1.

This coolant activity level would correspond to approximately 17 + 3 leaking fuel rods. As' can be seen from Table II.t.15.2, the radionuclide concentrations measured in reactor coolant samples following the reactor transient indicated that all the iodine and cesium nuclides exhibited increased activity which is characteristic of the fission product spiking phenomena which occurs during normal shutdowns when a plant is operating with leaking fuel. The spike factnrs (after correcting for decay but without correcting for dilution) appear to be between 5 and 6 for the iodine and cesium nuclides. Based on about 47,000 gallons being pumped from the BWST into the reactor coolant system, the dilution factor is 1.89 (or 53% of the activity remains in the RC system) after appropriately correcting for density differences and asstaning a well-mixed reactor coolant loop (which seems to be a reasonable assumption during natural circulation with HPI flow). Thus, after adjusting for dilution, the spike factors become 9.5 to 11.5 which is a realistic range. Using the total water inventory change during the incident and assuming that 72% is discharged into the reactor building and 28% is discharged into the bleed holdup tanks, the measured activity in a containment atmosphere agrees with the activity release associated with the reactor coolant discharge and with the pressurizer steam space discharge (see Table II.A.15.3). Since there is no evidence to support 28% of the coolant being discharged into the bleed holdup tanks, the most reasonable explanation of the disparity is that the reactor building atmosphere is not completely well-mixed and that the actual concentration might be 28% higher than the sample results in Table II.A.15-3 would indicate. -

Despite the reasonably good agreement between the measured airborne activity in the reactor building and the activity relsase assuming no additional fuel failures, one might choose to disregard the concept of fission, product spiking and claim that the activity increase shown in Table II. A.15-2 is due to the failure of additional fuel rods. An equivalent increase could have been produced by small leaks in approximately 10 to 20 fuel rods which were previously intact. Since i there are 36,816 fuel rods in the core,10 to 20 rods represent only 0.03 to 0.06 percent of the total. Figure III-42 shows the dose rates measured by the radiation monitors in the reactor. building. Calculations were made to see if the release of reactor coolant into the containment would give a similar response. Based on releasing the coolant activity into the reactor building atmosphere and assuming that the atmosphere was well-mixed, the dose rate would be between 1.5 and 3.0 R/hr in any unshielded area; however, since the calculation neglected many short half-life nuclides, it might have been possible to have a slightly higher dose rate for a short time. The peak reading of 60 R/hr reading on the dome monitor (RM-G19) and its rapid decrease to under 5 R/HR (which is the minimum reading) can best be explained by assuming that the - activity was not well-mixed with the entire building atmosphere for the first several hours. This explanation is further supported by the great disparity between the four radiation monitors in the reactor building. Apparently, the steam and hot gases rose rapidly to the dome and tended to accumulate there until the steam release was terminated. The monitors that read the highest were generally located at higher elevations in the building and on exterior walls which induced convection downdrafts that swept the activity from the dome to the monitor. The reactor building isolation was effective in containing the radioactivity within the building. Thus,-there was no amount of activity detected above background in the environment external to the Reactor Building.

16. Sodium Contamination of RCS The low pressure injection system (LPI) was autoinatically actuated, as designed, when RCS pressure decreased to 1500 psi. The pressure in the RCS was high enough that the only flow in the low pressure injection system was the recirculation around the pumps. When the i

reactor building pressure increased to 4.0 psig, the valve on the 11% sodium hydroxide tank opened allowing caustic to flow Into the LPI System. Tank readings indicated 1.8 feet or 615 gallons of 11% l caustic left tt.t tank before the tank outlet valve was closed. The reactor coolant did not become contaminated with the sodium hydroxide because the Reactor Coolant System pressure was higher than the dead head pressure of the LPI pumps as indicated by an analysis of an RCS l sample taken at 05:15 on February 27, 1980. This sample indicated that sodium level was only 1.8 ppm. . l

TABLE II.A.15-1 COMPARISON OF REACTOR COOLANT iCTIVITIES PRIOR TO TRANSIENT WITH DESIGN BASIS x"TIVITY IN F5AR Activity Measured FSAR Design in Reactor Coolant Fraction of Basis Activity (1) Before Transient (2) Design Basis Nuclide ( uci /ml ) ( uCi /ml ) Activity in Coolant I-131 3.17 0.184 0.058 I-133 3.75 0.174 0.046 I-135 1.92 0.072 0.037 XE-133 250.00 9.200 0.037 Average = 0.045(3) (1) FSAR Table 11-2 for 1% defective fuel (368 leaking fuel rods) based on hot coolant density. (2) Based on reactor coolant samples taken on February 26, 1980, at 00:10 (liquid) and 10:15 (stripped gas) corrected to hot coolant density by multiplying by ( h/ c) = (46.5/62.2) = 0.748. (3) $quivalent to 17 + 3 leaking fuel rods. , l

TABLE II.A.15-2 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT SAMPLES Radionuclide Concentration in uti/ml" Date Time XE-133 I-131 I-133 1-135 CS-134 CS-136 CS-137 0.246 0.233 0.096 0.018 02/26/80 00:10 10:15 12.3 0.454 0.429 0.180 0.045 0.007 16:25 0.03? 0.5438 0.158 0.058 0.013 21:35 1.190 0.884 0.215 0.089 02/27/80 00:58 0.924 0.574 0.100 0.085 0.039 06:15 08:55 0.'833 0.506 0.073 0.1 04' 0.101 0.077 0.036 0.083 12:55 0.818 0.437 17:55

0. 94 9 0.416 0.031 0.094 0.035 0.088 0.736 0.288 0.016 0.071 0.024 0.066 21:10 02/28/80 08:15 0.575 0.160 0.004 0.068 0.022 0.067

• Based on coolant density at room temperature.

l TABLE II.A.15-3 GASEOUS ACTIVITY IN THE REACTOR BUILDING Msasured Release Released Total Measured (1) Conc. In with with Pzr. Release Conc. in RB Curies in Coolant (3) Coolant (4) Steam (5) (C1) Nuclide (uci/mi) RB Atmos.(2) (uci/ml) (uci/ml) (Ci) from RCS Xe-133 2.94x10-2 1582 12.3 1075 504 1578 Xe-135 7.0lx10-3 377 2.86 250 117 367 Kr-88 1.80x10-3 97 0.648 57 27 84 I-131 1.17x10-6 0.063 0.55 68 (6) 68 (1) Based on containment gas sample taken 2/26/80 at 21:55 (except I-131 comes from RM-A6 charcoal 2/27/80 at 04:05) and decay corrected back to the time of trip. 3 6 ft, (2) Assumes a reactor building free volume of 1.9 x 10 (1) Based on reactor coolant sample taken 2/26/80 at 10:15 and stripped of gas .(except iodine activity which was estimated from Table II.A.15-2). (4) Based on 32,868 gallons of cold coolant in RB and 12,782 gallons of cold coolant in bleed holdup tanks. 4 (5) Based on a Henry's Law constant of 8.5 x 10 psia per mole fraction. (6) Corresponds to a partition factor of about 1080.

1

l' 2 Before the RCS was placed in cold shutdown with the LPI system operating, the LPI System was recirculated to the borated water storage tank (BWST). The sodium levels in the BWST increased up to approximately 60 ppm. After the LPI System was connected to the RCS for cold shutdown cooling, sodium in a sample of the LPI l loop indicated that the sodium level was about 45 ppm. I Sodium in the BWST is being reduced by processing the water through a spent fuel demineralizer. The LPI and RCS water is also being processed through a demineralizer system to reduce sodium. l As of March 1, the sodium was in the range of 10 to 20 parts per million in the Reactor Coolant System. This sodium concentrated is not expected to harm any of the system components. B. Safety Implications The loss of electric power is an analyzed design basis event for the CR-3 FSAR. The following criteria for reactor protection are required for this event: a) Fuel damage will not occur from an excessive power-to-flow rate. b) Reactor coolant system pressure wil not exceed code pressure limits. c) The resultant doses are within 10CFR100 limits. The February 26, 1980, transient involved a partial loss-of-electrical power. The applicable protection criteria appear to have been met throughout the event. The sequence of events following the power supply failure in the NNI "X" cabinet is similar in nature to a loss of feedwater type transient whose consequences have been bounded for the FSAR analysis by the design basis event of a Feedwater Line Break. The reactor protection criteria for the Feedwater Line Break are: a) The core shall remain intact for effective core cooling. b) The reactor coolant system pressure shall not exceed code pressure limits of 110% of 2500 psig; i.e., 2750 psig. The major difference between the February 26, 1980, event and the bounding Feedwater Line Break event analyzed for the FSAR are itemized below. 1. The main feedwater valves were closed over a 25-30 second time interval rather than the conservative FSAR assumption of an instantaneous rupture of the main feedwater header at the steam generator inlet nozzles. 2. The combination of the opening of the PORY, the ICS demand for 103% full power and feedwater runbacks caused the reactor trip on i j high pressure to occur at 10-25 seconds into the event ratehr than i at 11.8 seconds for the FSAR event. --

3. The opened PORV allowed a rapid depressurization tn 1500 psi such that the ESFAS was actuated and two HPI pumps wert riarted. The PORY and spray block valves were closed about 5 minutes into the event and one pressurizer code safety valve actuated about 10 minutes into the event. This was much slower than 22.3 seconds for the FSAR events. 4. The operator initiated emergency feedwater at approximately 9 minutes into the event rather than the 15 minutes which was demonstrated for the FSAR as an acceptable time for operator action to prevent core or reactor coolant boundary damage. ~ 5. Main feedwater was supplied "B" OTSG at about 4 minutes (approximately 1400 gpm) rather than being totally lost as analyzed in the FSAR. 6. The maintenance of full HPI flow for 32 minutes provided a significantly greater heat removal capacity than assumed in the FSAR where only relief through the safety valves of the expanding RCS system water is considered. For both events, core coverage was maintained, no fuel damage occurred and the reactor coolant system pressure remained within code allowable limits. The safety evaluation criteria were met. Although apparently initiated by a single fault, this transient involved a series of multiple, propagating abnormalities through the interaction of the various sytstems which are affected directly or indirectly by the interruption of the NNI "X" power source. These interactions led to: (a) Loss of many instrumentation and annunciator indications normally available to the plant operator. Presentation of some confusing and invalid information. (b) Presentation of some invalid plant signals to the Integrated Control System, which acted upon them to increase reactor power, decrease feedwater flow and increase steam flow. (c) The pressurizer PORY was opened and latched at a system pressure below its setpoint. (d) Automatic initiation of emergency feedwater on low OTSG level did not occur. The maintenance of plant conditions within the safety criteria required the action of automatic safeguards (reactor trip, high pressure injection, reactor butiding isolation), safeguards features (safety relief valve), and/or operator action (start emergency feedwater) in response to items (b), (c), and (d). Item (a) reduced the operator's ability to respond. l l ~ l l l

l l C. Conclusions and Recommendations l The conclusions and recomendations are preliminary and are subject to l change. I 1. Conclusions (a) The PORY was opened by an electricsl upset on the NNI "X" 24 VDC bus. This upset also partially opened the pressurizer spray valve. (b) The loss of the NNI "X" 24 VDC bus caused failure of input signals to the ICS, the computer, alarms, and control room indications. (c) The ICS, with available input information, acted in a predictable manner. (d) The Safety Systems actuated as designed. (e) There was no known radiation released outside the reactor building. (f) The reactor core remained covered, and for the greater part of the event subcooling margin was greater than 500F. (g) The operators took the correct actions, in the safe direction. (h) Auxiliary feedwater (AFW) was not initiated automatically on low OTSG level because of OTSG startup level indication failure on loss of power. (i) Feedwater was lost to the "A" steam generator for about 30 minutes. Operators followed procedures in re-establishing flow and level to the "f.' steam generator. (j) There is no indication of fuel failure due te this incident. (k) Pressurizer safety valve RCV-8 opened at about 2400 psig, reseated at about 2300 psig, and was subjected to water flow of about 700 gpm. (1) RC System cooldown after RCV-8 reseated was a normal operation with two RC pumps running. 1 (m) Incident review would have been more accurate and timely if the data logging and transmission facilities had been functioned properly. i 2. Short Term Recommendations (a) Complete review of transient effects on RCS components, including fuel. See Section IV for detailed recomendations. (b) Determine cause of the power fault which interrupted "NNI-X" +24 VDC power supplies and take appropriate corrective action on the basic cause of the fault. (c') Review PORY circuitry and modify to assure that credible power failures do not cause the PORY to open when it is not required to open. (d) Review power supply independence between PORY and PORY isolation block valve and modify, if necessary, to assure that a failure which would affect PORY does not eliminate the possibility of PORY isolation block action. (e) Review emergency feedwater automatic initiation system and modify to assure that no power supply failure can inhibit. automatic initiation of the emergency feedwater on low OTSG level. (f) Review availability of vital signals to the control room operator in the event of power supply failures such as this one and provide for vital signal availability to the operator. (g) Review guidance and training provided to operator for managing the transient when confronted with loss of control room instrumentation which might accompany credible power supply failures. Revise as appropriate considering this transient. l l l i

III. EVENT DETAILS AND INPUT DATA A. Initial Plant Conditions The following table of data summarizes the initial plant conditions. Time of Reactor Trip Approximately 14:23 February 26, 1980 Reactor Power 98.6 Full Power RCS Temperature (Tave) 578'F RCS Pressure 2157 psig Pressurizer Level 202 inches Number of RC Pumps Operating 4 Steam Pressure Loop A - 911 psig loop B - 909 psig Number of Main Feedwater Pumps Operating 2 Tests in Progress None ICS Mode Automatic B. Plots of Major Parameters Several of the recorded parameters were selected for plotting to show major trends. These plots are shown in the following pages and are listed below: Figure III-1 Reactor Power from NI-5 (-360 to 840 seconds) Figure III-2 Control Room Strip Chart of Reactor Power from i NI-5 Figure II,I-3 Reactor Coolant System Pressure (-2 to 16 minutes) Figure III-4 Reactor Coolant System Pressure (-4 to 72 minutes) Figare III-5 Control Room Strip Chart of RC Pressure - Loop B (Narrow Range) Figure III-6 Control Room Strip Chart of RC Pressure (Wide Range) Figure III-7 Reactor Coolant Pressure (-2 to 42 hours) Figure III-8 Core Outlet Temperature and Reactor Coolant Tsat (-4 to 72 minutes) i '

Figure III-9 Control Room Strip Chart of RC Unit Average Temperature Figure III-10 Control Room Strip Chart of RC Outlet Temperature (Narrow Range) Figure III-ll Incore Thermocoule (Maximum) (-2 to 42 hours) Figure III-12 Reactor Coolant Cold Leg Temperatures (-2 to 42 hours) Figure III-13 Reactor Coolant Flow (-360 to 840 seconds) Figure III-14 Control Room Strip Chart of Reactor Coolant Total Flow Figure III-15 Control Room Strip Chart of Pressurizer Level Figure III-16 Pressurizer Level Indication (-360 to 840 seconds) Figure III-17 Pressurizer Level Indication (-2 to 42 hours) Figure 111-18 Pressurizer Temperature (-2 to 42 hours) Figure III-19 Control Room Strip Chart of Makeup Tank Level Figure III-20 HPI Flow During CR-III Incident Figure III-21 Control Room Strip Chart of Reactor Coolant Pump Seal Injection Temperature Figure III-22 Control Room Strip Chart of Megawatts Electric Generated Figure III-23 Main Steam Pressure (-360 to 840 seconds) Figure III-24 Control Room Strip Chart of Main Steam Temperature Figure III-25 Steam Generator Pressure (-360 to 840 seconds) Figure III-26 Steam Generator Pressure (-4 to 72 minutes) Figure III-27 Control Room Strip Chart of Steam Generator A Outlet Pressure Figure III-28 Control Room Strip Chart of Steam Generator B Outlet Pressure Figure III-29 Steam Generator Pressure (-2 to 42 hours) Figure III-30 Control Room Strip Chart of Steam Generator A Level Figure III-31 Control Room Strip Chart of Steam Generator B Level Figure III-32 Steam Generator Levels (-2 to 42 hours) Figure III-33 Feedwater Flow (-360 to 840 seconds) Figure III-34 Feedwater Flow (-4 to 72 minutes) Steam Generator Operate Level (-4 to 72 minutes) Steam Generator Pressure (-4 to 72 minutes) Figure III-35 Control Room Strip Chart of Feedwater Flow to Steam Generator A Figure III-36 Control Room Strip Chart of Feedwater Flow to Steam Generator B Figure III-37 Feedwater Pump Speed (-360 to 840 seconds) l l, l l l l

Figure 111-38 Feedwater Pump Discharge Pressure (-360 to 840 seconds) Figure III-39 Deaerator Feed Tank Level (-360 to 840 seconds) Figure III-40 Control Room Strip of - Reactor Building Sump Level - Borated Water Storage Tank Level - Reactor Building Pressure Figure 111-41 Letdown Monitor Readings on February 26, 1980 Figure III-42 Radiation Monitor Readings in Containment on February 26, 1980 Figure III-43 Reactor Building Pressure (-360 to 840 seconds) Figure III-44 Control Room Strip Chart of Reactor Building Temperatures Figure III-45 Control Room Strip Chart of Reactor Building Purge Exhaust Flow l l l l 1 l1l1l lI 0 4 8 I 08 7 W 0 I E 2 S I 7 O V H E A i R L P 0 8 I 6 R 6 i. T M5 - T DI S NN O 0 A P 0 M 6 ,O R l 7R E - F I U 1NA P 0 1 M 1 ,A O 5 l6D C i0 5 N1 5 8 0 iML 8 OA = i 4 RC i E Fl i t. C l f A R TE U 0 AD O 3 DI S 4 5 A T I E A N T D 0 O 6 N 3 M OR C F 0E 0S 3 R E E H W l O T 0 P 4 2 R O T 0 C 8 A 1 E R 0 2 1 1 I 0 I 6 I E RUG s 0 I F 0 6 0 2 1 0 8 1 0 4 2 l l 0 Il 0 3 R ( il l i' dE ht'L :: g52 "bE 0 [ 6 3 0 0 0 0 0 0 0 0 0 0 0 0 9 8 7 6 5 4 3 2 1 1 4

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i FIGURE III-33 FEEDWATER FLOW 6000 5500 LOOP B / l .. d. ..~. --{ 5000 y / / 4500 LOOP A {; 4000 31 h 3500 Eiy; INVALID DATA, LOOP A INSTRUNENT FAILED HID SCALE 3000 5d 2500 t C3 2000 E Dd DATA SOURCE = 855 COMPUTER POST TRIP REVIEW 15M vs l 1000 e,J 500 iI:7- '3 k -.s LOOP B .l 360 -300 -240 -180 -120 -60 0 60 120 180 240 300 360 420 480 540 600 660 720 780 840 Tgpr. SEC

l FIGURE III-34

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• 20

= 36-4 0 120e STEAM GENERATOR PRESSURE 35 0 q } 1000 i != L= 2

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. 3p) ' s 1000 I:N

---- 2 500

( [LM W 0 4 -360 -300 -240 -180 -120 -60 0 60 120 180 240 300 360 420 480 540 600 660 720 780 840 ilHE, SEC

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f. FIGURE III 41 LETOOWN MONITOR READINGS ON FEBRUARY 26, 1980 SX104 2O 5X104 O = 6 3 4X10 4 C a E. 4 j 3X10 5 ll 3 a 4 2X10 I I f I e i 14: 00 15:00 16:00 17:00 18:00 19:00 20:00 Time of Day FIGURE ll1 42 RADIATION MONITOR REA0lNGS IN CONTAINMENT ON FEBRUARY 26, 1980 102 00ME MONITOR (RM G19) I EL.211' 10; y i 1 1 I INCORE TANK MONiiOR 100 (RM-G18) EL.161' -= F.H. BRIDGE MONITOR 6 (RM GIS) EL.165' m / 10 I T,', ' N - - - ----- - -- a PERSONEL HATCH MONITOR I (RM G17) i EL.124' i 10-2 l I w I t di I E "E I F F NE l -m a m e 0: 10 3 E' }' 'i FEB 26 2:00 PM 3:00 PM 4:00 PM 5:00 PM 6:00 PM 7:00 PM 8:00 PM Time of Day 3

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.m l .if.. a,: e,. v i are t l

A.'*:

c.a..... a;: t;,; i. . gat i... c a ; s 1L ,09L.s 1 s.. - .w ,.: i.. t i s ...t;

. i,...

l ;;,. ..i ,r.. ,6, s we I ....a.. r. t .i. n.- t, *.. ;

r. *

..). s,p p* -a..i, e s .~p. 1

n-e o F FIGURE III-45 CONTROL ROOM STRIP CIMRT OF REACTOR BUILDING PURGE EXIMUST FLOW m 60 - 60 60 - 60 - i g 50 - - ~: h 50 - 50 - 50 - T

1. v A/4 + N94,4%emi CFM x163 CFM x 10 8 8

W' ~~ ~~~ 47ypab f%.. [5g (We kl k M% 30-10 - 30 - 30 - 20 - 20 - 20 - 20 - g_ l I -1 0-0-~ . " Z. 0 - Time =

IV. COMPONENT TRANSIENT ASSESSMENT A preliminary evaluation has been conducted for major RCS components which - were subjected to this transient. These evaluations focused on the component structural integrity and their acceptability for continued operation. Additional structural fatigue analyses would be necessary in order to determine the overall effects of this transient on the 40-year design life of the plant. A. Reactor Vessel The characteristic of this transient most affecting reactor vessel structural integrity was the change in reactor coolant inlet temperature which developed high thermal stresses in the inlet nozzle and shell areas. This will only affect the fatigue life of the vessel and does not affect the short term safe operation of the plant. A fatigue analysis would be necessary in order to determine the impact of this transient on the 40-year design life of the reactor vessel. Brittle fracture is not a concern as a result of this transient since the minimum inlet downcomer coolant temperature based on calculations was 2500F Fracture mechanics analysis has demonstrated that crack initiation of flaws up to and including the standard postulated flaw size did not take place and the existing pressure temperature limitations curv'es are still valid. B. Reactor Vessel Internals The Reactor Vessel Internals were not adversely affected by this transient. Excessive thermal stresses were not experienced and positive clamping force was maintained between the reactor vessel and internal s. C. Steam Generators The maximum steam generator tube-to-shell differential temperature experienced was conservatively estimated to be 1380F This is less than the 1860F temperature differential which has previously been analyzed for the Rancho Seco rapid cooldown event of March 20, 1978. The resulting axial tube loads are bounded by this previous analysis and, therefore, are acceptable. The stresses developed in the shell-to-tubesheet and lower head-to-tubesheet regions are al so bounded by the previous analysis of the Rancho Seco event. These stresses were within code allowables and the effect on usage factor is negligible. The effect of the temperature change on the support skirt-to-lower head attachment is bounded by conditions already j analyzed in the Crystal River-3 stress report. , t i

Prior to the February 26, 1980, incident, a primary-to-secondary leak of approximately 0.4 gallons per day existed at CR-3. This leakage is thought to be caused by tube-to-tubesheet weld damage in the B-steam generator resulting from the burnable poison rod assembly which separated from the fuel assembly in March,1978. If an increased leakage rate is detected upon the plant's return to operation, it is more likely indicative of the general condition of the previously damaged tube-to-tubesheet welds rather than specifically related to the recent incident. The structural integrity of the steam generators has not been adversely. affected, and the steam generators are acceptable for resuming operation. Since the lower portion of the steam generator support skirt was potentially imersed in a pool of reactor coolant, it is recomended that the lower portion of the steam generator support skirts and anchor bolts be visually inspected during the next refueling outage and any corrosive residue removed. D. Pressurizer The pressure and temoerature transient within the pressurizer, spray and surge line nozzle were not severe enough to adversely affect the continued short term operation of the pressurizer. This qualitative assessment is based on a comparison of this transient to design transients prescribed in the CR-3 functional specification. The effect on the fatigue life of the pressurizer must be quantitatively assessed later. The pressurizer heaters (upper bundle and half the middle bundle) may have been subjected to saturated steam conditions while energized. This can lead to accelerated failure of heater elements, deterioration of the sheath to diaphram weld, deterioration of the heater bundle diaphram seal welds and possible stressing of the heater sheath due to axial growth of the heater element. An evaluation has been conducted for each of these areas of concern. It was concluded that the diaphram to sheath weld was not adversely affected and the heater sheath was not subjected to stresses as a result of heater growth. The heater bundles should be inspected for seal leakage and heater elements electrically checked for continuity prior to their return to operating service. A fatigue analysis should be performed to demonstrate that the 40-year design life was not adversely impacted by this transient. Loadings in excess of the original design loads may have been induced on the pressurizer relief nozzles during relief valve operttion. The relief valve loadings which occured during this transient should be determined and the effect on the pressurizer relief valve nozzles assessed later. (See Se: tion IV.H).,

E. Reactor Coolant Piping The design rapid depressurization transient in the Crystal River Unit 3 Functional Spec and for the rapid depressurization transient considered in the Rancho Seco Stress Report were either representative or conservative for all areas of the RC piping except the A-loop cold leg. Analyses of the A-loop cold leg pipe and HPI nozzle were performed and resulted in acceptable stresses and usage factors. It is concluded that this tr6nsient event had no significant effect on reactor coolant piping and the RC piping is suitable for continued operation. F. Reactor Coolant Pumos Evaluation of the effect of this transient on reactor coolant pumps included a review of the pump environment during the transient and a review of the performance parameters. The significant environmental conditions considered were system pressure, temperature, service water temperature and reactor building humidity, temperature and pressure. The pump performance parameters reviewed were pump mechanical performance, seal leakage rate and seal staging pressures. There is no indication the reactor coolant pumps sustained any damage during this transient which would affect their. performance. G. Control Rod Drive Mechanism (CRDM's) This transient had no impact on the ability of the CRDM's to perform a safety function (trip or maintain integrity of RCS pressure boundary). The maximum RCS pressure experienced and loss of component cooling water did not adversely affect the operation of the CRDM's. The high humidity in the Reactor Building may have, however, degraded stator insulation resistance. The majority of past CRDM failures have been attributed to moisture. All CRDM's should be checked for proper insulation resistance prior to their return to service and, if necessary, dried out to achieve acceptable readings. H. Pressurizer Relief Valve and Relief System The pressure, temperature, and flow conditions experienced by the power operated relief valve (PORV) were not outside of its design parameters. To the best of our knowledge, the PORY valve performed as expected and should not have been substantially affected by the transient conditions. , l l l

i Although the pressurizer code safety valve (s) lifted on steam, it experienced water flow for a considerable length of time. The relief of water through the code safety valve is not a normal condition and l may have had some detrimental effect,on the valve internals. j Reaction forces imposed by the valve on the pressurizer nozzle due to the flow conditions (i.e., water versus steam) should not have been any greater than the normal design loads. However, the several hour duration of flow through the discharge piping may have resulted in excessive expansion of the piping and imposed loadings on tne valve and pressurizer nozzle in excess of that considered in the original desi gn. Although the conditions experienced by the PORV and code safety valves should not have substantially affected their structural integrity, or their ability to perform their intended relief function the following actions are recomended prior to their return to operating service: 1. Perform a visual inspection of the pressurizer relief system (i.e., PORY, both code safety valves, and the discharge piping). Inspection of the discharge piping system including hangers should be performed to ensure that no gross distortions have occurred. 2. Confirm by calculation that the axial loads and moments imposed on the valves and pressurizer as a result of the extended period of discharge to the quench tanks are no greater than those considered in the original design evaluation. 3. Disassemble, inspect, and refurbish as needed the pressurizer code safety valves and PORV. I. Fuel The conditions experienced by the fuel should not have failed any sound fuel rods. The activity seen is consistent with the predicted 15 to 20 leaking rods present prior to the incident. No fuel technical specifications limits were violated. The transient pressure - temperature conditions did exceed the B&W recommended conditions provided under "Operat'ng Limits and Precautions," to ensure the fuel is maintained in compression. However, the recommended envelope was based upon very conservative EOL fuel pin pressure calculations assuming burn-ups greater than 40,000 NWD/MTV. The Crystal River plant is currently only 160 EFp0 into Cycle 2 and the estimated hot pin burn-up is less than 24,000 MWD /MTU. A preliminary assessment of the transient indicates that the fuel was maintained in compression at all times. V. REFERENCES 1. Crystal River-3 Model 855 Ccmputer Alarm Log - February 26, 1980 2. Crystal River-3 Sequence of Events Monitor - February 26, 1980 3. Control Room Charts - February 26, 1980 i 4. Crystal River-3 Model 855 Computer Post Trip Review - February 26, 1980 5. ASME Steam Tables - 1967 1 ! f i i-i

APPENDIX A INSTRUMENT STATUS The loss of the NNI "X" +24VDC power caused many instruments to give incorrect responses. This section identifies the affected instrumentation and alarms. This review is based upon the following: a. No loss of 120 VAC "X" power to and/or in the NNI "X" cabinets and peripheral components. b. Loss of +24 VDC and -24 VDC in the NNI "X" cabinets due to tripping of breakers Sj and S2 as shown on Bailey drawing D80340420. c. No loss of 120 VAC, +24VDC, or -24VDC to or in the NNI "Y" cabinets. d. All instrument strings were operable prior to the power loss in "b" above and were wired as shown on Bailey drawings listed below, but not included in this report. Bailey Drawings 080340420 NNI-X Power Dist. Sheet 1 of 4 08034042D NNI-X Power Dist. Sheet 2 of 4 08034042C NNI-Y Power Dist. Sheet 3 of 4 080340420 NNI-Y Power Dist. Sheet 4 of 4 08034033C R.C. Sys. Schematic Sheet 1 of 7 08034033C R.C. Sys. Schematic Sheet 2 of 7 08034033E R.C. Sys. Schematic Sheet 3 of 7 D8034033E R.C. SYS. Schematic Sheet 4 of 7 080340338 R.C. Sys. Schematic Sheet 5 of 7 08034033C R.C. Sys. Scheinatic Sheet 6 of 7 080340330 R.C. Sys. Schematic Sheet 7 of 7 C8034034C SP Sys. Schematic, Sheet 1 of 4 08034034C SP Sys. Schematic Sheet 2 of 4 D80340340 SP Sys. Schematic Sheet 3 of 4 08034034C SP Sys. Schematic Sheet 4 of 4 D8034035C CA Sys. Schematic 08034036C RB Spray Sys. Schematic N D8034038C C.F. Sys. Schematic 08034037E DH Sys. Schematic A-1

APPENDIX A (Cont'd) 08034039D Makeup Sys. Schematic Sheet I of 5 080340390 Makeup sys. Schematic Sheet 2 of 5 080340390 Makeup sys. Schematic Sheet 3 of 5 080340390 Makeup Sys. Schematic Sheet 4 of 5 080340390 Makeup Sys. Schematic Sheet 5 of 5 D8034040E Spent fuel Sys. Schematic 08034041C Condensate Flow Sys. Sheet 1 of 2 08034041B Condensate Flow Sys. Sheet 2 of 2 Cabinet Module arrangement Sheets Rev B dated 12/10/71. 3 Table 11-1 contains a tabulation of non-nuclear instrument parameters and provides the following information, assuming the failure of the NNI "X" +24VDC powe,r: a. Source of Power b. Validity of Signal to Computer c. Validity of Signal to Integrated Control System (ICS). d. Validity of Indication to the Control Room Operator. e. Validitj of Indication to the Hot Shutdown Panel Operator. Table II-2 contains a tabulation of non-nuclear instrumentation alarms for the reactor coolant, secondary plant, makeup & purification systems and provides the following information: a. Source of Power b. Validity of signal for actuation of High Alarm on Control Room Annunicator (assuming failure of the +24VDC NNI "X" supply). c. Validity of Signal for actuation of Low Alarm on Control Room Annunciator. A-2

APPENDIX A (Cont'd) Table II-3 contains a tabulation of non-nuclear instrumentation system control or interlock functions for the reactor coolant, secondary plant makeup and purification systems and provides the following information: a. Parameter b. Interlock or Control Function c. Source of Power d. Validity fo Signal for Actuation e. State of Interlock or Control function with no valid signal. The following definitions of table headings are provided to assist in understanding the information provided in table II-3. Definitions o " Parameter": The process variable being monitored. e " Interlock / control function": The function to be performed by contact actuation. e " Cabinet power source": The source of power for the device listed which provides the contact action. e " Valid signal available for implementation": The + 24VDC power and/or the Valid signal available for implementation - The + 24VDC power and/or the module (s) electrical signal representing the process variable being monitored does not exist or is in error, e "Was interlock / control action implemented without a valid signal": The + 24VDC power was not available to energize signal monitor relays ta change contact state to implement control action or interlock. The following is the status of the non-nuclear instrument system auto / manual controllers for the reactor coolant, makeup, and letdown systems in the event of NNI "X" + 24 VDC power failure. a. Reactor coolant pump seal injection flow control - No valid signal. b. Reactor coolant makeup flow control - No valid signal. l c. Pressurizer heater SCR control - No valid signal. d. Letdown flow control - Valid Signal. l A-3

APPENDIX A (Cont'd) The following is a listing of NNI control room indications not addressed in Table 11-1. The signals to these indications are derivec from various signal processing modules in the NNI. a. Total RC Flow Recorder - No valid signal b. Reactor Coolant Loop A&B aTc Indicator - No valid signal c. Reactor AT Indicator - No valid signal d. Reactor Th Recorder & Indicator - Signal valid only if loop A or B Th selected by RC15-MS. e. Reactor Tave' Recorder - No valid signal f. Reactor Tove Indicator - No valid signal g. Loop A&B AT Indicator - No valid signal h. Loop AAB Tave Indicator - No valid signal The following is a summary listing of NNI to ICS Signals taken from Table 11-1. These signals were not valid to the ICS. a. RC Loop A Flow b. Controling Tave c. Total RC Flow d. Reactor ATc

e.. Steam generator A Startup Level f.

Steam generator B Startup Level g. Steam generator A Feedwater Valve P h. Steam generator A Startup Feedwater Flow The following signals to the ICS would have been valid if the sensor signal select switch had been in the position identified in Table II-l. The presumed status of key instruments listed below is discussed later in Section II.A.12. I l i A-4 i i

APPENDIX A (Cont'd) a. RC Loop B Flow b. Reactor Th c. Steam Generator A Pressure d. Steam Generator B Pressure e. Steam Generator B Feedwater Valve LP f. Steam Generator A Operate Level g. Steam Generator B Operate Level h. Turbine Throttle Steam Pressure r

i. Steam Generator B Startup Feedwater Flow

j. Steam Generator A Feedwater Temperature k.

Steam Generator B Feedwater Temperatur 1. Steam Generator A Main Feedwater Flow m. Steam Generator B Main Feedwater Flow In most cases, instrumentation affected by the power loss will provide a mid-span (50% of ranae) indication on loss of +24'!DC power. A-5 l l i

Accendix A Status or NNI Signals Table II-1 Signal Validity to Computer, Control Room Indicator. Hot Shutdown Panel, and ICS l

• • • • Valid Signal Available To Cabinet Shuu ct:n f Power Con:rol P.ce=

Hot Para.neter Source Cctru:er Operator Panel Oper. I Its Pressurizer Level RC-LT1 X No No Yes-Indicator fl. A. RC-LT2 X No No No N.A. RC-LT3 Y Yes No Yes-Indicator N.A. Pressurizer Temp RC2-TE1 X No No N.A. N.A. RC2-TE2 Y No

• Yes-Indicator N.A.

N.A.

• 0nly if' RC2-TE2 Selected RC Flow Loop A From NI/RPS N.A.

No N.A. fio RCl4A-dPT 1 or 2

• Receive Flow Sianal From NI/RPS NNI Modules Powered from NNI X Trans mitterNI/RPSfpwer RC Flow Loop B Y

N.A. No N.A.

• • Yes From NI/RPS
  • 0nly if RC48-TE4 Sele cd RCl48-dPT 1 or 2 RC Loop A Th RCAA-TE1 X

No No N.A. fl. A. RC4A-TE4 Y No

• Yes-Indicator
• Yes-Indicator
  • Yes
• 0nly if RC4A-TE4 Selected
  • 0nly if RC4A-TE4 Selecte d & RC15-MS Sel ected to Loop A RC Loop B Th RC48-TE1 X

No No N.A. No RC48-TE4 Y

• Yes
  • Yes-Indicator **Yes-Indicator ***Yes
• If RC48~TE4 Not Selectec
  • 0nly if RC48-TE4 Selected
    • 0nly if RC48-TE4 Selected and RC15-MS Selected to Loop B

RC Loop A Tc Narrow RC5A-TE1 X No No N.A. No RC5A-TE3 Y Yes

• Yes-Indicator N.A.

Nc

• 0nly if RC5A-TE3 Selected l

l 1 I I l ( **** Assuming failure of the + 24 VOC NNI-X Power Supply)

Table II-1 Vclid Si;;n.1 Available To Cabinet Potter C.mncrol Rec: lEo: Shutdcr. Parc=eter Secrec l Cc=puter Operator l Panel Oper. IC3 RC Loop B Tc Narrow RC58-TE1 X No No N.A. No RC58-TE3 Y Yes

• Yes-Indicator N.A.

No I

• 0nly if RC58-TE3 Selected RC Loop A Tc Wide RC5A-TE2 X

No No N.A. N.A. RCSA-TE4 Y No

• Yes-Indicator N.A.

N.A. 'Only if RCSA-TE4 Selectied RC Loop B Tc Wide RC58-TE2 X-No No N.A. N.A. RC5B-TE4 Y

• les
  • Yes-Indicator N.A.

N.A. I

• 0nly if RCSB-TE4 Not Se ected
  • 0nly if RC58-TE4 Selectito RC Press Loop A Yes-Indicator Wide Range X

NNI N.A. Yes 'lecorder Yes N.A. Signal from ESFAS RC3A-PT3 RC Press Loop B Wide Range Signal from ESFAS RC38-PT3 X NNI N.A. Yes. N.A. N.A. I lPage 2 i 1 l t I i l l

.? Table II-1 Valid Signal Available To Cabinet Power Control Rocs Hot Shutdown Parameter Source Computer Operater Panel Oper. Ics RC Pump #2 Seal Cavity Pressure RC10A-PT1 X N.A. No N.A. N.A. RC10A-PT2 X N.A. No N.A. N.A. RC108-PT1 X N.A. No N.A. N.A. RC108-PT2 X N.A. No N.A. N.A. RC Pump #3 Seal Cavity Pressure RC19A-PT1 X N.A. No N.A. N.A. RC19A-PT2 X N.A. No N.A. N.A. RC198-PTl X N.A. No N.A. N.A. RC198-PT2 X N.A. No N.A. N.A. 1 Page 3 l t l l

Table II-1 Valid Signal Av:11sble To Cabine: "*r Centrol Roc = Hot Shutdown ,aramecc: Source Computer Operator Panel Opsr. Ics Start-up Stm. Gen. Loop A Level SPIA-LT4 X No No Yes No SP1A-LT5 Y No No N.A. No Start-up Stm. Gen. Loop B Level SPIB-LT4 Y

• Yes No Yes No SP18-LT5 X
  • Yes No N.A.

No

• 0nly if SPIB-LT4 Not Se' ected
  • 0nly if SPIB-LT4 Selected Stm. Gen. Outlet Press. Loop A SP6A-PT1 X

No No No No SP6A-PT2 Y Yes Yes-Recorder N.A.

• Yes
• 0nly if SP6A-PT2 Selected Stm. Gen. Outlet Press. Loop B SP6B-PT1 Y

Yes Yes-Recorder Yes ~*Yes SP6B-PT2 X No No No No

• 0nly if SP6B-PT1 Selected Stm. Gen. Level Full Range Loop A SP1A-LT1 X

No Yes N.A. N.A. Stm. Gen. Level Full Range Loop B SPIB-LT1 X No Yes N.A. N.A. Main Steam Temp. Loop A Sf4A-TE X No No N.A. N.A. l i Page 4 I l

Table 11-1 Valid Sige.a1.Available To Cabine: Pc ::: Cen:rcl P.oen I Hot Shutdown Parameter Source Cc=pu:er opers:e; l Par. eloper. CS Main Steam Temp Loop B SP48-TE Y Yes Yes-Indicator N.A. N.A. SGA FW VLv aP SP12A-dPT X N.A. N.A. -N.A. No SGB FW VLv aP SPl?.B-dPT Y N.A. N.A. N.A. Yes SGA Lower Down Comer Temp SP3A-TE1 X No No N.A. N.A. SP3A-TE2 Y No

• Yes-Indicator N.A.

N.A.

• 0nly if SP3A-TE.2 Selected SGB Lower Down Comer Temp SP38-TE1 X

No No N.A. N.A. SP38-TE2 Y

• Yes
  • Yes-Indicator N.A.

N.A.

• 0nly if SP38-TE2 Not Se :ected
  • 0nly if SP38-TE2 Selected SGA Operate Level SPIA-LT2 X

No No N.A. No SP1A-LT3 Y No

• Yes-Recorder N.A.
• Yes
• 0nly if SP1A-LT3 and SP1A-TE2 Selected 9

Page 5 i

Table II-1 Valid Signal Available To Cabinet Pcuer Control Roc: Hot Shu:deten Parameter Source Cc=pu:er Operate Panel Oper. ICS SGB Operate Level SPIB-LT2 X No No N.A. No SPIB-LT3 Y

• Yes
  • Yes-Recorder N.A.
  • Yes
• 0nly if SPIB-LT3 Not Selected
  • 0nly if SPIB-LT3 and SP30-TE2 Selected Turbina Throttle Steam Pressure (SGA) SP10A-PT1 Y

Yes

• Yes-Recorder N.A.
• Yes (SGA) SP10A-PT2 X

No No N.A. No (SGB) SP108-PT1 Y Yes

• Yes-Recorder N.A.
• Yes (SGB) SP108-PT2 X

No No N.A. No

• 0nl'y if this Transmitter Selected SGA FW Temp SP5A-TE1 X

No No N.A. No SPSA-TE2 Y No

• Yes-Indicator N.A.
• Yes
• 0nly if SP5A-TE2 Selected SGB FW Temp SP5B-TE1 X

No .No N.A. No SPSB-TE2 Y

• Yes
  • Yes-Indicator N.A.
  • Yes
• 0nly if SP5B-TE2 Not Sel ected
  • 0nly if SP58-TE2 Selected i Page 6 i

i l l l

Table 11-1 Valid Signal N ailable To Cabinet Power Control Roo: Hot Shutdcun Paraceter Source Co puter Operator Panel Oper. ICS SGA Startup FW Flow SP7A-dPT X No No N.A. No SGB Startup FW Flow SP7B-dPT Y Yes

• Yes-Indicator N.A.
• Yes
• 0nly if iP58-TE2 Selected SGA Main FW Flow SP8A-dPT1 X

No N.A. N.A. No SP8A-dPT2 Y No N.A. N.A.

• Yes
• 0nly if iP8A-dPT2 and SP0A-TE2 Selected SGB Main FW Flow SP88-dPT1 X

No N.A. N.A. No SP08-dPT2 Y

• Yes N.A.

N.A.

• • Yes
• 0nly if $P88-dPT2 flot Sel ected
  • 0nly if 3P88-dPT2 and SP08-TE2 Selected I

l Page 7 i

Table 11-1 Valid Sigt.a1 Available To Cabine: Potter Con:rol Ro:: Hot Shu det'n Parameter Source Computer oper.:or Panel Oper. ICS RCP Total Seal Flow MU27-dPT X No No N.A. N.A. HPI Flow l MU23-dPT1 X N.A. No N.A. N.A. -MU23-dPT3 Y N.A. Yes N.A. N.A. MU23-dPT2 X N.A. No N.A. N.A. MU23-dPT4 Y N.A. Yes N.A. N.A. Letdown Flow MU4-d?T Y N.A. Yes N.A. N.A. MUS-TE X N.A. Yes N.A. N.A. Makeup Tank Level MU14-LTI X No No N.A. N.A. MU14-LT2 Y No No N.A. N.A. MU Pump Pressure MU2-PT Y No Yes N.A. N.A. MU Tank Pressure MU17-PT X N.A. No N.A. N.A. MU Filter P MU18-dPT Y N.A. Yes N.A. N.A. Makeup ' Flow MU24-dPT X N.A. No N.A. N.A. I Page 8 i 1 l l l

l Table II-1 Valid Sigr.al krailable Te Cabince l Pover Cer. trol Rec. Hot Shu:detm l Cer.puter Opera:Or Par.el Oper. l Ics Para:neter Scurce RC Pump Seal Flow MU7-dPT1 X N.A. No N.A. N.A. MU7-dPT3 X N.A. No N.A. N.A. MU7-dPT2 X ft.A. No N.A. N.A. MU7-dPT4 X N.A. No N.A. N.A. RC Pump Seal Return Flow MU31-FT1 X No No N.A. N.A. MU31-FT2 X No No N.A. N.A. MU31-FT3 X No No N.A. N.A. MU31-FT4 X No No N.A. N.A. i l l Page 9 i i + i i

Table II-I Valid Signal Avsilable To I Power Centrol Room Hot Shutdcen Parameter Source Computer operage; pzP81 0Fer-ICS DH Removal Flow DH1-dPT1 X N.A. No N.A. N.A. OH1-dPT2 y N.A. Yes-radicator N.A. N.A. DH Injection Temp DH2-TE1 X No Yes-Indicator N.A. N.A. CH2-TE2 Y Yes Yes-Indicator N.A. N.A. DH Cooler Inlet Temp DH6-TE1 X N.A. Yes-Indicator N.A. N.A. OH6-TE2 Y N.A. Yes-Indicator N.A. M.A. L Page 10 I i l I l.

Table II-I Valid Signal Available To Cabinet I Power Control P.ocu Hot Shutdcun Paraceter Source Computer Operator Panel Oper. ICS Core Flood Tank A Pressure CF1-PT1 X N.A. No N.A. N.A. CF1-PT2 Y N.A. Yes-Indicator N.A. N.A. Core Flood Tank B Pressure CF1-PT3 X N.A. No N.A. N.A. CF1-PT4 Y N.A. Yes-Indicator fi. A. N.A. Core Flood Tank A Level CF2-LT1 X N.A. No N.A. N.A. CF2-LT2 Y Yes Yes-Indicator N.A. N.A. Core Flood Tank B Level CF2-LT3 X No No N.A. N.A. CF2-LT4 Y N.A. Yes-Indicator N.A. N.A. Page 11

Table II-1 Valid Signal Avcilable != Cabinet Pouer Centrol P,oc= Ect Shutdo.cn Parameter Source Computer Operator Fattel CPer-ICS RB Spray Flow BSI-dPT1 X ti. A. No N.A. N.A. .BSI-dPT2 Y N.A. Yes-Indicator N.A. N.A. Sodium Thio. Tank Level l BS3-LT X N.A. No N.A. N.A. Sodium Hydroxide Tank Level BSS-LT -X N.A. No N.A. N.A. Sodium ' Thio. Tank Temperature BS7-TE X N.A. Yes-Indicator N.A. N.A. Sodium Thio. Tank Pressure BS15-PT X N.A. No N.A. N.A. Sodium Hydroxide Tank Temperature BS8-TE X N.A. Yes-Indicator N.A. M.A. Sodium Hydroxide Tank Pressure BS14-PT X N.A. No N.A. N.A. I 1 i 'Page 12 3 j l

Table II-1 Valid Signal Available To Cabine: Pouer Control F.ccu F.o: Shutd: n ; Parameter Source Cc puter Operater Panel Oper. ICS Boric Acid Tank #1 Temperature CA10-TE X No N.A. Yes N.A. Boric Acid Tank #1 Level Call-LT X No N.A. No N.A. Boric Acid Pump PIA & PIB~0ischarge Pressure .CA14-PT X N.A. No No N.A. Boric Acid Tank #2 Temperature CA12-TE X No N.A. Yes-Indicator N.A. Boric Acid Tank #2 Level CA13-LT X No N.A. No N.A. ' Page 13 e i

Table II-1 Valid Signal Available To Cabinet Power Control Recs Het Shutdown Pars:cter Source Co pu:ar operster Panel Oper. ICS S. F. Storage Pool "A" Leve! SF1-LTI Y N.A. fes-Indicator N.A. N.A. j S. F. Storage Pool "B" Level SF1-LT2 Y N.A. Yes-Indicator N.A. M.A. 1 l 14 l Page la i l

Table 1I-1 Valid Signal Available To Cabinet Power Con:rel Roen Eot Shu:dewn Parameter Source Ce puter Opera:cr Panel Oper. ICS CD Deaerator Level CD61-Lt Y Yes Yes-Recorder N.A. N.A. CD Total Cond. Flow CD15-FT Y N.A. Yes-Recorder N.A. N.A. r i I Page 15

Tablo TI-2 Valid Operable Alarms Cabinet Power Valid Sianal Available For Actuation of Parameter Source Hi Alarm Lo Alarm Prossurizer x Level (RCl-LR/L.92) No No Pressurizer x Level (RCl-LSI) No No Loop A RC Flow x (RC14A-FS) N.A. No Loop B RC Flow x (RC148-FS) N.A. No Total RC Flow x (RC13FR/FS) N.A. No Loop A/B ATC x (RC8-dTS) No No Reactor TH x (RC4-TR/TS)

• yes N.A.
• 0nly if Loop A or B TH Selected by RC 15-MS Loop A RC Pressure x

wide range (RC3A-PR2/ PS ) N.A. Yes RC Pressure Vs. x Core Flood Valve (RC3A-PS2 ) No No Position Alarm RC Pressure x High Press Inj. (RC3A-PS6) No N.A. Not by Passed RC Pressure x Low Press Inj. (RC3A-PS S) No N.A. Not by Passed RC Pump #2 Seal x Cavity Pressure (RC10A-PSI) No N.A. (All 4 Pumps) (RC10A-PS2) (RC10B-PS 1) (RC10B-PS 2) RC Pump #3 Seal x Cavity Pressure (RC19A-PS I) No N.A. (All 4 Pumps) (RC19A-PS2) (RC198-PSI) (RC198-PS2) SGA Startup level x (SP1A-LS2) N.A. No SGA Startup leve' x (SP18-LS 2) N.A. No SGA Operate Levt x (SPIA LR/LSI)

• Yes N.A.
• Only if SP1A-LT3 and SP3A-TE2 is selected.

SGB Operate Level x (SP1B-LR/LS1 ).

• Yes N.A.
• only if SP18-LT3 and SP38-TE2 is selected.

Table f!-2 Valid Signal Available For Actuation of Cabinet Power ~~ Parameter Source Hi Alarm to Alarm ~ Total RC Pump Seal Inject. x Flow (MU27-FS) No No High Pressure Injection Flow x RC Loop A (MU23-FS2) No No Y (MU23-FS4) Yes Yes High Pressure Injection Flow x RC Loop B (MU23-FS1) No No Y (MU23-FS3) Yes Yes Let down Temp x (MUS-TS) No N.A. Makeup-Tank Level x (MU14 LR/LSI) No No Makeup Tank Press x (MU17-PS) No No RC Makeup Flow x (MU24-FS) No No RC Pump Seal Injection Flow x (all 4 pumps) (MU7-FS1) No No x (MU7-FS2) No No x (MU7-FS3) No No x (MU7-FS4) No No RC Pump Seal Return Flow x (all 4 pumps) (MU31-FS1) No N.A. x (MU31-FS2) No N.A. x (MU31-FS3) No N.A. x (MU31-FS4) No N.A. 'I Turbine Throttle x Pressure (SP10-PR/ PSI)

• Yes
• Yes
• 0nly if SP10A-P T1 or SP108-PT1 is selected Page 2

Table 11-3 Validity of NNI, Digital Interlock and Control Functions For Reactor Coolant Makeup and Purification Systems Cabinet Valid Signal as In erl ck/ Control Interlock / Control Power Available for Parameter Function Source Luplementa tion fifthout lid Signal ~ Pressurizer De Energize Pressurizer x Level lleaters on Low Level (RCl-LSI) No No Loop A RC Flow Transfer Controlling Tave x To Loop B on Low Loop A (RCl4A-FS) No No Flow Loop B RC Flow Transfer Controlling Tave x fo Loop A on Low Loop B (RC148-FS) No No flow RC Loop A Tc Prevent Start of 4th RC x (wide range) Pump when Tc less than (RCSA-TS) No No Temp. Set Point RC Loop B Tc Prevent Start of 4th RC x Pump when Tc Less than (RCSB-TS) No No Low Temp. Setpoint ~ RC Pressure Open & Close Pressurizer x Electro Matic Relief (RC3-PSB) No Na RC Pressure Open/Close Pressurizer x Spray Valve (RC3-PS3) No No i RC Pressure On/Off Pressurizer x llenter Bank #3 (RC3-PS6) No No RC Pressure On/Off Pressurizer x lleater Bank #4 (RC3-PS7) No No RC Pressure Decay lleat Valve (Dit-VI) x Prevent opening when (RC3A-PS3) No No RC Pressure liigh RC Pressure liigh RC Pressure Interloc ( x i To Customer Spray Valve (RC3A-PS7) No No (PZR) RCV-53F Letdown Temp Close MU-V3 on High x (MUS-IS) No No Makeup Tank Terminate feed Bleed x On Low MU Tank Level. (MU14-LS2) No No

-~ Interlock / Control P A> abl r i a e d Paraseter function Source Impleuentation Without Valid Signal l i

RC Pump Seal Close RC Pump Seal x

, Injection flow Return Valve on Low (MU7-FS1) No No'

(all 4 pumps)

Seal injection flow and x j Prevent Start of (MU7-FS2) No No i RC Pump with Low Seal Inject. Flow (MU7 FS3) No No X (MU7-FS4) No No I i 1 I i I 1 Page 2 - ?

i ) APPENDIX B j i-ANALYSIS OF CORE SIGNALS 1. Alam Printout Data The alarm printout from the February 26, 1980 Trip at Crystal River-3 has been examined to determine if SPND alarms were indicative of any significant occurrences in the ccre. Based on this analysis and the lack of any i evidence to the contrary, it is believed that the alarms were caused by l normal gama-induced background currents rather than neutron or temperature induced currents. An alarm is printed whenever a parameter changes status from " normal" to " bad" or vice versa. A bad signal is one that is outside of some setpoint range and may be high or low. The printed message does not indicate whether j the parameter is high or low although the value in question is printed when it returns to normal. It appears that the low setpoint for SPND's at CR-3 i s -10 nanoags. The SPND's are scanned once every 60 seconds; their status is printed only if it has changed since the last scan. J The first SPND alarm occurred at 1430 at location 8-F, level 7 (string 4). Though other SPND's alarmed thereafter, most at level 7 (top of the core), several SPND's from levels 6 and 5 also alarmed. In addition, one level 2 SPND (string 6 in location 7-F) alarmed although it returned to normal within one minute (at 1515). Frequently, SPND's w:uld return to " normal" several minutes after alarming and might go through the cycle of alarming i and returning to normal several times in a period of 2-1/2 hours following the trip. Because the current was always -10 to 0 nanoamps when the SPND's returned to " normal" it is felt that all of the alarms were " low". Figure 1 and 2 show the time of alarms for levels 6 and 7 detectors between 1430 and 1437. The mechanism causing the aTarms is background current attributable to gama-induced electrons.. The positive component of the signal arising from neutron reactions decreases after the trip with a 60-second half-life characteristic of Rh-104 decay. Thus, after a period of five to eight minutes following the trip, the neutron portion approaches zero. The gama portion remains relatively high due to fission product decay and causes a i negative current. The gama currents are -5 to -15 nanoamps in magnitude for level 7 SPND's and smaller currents (in absolute magnitude) are j generated for detectors lower in the core since their leadwires are shorter. l Therefore, it is expected that more level 7 detectors would alarm than would detectors at other levels and that the same SPND would alarm repeatedly as i the random background noise rose above and fell below -10 nanoamps. This explanation is verified by the alarm printout for a " normal" CR-3 trip (one without loss of coolant) (Ref.1). The trip occurred when the reactor was at 75".FP at BOC-1, and several level 7 SPND's began alarming off-scale, low, approximately five minutes following the trip. B-1

\\ +/ %4 bb\\\\ IMAGE EVALUATION NNN\\ TEST TARGET (MT-3) 1.0 5 8E E=a e m p_ L , m _aa I.I L"'lM \\ l 1.8 1.25 1.4 1.6 4 6" MICROCOPY RESCLUTION TEST CHART //// 4% +%p // l* ff g;;4f\\9 S 5,, 7 1 'j i .. _. mi.G. 4

APPENDIX B (Cont'd) 1 It has been speculated by others that the SPND alarms were caused by thermionic-induced currents. To produce a current of sufficient magnitude to initiate an alarm, a temperature in excess of 1000*F is required. Based on all other indications, such temperatures were not reached during this transiesnt. The behavior of zircaloy leadwires differs from thht of inconel in that the gamma-induced current is positive and much smaller (+3 nanoamps [zircaloy) vs. -80 nanoamps [inconel] for level 7 at full power). It should be noted that all SPND strings in CR-3 have inconel leadwires. Most of B&W's other reactors, including TMI-2, predominantly have zircaloy leadwire strings and this phenomenon of large negative background currents following reactor trips will not be observed. 2. Analysis of Back-Up Recorder Data Two L&N recorders provide continuous incore detector output as a "back-up" to the plant computer. Each recorder has the ability to monitor 24 individual signals. Figure 3 illustrates one of the backup recorder charts after the trip. One channel indicated a signal corresponding to 100%FP while all other channels were zero. It was indicated by M. Co. lins (Reactor Engineer for CR-3) per a telecon (3/6/80) that only the first 18 channels are used on each recorder. The signal in question was transmitted by channel 24 and no incore detector was connected to this channel. This was further verified by SP-433. In-Core Neutron Detectors Channel Check, which listed only the first 18 channels for each recordar. Mr. Collins also indicated that this channel responded in the same manner during a reactor trip on 12/21/79. Furthermore, the channel in question was responding in the same manner on 3/6/80, over a week after the trip. 3. Analysis of Source Range Detector Response Source range signals ar. provided by two BF3 proportional counters located on opposite sides of the core. Figures 4 and 5 illustrates the response of each source range detector subsecuent to the trip. Both traces appear to be normal and decrease smoothly from an initial value of 1200 CPS. This response is consistent withan earlier trip on 9/21/78 shown in Figure 6. B-2 l

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' APDENDIX C - SEQUENCE OF EVENTS REPORTED ^ 3Y FLORIDA POWER CORo0 RATION Rev, 5 REVISION 5 ?*8* :- SEQUINCI (AS OF 2300 3/1/80) 26 Februa:7 *:ansian: C2-3 _EVINT SU OPS *S A: 14:23 cm Febntary 26, 1980 Crystal River -3 Nuclear 5:a:1cn experienced a reactor : rip from approxi=ataly 100% full power. A svuopsis of key events and parameters was obtained from.he plant computer's post-:: p review and plan alarm summary, the sequence of events monitor, control room s::1p charts, and the Shif: Superviser's log. The reacton was opera:ing at a-proxima:aly 100 : full power w1:h ' Integrated Control System (ICS) in auto =atic. No tests were in progress and =inor main-was being performed in he Non-Nuclear Inst =nentation (NNI) tenance cabina: "Y". Time Ivent Causa/C.~. ents 14:23:00 The following is a summa:7 of plan: cond1:1ons prior to the : rip nux 98.6% 'S 1C ?:sssure 2157 psig \\ \\ P:'.1 level 202 inches 9 Q MU tank level 71 inches "A" 599*F. Q "3" 600*F. 9 'O "A" 557*F. .C "3" 556*F. T d now "A" 73 I 106 lbs/L RC now "3" 73 I 106 lbs/h: Latdown n ov 48 gym CTSG "A" Iv1 (OP) 67* OTSG "3" Iv1 (OP) 65: OTSG "A" FRI.7 242 inch is CTSG "3" FILY 254 inchss OTSG "A" P:ssaure 911 p is OTSG "3" pressure 909 psig Main Steam ?: essure 394 psig Main Steam Tr g. 389 F. Condenser Vact um 1.76 Generated MEI 834 DTI level 12.7 ft. Feed.now "A" 3 I 106 lbs/h Feed now "3" 5 I 106 lbs/h: Feed Pressure "A" 970 psig Feed Pressure "3" 968 psig 14:23:21 +24 Vol: Bus Failure (NN: Cause still unknown. Apparently. Power loss "I" supply) the positive 24 VDC bus shorted i dragging the bus voltage down to a l C-1

.'I Rev. 5 Page : Ti:ne Event Cause/Com=ents lw voltage trip condition. There is a built-in k to y second delay at which time all power supplies will trip. There was no :1p indication on negative (-) voltage. This event was missed by the annunciator. Following the NNI power failure, much of the control room indication was lost. Of the instrum- 'entation that remained operable transient conditions made their indic-cation questionable to One ocerators. 14:23:21 PORY and Spray Open When the positive 24 VDC supply was lost due to the sequence discussed above the signal monitors in NNI changed state causing PORV/ Spray valves to open. De PORV circuitr; is designed to seal in upon actuation and did so. ne resultant loss of the negative 24 VDC halted spray valve motor operator and prevected PORV seal in from clearing on low N9 \\ pressure. It is postulated that the PORV opened fully and the spray valve stroked D for approximately b second. ne 40% open g \\j indication on spray valve did not actuate, therefore, the spray valve did not exceed 40~. open. 14:23:21 Reduction in Feedwater As a result of the "f' power supply failure many primary plant control signals responded erroneously. Tcold failed to 570*F (normal indication was 557*F) producing several spurious alar =s. Tave failed to 570*F (decreased). "'he rasultant Tave error nodified the reactor denand such that control rods were l l w'ichdrawn' to incrase Tave and reactor l power. D e power increase was te:minated l at 103* by the ICS and a " Reactor Demand l High Limit" alarm was received. Bot failed to 970*F (low) and RC flow failed E to 40 I 10 lbs/hr in each loop (low). Both these failures created a 3TU alarm and limit on feedwater which caduced i feedwater flow to both OTSG's to essentially tero. Turbine Header Pressure l failed to 900 psig (high) which caused the turbine valves to open slightly to s C-2

Rev. 5 Page 3 9 Time Event Cause/Cc=ments regula:e header pressure.hus incressing generated megava::s. 3ese combined failures resulted in a loss of hea: sink to :he reactor initiating an excessively high RO pressure condi:1on. 14:23:25 Reac c *:1p/Turbi=e 2x ::1p caused by high RCS pressure a: 2300 psi Trip Turbine was ::1pped by the reactor. 14:24:02 Ei Pressure Inj. This was a e m,,4ter pris.ou: and indicates Reg. (Flag) <50* subecoling.* See at ached graph of RC Pressure /*emp. vs. Tiae. This g aph is based on Pos: Trip data and ac:ual incore.hermo-couple data. From.he reac:c: ::1p point (14:23 to 14:33, core exit temperature da:a was obtained by extrapola: ion and calculated da:a. This is supported by.tvo alarm data points plotted a: 18* and 21* of subcool1=g during 3 chis period from the cesputer. It is 1:opor:an: 1 to note that loves: level of subcooling was 8*F for a very sher: period of time. 9 9

• NOTE: nis computer program was initia:ed as a result of the TMI incident.

14:24:02 loss of 3oth Suspect condensa:c pu=p ::1pped due to high Condensate Pumps DF" level. 2 1s is verified by 7777 printed by compu:er, indica:izig the level instrument was over ranged as well as a low flow indication in the gland stesa condense; as also indicated by computer. 14:25:50 PORY Isolated At this time a high 2C Orain Tank level ala:m was received. This was resultant from the PORY :==dning open and was positive indication that the PORY vas open. At this time, the operator closed the PORY block valve due to RCS pressure decreasing and high RC:r; level. 14:25:41 HPI Auto Initiation HPI initiated automatically due to lov ECS pressure of 1500 psig. The icv pressure condition was ' resultant from the PORY :==d*'ag full open while the pla=: vas tripped. Tull HPI was initiated vi:h 3 pumps resulting in approximately 1100 gym flow to the RCS. A: this :ime, all :==in*ng non-essential 1.3. 1 solation valves C-3

Fa 4 Ti:na Event Cause/ Comments sere closed per TMI Lessons-learned Guidelines. 14:26:54 RC Pumps Shutdown Operator turned RC pumps off as required by the applicable emergency procedure and 3 & *J small break guidelines. 14:27:20 R3 Pressure Increasing This is first indication that RCDT rupture disc had Yuptured. R3 pressure increase data was obtained from Post Trip Review and Strip Char: indication. 14:31:32 R3 Pressure High This alarm was initiated by 2 psig in R3. This is attributed to steam release from RCDT. Code safeties had not opened at this time based upon tail pipe temperatures recorded at 14:32:03 (Computer). 14:31:49 OSTG "A" Rupture Matrix his occurred due to <600 psig in OTSG "A. Actuation The low pressure was caused by OTSG "A" boiling dry which was resultant from the 3TU limit and failed OTSG 1evel transmitter. This resulted in the closure of all feedwater and steam block valves which service OTSG "A". 14:31:59 Main Feedwater Pump 1A Caused by suction velve shutting due to Tripped matrix actuation in previous step. 14:32-14:41 Es A/3 Sypass Manually bypassed and HPI balanced between all 4 nozzles (Total flow approximately 1100 gym -s==11 break operating guidelines). 14:32:35 Started Steam Driven Started by opera ~ tor to ensure feedvatar was Emergency Feedwater Pump avs41=hl.a to feed CTSG's. 14:33 Core Exit Temp. Verified The core exit incore thermoccupies indicated the highest core outlet temperature value was 560*F. RCS pressure was 2353 psig atthis time, I therefore, the subcooling nargin at this cine was 100*F. .v' ni nn ne subcooling nargin for the entire transient was 8* F. It is postulated that some locali:ed boiling occurred in the core at this point as indicated by the self powered neutron detectors. j 14:33-14:44 Started Motor Driven Emer-Same discussion as " Started Steam Driven Ener-gancy Feedvater Pump gancy Peedvater Pu=p." 14:33:30 RC Pressure High (2395 psig) At this point, pressuri:er is solid and code l safety lifts (RCT-8). This is the highest RCS pressure as recorded on Post Trip Review. 1 Apparently, RC7-8 lifced early due to seat

V n-e /rQ O u Rev. 5 D Page 5 o-Time Event Cause/Costents leakage prior to the transient and RCV-9 did not lift. 14:34:23 RB Dome Hi Rad Level RMC-19 alarmed at this point. Eighest level indicated during course of incident was 50 R/hr. Eigh radiation levels in RB caused by release of non-condensable gases in the press-urizer and coolant. 14:35:33 Attempted NNI Repover With-This rasulted in spikes observed du de-ener-i out Success gized strip charts. 14:36:50 Computer overload Caused i.:7 overload of buffer. Resulting in no further coeputer data until buffer catches up with printout. 14:38:15 TWV-34 Closed This valve was closed to prevent overfeeding OTSG "3" beyond 100% indicated Operating Ra=ge. 14:44:12 NNI Power Restored Success-NNI was restored by removing the X-NNI ?over fully Supply Monitor Module. This allowed the breakers to be reclosed. At this time, it was observed that the "A" OTSG was dry, the press-urizer was solid (Indicated off scale high), RC outlet temperature indicated 556*F (Loop A & B average), and RC average temperature indi-cated 532*F (Loop A & 3). The highest core exi-thermocouple temperature at this time was $31*F RSC pressure was 2400 psig (saturation temp. at this pressure is 662*F.). This data verified natural circulation was in orceress and the plaar subcooling margin was 131*F. (based on core exit thermocouples). 14:44:31 RB Isolation and Cooling Actuation At this time, R3 pressure increased to 4 psig and initiated R3 Isolation. The operator verified all immediate actions occurred proper 1: for HPI, LPI, and R3 Isolation and Cooling. The increasing R3 pressure was resultant from RC7-8 oassing HPI at this time. 14:46:10 Bypassed HPI, LPI and RB These "ES" systems were bypassed at this ti:se Isolation and Cooling to again balance EPI flow and restore cooling water to essential auxiliary equipment (i.e., RCP's, letdown coolers, CEM's etc.). C-5 i 6

D Rev. 5 O ? age 6 Time Even: Cause/ Comments 14:51:57 Rupture Matrix Actuation on The actuation was resultant from a deg-OTSG-3 radation of OTSG-B pressure. Cold emer-gancy feed was being injected into the OTSG at this time. This matrix actuation isola:ed all feedvater and steam block valves so the 3-0TSG and tripped the "T' main FW pu=p. Both Emergency W purps were already in operation at this time. 3-0TSG 1evel at this ti== vas 70: (Operation Range). 14:52 HPI Throttled and RCS At this time, the mmvdmum core exit ther=o-Pressure Reduced to 2300 couple tempera:ure was $15*7, RCS pressure psig was 2390 psig. Therefore, the subcooling margin was 147'F. Natural circulation was in effect as verified previously. All con-dicions had been satisfied to throt le EPI. Therefore, flow was throt. led down to approx-imately 250 gpm to reduce RCS pressure to j 2300 psig in order to attempt ':o reduce the l flow rate through RGV-8 and it.to the RB. 14:53 Reestablished Leedown A: :his time, the operator was at:empting to establish RCS pressure control via nor=al RC makeup and letdown. 14:56 Opened MU Pump Recire. This was done to assure the MU pumps vould Valves have minimum flow at all times to prevent possible pump damage. 14:56:43 Bypassed the A-0TSG Rupture Feedvater was slowly admi:ted Matrix and Reestablished - to the A-0TSG which was dry up to this point. i Feed to the A-0TSG Teodwater was admitted through the Auxiliary FW header via the EFW bypass valves. The feedrate was very slow in order to minimire thermal shock to the OTSG and resultant depres-i suriration of the RCS. RCS pressure control I was very unstable at this time. It is postulecec that some localized'beiling.occured in core at this point' as indicated by.self neutron l detectors. e C-6 1 1

i r. r e7 9tt T6e Event cause/ Comments 14:57:09 '3ypassed the 3-0TSG This was done to regain W control of the Rupture Mat:-iz 3-0TSG. Level was still high in this OTSG (approximately 65 Operating Range). Therefore, feed was not necessary at this ti:ne. The Main Stesa Isolation valves were open in preparation for bypass valve operation (when necessary). 14:57:15 Established RC Pump This was done 6 ;n paration for a RC? star Seal Return (when necessary) ant. to m4M m4:e pump seal degradation. t 15:00:09 Reestablished Level This verified feedvater was being ad=1tted to In A-0TSG the OTSG and nade it available for core cooling via natural circulation. Feed to this generator was continued with the intent of proceeding to 95% on the Operating Range. 15:00:09 77*.? Subcooled "1" Loop This value was based upon "A" RCS loop parameters at this cine. The "A" loop was being cooled down at this ti=a by the A-0TSG fill and t'se operator was attempting to equalize loop temperatures. 15:15 23*F Delta-T/ Manned the At this time, loop temperatures were nearing Technical Support Center equalization. This delta-T was calculated from loop A & 3 T 's and core exit therse-e couples. 15:17 Declared Class "B" Emergency This was done based on the fact there was a loss of coolant through RC7-8 in the ~ containment and HPI had been initiated. All non-essential Citi i personnel were directed to tvacuate and;c'ontact off-site agencies *be- 'gan. Survey" team was'sent to Auxiliary Building. 15:19 Opened Emergency N 31ock M this point the A-0TSG 1evel was increasing to 3-0TSG and the decision was made to commence filling the 3-0TSG simultaneously. The intent was to go 95% on both OTSG's without exceeding RCS cooldown W s (10&'F/hr) while naintaining RCS pressure control. i l l C-7 \\ ,--m.-. y, __,r

~ 9 m Rev.5 D Page 3 e @g Time Event Cause/C==ents 15:26 La Level Alarm in Sodium This was recultant from the tank supply valve Hydroxide Tank opening when the 4 psig R3 isolation and cool-ing signal actuated. The sodium hydroxide was released to both LPI trains. Sodium Eyd::ixide was - Mtted to the RCS via.H?! from.the 3*4ST. (Approximately 2. ppm injected into the RCS.) 15:50 Terminated HPI At this time, all conditions had been satis-fled (per small break operating guidelines) to terminate HPI. RCS pressure control had been established using normal makeup and latdow :. HPI was terminiated and essentially all releases to the R3 were discontinued. 16:00 Commenced Pressuri:er At this time, RCS pressure and temperature Heatup were well under control. Natural cir:.ulation was functioning as designed (approxinately 23*? delta-T). RCS temperature was being =aintained at approximately 450*. RCS pressure was approx-imately 2300 psig. The decision was made at this point to commence pressurizar heacup in preparation to re-establish a steam space in the pressurizer. 16:07 Survey Tesa Report The Emergency Survey Team reported no radiation survey results taken offsite were above back-ground. 16:08 :04 Shutdown Steam Drive Emergency N Pump The motor driven Emergency W pump was twMng, therefore, the steam driven pump was not needed The plant remained in this condition for app-roximately 2 hours, while heating up the press-urizar to saturation temperature for 1800 psig. 16:15 Press Release helf s was notified of plant status. 18:05 Established Steam Space Pressurizer At this point, pressurizer temperature was approximately 620*?. Pressurizer level was brought back on scale by increasing letdown. From this point pressurizar level ves reduced to normal operating level and normal pressure was established via pressure heaters. 18:30 Terminate.1 Class 3 Emergency State and Federal Agencies notified. C-8 i

/ p' e9 Time Etent Cause/ Comments 21:07 Forced Flow Initiated The decision was =ade to re-establish forced in RCS flow cooling in the RCS at this time. 3&W and NRC were consul:ed. RC?-I.3 and ID vere started. A: this point, RCS parameters were stabilized and =a:.ntained at RC pressure-2000 psig, RCS tempera:ure-l.20*?. Pressurizar level-235 inches. The plan: vas considered in a nor=al configuration. i l l i l C-9 .}}