ML20211E032
| ML20211E032 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, Rancho Seco, 05000000 |
| Issue date: | 09/26/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19292G087 | List: |
| References | |
| NUDOCS 8610220338 | |
| Download: ML20211E032 (24) | |
Text
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TRANSIENT ASSESSMENT PROGRAM REPORT FOR RANCHO SEC0 REACTOR TRIP ON JUNE 17, 1981 PREPARED AND RELEASED BY SACRAMENTO MUNICIPAL UTILITY DISTRICT TAP NUMBER - RS-82-04
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8610220338 860926 2 ADOCK 0500 Transient Assessment Program gDR 12-1132714-00
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SUMMARY
A. Event Description Rancho Seco was operating at 80% full power (FP) on June 17, 1981.
All Integrated Control System (ICS) stations were in automatic mode.
Two reactor runbacks occurred at 14:52 and 15:05 due to " loss of feed-pug " signals. Following the second runback, the "A" main fee &ater pump (MFP) tripped on overspeed. The "B" MFP did not pick up the load fast enough, and the resulting loss of feedwater caused the reactor )
l to t ri p on hi gh p ress u re. Following the reactor trip, a modulating atmospheric dump valve (MADV) failed to rescat properly, causing a loss of steam pressure until the MADV was manually isolated. Because of the stuck-open MADV, the RCS stabilized at a lower temperature than no rmal for a reactor trip. A turbine bypass valve control linkage fell of f 19 minutes af te r the t rip, causing a further loss of steam pressure and lowe r RCS tempe rature until the linkage was re conne ct e d .
B. Initiating Event An in te rmi tten t failure in a control module set the "A" MFP to minimum s pee d, initiating a reactor runback. The module subsequently corrected i tsel f, and the pump t ripped on ove rspeed when it tried to respond to the exis ting hi gh-demand s i gnal . The resul ting loss of feedwater caused reacto r pressure to increase, reaching the high pressure trip setpoint of 2290 psig.
C. Root Cause The root cause of this transient was an intermittent failure in a modul e i n the "A" M FP con t rol l e r.
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o II. TRANSIENT ASSESSMENT A. Sequence of Events This is an abbreviated sequence of events taken from the Event Log, Alarms Typer, Shi f t Supervisor's Log, and Control Room Log. The re is a discrepancy between the times recorded by the Bailey and MODCOMP comp u te rs . Since the MODCOMP agreed rest nearly with the wall clock, as recorded in the Shi ft Supervisor's and Control Room Logs, the times recorded by the Bailey have been adjusted to the MODCOMP recorded trip t i me o f 15:08:38.
N14:52 Runback due to " Loss of Feed Pump" signal N14:58 Increasing Powe r to 80%
sl5:05 Runback due to " Loss of Feed Pump" signal l
l 15:05:20 RPS RC Loop ASB High Pressure Alarm -
15:05:50 RPS RC Loop A&S Pressure normal l- 15:07:42 BFP A Trip, on overspeed (as evidenced by review i data and operator logs) 15:08:36 RPS RC Loop A&B High Pressure Alarm 15:08:38 RPS Chan A&C High Pressure Trip Reactor / Generator Trip
. 15:08:39 RPS Chan D High Pressure Trip
! 15:08:48 OTSG A&B Outlet High Steam Pressure Alarm l
15:08:50 RPS' RC Loops A&B Low F ressure Ala'rm
! 15:08:54 BFP A Trip (According to Events Log) r 15:08:56 HPI Pug B Started (Manually) i i
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15:09: 18 RPS Chan B Low Pressure Trip 15:09:40 Pressurizer Low Level Alarm
( 15:10:33 OTSG A&B Outlet Low Steam Pressure Alarm 15: 14:13 BFP B Trip 15:15:01 OTSG B Low Level Alarm ,
$15: 15 MADV isolated 15:16:26 OTSG B Low Level Alarm Clear 15: 17:55 HPl Pump B Shut OFF N15:20:30 Pressurizer Level Normal c.24:01 BFP 8 Started 15:27: 16 8FP A Started sl5:27 Turbine Bypass Valve Linkage Fell Off 15:27:34 OTSG B Low Level Alarm 15:29:33 OTSG B Low Level Alarm Clear 15:30:50 Heat / Cool High Rate Alarm 15:33: 16 RC Pump D Tripped 15:35:20 HPl Pump B Sta rted
$15:37 Turbine Bypass Valve' Linkage Reconnected 15:38:50 Heat / Cool High Rate Alarm Clear 15:40:47 HPI Pump B Shut Off B. PLAHT PERFORMANCE
- 1. P re-Trip Review Rancho Seco was operating at 80% full power on June 17, 1981, with a generator output of 720 MWe. All ICS stations were in automatic mode except A T . A new feedpump control . system 1ad been Installed during the recent refueling outage. The MFP's had been operated in the manual mode earlier in the day due to pump control problens encountered when in the automatic mode.
They were returned to the automatic mode several hours before the trip.
- 2. Initiating Event Two power runbacks occurred prior to the trip (at 14:52 and 15:05 due to loss-of-feedpump signals). Both runbacks were caused by an i n te rmi t ten t f ai l u re i n a con t rol module fo r the "A" MFP which set the pump to i ts minimum speed. in both cases
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.- I the module corrected itself after a short period of time (approximately two minutes) . After the fi rst runback, the plant recovered; and at 14:58 the operators began to return from 76% to 80% pcwer. It was during recovery from the second runback that the trip occurred.
When the faulty control module corrected itsel f, the "A" MFP received a high-demand signal due to the mismatch between FW demand and FW flow. This caused the MFP turbine governor valves to open fully. Shortly thereaf ter, the pump t ripped,-
apparently on an overspeed" signal .* The "B" MFP di d not respond in time to maintain adequate feedwater f' low. The insufficient FW flow caused inadequate heat removal, al l owi ng RCS temperature and pressure to increase. The reactor tripped when the high pressure trip setpoint of 2290 psig was reached.
- 3. Pos t-Trip Response The plant response to this transient is acceptable. Af te r the reacto r t rip, a Modulating Atmospheric Dump Valve (MADV),
PV-20562C on the B main steam line, stuck open--causing the steam generators to depressurize. This caused a greater than normal RCS cooldown. OTSG "A pressure dropped below 747 psig and OTSG "B" pressure dropped below 732 psig. The operators isolated the MADV between 6 and 7 minutes af ter the trip, halting the steam pressure loss. The RCS hot leg temperatures stabilized at 520 F after the MADV was isolated and 0TSG pressures had stabilized. This is below the normal post-trip window (s545 F) .
The minimum subcooling margin following the trip was 47 F.
l The pressurizer level responded typically to the trip. The level dropped rapidly as the RC volume shrank due to decreasing te mpe ra tu re. HPl was manually started into RCS Loop A (Nozzlel) shortly after the trip, which slowed the level decrease. A minimum level of 37 inches occurred 6 minutes after the trip,
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- Either high speed or high discharge pressure can cause an MFP trip.
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i, before recovering to the normal level of 150 ISO Inches.
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.Approximately 6 minutes after the reactor trip, the "B" MFP tripped and auxiliary feedwater started automatically. The cause of this trip is not known. Sufficient feedwater flow was maintained to prevent the OTSG's f rom drying out.
A second cooldown began about 19 minutes af ter the trip when a linkage to a turbine bypass valve (PV-20564) fell off. The linkage was reconnected and the minimum RCS tem-perature due to this cooldown was 496 F. The plant was then returned to 527 F. The total cooldown did not exceed the Tedi Spec limit of 100 F/hr.
- 4. Operator Actions and P rocedurs Adequacy Following the reactor trip, the operators followed the Reactor /
Turbine Trip Procedure (D.2) of the Plant Operations Manual Emergency Procedures. Reactor t rip and decreasing neut ron fl ux we re ve ri fied. The operator observed steam pressure below the Turbine Bypass System setpoint of 1000 psig, located a partially-open MADV, and manually Isolated the valve. Because of dropping RC pressure, the operators manually started the "B" HPl pump. The pressurizer level remained on-scale and returned to the normal operating range af ter the MADV was isolated.
After the reactor trip, the "B" MFP tripped and Auxiliary Feed-water (AFW) started automatically. AFW was veri fied to be con-trolling 0TSG level at 20 to 24 inches.
After plant conditions had begun to stabilize, a turbine bypass valve linkage fell of f--causing a further cooldown. An operator was dispatched and reconnected it. The operator secured reactor coolant pump D when RCS temperature reached 507 F. (The minimum temperature for 4 RC pump operation is 500 F.)
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C. SAFETY CONSIDERATIONS There were no safety implications associated wi,th this transient.
The F4 actor Protection System functioned as designed.- P ressuri ze r level remained on-scale and adequate subcooling margin was main-tained at all times. The PORV was not- challenged and the 3FAS was not initiated. Steam generator levels were not lost. This transient is classifled under 8C in the General Functional Specifications for Reactor Coolant System Congonents.
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9, TRANSIENT ASSESSENT I REPORT b.e 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 A5 AVAILABLE TO US THROUGH MARCH 9, 1980. THE INFORMATION AND CONCLUSIONS HEREIN ARE PROVIDED SUBJECT TO FURTHER YERIFICATION. 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
- f Prepared By: _i im .aA
^
R iewed By- eviewed By: b~ v Reviewed By: 41 (( ' ' / ' Reviewed By: / es J #8
.~/ ~ f y
/
Babcock & Wilcox Company l
( _\ TABLE OF CONTENTS Page ( I. EVENT SYNOPSIS 1
- II. PERFORMANCE EVALUATION AND RECOMMENDATIONS S 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 HP! 17
- 7. RC Pump Trip 17 r, 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
- i. A. Reactor Vessel
- 82 B. Reactor Vessel Internals 82 C. Steam Generators 82 D. Pressurizer 83 l
E. Reactor Coolant Piping 84 F. Reactor Coolant Pumps 84 G. Control Rod Drive Mechanisms 84 s .. H. Safety Valves 84
- 1. Fuel 85 V. REFERENCES 86 e
. . _ _ _ . _ _ _ - _ . _7_____.-____, ,_ . . _ __
( \
- TABLE OF CONTENTS (Cont'd)
Page _( i APPENDIX A - INSTRUMENT STATUS A-1 THRU A-5 TABLE II SIGNAL VALIDITY TO C0t9 UTER, CONTROL
~
. ROOM INDICATOR, HOT SHUTD0WN PANEL, AND ICS TABLE 11 VALID OPERABLE ALARMS I' TABLE II VALIDITY OF NNI, O!GITAL INTERLOCK AND la CONTROL FUNCTIONS FOR RC MAKEUP 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 i.
t e I t 0 t e
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.. GLOSSARY OF ABBREVIATIONS Differential pressure ,'['
dp delta p
. - - delta T Differential Temperature AT AFW Auxiliary Feedwater ._ EFW Emergency Feedwater -- ESFAS Engineered Safeguards Features Actuation System FSAR Final Safety Analysis Report 'I 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 p ressure Injection MFW Main Feedwater a MT Main Turbine NNI Non-Nuclear Instrumentation OTSG Once Through Steam Generator PORY Power Operated Relief Valve RB Reactor Building . RC Reactor Coolant
- i. .
RCS Reactor Coolant System l RCV Reactor Coolant Valve
- RPS Reactor Protection System SEM Sequence of Evants Monitor Tave Average Temperature of the Reactor's Inlet and Outlet Temperatures Tc Reactor Inlet Temperature Tcold T
e
. s
~
.. h
,, GLOSSARY OF ABBREVIATIONS (Cont'd)
Th Reactor Outlet Temperature Thot X Power bus which supplies 120 VAC and plus and minus 24 VDC to NNI instrianentation and control systems Y Power bus which supplies 120 VAC and plus and minus 24 VDC to NNI instrissentation l F
- h, i
l t e 9 4 e e e e 5
e t
- 1. EVENT SYN 0PSIS
/
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 computer'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. l, Prior to the incident, the reactor was operating at approximately 1005 FP with
~
Integrated Control System (ICS) in automatic. No tests were in progress and minor maintenance was being performed in the Non-Nuclear Instrimentation (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 r.
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. Alarms recorded at this time indicated that the reactor inlet temperature was erroneously indicating approximately 5700F (instead of a normal 5550 F) and that Reactor Coolant System (RCS) flow indication in one loop was erroneously indicating 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 NN! "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 both 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 s.. 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.
Computer alarm printer. Time
**difference Times are from the between t 8@he Sequence of Events Monitor and
855 Computer is + 5 seconds.
*** Times related to reactor trip
_ _ - - . . . _ _ _ _ _ . - . _ _ . _ _.-.____.._.-,.,_w. . _ . _ . _ _ _ - - . - . _ _ . . ~ - , - , . . , , _ - _ _ - .
e t
,s Event Synopsis
~'t 14:23:45** The reactor tripped on high RCS pressure at 2300 psig. 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 PORV. 14:25:50* The Reactor Coolant Drain Tank level alarmed high due to the
'..'. (65 sec) accismulation of RCS coolant being released through the open PORY.
s.. 14:26:41* RCS pressure decreased to 1500 psig and the Emergency (176 sec) Safeguards System activated the High Pressure Injection (HP1) f.' System which started two additional HPI/ makeup pumps. The full flow of three pumps was now being injected to the RCS. I 14:27:04** The four Reactor Coolant Pumps (RCP's) were turned off in k- 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 soray line isolation valve were
. (255-495 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 *A. 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 sec)
{ 14:33:15 RCS code safety relief valve opened (time approximate). L. 14:33:30 (570-585 sec) 14:34:33* Reactor building dome high radiation alarm. (648 sec) e (Note 1) At end of this section s _
r e t Event Synopsis 14:44:12* NNI "X" 24 VDC power supply re-energized. This provided I' (1227 sec) reliable indication of the primary. and secondary parameters. 14:44:31* Reactor butiding 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 (app rox. ) alarms. B&W established emergency communications and support i, to CR-3 control room from Lynchburg.
14:46:10 The reactor building isolation was bypassed to allow injection l (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. ) I 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 tatrix and feedwater to OTSG A was (app rox. ) 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) 15:15 Verified that natural circulation cooldown had been (app rox. ) established on both OTSG's with approximately 230F reactor differential temperature. The Technical Support Center was manned. 6d
%e
j
. i i
l Event Synopsis J 15:17 A Class "B" accident was declared and evacuation of r' ( app rox. ) non-essential personnel from the site was initiated.
'~
15:19 Connenced 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 (app rox. ) Terminated HPI, flow and established Makeup and Letdown control of RCS pressure. At this time the RCS and pressurizer were completely filled with liquid (" solid" system). (, 16:00 Connenced 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 1D. RCS pressure 4 (appro x. ) was approximately 2000 psig, Tave was was approximately 4200F and pressurizer level was 235". This returned the
< plant to a normal shutdown condition and terminated the abnormal transient.
NOTE 1: The time of the PORV 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 into the
! RC drain tank for at 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.
%e
_ - , , . , ,..--,---w,. ,,,- - - - . - , - - - , - - - - - - - - - - -
, o t I II. PERFORMANCE EVALUATION AND RECO M NDATIONS
. . .i A. Expected Plant Performance and Deviations a
{, 1. Initiating Cause Assessment
- The distribution of essential AC electrical power at the Florida i
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, IWI-X,
- ,, and NNI-Y. Instrumentation, transmitters, recorders, indicators, i
etc., is assigned to one of these two sources as discussed in ! s Section II. A.2. 5- The Non-Nuclear Instrumentation (NNI) power distribution system for NNI-X channel is shown on Figure 11-2. A similar power [~ distribution system exists for NNI-Y and is shown on Figure 11-3.
... The power distribution system provides + 24VDC as required for -
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 +24VDC 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 Sl 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 3D.
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 i- 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 , the NNI since the diode auctioneering circuit will allow the other j ,, power supply to continue to function. l l (. t . I
, t
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 s
inputs are used to initiate an alarm relay contact to annunciate
~
to the operator either a loss of one of the AC 118VAC 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 S2 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 CAI 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 S1 and S2. When they could not reset S1 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 51 and S2. 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-u appears to have been located in a Loop B wide' power range supply 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 afgs and without the power supply monitor this current could burn through a shorted component.
L. In summary, the NNI-X power supplies and power supply monitor all functioned as expected during a short circuit event. i
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- 2. Loss of Instrumentation Function
.)
This section identifies which displays, controls, ennunciators, P interlocks, and signals were valid and/or available to the L operator during the time that the NNI "X" channel DC power supplies were tripped. Appendix A provides the detati results of p' an evaluation of these signals and contains the following g, tables: Table II-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 p control room annunciator for the Reactor Coolant, Secondary Plant and Makeup and Purification Systems. Table II-3 addresses the digital interlock and control functions f performed by the NNI for the Reactor Coolant and Makeup and L 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 I. 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 c Features Actuation System.
s l- Valid RC wide range (0-2500 psig) pressure indication for Loop A was available at the Hot Shutdown Panel. The I information is displayed on an indicator and is derived from L the Engineered Safety Features Actuation System. I' w. s. m i S9 [
l . o l e, e t' Valid RC outlet temperature (520-6200 F ) indication for Loops
, A and B was available in the Control Room and at the Hot Shutdown Panel if the sensors powered by the "Y" cabinet were r- selected. The information is displayed on indicators.
i Valid RC inlet temperature (520-6200 F ) 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. r- Valid RC inlet temperature (50-6500F) indication for Loops A i l' and B was available in the Control Room if the sensors powered by the "Y" cabinet were selected. The information is displayed on indicators.
- i. , Valid uncompensated pressurizer level was available in the Control Room from RC1-LT3 via the computer.
r, j' Valid uncompensated pressurizer level was available at the hot shutdown panel from RCl-LT1 and RCl-LT3 and is displayed on
,., indicators.
i . 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 t 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. s f 6 6. I s., L t L.
\e 's
9* %
- 3. Opening of the P0tV and Pressurizer Spray 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
, s was removed ^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 command to the PORY (see Figure II-4) and, until power is mstored, the circuit not be able to send a close l command. The NNI would also send an open command to the pressurizer
- t. spray valve (see Figure 11-5) for the period of time from the loss of the +24VDC to the loss of -24VDC (approximately 0.5 seconds). A close i
r- 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 g state was monitored. The +24VDC power was momentarily removed and restored. The -24VDC was then removed and restored. The results of
, these tests were: ' '
L. 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 I
signal converters and " " first on newer models). r. I . I I L f w. O t
.e _ , , . . ,.n., , ._ - - _ , - . - . - - - - ,. - . _ , , , - ,,----~~---r -- - --
~
, o m qPl! PORY Activaticu circuit SIGNAL NDPITOR g
" During nomal conditions when the sisNAL NOT usto
. RC Narrow Range Pressure exceeds M NITOR 22 RC NARROW the high setpoint, the "H" con-N/L
" tact is Closed. When the pres-RANGE PRESSURE g 27 sure is reduced, the "H" contact opens, ht When the pressure decreases be-low the low setpoint, the "L"
-24(DCPower 'a contact opens. When pressure
" from NNI-X -
returns to above the setpoint, NOT usEn
- t. T the contact closes restoring (a) PORV Activation Circuit Signal flonitor the control circuit.
cr. r P,efore the loss of the +24 VDC ( H :: ;dL ;f 20 power supply, the "H" contact was open and the "L" contact was N- closed. The K20 relay and PORV ( e [y/ / 20 solenoid were deenergized.
/
r PORV s0:.EPOID ,
; RELAY (n0T ENERGl2ED)
(NOTENERGt2ED)
' Simplified PORV Control Circuit
. (b) n efore Loss of NNJ "X" Rp$$
f l At the time the +24 VDC power
, N ;d pdL pg,20 supply was shorted, the PORV signal monitor closed the "H" M20
/ relay. The K20 relay ener-V gized closing contacts to the f
' PORV solenoid and a " seal-in"
/ contact in'the K20 circuits
{ E AY E RG D g
. (ENERGlZED)
AT THE TlHE OF Loss CF NNI 1 +24 VOLT Buss (c) Simplified PORV Contr5TWcDit
]L h
After the $24 VDC power supply NC pdL % 20 was tripped, the "H" opened, but
/ . the K20 relay and the PORV sole-g20 / noid remained energized because
- v. Y the "L" contact could not open.
t /
/
R20 PORY SOLEN 0ID i - RELAY (EMERGl2ED) l (ENERGl2ED) AFT FIGURE 11-4
.(d) ER Los5 0F kNI X BU$s._ _ Simplified PORV Page Control 13 Circuit
( PRES $URIZER SPRAY VALVE During nonnal operation when the RC
" Narrow Range Pressure exceeds the
~
s.- NOT USED high setpoint, the "H" contact closes g
# which causes the pressurizer spray r- 9 motar valve to energize in the open
' slanAL
~~
H direction. As pressure decreases, RC uARRW RAn{E NOMITOR T the "H" contact opens and the spray yalve remains at that position. I, , ' PRESSURE aft When pressure goes below the low 1 y _ NOT usEn setpoint, the "L" contact closes and the spray valve's motor is en-i' y ergized to close the valve. As the (' pressure raises above the setpoint,
~2 r -
L the "L" contact opens and the sys-f y 7, tem is restored. a -r---------------- SIMPLIFIED PRES 3URIZER SPRAY MOTOR OPERATED VALVE CONTROL r* (.
--H _, L I - Before the loss of 424 VOC, the I spray val.ve was shut and the motor
${, CONTROL CLOSE operated valve control was de-
'
- energized.
(DE-ENERGl2ED) (DE-EMERGl2ED)
/
BEFORE LCSS of Mul X BUS t At the time of the loss of flNI's "X" +24VDC Bus, the "H" contact
"[ H ]L Closed which energized the Motor Operated Valve (f10V) open control.
The valve started to open. MOV C4NTROL CLOSE {
\/ RG12[D) / (DE-EMERGlZEO
( i-
, AT THE TIME 0F LO33 0F MNI "X" 24Y BUS s
! After the 424VDC power was removed, L the "H" contact opened. This ac-
-H --L tion deenergized the MOV open con-
~~ -- -
trol and stopped the pressurizer spray valve in a partial open posi-
'- tion.
$[, CONTROL CL0$E (DE-tMinGizEo (DE-EMERGl2ED)
AFTER L0ss er hul x 2sv Bus pAGE 14 FlCURE 11 5
I 1.
- 4. Control System Actions Before Reactor Trip
/
- 1. ICS Signal Status r'
s The interruption of the NNI "X" power supply caused several NSS process signals to fail which are inputs to the ICS. Among the r- signals important to the ICS performance are: i
' RCS Hot Leg Temperature (Th I Steam Generator Startup Levels I' RCS Average Temperature (Tave) calculated from RCS hot and
,. cold leg temperatures Steam Generator Operate Range Levels p Feedwater Control Valve Differential Pretsure Steam Generator Outlet Pressure (AaB)
,, Total RC Flow - calculated from individual loop measurements I
- t. 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:
f Setpoint or Failed
, Signal Normal Value Value Th 600 570 Tc 556 570 t
;, 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%
.. WW Control Valve' Differential Pressure 50 psi 50 psi , Steam Generator Outlet Pressure 1000 psi 600 psig I
- Low load or post trip level setpoint, RC pumps on
** Post-trip level setpoint with RC pumps off 15-(
- s. :
n In the period before and after the reactor trip (up until NNI "X" power was tj restored), some, or all of the NNI signals listed above, reached the failed value. The preliminary data assessment indicated the following: O Signal Loop A Loop B t.) Th Failed OK G Failed Failed i.} Tc q ' Tave -----Failed---------- Total RC Flow Failed OK Loop RC Flow Failed
'l OK Turbine Header Pressure -----Probable Failure-----
Steam Generator Outlet Pressure Failed OK SG Startup Level Failed Failed l SG Operate Level Failed OK l 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:
k. (a) A rapid reduction in feedwater flow to the A and B generators, f responding to the indicated drop in Thot and the indicated
; RC flow reduction.
, (b) A control rod withdrawal to bring Tave back toward its setpoint.
Power increase was limited within the ICS to 103%. 1 (c) Opening of the turbine control valves in response to the indicated
- rise in turbine header pressure.
This response was consistent with expected ICS performa'nce, given the r invalid input signals. L L. f'
% .a i
% d e
1
I'
- i .1 C
L- 5. Reactor Shutdown t' The control system's actions cau. sed an increase in the actual RCS
, Ta , and the RC system pressure, although the PORY was open, wnf!h led to a high pressure reactor trip at about 2300 psig some 10 g to 25 seconds af ter the NNI "X" electrical upset. Pressure response is indicated on the wide range control room recorder chart, Figure b 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 7, Following reactor trip, the relief of steam through the open PORY
! caused RC Pressure to fall rapidly as shown in Figure III-3.
i. The Emergency Safeguards Features Actuation System (ESFAS) initiated high pressure injection automatically as designed, when the RC pressure fell to approximately 1500 psig. j- 7. RC Pump Trip B&W instructions specify that the reactor coolant pumps (RCP's) shall be tripped inneidately after initiation of ESFAS by low reactor { coolant (RCS) pressure. Low RCS pressure alarms were received at
- t. 14:24:43, alerting the operators ^ hat RCS pressure was decreasing.
The low pressure ESFAS initiation alarm was annunciated at 14:26:41. r 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 i 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 b- These actions are specified to isolate a small break, if the break is downstream of the PORY block valve or the letdown isolation valves or
,f ' between the spray block valve and the spray control valves.
i .. The operators had received a low RCS pressure alarm at 14:24:43,
, following the high pressure trip. A low pressure ESFAS actuation I 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
... reactor trip in accordance with applicable procedures. The operator then manually closed the PORV block valve and the spray block valve in accordance with the Small Break instructions.
i.
L
- 9. Control of High Pressure Injection The High Pressure Injection (HPI) system was actuated automatically
[ on the low pressure ESFAS signal. Three HPI pumps functioned, L providing a flow rate in excess of 1000 gallons per minute at the airimum RC pressure of 1325 psig. Pump suction was from the Borated P Water Storage Tank (BWST). The HPI water was injected through the i~ normal safety injection nozzles, in the four cold legs between the reactor coolant pisap discharge and the reactor vessel inlet.
~
Initially, the operator chose not to throttle HPI flow. The operator
.- apparently could not immediately 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 p::wer loss) r the operator could assess plant conditions. They were as follows: e HPI cooling in effect, and RCS filled with water (pressurizer level off-scale high, RCS pressure oscillating at about 2400 I~ 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).
l
'HPI injection was then throttled to about 250 gpm. This is i consistent with instructions allowing throttling when pressurizer level is high and subcooling margin is greater than 500F.
Natural circulation was then established in the A steam generator by raising the secondary water level and observing the proper differential temperature ( 30*F) across the generator.
- i. The control room pressure trace shows RCS pressure oscillating between 2250 and 2400 psi. This indicated that the pressurizer code i safety valve was cycling as HPI was adding water. HP! 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. I Some point HPI flo.s rates were read by the operators during HPI I L- injection, but no continuous recording of HPI flow vs. time is available. Based on 8WST levels before and after the incident, about i l 47,000 gallons of HPI water was injected. About 43,000 gallons were l L. discharged through the PORY and code safety valve to the RC drain tank and thence to the containment building basement. The remaining i 4,000 gallons filled the pressurizer steam space, and was discharged ! to a reactor coolant bleed tank when the pressurizer steam space was l established.
~
I j .. l
~
I I l. r~ [, 10. PRESSURIZER RELIEF VAL'VES r, 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 f' the seat leakage experienced by this valve prior to lifting. Seat L leakage causes the thermal expansion of the valve internals, and the thermal expansion increases the areas exposed to the escaping steam e- 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.
I' Ouring the two hours that the safety valve appeared to be open, it i, passed saturated steam, two-phase flow and water. The pressure and temperatures seen by the valve are ricorded in Section 111. The j- valve appears to have resented at approximately 2300 psi or less than (, a 55 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. Af ter approximately 5 minutes, the PORY isolation valve was manually closed. During the time the PORV was open, it is likely that the PORV passed only saturated steam, based on event simulation. The pressure range experienced was approximately 2320 to 1350 psig as shown in Figure
!!l-4.
The PORY appears to have performed as expected.
- 11. Steam Generator Cooling (Reactor Trip 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"
- i. level on steam generator startup range instrtsnentation (50% on operate range if no RC pumps are operating).
r 4 (b) Transfer turbine bypass setpoint from 885 psig to about 1000
'- psig for steam generator pressure control.
l (c) Transfer steam generator pressure control signal from turbirie
- 6. header pressure to steam generator outlet pressure.
~.
j
' As a result of the input signal failures, a preliminary evaluation J indicates:
P (1) The selected startup range level indications for both steam generators failed to mid-position, indicating a level of j ', 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 mactor coolant I pumps were tripped at about 3-1/2 minutes into the event, the ICS
- 1. . 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 generator, 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
., feedwater 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
,, initiated and controlled by the operator at about 10 minutes into i the event. The B steam generator level was increased to about
- 1. 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 f flow was established to bring the A steam generator level to 50% [, on the operate range.
,. (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--indicating a valid pressure measurement), and invalid in the A steam generator (e.g., loss of pressure such that the rupture matrix was activated leading to isolation from l ,, main feedwater).
- 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 !!!-34.
s t l t
l I
- f. .
Automatic actuation of the auxiliary feedwater flow may have been _ prevented by the mid-range failure of the sts.am generator startup range level signals. A valid low level signal is required from
- both steam generators to actuate the auxiliary feedwater system, i 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 feedwater was supplying feedwater through the startup valve I at this time. After AFW initiation steam generator flow control L, 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 B 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 I. was maintained until RC Pump restart at 21:07. 1.
- 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 af ter 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 co e outlet temperature data during the first ten minutes of the transie'.t are incomplete, the highest core outlet temperature f' recorded in the data is 600'F. The minimum subcooling appears to have L been reached at approximately six minutes after reactor trip. From this point on, subcooling margin increased as shown in Figure !!!-8.
Heat transfer to the RCS fluid from the core, primarily in the thermal convection modt, appears to have continued throughout the transient.
- In these circumstances no fuel cladding failures would be expected.
N J l
. _ _ . _ - -,,.__.,,--_._-,m .
- l. -
,., - 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 resumed by the restart of two reactor coolant pumps.
_ The evidence from measured radiation levels in the containment butiding 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 as follows: e 1. The radiation montfor on the letdown line, RM-L1 (which is recorded on RM-RS) showed a continuously decreasing activity level following reactor trip. Figure 111-41 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
- k. 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
- i. 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 I 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 !!.A.15 presents the technical basis for this conclusion. These data do not preclude that a small number of fuel mds (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.
f.. e L.
~
l W
i.
~
To further demonstrate adequate core cooling, the responses of a various neutron detectors at Crystal River-3 (CR-3) have been analyzed to determine if they are indicative of any localized boiling F. or voiding as was seen at TMI-2. The sources of these data were from l ,' the alare 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.
>Q
~
No localized boiling can be inferred from the core instrimentation responses. A description of the core instrumentation indications and
. their meaning is provided in Appendix 8. -
- Based on these analyses, we conclude that adequate core cooling continued at all times throughout the transient.
- 13. Restoration of Pressurizer Pressure Control
{,
- l. After high pressure injection was terminated, pressure control was achieved by balancing letdown and makeup. RCS temperature was j- controlled using natural circulation and steam generator pressure i
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 ; temperatures. The plant was stable, with about 1400F subcooling. A
- i. 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
- f. radiation monitor readings. Further, the condenser air ejector i
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
- k. e Reestablish normal pressurizer control, then cooldown by either natural circulation or forced flow.
l 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 j on the decay heat removal system. 6. 6 b. t t 4
[.
! 1.'
i ' ' Pressurizer water temperatum 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 l c 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 1 g anticipated, and none were encountered. b Pressurizer water temperatum increased about 400F per hour with all
- heaters activated. When water tesperature reached about 6200F ,
~
letdown flow was increased (to a bleed storage tank, to maintain i ,, proper makeup tank level control) and a steam space was formed. There j were no unexpected pressure or temperature swings during this
- p evolution.
E- 14. Restart of Reactor Coolant System Pumps l {,, RC Pump seal data had been retrieved on an hourly basis and reviewed by B4W. 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 O pump was chosen for startup in loop 2.
t l There was a differential temperature of about 25'F across the core, { i, before RC pump restart. After pump start, this differential i 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 j , taps. The actual system pressure would not spike. A pressure
; i decrease of about 150 psi, and a presurizer level decrease of about 20 l' i inches, was shown due to bulk average temperature decrease. Neither condition was detrimental to the plant, but the expected response was l
discussed thoroughly with the Crystal River Unit 3 Operations
,' Technical Advisor so that the operator would recognize such changes as ,
normal. '
!suediately before planned pump start, low oil level alarms were
- received on the upper oil pots on RC pump B40 motors. 84W evaluated these alarms as thermal contractions of the oil volume, and not detrimental to pump restart. This was discussed with Crystal River L Three personnel and the alarms were bypassed. The operators were instructed to closely monitor the motor upper thrust bearing
. tesperatures, and trip the pump if this temperature exceeded 185'F.
G
- o ad i
_,_-----,.,__-,,,_,._,,,_,_,,_~,,-..n.,,- , - , _ , , n,-,,,,,_ -,.-_-,--,,,n._--
pue l. AC pump B was started first, without incident. The pressurizer level L. only decreased five inches, and little pressure change occurred. The core differential temperature decreased to about 2*F, as expected.
.p Cold leg temperatures, core outlet temperatures, system pressum, and
(, pressurizer level were all normal. The D pump was started without incident. Upper thrust bearing temperatures stayed below 130*F, and n oil level alarms cleared. The plant was in normal cooldown configuration. { High letdown was continued until the pressurizer level was about 210 inches. L -
- 15. Release of psdiation into the Reactor Building r.
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 !!.A.15-1. This coolant l... activity level would correspond to approximately 17,+ 3 leaking fuel rods, t* As can be seen from Table !!.A.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 factors (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 531 of the
.'. activity remains in the RC system) after appropriately correcting for i density differences and assuming a well-mixed reactor coolant loop (which seems to be a reasonable assumption during natural circulation with HP! 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 721 is discharged into the reactor building and 28% is 5
discharged into the bleed holdup tanks, the measured activity in a containment atmosphere agrees with the activity mlease associated l l with the reactor coolant discharge and with the pressurizer steam ! i, space discharge (see Table !!.A.15.3). Since there is no evidence to support 281 of the coolant being discharged into the bleed holdup j f tanks, the most reasonable explanation of the disparity is that the
- j* reactor building atmosphere is not completely well-mixed and that the
- . actual concentration might be 285 higher than the sample msults in j ,
Table !!. A.15-3 would indicate. l . L. i . M L
r-
!s.,-
I Despite the reasonably good agmement between the measured airborne U activity in the reactor. building and the activity release 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 !!.A.15-2 is due to the failure of additional fuel rods. An equivalent increase could have been produced by small leaks in n approximately 10 to 20 fuel rods which were previously intact. Since 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 !!!-42 shows the dose rates measured by the radiation monitors
,. in the reactor butiding. Calculations were made to see if the release of reactor coolant into the containment would give a similar msponse.
- c. 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 f
- i. 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 r* the minimum reading) can best be explained by assuming that the I 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 C 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 4 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 automatically 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 LP! r System. Tank readings indicated 1.8 feet or 615 gallons of lit i '. caustic left the tank before the tank outlet valve was closed. The l
'* reactor coolant did not become contaminated with the sodium hydroxide
) , because the ReactoriCoolant System pressure was higher than the dead
'- head pressure of the*LPI pumps as indicated by an analysis of an RCS
! t. sample taken at 05:15 on February 27, 1980. This sample indicated I that sodium level was only 1.8 ppm. l ! 1. l .. L
s . f' TABLE II.A.15-1
- i. .
ra COMPARISON OF REACTOR COOLANT ACTIVITIES PRIOR TO TRANSIENT l WITH DE5IGN BASIS ACTIVITY IN F5AR
,.. Activity Measured j FSAR Design in Reactor Coolant Fraction of
\- Basis Activity (1) Before Transtent(2) Design Basis Nuclide ( uct /al ) ( uC1/ ell Activity in Coolant o';
[j I-131 3.17 0.184 0.058
-. 1-133 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
,e 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.
I (3) Equivalent to 17 f,3 leaking fuel rods. I i e l . be e
%e 4.
I U I TABLE !!.A.15-2 (,, n RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT SAMLES
- t, I Radionuclide Concentration in uC1/m1*
9 Date Time XE-133 I-131 1-133 I-135 CS-134 CS-136 CS-137 L 0.018 02/26/80 00:10 -- 0.246 0.233 0.096 -- -- o
. 10:15 12.3 -- -- -- -- -- --
[ 16:25 -- 0.454 0.429 0.180 0.045 0.007 -- l: 21:35 -- 0.664 0.5438 0.158 0.058 0.013 -- 02/27/80 00:58 -- 1.190 0.884 0.215 0.089 -- -- 06:15 -- 0.924 0.574 0.100 0.085 0.039 --
. 08:55 -- 0.833 0.506 0.073 0.1 04 -- 0.101 12:55 -- 0.818 0.437 -- 0.077 0.036 0.083 17:55 -- 0.949 0.416 0.031 0.094 0.035 0.088
, 21:10 -- 0.736 0.288 0.016 0.071 0.024 0.066 02/28/80 08:15 -- 0.575 0.160 0.004 0.068 0.022 0.067
- Based on ' coolant density at room temperature.
4 r 4 . k
\.
i
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s,. )
F ( .'
.- TABLE II.A.15-3 GASE0US ACTIVITY IN THE REACTOR BUILDING
(* Measured Release Released Total
,. Measured (1) Conc. In with with Pzr. Release l Conc. in R8 Curies in Coolant (3) Coolant (4) 5 teas (5) (C1) ' Nuclide (WC1/ml) RB Atmos.(2) (uC1/ml) (uC1/ml) (01) from RC5 Xe-133 2.94x10-2 1582 12.3 1075 504 1578 Xe-135 7.01x10-3 377 2.86 250 117 367 Kr-88 1.80x10-3 97 0.648 57 27 84 1-131 1.17x10-6 0.063 0.55 60 --
(6) 68 k, . (1) Based on containment gas sample taken 2/26/80 at 21:55 (except 1-131 comes
- from RM-A6 charcoal 2/27/80 at 04:05) and decay corrected back to the time of trip.
(2) Assumes a reactor building free volume of 1.9 x 106 ft3. (3) Based on reactor coolant sample taken 2/26/80 at 10:15 and stripped of gas (except iodine activity which was estimated from Table !!.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 e (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. h i. L. f 29-
- . - . _ , , .---.-,,-,.-n - -- - - , , , , - - - - - -.. . - . - - - - -
d Before the RCS was placed in cold shutdown with the LPI system operating, the LPI System was recirculated to the borated water storage tank (WST). The sodium levels in the WST increased up {, to approximately 60 ppe. After the LPI System was connected to ( the RCS for cold shutdown cooling, sodium in a sampie of the LPI loop indicated that the sodium level was about 45 ppe. o. j' Sodium in the WST is being reduced by processing the water through a spent fuel domineralizer. The LPI and RCS water is also i ,.. being processed through a domineralizer system to reduce sodium. l As of March 1, the sodium was in the range of 10 to 20 parts per U million in the Reactor Coolant System. This sodium concentrated is not expected to harm any of the system components. r. I (, 8. 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 h- for this event: ) , a) Fuel damage will not occur from an excessive power-to-flow rate. i b) Reactor coolant system pressure wil not exceed code pressure limits.
, c) The resultant doses are within 10CFR100 limits.
I- The February 26, 1980, transient involved a partial loss-of-electrical power. The applicable protection criteria appear to have been met i 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 l 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 ;'f00 psig; i.e., 2750 psig. '
i ( The major difference between the February 26, 1980, event and the l ,. bounding Feehater Line Break event analyzed for the FSAR are itemized below.
}
- 1. The main feedwater valves were closed over a 25-30 second time j interval rather than the conservative FSAR assumption of an j t, instantaneous rupture of the main feedwater header at the steam i
generator inlet nozzles. - i 2. The combination of the opening of the PORV, the ICS demand for 1 L 1035 full power and feehater runbacks caused the reactor trip on
- high pressure to occur at 10 25 seconds into the event ratehr than i i at 11.8 seconds for the FSAR event.
j <. [ ,
- se
L^
- 3. The opened PORV allowed a rapid depressurization to 1500 psi such that the ESFAS was actuated and two HPI pumps were started. The PORV and spray block valves were closed about 5 minutes into the C'
- event and one pressurizer code safety valve actuated about 10
(; minutes into the event. Tnis was much slower than 22.3 seconds for the FSAR events. f 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 c- (approximately 1400 gpe) 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
- f. 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.
L I 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:
f (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.
t (c) The pressurizer PORV was opened and latched at a system pressure
,'- below its setpoint.
I
- t. (d) Automatic initiation of emergency feedwater on low OTSG level did not occur.
f The maintenance of plant conditions within the safety criteria ' required the action of automatic safeguards (reactor trip, high pressure injection, reactor building 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.
9 8 e n 9
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c.I i' III. EVENT DETAILS AND INPUT DATA C A. Initial Plant Conditions The following table of data summarizes the initial plant conditions. Time of Reactor Trip Approximately 14:23 7, g February 26, 1980 Reactor Power 98.6 Full Power r' (, 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 l
Number of Main Feedwater
. Pumps Operating 2 4 Tests in Progress None c ICS Mode Automatic-i B. Plots of Major Parameters I
Several of the recorded parameters were selected for plotting to show
- major trends. These plots are shown in the following pages and are i listed below:
i 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 III-3 Reactor Coolant System Pressure (-2 to 16 minutes)
Figure III-4 Reactor Coolant System Pressure (-4 to 72 i minutes) {. Figure III-5 Control Room Strip Chart of RC Pressure - Loop B (Narrow Range) 3 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 l : (-4 to 72 minutes) L. L.
; .- x k ,e 4
r.: i' i Figure III-9 Control Room Strip Chart of RC Unit Average Temperature
- f. Figure III-10 Control Room Strip Chart of RC Outlet Temperature (Narrow Range) l Figure III-11 Incore Thermocoule (Maximum) (-2 to 42 hours)
Figure 111-12 Reactor Coolant Cold Leg Temperatures (-2 to 42 i' hours)
- 1. 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) I' Figure III-17 Pressurizer Level Indication (-2 to 42 hours) Figure III-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 s Figure III-23 Main Steam Pressure (-360 to 840 seconds)
' Figure III-24 Control Room Strip Chart of Main Steam Temperature i 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 l Figure III-28 Control Room Strip Chart of Steem Generator B Outlet Pressure i 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 I Figure 111-32 Steam Generator Levels (-2 to 42 hours) Figure 111-33 Feedwater Flow (-360 to 840 seconds) i Figure III-34 Feedwater Flow (-4 to 72 minutes) ! 7. 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 j Steam Generator B (. Figure III-37 Feedwater Pump Speed (-360 to 840 seconds) i J ed O e w n - &
F ( _.
~ \' Feedwater Pump Discharge Pressure (-360 to 840
(,' Figure III-38 seconds) J' Figure III-39 Deaerator Feed Tank Level (-360 to 840 seconds) l Figure 111-40 Control Room Strip of
- Reactor Building Sump Level I - Borated Water Storage Tank Level I' - Reactor Building Pressuit Figure III-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 111-44 Control Room Strip Chart of Reactor Building Tenperatures Figure III-45 Control Room Strip Chart of Reactor Building h Purge Exhaust Flow f s t L. (' ( i k L. k i (~ L f L i L. ( , i .- i" .
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i s. IV. COMPONENT TRANSIENT ASSESSMENT, [~ L, A preliminary evaluation has been conducted for major RCS components which r- were subjected to this transient. These evaluations focused on the i' 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. l A. Reactor Vessel r-
- j. The characteristic of this transient most affecting reactor vessel structural integrity was the change in reactor coolant inlet
.. temperature which developed high therma' 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 t.peration of the plant. A fatigue analysis would be necessary in or ter to determine the impact [* cf this transient on the 40-year design life of the reactor vessel. i 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 curves 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 also 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 5
analyzed in the Crystal River-3 stress report.
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'- 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 c- leakage rate is detected upon the plant's return to operation, it is l more likely indicative of the general condition of the previously damaged tube-to-tubesheet welds rather than specifically related to the 7 recent incident. - The structural integrity of the steam generators has not been adversely affected, and the steam generators are acceptable for resuming r- operation. Since the lower portion of the steam generator support j' skirt was potentially innersed in a pool of reactor coolant, it is recommended 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.
l D. Pressurizer The pressure and temperature 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 i transients prescribed in the CR-3 functional specification. The effect on the fatigue life of the pressurizer must be quantitatively assessed later.
{I 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 j on the pressurizer relief nozzles during relief valve operation. The g, relief valve loadings which occured during this transient should be determined and the effect on the pressurizer relief valve nozzles
, assessed later. (See Section IV.H).
e 1
l E. Reactor Coolant Piping The design rapid depressurization transient in the Crystal River Unit 1 ['i 3 Functional Spec and for the rapid depressurization transient considered in the Rancho Seco Stress Report were either representative i or conservative for all areas of the RC piping except the A-loop cold p leg. Analyses of the A-loop cold leg pipe and HPI nozz!e were performed and resulted in acceptable stresses and usage factors. it { is concluded that this transient event had no significant effect on y, reactor coolant piping and the RC piping is suitable for continued operation. F. Reactor Coolait Pumps
! I j 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
- j conditions considered were system pressure, temperature, service water i 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 sustaind any damage during this transient which would affect their perforiarce.
! G. Control Rod Drive Mechanism (CRDM's)
This transient had no impact on the ability of the CROM'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 i . humidity in the Reactor Building may have, however, degraded stator l* insulation resistance. The majority of past CRDM failures have been attributed to moisture. All CROM's should be checked for proper
- insulation resistance prior to their return to service and, if necessary, dried out to achieve acceptable readings.
l H. Pressurizer Relief Valve and Relief System i . 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, i
L. L l 1 I i
i. l' [" 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 ,j' may have had some detrimental effect on the valve internals. .
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 I
, greater than the normal design loads. However, the several hour duration of flow through the discharge piping may have resulted in
- c. excessive expansion of the piping and imposed loadings on the valve and pressurizer nozzle in excess of that considered in the original design.
Although the conditions experienced by the PORY and code safety valves { should not have substantially affected their structural integrity, or their ability to perform their intended relief function the following r- actions are reconnended 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 s 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.
- 1. 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 reconnended conditions provided under " Operating Limits and Precautions," to ensure the fuel is maintained in compression. However, the reconnended envelope was based upon very conservative EOL fuel pin pressure calculations assuming burn-ups greater than 40,000 MWD /MTU. The Crystal River plant is currently only 160 EFP0 into Cycle { 2 and the estimated hot pin burn-up is less than 24,000 MWO/MTU. A
- t. preliminary assessment of the transient indicates that the fuel was maintained in compression at all times.
r 5+
-85
l (.. j' V. REFERENCES f, 1. Crys.t al River-3 Model 855 Computer Alarm Log - February 26, 1980 i 2.f Crystal River-3 Sequence of Events Monitor - February 26, 1980 l, . 3. . Control Room Charts - February 26, 1980
,- 4. Crystal River-3 Model 855 Computer Post Trip Review
- February 26, 1980
- 5. ASME Steam Tables - 1967 f ,
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e L* APPENDIX A f: INSTRUMENT STATUS r' 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 e24 VDC in the NNI "X" cabinets due to tripping of breakers Sj and S2 as shown on Bailey drawing 080340420.
- c. No loss of 120 VAC, +24YDC, or -24VDC to or in the NNI "Y" cabinets.
f
- 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 i 080340420 NNI-X Power Dist. Sheet 1 of 4 D8034042D NNI-X Power Dist. Sheet 2 of 4 D8034042C NNI-Y Power Dist. Sheet 3 of 4 D8034042D NNI-Y Power Dist. Sheet 4 of 4 D8034033C 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 Sneet 4 of 7 D8034033B R.C. Sys . Schematic Sheet 5 of 7 D8034033C R.C. Sys. Schematic Sheet 6 of 7 D8034033D R.C. Sys. Schematic Sheet 7 of 7 j D8034034C SP Sy:. Schematic Sheet 1 of 4 08034034C SP Sys. Schematic Sheet 2 of 4 D8034034D SP Sys. Schematic Sheet 3 of 4 08034034C SP Sys. Schematic Sheet 4 of 4 D8034035C CA Sys. Schematic i i i [. D8034036C RB Spray Sys. Schematic D8034038C C.F. Sys. Schematic ' 08034037E DH Sys. Schematic l
l
/
A-1 l
o I l i l APPENDIX A (Cont'd) i p . L D80340390 Makeup Sys. Schematic Sheet 1 of 5 D80340390 Makeup sys. Schematic Sheet 2 of 5 [' 080340390 Makeup sys. Schematic Sheet 3 of 5 ( 080340390 Makeup Sys. Schematic Sheet 4 of 5 D80340390 Makeup Sys. Schematic Sheet 5 of 5 D8034040E Spent fuel Sys. Schematic f 08034041C Condensate Flow Sys. Sheet 1 of 2 D8034041B Condensate Flow Sys. Sheet 2 of 2 Cabinet Module arrangement Sheets Rev B dated 12/10/71. I Table II-1 contains a tabulation of non-nuclear instrument parameters and provides the following information, assuming the failure of the NNI "X" +24VDC power: f a. Source of Power
- b. Validity of Signal to Computer
(
- c. Validity of Signal to Integrated Control System (ICS).
i d. Validity of Indication to the Control Room Operator.
- e. Validity of Indication to the Hot Shutdown Panel Operator.
Table 11-2 contains a tabulation of non-nuclear instrumentation alarms for the reactor coolant, secondary plant, makeup & purification systems and prcvides 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.
l o i L. t A-2
' APPENDIX A (Cont'd)
Table 11-3 contains a tabulation of non-nuclear instrumentation system control
' or interFock functions for the reactor coolant, secondary plant makeup and p purification systems and provides the following information:
- a. Parameter i' b. Interlock or Control Function
(
- c. Source of Power f'
l d. Validity fo Signal for Actuation
- c. 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 11-3. Definitions
, o " Parameter": The process variable being monitored.
o " Interlock / control function": The function to be perfomed by ccr. tact 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. p "Was interlock / control action implemented without a valid signal": The + 24VDC power was not available to energize signal monitor relays to change contact state to implement control action or interlock. The following is the status of the non-nuclear instrument system auto / manual j controllers for the reactor coolant, makeup, and letdown systems in the event of NNI "X" + 24 VDC power failure. l {. _
- a. Reactor coolant pump seal injection flow control - No valid signal.
l [. b. Reactor coolant makeup flow control - No valid signal.
- c. Pressurizer heater SCR control - No valid signal.
- d. Letdown flow control - Valid Signal.
l ( A-3
i L. APPENDIX A (Cont'd)
'P The following is a listing of NNI control room indications not addressed in l
(, Table II-1. The signals to these indications are derived from various signal processing modules in the NNI. r-p a. Total RC Flow Recorder - No valid signal
. b. Reactor Coolant Loop A&B ATc Indicator - No valid signal t 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 Tave Indicator - No valid signal
- g. Loop A&B AT Indicater - No valid signal
- h. Loop A&B 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 Flo,
- 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 I The following signals to the ICS would have been valid if the sensor signal
[ select switch had been in the position identified in Table 11-1. The presumed status of key instruments listed below is discussed later in Section II. A.12. L.
- A-4
c' APPENDIX A (Cont'd) ( r, a. RC Loop B Flow
- b. Reactor Th
(' c. Steam Generator A Pressure k ..
- d. Steam Generator B Pressure g
- e. Steam Generator B Feedwater Valve AP
- f. Steam Generator A Operate Level 9 Steam Generator B Operate Level
- h. Turbine Throttle Steam Pressure
- 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 range) indication on loss of +24VDC power.
1 s 5 8 5 i . A-5 (s l l
F L - r- Appendix A Status or NNI Signals
.(- Table II-1 Signal Validity to Computer, Control Room Indicator. Hot Shutdown Panel, and ICS
**** Valid Signal.Available To Cabinet r P wer Control Room Hot Shutdown Parameter Source Computer Operator Panci Oper. ICS Pressurizer Level t RC-LT1 X No No Yes-Indicator N.A.
RC-LT2 X No No No N.A. p- RC-LT3 Y Yes No Yes-Indicator N.A. Pressurizer Temp p RC2-TE1 X No No N.A. N.A. j 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. No RC14A-dPT 1 or 2
- Receive Flow Signal Fron NI/RPS NNI fiodules Powered from NNI X Trans mitter NI/RPS Fower RC Flow Loop B Y N.A. flo N.A. **Yes From NI/RPS **0nly if RC4B-TE4 Selected RCl48-dPT 1 or 2
, RC Loop A Th RC4A-TE1 X No No fl . A . t N.A.
RC4A-TE4 Y No *Yes-Indicator *Yes-Indicator' **Yes
*0nly if RC4A-TE4 Selecte d
**0nly if RC4A-TE4 Selecte d & RC15-MS Sel ected to Loop A
. RC Loop B Th RC4B-TEl X No No N.A. ' No RC48-TE4 Y *Yes **Yes-Indicator **Yes-Indicator ***Yes
*If RC4B-TE4 Not Selectec
**0nly if RC4B-TE4 Selected i
***0nly if RC4B-TE4 Selected and RC15-MS Selected to Loop B I
RC Loop A Tc Narrow RCSA-TE1 X No No N.A. No i RC5A-TE3 Y Yes *Yes-Indicator N.A. No
*0nly if RC5A-TE3 Selected L
i l
, : 8 I i
(**** Assuming failure of the + 24 VDC NNI-X Power Supply)
i' i {' Table 11-1 L,; - I' Vclid Signal Available To Cabinet Power Control Room Hot Shutdown f' Parameter Source Computer Operator Panel Oper. ICS (. RC Loop B Tc Narrov [" RCSB-TE1 X No No N.A. No (. RCSB-TE3 Y Yes *Yes-Indicator N.A. No
*0nly if RC58-TE3 Selectied 7
j RC Loop A Tc Wide RCSA-TE2 X No No N.A. N.A. RC5A-TE4 Y No *Yes-Indicator N.A. N.A.
*0nly if RCSA-TE4 Selectied
- f. ,
RC Loop B Tc Wide RC58-TE2 X No No N.A. N.A. { RCSB-TE4 Y *Yes **Yes-Indicator N.A. N.A.
*0nly if RCSB-TE4 Not Se~ected , **0nly if RCSB-TE4 Selected RC Press Loop A Yes-Indicator Wide Range X NNI N.A. Yes-Recorder 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. k e L (. i
- k. .
Page 2 l l I
L: Table 11-1 t,., e, Valid Signal Available To l '~ Cabinet 7, Power Control Room Hot Shutdown Parameter Source Computer Operator Panel Oper. ICS RC Pump #2 Seal f Cavity Pressure L 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.
- t. RC108-PT2 X N.A. No N.A. N.A.
- RC Pump #3 Seal I Cavity Pressure RC19A-PT1 X N.A. No N.A. N.A.
f RC19A-PT2 X N.A. No N.A. N.A. ( RC198-Pil X N.A. No N.A. N.A. RC198-PT2 X N.A. No N.A. N.A. l l 9
\
( ' i. f r ( iu. 4 Page 3
I l. f Table 11-1 t ., i' g Valid Signal Available To Cabinet e- Power Control Room F.ot Shutdo.m Parameter Source Computer Operator Panel Oper. ICS { r Start-up Stm. Gen.
; Loop A Level SPIA-LT4 X No No Yes No e, SP1A-LTS Y No No N.A. No i
l Start-up Stm. Gen.
- Loop B Level f' SPIB-LT4 Y *Yes No Yes No
( SPIB-LTS X **Yes No N.A. No
*0nly if SPIB-LT4 Not Se' ected r **0nly if SP18-LT4 Select <td Stm. Gen. Outlet Press. Loop A
{ SP6A-PT1 X No No No f;o
- i. SP6A-PT2 Y Yes Yes-Recorder N.A. *Yes
*0nly if SP6A-PT2 Selectitd i
Stm. Gen. Outlet Press. Loop B SP6B-PTl Y Yes Yes-Recorder Yes *Yes
-SP6B-PT2 X No No. No No . *0nly if SP68-PT1 Selected Stm. Gen. Level Full Range Loop A SP1A-LT1 X No Yes N.A. N.A.
I Stm. Gen. Level - s Full Range Loop B SPIB-LT1 X No Yes N.A. N.A. i t" Main Steam Temp. Loop A Sf4A-TE X No No N.A. N.A. i.. i Page 4
i' i Table 11-1 (. [' Valid Signal Available To Cabinet Power ':entrol P.oom Eot Shutdown [ Parameter Source Computer Operato; Panel Oper. ICS ( .. Main Steam Temp [' i. Loop B SP48-TE Y Yes Yes-Indicator N.A. N.A. j- SGA FW VLv aP SP12A-dPT X N.A. N.A. N.A. No SGB FW VLv aP { SP12B-dPT Y N.A. N.A. N.A. Yes i 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-TE2 Selected
- SGB Lower Down Comer Temp SP3B-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 SP3B-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 SP3A-TE2 Selected f .
s t 4. l ik* 9 L Page 5 5 i l l l
f' i r Table 11-1 L Valid Signal Available To L-Cabinet Power Control Rocn llot Shutdotin
' Computer Panel Oper. ICS Parameter . Source Operator
( SGB Operate Level {. SP1B-LT2 X No No N.A. No
- t. SP18-LT3 Y *Yes **Yes-Recorder N.A. **Yes
*0nly if SPIB-LT3 Not Sel ected r **0nly if SP18-LT3 and SP: B-TE2 Selected Turbine 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) SP10B-PT2 X No No N.A. No
; *0nly if this Transmitter Selected
, SGA FW Temp SP5A-TE1 X No No N.A. flo
' *Yes-Indicator N.A. *Yes SP5A-TE2 Y No
*0nly if SPSA-TE2 Selecte d i
, SGB FW Temp SPSB-TE1 X No No N.A. No
, SP5B-TE2 Y *Yes **Yes-Indicator N.A. **Yes
*0nly if SP5B-TE2 flot Sel ected
**0nly if SPSB-TE2 Selected e .
s f . L f f i
- k. .
I t. u Page 6
f' i Table 11-1 i. P Valid Signdl Available To 1 Cabinet Power Control Room Ilot Shutacwn f" Parameter Source Computer Operator Fanel Oper. ICS ( SGA Startup I. . FW Flow . 1 SP7A-dPT X No No N.A. No r ' SGB Startup {' FW Flow SP78-dPT Y Yes *Yes-Indicator N.A. *Yes
-, *0nly if iPSB-TE2 Selecte<1 -
SGA Main FW Flow
' SP8A-dPT1 N.A.
X No N.A. .No i SP8A-dPT2 Y No . N.A. N.A. *Yes
*0nly if .iP3A-dPT2 and SP!iA-TE2 Selected SGB Main FW Flow SP88-dPT1 X No N.A. N.A. No SP88-dPT2 Y *Yes N.A. N.A. **Yes
, *0nly if :iP88-dPT2 Hot Sei ected
**0nly if liP88-dPT2 and SP!;B-TE2 Selected t
t o. L w. t. L Page 7 , 6
Table 11-1 r . (. . f Valid Signal.Availt.ble To i Cabinet Potter Control Room I!ot Shutdom {' (. Para:neter Source Computer Operator Panel Oper. ICS RCP Total Seal Flor F MU27-dPT X No No N.A. N.A. ( HPI Flow I 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. r I
- 1. Letdown Flow MU4-dPT Y N.A. Yes N.A. N.A.
r MUS-TE X N.A. Yes N.A. N.A. Makeup Tank Level MUl4-LT1 X No- No ~ N.A. N.A. t MUl4-LT2 Y No No N.A. N.A. t 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 t.P MU18-dPT Y N.A. Yes N.A. N.A. i Makeuo Flow MU24-dPT X N.A. No N.A. N.A. ( . 1 k I ( e k i. s b
- i. : Page 8 ,
]
'J
- Table 11-1
.).
f Valid Si;;nal Available To Cabinet Power Control Room Hot Shutdotm [ Parameter Source Cosputer Operator Panel Oper. 'ICs
!)
P RC Pump Seal Flow
- l. MU7-dPT1 X N.A. No N.A. N.A.
MU7-dPT3 ~X N.A. No N.A. N.A. MU7-dPT2 X N.A. No N.A. N.A.
,. MU7-dPT4 X N.A. No N.A. N.A.
I~ RC Pump Seal L Return Flow MU31-FT1 X No No N.A. N.A. r 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 t 9 e t L. r 1 Page 9
F l
- Table 11-1
- i. ,
r Valid Signal Available To
'- Cabinet Power Control Roo::: Hot Sbutdovm '~
Para:neter Source- Co:::puter Operator Penel Oper. ICS l DH Removal Flow I' DH1-dPT1 X N.A. No N.A. N.A.
- l. DH1-dPT2 Y N.A. Yes-Indicator N.A. N.A.
p- DH Injection Temp
. DH2-TE1 X No Yes-Indicator N.A. N.A. ' Yes-Indicator N.A. N.A.
DH2-TE2 Y Yes f" DH Cooler Inlet l Temp DH6-TE1 X N.A. Yes-Indica tor N.A. N.A.
, DH6-TE2 Y N.A. Yes-Indicator N.A. N.A.
i i i t i t i k. E Page 10 l
- ~-
F l. Table 11-1 a-Valid Signal Available To Cabinet Four Control Poon Hot Shutdotm Parameter Source Computer Operator Panel Oper. ICS j r Core Flood Tank A j' Pressure CF1-PT1 X N.A. No N.A. N.A. CF1-PT2 Y N.A. Yes-Indicator N.A. N.A.
- i. Core Flood Tank B Pressure CF1-PT3 X N.A. No N.A. N.A.
CF1-PT4 Y N.A. Yes-Indicator N.A. N.A. Core Flood Tank A Level i CF2-LT1 X N.A. No N.A. N.A. CF2-LT2 Y Yes Yes-Indicator N.A. N.A.
- 3. 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. f L. t P l f l , l l ' Page 11
a
~
l i.
,- Table 11-1 Valid Signal Avcilable To Cabinet Power Control Roo: Eot Shutdown Parameter Source Computer Operator Panel Oper. ICS
- 1. .
, - . RB Spray Flow BSI-dPT1 X N.A. No N.A. N.A.
1 I BSI-dPT2 Y N.A. Yes-Indicator N.A. N.A. r Sodium Thio. Tank
, Level BS3-LT -X N.A. No N.A. N.A.
r-Sodium Hydroxide 8 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 BSS-TE j. X N.A. Yes-Indicator N.A. f!. A . Sodium Hydroxide Tank Pressure
. BS14-PT X N.A. No N.A. N.A.
1 Page 12
._ l l
Table II-I Valid Signal Available To Cabinet Pouer Control Roou Hot Shistd wn I' Parameter Source Computer Operator Panel Oper. ICS l i-Boric Acid Tank #1
- Ternperature
!- CA10-TE X No N.A. Yes N.A.
r-Boric Acid Tank #1 [. Level - Call-LT X No N.A. No N.A.
- g. .
Boric Acid Pump ( PIA & PIB Discharge Pressure CA14-PT X N.A. No No N.A. Boric Acid Tank #2
, Tempera ture ! CA12-TE X No N.A. Yes-Indicator N.A.
Boric Acid Tank *2 Level i CA13-LT X No N.A. No N.A. f e l t l I - t-i s.. f i l Page 13 l
F f- Table II-I
.. .e r Valid Signal Available To f
Power Control Room Pct Shutdown r' Parameter Source Coeputer Operstor Panel Oper. ICS t. S. F. Storage Pool r' "A" Level l SF1-LTI Y N.A. Yes-Indicator N.A. N.A. S. F. Storage Pool ,
"B" Level SF1-LT2 Y N.A. Yes. Indicator N.A. ft . A.
. I' , i l t
?
I i s e 0 e L 8 a f 4 .e Page 14 ,
+ - - ,
c ,, ,,,.m,.rw-- - - - - - - - - - - - - - > - - - - - - --
. l l l I L <
,, Table 11-1 li m.
I~ Valid Signcl Available To L Cabinet ,
~~
Power Control Roop Hot Shutdown Parameter Source Cemputer Operater Panel Oper. ICS i
~~
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.
- k. .
f' I h. L.
,. *h L.
- Page 15 i
I .
r "able 11-2 Valid Operable Alarms Cabinet Power Valid Signal Available For Actuation of "arameter Source Hi Alarm to Alarm Pressurizer x Level (RC1-LR/L.52) No No
Pressurizer x
- Level (RCl-LS1) No No
,., Loop A RC Flow x (RC14A-FS) N.A. No Floop 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 cr B TH Selected
/' by RC 15-MS Loop A RC Pressure x wide range (RC3A-PR2/ PS ) N.A. Yes il RC Pressure Vs. x Core Flood Valve (RC3A-PS3 ) No No
,.P ' tion Alarm I
Rt. eressure x High Press Inj. (RC3A-PS6) No N.A.
,..Not by Passed RC Pressure x Low Press Inj. (RC3A-PS S) No h.A.
, . Not by Passed RC Pump #2 Seal x Cavity Pressure (RC10A-PSI) No N.A.
, (All 4 Pumps) RC10A-PS 2
! RC10B-PS I i RC108-PS 2
. RC Pump #3 Seal x
, Cavity Pressure (RC19A-PS 1 (All 4 Pumps) No N.A*
(RC19A-PS2 (RC19B-PS1
; (RC19B-PS2
- SGA Startup level x (SP1A-LS2) N.A. No
, SGA Startup level x (SPIB-LS 2) N.A. No SGA Operate Level x (SPIA LR/LSI) *Yes N.A.
- Only if SP1A- LT3 and SP3A-TE2 is selected.
SGB Operate Level x
*Yes N.A.
(SP18-LR/LSI )
*only if SPIB- LT3 and SP3B-TE2 is selected.
L
E Table 11-2
'- Valid Signal Available For Actuation of F.
Cabinet Power
,_ Parameter Source Hi Alarm Lo Alarm 1
\
Total RC Pump Seal Inject. x c Flow (MU27-FS) No No L.High Pressure Injection Flow x RC Loop A (MU23-FS2) No No I' y (PtJ23-FS4) Yes Yes High Pressure Injection Flow x RC Loop B (MU23-FSI) No No Y (MU23-FS3) Yes Yes f.. g Let down Temp x (MUS-TS) No N.A.
Makeup Tank Level x
{ (MU14 LR/LSI) No No Makeup Tank Press x j (MU17-PS) No No I-RC Makeup Flow x (MU24-FS) No No Rt Pump Seal Injection Flow x
' (all 4 pumps) .(MU7-FS1) No No x
$ (MU7-FS2) No No x
(MU7-FS3) No No x l (MU7-FS4) No No
, RC Pump Seal Return Flow x i
l (all 4 pumps) (MU31-FSI) No N.A. x (MU31-FS2) No N.A.
- b. x (MU31-FS3) No N.A.
{ x (_ (MU31-FS4) No N.A. Turbine Thrott14 x Pressure (SP10-PR/ PSI) *Yes *Yes
*0nly if SP10A-P T1 or SP108-PT1 is selected t.
Page 2
7.-. ( -- ~ . . - . . -. . . . . . _, , .. ,_ t Table II-3 Validity of NNI, Digital Jrlock and Control Functions For Reactor Coolant Makeup and Purification Systems . Cabinet Valid Signal Interlock / Control Power Available For '$ " " Parameter Function Source Implementa tion cti n e Cjrol nt Without Valid Signal Pressurizer De Energize Pressurizer x Level Heaters 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 To Loop A on Low Loop B (RCl40-FS) No No 1 Flow RC Loop A Tc Prevent Start of 4th RC x (wide range) Pump when Tc less than (RC5A-TS) No No i
Temp. Set Point RC Loop B Tc Prevent Start of 4th RC x Pump when Tc less than (RC5B-TS) No No Low Temp. Setpoint RC Pressure Open & Close Pressurizer x Electro Matic Relief (RC3-PS8) No Nn RC Pressure Open/Close Pressurizer x l Spray Valve (RC3-PS3) No No RC Pressure On/Off Pressurizer x Hester Bank #3 (RC3-PS6) No No RC Pressure On/Off Pressurizer , x Heater Bank f4 (RC3-PS7) No No RC Pressure Decay Heat Valve (DH-V1) x Prevent opening when (RC3A-PS3) No No RC Pressure High RC Pressure High RC Pressure Interloc k x - To Customer Spray Valve (RC3A-P57) No No (PZR) RCV-53F Letdown Temp Close MU-V3 on High x (MUS-TS) No - No Makeup Tank Terminate Feed Bleed x On Low MU Tank Level. (Mul4-LS2) No , No
. p . ,. . ,_ - - _. _. ._. _ _ . - . ._. --.
.. _ _ ) .
> inuie In-a .,
g ' Cabinet lid Signal Interlock / Control Power Available For Was Interlock / Control i Parameter Function Source Implementation Action Implemented Without Valid Signal
! RC Pump Seal Close RC Pump Seal x .
Inj:ction Flow Return Valve on Low (MU7-FS1) No No' * (all 4 pumps) Seal Injection flow and x Prevent Start of (MU7-FS2) No No , RC Pump with Low l Seal Inject. F1 m (MU7 FS3) No No X (MU7-FS4) No No ) ,
. i 4
e 9 I 1
- 4 I
i
- Page 2
_ _. . - . . _ _ . - ~ _ - - APPENDIX B ; ANALYSIS OF CORE SIGNALS I
- 1. Alam Printout Data g,
The alarm printout from the February 26, 1980 Trip at Crystal River-3 has b been examined to determine if SPND alams were indicative of any significant - 4 occurrences in the core. Based on this analysis and the lack of any I evidence to the contrary, it is believed that the alarms were caused by i c normal gamma-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
{' 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 nanoamps. The SPND's are scanned once every 60 seconds; their status is printed only if it has changed since the last scan. 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), i several SPND's from levels 6 and 5 also alarmed. In addition, one level 2 l SPND (string 6 in location 7-F) alarmed although it returned to normal within one minute (at 1515). Frequently, SPND's would 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 alarms is background current attributable to
; gama-induced electrons. The positive component of the signal arising from neutron reactions decreases af ter the trip with a 60-second half-life
. characteristic of Rh-104 decay. Thus, after a period of five to eight l minutes following the trip, the neutron portion approaches zero. The gama i
portion remains relatively high due to fission product decay and causes a negative current. The gama currents are -5 to -15 nanoamps in magnitude
- i for level 7 SPND's and smaller currents (in' absolute magnitude) are
- generated for detectors lower in the core since their leadwires are shorter.
Therefort, 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 I-explanation is verified by the alarm printout for a "ncrmal" CR-3 trip (one without loss of coolant) (Ref.1). The trip occurred when the reactor was j at 75%FP at 80C-1, and several level 7 SPND's began alarming off-scale, L. low, approximately five minutes following the trip.
B-1 __,r-__., -_..,,y n -,.--_w-,, - -- _.._,,,._._y.---_-.-,_,.w . . . - _ , ,,. __,__ . . - - , . . - , ,_.y. , . .
APPENDIX B (Cont'd) l' 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 th&t of inconel in that the j' gama-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. f 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 af ter the trip. One channel indicated a signal corresponding to 100*.FP while all other channels were zero. It was indicated by M. Collins (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 ; i further verified by SP-433, In-Core Neutron Detectors Channel Check, which 8
listed only the first 18 channels for each recorder. 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, i 3. Analysis of Source Range Detector Response
- Source range signals are provided by two BF3 proportional counters located on opposite sides of the core.
i Figures 4 and 5 illustrates the response of each source range detector subsequent 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. t [ 4 L B-2 l '. f
I * (, FIGURE 1 THE DADCOCK C WILCOX CO. . CR-3 2/26/80
'"t Time of SPfiD Level 6 Alarms i'
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- APPENDIX C - SEQUENCE OF EVENTS REPORTED BY FLORIDA POWER CORPORATION Rev. !
REVISION 5 Page i
,] SEQUENCE (,AS OF 2300 3/1/80) 26 February Transient CR-3 l.
I- EVENT STNOPSIS { At 14:23 on February 26, 1980 Crystal River -3 Nuclear Station experienced a [ reactor trip from approximately 10C: fun power. A synopsis of key events and parameters was obtained from the plant computer's post-trip review and plant i-alara summary, the sequence of events monitor, control room strip charts, and the Shift Supervisor's log. The reactom was operating at approximataly 100 : full power with "Integrateci { Control System (ICS) in automatic. No tests were in progress and minor main-tenance was being performed in the Non-Nuclear Instrumentation (NNI) cabinet "Y". I
! Time Event Cause/Com=ents 14:23:00 The following is a s a ry of
{ plant conditions prior to the trip nux 98.6% RC ?ressure 2157 psig P:R level 202 inches
}C tank level 71 i=ches T
H "A" 599'F. t "B" 600*F.
' TH "A" $57*F.
TC "B" 556*F.
- Rb now "A" 73 I 106 lbs/hr RC now "B" 73 I 106 lbs/hr Letdown now 48 gym OTSG "A" lvi (CP) 67:
OTSG "B" Iv1 (CP) 65:
, OTSG "A" FRLY 242 inches OTSG "B" FRLV 254 inches
,. OTSG "A" Pressure 911 psig OTSG "B" pressure 909 psig Main Steam Pressure 894 psig Main Steam Temp. 589*F.
l Condenser vacuum 1.76
!. Generated }$1 834 D7T level 12.7 ft. # Feed .now "A" 5 I 10 lbs/hr i Feed now "B" 5 I 10 lbs/hr Feed Pressure "A" 970 psig Feed Pressure "B" 968 peig I
14:23:21 +24 Volt Bus Failure (NNI
' Cause still unknovu. Apparently, power loss."I" supply) the positive 24 VDC bus shorted dragging the bus voltage down to a C-1
. . . . - . . . = . - - - - - -- - - -
O
a o Rev. 5
, Page 2 i l Time Event Cause/ Comments-
.j low voltage trip condition. There is a built-in k to h second delay
. at which time all power supplies will trip. There was no trip indication on negative (-) voltage.
This event was missed by the r-annunciator. Following the NNI power f failure, such of the control room indication was lost. Of the instrum-r 'entation that remained operable transient conditions made their indic-cation questionable to the operators. c 14:23:21 PORV and Spray Open When the positive 24 VDC supply was lost due to the sequence discussed abeve f the signal monitors in NNI changed state causing PORV/ Spray valves to open. The PORV circuitry is designed to seal in upon actuation and did so. The resultant
', loss of the negative 24 VDC halted
; spray valve motor operator and prevented PORV seal in from clearing on low pressure. It is postulated that the PCRV opened fully and the spray valve stroked for approximately h second. The 40% open indication on spray valve did not I
actuate, therefore, the spray valve did
! not exceed 40% open.
< 14:23:21 Reduction in Feedwater As a result of the "X" power supply
; failure many primary plant control signals responded erroneously. Teold
, failed to 570*F (normal indication was l
557'F) producing several spurious alar =s L Tave failed to 570*F (decreased) . The resultant Tave error modified the reacto
-I demand such that control rods were t w'ichdrawn' f to incrase Tave and reactor power. The power increase was terminated
, at 103% by the ICS and a " Reactor Demand
! High Limit" alarm was received. That failed to g70* F (low) and RC flow failed to 40 I 10 lbs/hr in each loop (low).
f . Both these failures created a BTU alarm
' and limit on feedwater which reduced feedwater flow to both OTSG's to
, essentially zero. Turbine Header Pressur failed to 900 psig (high) which caused
' the turbine valves to open slightly to s
C-2
( Rev. 5 L.t Page 3 t-0 { Time Event Cause/ Comments regulate header pressure thus increasing generated megawatts. These combined failures resulted in a loss of heat sink to the reac:or initiating an excessively high RC pressure-condition. 14:23:35 Rasctor Trip / Turbine ' Rx trip caused by high RCS pressure at 2300 pst r Trip Turbine was tripped by the reactor. 14:24:02 El Pressure Inj. This was a computer printout and indicates Raq. (Flag) <50* subcooling.* See attached graph of RC C Pressure / Temp. vs. Time. This graph is based on Post Trip data and actual incere thermo-couple data. From the reactor trip peint (14 :2* to 14:33, core exit temperature data was f i obtained by txtrapolation and calculated data, i This is supported by .:vo alar = data points plotted at 18' and 21* of subcooling during this period frem the ce=puter. It is i=portant i to note that lowest level of subcooling was 8*F for a very short period of time. e
- NOTE: This computer program was initiated as
{ a result of the TMI incident. 14:24:02 Loss of Both Suspect condensato pump tripped due to high
', Condensate Pumps DFT level. This is verified by ???? printed
'by computer, indicatirig the level instru=en:
was over ranged as well as a low flow indication in the gland steam condenser as alsc indicated by computer.
' 14:25:50 PORY Isolated At this time a high,RC Drain Tank level alarm vas received. This was resultant from the
{ PORY r-==4a4ag open and was positive indicatiot that the PORY was open. At this time, the operator closed the PORY block valve due to
- RCS pressure decreasing and high RCDT level.
l l 14:26:41 HPI Auto Initiation HPI initiated automatically due to low RCS l { pressure of 1500 psig. The low pressure condition was. resultant from the PORY remaining c
. full open while the plant was tripped. Full l . HPI was initiated with 3 pumps resulting in l ' approximately 1100 gym flow to the RCS. At this time, all remaining non- essential R.B.
- isolation valves l r l
C-3
. _ . _. .. - _ . = ...
I
c> u Rev. 5 , c. Page 4 Time Event Cause/ Comments
- were closed per TMI Lessons' Laarned Guideline g .
e . I L g ., 14:26:54 RC Pumps Shutdown Operator turned RC pumps off as required by the applicable emergency procedure and B & W L small break guidelines. p- 14:27:20 RB Pressure Increasing This is first indication that RCDT rupture disc had Yuptured. RB pressure increase data was obtained from Post Trip Review and Strip Char
- indication.
14:31:32 RB Pressure High This alarm was initiated by 2 psig in RB. Th: is attributed to steam release from RCDT. Co< ( safeties had not opened at this time based up< ( tail pipe temperatures recorded at 14:32:03 (Computer). f 14:31:49 OSTG "A" Rupture Matrix This occurred due to <600 psig in CTSG "A. 5 Actuation The low pressure was caused by OTSG "A" boili: dry which was resultant from the BTU limit anc failed OTSG level transmitter. This resulted in the closure of all feedwater and steam blot i valves which service OTSG "A". 14:31:59 Main Feedwater Pump 1A Caused by suction valve shutting due to
' Tripped matrix actuation in previous step.
I 14:32+14:41 ES A/B Bypass Manually bypassed and HPI balanced between al;
; 4 nozzles (Total flow approxinately 1100 gpm
-small break operating guidelines).
t
- 14:32:35 Started Steam Driven . Started by operator to ensure feedwater was i
Emergency Feedvater Pump available to feed OTSG's. I 14:33 Core Exit Temp. Verified The core exit incere thermocouples indicated L the highest core outlet cesperature value was. 560*F. RCS pressure was 2353 psig methis time { therefore, the subcooling margin at this time I was 100*F. Minimum subcooling margin for the entire transient was 8*F. It is
, , postulated that some localized boiling l , occurred in the core at this point as indicated by the self powered neutron detectors.
{ 14:33+14:44 Started Motor Driven Emer- Same discussion as " Started Steam Driven Eme gency Feedwater Pump gency Feedwater Pump." 14:33:30 RC Pressure High (2395 psis) At this point, pressuri:er is solid and code l 1 safety lifts (RCV-8). This is the highest RCS pressure as recorded on Post Trip Review. i i Apparently, RCV-8 lifted early due to seat I C-4 k
i
,e
- j l
l
. l i
Rev. 5
, Page 5 c-l Time Event 'Cause/ Comments t
leakage prior to the transient and RCV-9 did
- I not lift.
i 14:34:23 RB Dome Hi Rad Level RHG-19 alarmed at this point. Highest level
<-- indicated during course of incident was 50 R/hr. High radiation levels in RB caused by
' release of non-condensable gases in the press urizar and coolant.
14:35:33 Attempted NNI Repower With- This resulted in spikes observed en de-ener-out Success gized strip charts. 14:36:50 Computer overload Caused by overload of buffer. Resulting in ( no further computer data until buffer catches i up with printout.
, 14:38:15 FW-34 Closed This valve was closed to prevent overfeeding
.0TSG "B" beyond 100:l indicated Operating Rang
{ 14:44:12 NNI Power Restored Success- NNI was restored by removing the X-NNI Power I, fully Supply Monitor Module. This allowed the f breakers to be reclosed. At this time, it va observed that the "A" OTSG was dry, the press t urizer was solid (Indicated off scale high), RC outlet temperature indicated 556*F (Loop A
& B average), and RC average temperature indi 7 cated 532*F (Loop A & B). The highest core a:
thermocouple temperature at this ti=e was 531 l , RSC pressure was 2400 psig (saturation temp. . this pressure is 662*F.). This data verified natural circulation was in progress and the ( [ plant subcooling margin was 131*F. (based on core exit thermocouples). f 14:44:31 RB Isolation and Cooling L Actuation At this time, RB pressure increased to 4 psig and initiated RB Isolation. The operator verified all inunadiate actions occurred prope: { for HPI, LPI, and RB Isolation and Cooling.
- increasing RB pressure was resultant from RCV.
passing HPI at this time. k 14:46:10 Bypassed HPI, LPI and RB These "ES" systems were bypassed at this ti=e j f Isolation and Cooling to again balance HPI flow and restore cooling [' veter to essential auxiliary equipment (i.e., RCP's, letdown coolers, CRDM's etc.). L C-5 l t
o o i 4
~
Rev. 5 Page 6 -
, ,I .
b- Time Event Cause/Conunents ( 14:51:57 Rupture Matrix Actuation on The actuation was resultant from a deg-OTSG-B radatica of OTSG-B pressure. Cold emer-gency feed was being injected into the OTSG
, at this time. This matrix actuation isolated l all feedwater and steam block valves ao the L' B-0TSG and tripped the "B" sain W pump. Both Emergency W pumps were already in operation
[ at this time. B-0TSG level at this time was 70% (Operation Range). 14:52 HPI Throttled and RCS At this time, the maximum core exit ther=o-Ie , Pressure Reduced to 2300 cogle temperature was 515'F, RCS pressure I psig was 2390 psig. Therefore, the subcooling margin was 147'F. Natural circulation was i ( in effect as verified previously. All con-
; dicions had been satisfied to throttle EPI.
Therefore, flow was throttled down to approx-
, imately 250 gpm to reduce RCS pressure to 2300 psig in order to attempt to reduce the i flow rate through RCV-8 and into the RS.
( 14:53 Reestablished Letdown At this time, the operator was attempting ( to establish RCS pressure control via normal RC makeup and letdown.
! 14:56 Opened MU Pu=p Recire. This was done to assure the MU pumps would I
Valves have minimum flow at all times to prevent possible pump da= age.
, 14:56:43 Bypassed the A-0TSG Rupture
- Feedwater was slowly admitted Matrix and Reestablished to the A-0TSG which was_ dry up to this point.
g Feed to the A-0TSG Teedwater was admitted through the Auxiliary l N heeder via the EW bypass valves. The feedrate was very slow in order to minimize thermal shock to the OTSG and resultant depres l surization of the RCS. XCS pressure control
- t. was very unstable at this time. It is postulu that some localized' boiling.occured in core g at this point' as ' indicated by .self neutron
[ detectors. f, , k
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'ime q 14:57:09 -
M .' Bnassed Rupture Matchthe B-0Ts [E B4 k G (*F
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l.15:00:0 in 5 14:57:15 -(whe Pump Establish Seal This Rat (whet degra DReestalshed AW Level This
- 1 - the 01 coolin
, genera i 15:00:09 proceet
# yypSubcooled "A" Loop 4
This va
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J being et m 15:15 fill anc
, a equalize
" 23 F Delta-T/ Manned the T,chnical Support Center At this t equalizat 15:17 from loo couples.p Declared Class "B" Emerg ency This was d loss of co.
i containment non-essenti 15:19
- , , to' evacuate Opened Emargency W s' an.
to B-0TSG Surve:
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Block
, At this poit and l* the deci 1 , the B-0TSG s:
l l go 95% on bot
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cooldown limi i RCS pressure
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m . s Event l valve Time This was recultant from the tank supp yi n and l opening when the 4 psigSodium RB iso at oThe sodium Lo Level Alarm in Sodium Hydro'xide 15:7 A I Hydroxide Tank ing signal actuated. released to both LPI tr from.the BWST. RCS.) was admitted to the been RCS via.HPI(A satis-lines) Terminated HPI At this time, all conditions had 15:50 to terminate HPI. lly had been established using no.maHPI wa discontinued. letdown. all releases to the RB were ature At this time, RCS pressure and temperNatur (approximately 23*F 16:00 Commenced Pressuriser were well under control.was RCS temperature functionin was approx-Heatup ure at RCS pressThe delta-T).at approximately 450'. decision was na imately 2300 psig. urizer beacup in this point to commence press space in preparation to re-establish a steam the pressurizer. d no radiation The Emergency Survey Team reporteoff site wer Survey Team Report survey results taken 16:07 ground. as running, The motor driven Emergency Wfor pu=papp-wump w Shutdown Steam Drive therefore, the steam driven p the press-16:08 :04 Emergency W Pump The plant remained in this con urizer to saturation temperature Media was notified of plant status. Press Release 16:15 At this point, pressurizer ce=perature was Established Steam Space letdown. 18:05 approximately 620*F. educed Pressurizer brought back on scale by incr i to normal operating level an heaters.
' was established via pressure State and Federal Agencies notified. >
Terminated Class B Emergency f
' 18:30 b
- f. C-8 t
i t (' Rev 5 {, Page 9 I [' Time Event Cause/ Comments (
, 21:07 Forced Flow Initiated The decision was made to re-establish forced r- in RCS flow cooling in the RCS at this time. B&W I
and NRC were consulted. RCP-1B and 1D vere started. At this point,,RCS parameters were 7 stabilized and maintained at RC pressure-2000
-psig, RCS temperature-420*F. Pressurizar
( level-235 inches. The plant was considered in a normal configuration. ( l I o C-9
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