Information Notice No.80-10 – Partial Loss of Non-Nuclear Instrument System Power Supply During Operation
SSINS No.: 6870
Accession No.:
8002280640
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
March 7, 1980
Information Notice No. 80-10
PARTIAL LOSS OF NON-NUCLEAR INSTRUMENT SYSTEM POWER SUPPLY DURING OPERATION
Description of Circumstances:
This notice contains information regarding Crystal River Unit 3 response to
a loss of non-nuclear instrumentation (NNI) as a consequence of loss of the
+24 volt power supply to the NNI.
At 2:23 p.m. on February 26 with Crystal River Unit 3 at 100% power, the +24
volt power supply to the NNI was lost, due to a short to ground. This
initiated a sequence of events (detailed in the enclosure) wherein the PORV
opened and stayed open as a direct result of the NNI power supply loss. HPI
initiated as a result of depressurization through the open PORV, and with
approximately 70% of NNI inoperable or inaccurate, the operator correctly
decided that there was insufficient information available to justify
terminating HPI. Therefore, the pressurizer was pumped solid, one safety
valve lifted, and flow through the safety valve was sufficient to rupture
the RC Drain Tank rupture disk, spilling approximately forty-three thousand
gallons of primary water into containment.
The Crystal River 3 event is closely related to the November 10, 1979 event
at Oconee Unit 3 wherein the inverter supplying power to the Integrated
Control System (ICS) and to parts of the NNI failed. That event was the
subject of Information Notice No. 79-29 (November 16, 1979) which was
followed by IE Bulletin 79-27 (November 30, 1979).
The CR-3 event involved loss of only part of the power available from an
inverter, rather than the inverter itself, since the +24v supply is only one
of several power supplies drawing power from one inverter. The effects are
very similar, however, in that the ICS lost part of its input signals in both
events.
The +24 volt power supply short to ground has tentatively been identified by
the licensee to have occurred between knife edge connectors of a Bailey
Control Company Voltage Buffer Card. The voltage buffer card was misaligned
in its receptacle, and adjacent connectors carrying +24v and "common" were
bent such that they contacted one another. This short circuit cleared itself
during subsequent re-energizing of the power supply by burning through the
foil on a printed circuit card. Subsequent review by the licensee identified
a second voltage, buffer card which was also misaligned but had not caused
a short circuit. The specific circuit cards which were misaligned carried
part number 6624609L1. The connectors on these cards are slightly thinner
and appear to have a somewhat different angle than those found on similar
cards elsewhere in the NNI which carry part numbers 6624608A1 or 6624609A1.
The 6624609LI cards appear to be more subject to misalignment.
.
Information Notice No. 80-10 March 7, 1980
Page 2 of 2
The specific shorted voltage buffer card provided the signal to the NNI "x"
saturation meter.
Licensees which utilize Bailey Control Company Voltage Buffer Cards are
requested to carefully inspect the cards for possible misalignment and take
corrective actions if misalignments are identified. Specific instructions
for carrying out these inspections and providing any other information which
may be required to define appropriate corrective action is being prepared by
Baily Control Company for transmittal to purchasers of this equipment by
March 11, 1980.
Initial screening of IE Bulletin No. 79-27 responses indicates a range of
responses regarding depth and scope of review.
IE Bulletin No. 79-27 was intended to cause licensees to investigate loss of
individual power supplies as well as total loss of an inverter or vital bus.
An addendum to IE Bulletin No. 79-27 is planned to be issued in the near
future to reflect the CR-3 event.
This Information Notice is provided to inform licensees of a possibly
significant matter. It is expected that recipients will disseminate the
information to all operational personnel working at their licensed
facilities. (A meeting was held with B&W licensees in NRC Headquarters on
March 5, 1980 to review the event at Crystal River and to discuss proposed
corrective actions. Responses to specific questions have been requested of
the-B&W licensees.) If you have questions regarding this matter, please
contact the Director of the appropriate NRC Regional Office.
No written response to this Information Notice is required.
Enclosure:
Sequence of Events
.
26 FEBRUARY TRANSIENT CRYSTAL RIVER UNIT-3
SEQUENCE OF EVENTS
EVENT SYNOPSIS
At 14:23 on February 26, 1980, Crystal River-3 Nuclear Station experienced
a reactor trip. Nominal full power primary and secondary system parameters
were present. A synopsis of key events and parameters was obtained from the
plant computer's post-trip review and plant alarm summary, the sequence of
events, monitor, control room strip charts, and the Shift Supervisor's log.
The reactor was operating at approximately 100% full power with Integrated
Control System (ICS) in automatic. No tests were in progress.
Time Event Cause/Comments
14:23:21 +24 Volt Bus Failure (NNI power The positive 24 VDC bus
loss "X" supply) shorted dragging the bus
voltage down l to a low
voltage trip condition.
There is a built-in 1/4 to
1/2 second delay at which
time all power supplies
will trip. The trip
indication on negative (-)
voltage was missed by the
annunciator. Following the
NNI power failure, much of
the control room
indication was lost. Of the
instrumentation that
remained operable,
transient conditions made
their indications
questionable to the
operators.
14:23:21 PORV and Pressurizer Spray When the positive 24 VDC
Valve Open supply was lost due to the
sequence discussed above
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 PORV opened fully
and the spray valve stroked
for approximately 112
second. The "40% open"
indication on spray valve
did not actuate, therefore,
the spray valve did not
exceed 40% open.
.
26 February Transient CR-3 -2-
Time Event Cause/Comments
14:23:21 Reduction in Feedwater As a result of the "X"
power supply failure many
primary plant control
signals responded
erroneously. T-cold failed
to 570F (normal
indication was 557F)
producing several spurious
alarms. T-ave failed to
570F (decreased). The
resultant T-ave error
modified the reactor demand
such that control rods were
withdrawn to increase T-ave
and reactor power. The
power increase was
terminated at 103% by the
ICS and a "Reactor Demand
High Limit" alarm was
received T-hot failed to
570 F (low) and RC flow
failed to 40 x 106 lbs/hr
in each loop (low). Both
these failures created a
BTU alarm and limit on
feedwater which reduced
feedwater flow to both
OTSG's to essentially zero.
Turbine header pressure
failed to 900 psig (high)
which caused the turbine
valves to open slightly to
regulate header pressure
thus increasing generated
megawatts. These combined
failures resulted in a loss
of heat sink to the reactor
initiating high RCS
pressure condition.
14:23:35 Reactor Trip/Turbine Trip Rx trip caused by high RCS
pressure at 2300 psi.
Turbine was tripped by the
reactor.
14:24:02 Hi Pressure Inj. This was a computer
Reg. (Flag) printout and indicates <
50 subcooling. (The
lowest level of subcooling
was 8F for a very
short period of time, at
about 14:30)
14:24:02 Loss of Both Condensate Suspect that the condensate
Pumps pump tripped due to high
De-aeriating feed tank
level.
.
26 February Transient CR-3 - 3 -
Time Event Cause/Comments
14:24:02 This is verified by a
(Continued) series of questions marks
???? printed by computer
indicating that the level
instrument was over-ranged.
A low flow indication in
the gland steam condenser
was also indicated by
computer.
14:25:50 PORV Isolated At this time a high RC
Drain Tank level alarm was
received. This was
resultant from the PORV
remaining open and was
positive indication that
the PORV was open. At this
time, the operator closed
the PORV block valve due
to RCS pressure decreasing
and high RCDT level.
14:26:41 HPI Auto Initiation HPI initiated automatically
due to low RCS pressure of
1500 psig. The low
pressure condition was
resultant from the PORV
remaining full open while
the plant was tripped.
Full HPI was initiated with
3 pumps resulting in
approximately 1100 gpm flow
to the RCS. At this time,
all remaining non-essential
reactor building (RB)
isolation valves were
closed per TMI Lessons
Learned Guidelines.
14:26:54 RC Pumps Shutdown Operator turned RC pumps
off as required by the
applicable emergency
procedure and B&W small
break guidelines.
14:27:20 RB Pressure Increasing This is first indication
that RCDT rupture disc had
ruptured. RB pressure
increase data was obtained
from Post Trip Review and
Strip Chart indication.
.
26 February Transient CR-3 - 4 -
Time Event Cause/Comments
14:31:32 RB Pressure High This alarm was initiated
by 2 psig in RB. This is
attributed to steam release
from RCDT. Code safeties
had not opened at this time
based upon tail pipe
temperatures recorded at
14:32:03 (Computer).
14:31:49 OSTG "A" Rupture Matrix This occurred due to < 600
Actuation psig in OTSG "A". The low
pressure was caused by OTSG
"A" boiling dry which was
resultant from the BTU
limit and failed power
supply to OTSG "A" level
transmitter. This resulted
in the closure of all
feedwater and Steam block
valves which service OTSG
"A".
14:31:59 Main Feedwater Pump 1A Caused by suction valve
Tripped shutting due to rupture
matrix actuation in
previous step.
14:32:14:41 ES A/B Bypass Manually bypassed and HPI
balanced between all 4
nozzles (Total flow
approximately 1100
gpm-small break operating
guidelines).
14:32:35 Started Steam Driven Started by operator to
Emergency Feedwater Pump ensure feedwater was
available to feed OTSG's.
14:33 Core Exit Temp. Verified The core exit incore
thermocouples indicated
the highest core outlet
temperature value was
560F. RCS pressure
was 2353 psig at this time,
therefore, the subcooling
margin at this time was
100F. Minimum
subcooling margin for the
entire transient was
8F at 14:30. It is
postulated that some
localized boiling occurred
in the core at this point
as indicated by the self
powered neutron detectors.
.
26 February Transient CR-3 - 5 -
Time Event Cause/Comments
14:33:14:44 Started Motor Driven Emer- Same discussion as "Started
gency Feedwater Pump Steam Driven Emergency
Feedwater Pump."
14:33:30 RC Pressure High (2395 At this point, pressurizer
psig) is solid and code safety
lifts (RCV-8). This is the
highest RCS pressure as
recorded on Post Trip
Review. Apparently, RCV-8
lifted early due to seat
leakage prior to the
transient and RCV-9 did not
lift.
14:34:23 RB Dome Hi Rad Level RMG-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
pressurizer and coolant.
14:35:13 Attempted NNI Repower This resulted in spikes
Without Success observed on de-energized
strip charts.
14:36:50 Computer Overload Caused by overload of
buffer. Resulted in no
further computer data until
buffer catches up with
printout.
14:38:15 FWV-34 Closed This valve was closed to
prevent overfeeding OTSG
"B" beyond 100% indicated
Operating Range.
14:44:12 NNI Power Restored NNI was restored by
Successfully removing the "X"-NNI Power
Supply Monitor Module. This
allowed the breakers to be
reclosed. At this time, it
was observed that the "A"
OTSG was dry, the
pressurizer was solid
(Indicated off-scale high),
RC outlet temperature
indicated 556F (loop
A & B average), and RC
average temperature
indicated 532F (Loop
A & B). The highest core
exit thermocouple
temperature at
.
26 February Transient CR-3 - 6 -
Time Event Cause/Comments
14:44:12 this time was 531F.
(Continued) RCS pressure was 2400 psig
(saturation temp. at this
pressure is 662F.).
This data verified natural
circulation was in Progress
and the plant subcooling
margin was 131F
(based on core exit
thermocouples).
14:44:31 RB Isolation and Cooling At this time, RB pressure
Actuation increased to 4 psig and
initiated RB Isolation. The
operator verified all
immediate actions occurred
properly for HPI, LPI, and
RB Isolation and Cooling.
The increasing RB pressure
was resultant from RCV-8
relieving pressure due to
continued HPI.
14:46:10 Bypassed HPI, LPI and RB These "ES" systems were
Isolation and Cooling bypassed at this time to
balance HPI flow and
restore cooling water to
essential auxiliary
equipment (i.e., RCP's,
letdown coolers, CRDM's
etc.).
14:51:57 Rupture Matrix Actuation The actuation was resultant
on OTSG-B from a degradation of
OTSG-B pressure. Cold
emergency feed was being
injected into the OTSG at
this time. This matrix
actuation isolated all
feedwater and steam block
valves to the B-OTSG and
tripped the "B" main FW
pump. Both Emergency FW
pumps were already in
operation at this time.
B-OTSG level at this time
was 70% (Operation Range).
14:52 HPI Throttled and RCS At this time, the maximum
Pressure Reduced to 2300 core exit thermocouple
psig temperature was 515F,
RCS pressure was 2390
psig.
.
26 February Transient CR-3 - 7 -
Time Event Cause/Comments
14:52 Therefore, the subcooling
(Continued) margin was 147F.
Natural circulation was in
effect as verified
previously. All conditions
had been satisfied to
throttle HPI. Therefore,
flow was throttled to
approximately 250 gpm to
reduce RCS pressure to 2300
psig in order to attempt
to reduce the flow rate
through RCV-8 and into the
RB.
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 Pump Recirc. This was done to assure the
Valves MU pumps would have minimum
flow at all times to
prevent possible pump
damage.
14:56:43 Bypassed the A-OTSG Feedwater was slowly
Rupture Matrix and admitted to the A-OTSG
Reestablished Feed which was dry up to this
to the A-OTSG point. Feedwater was
admitted through the
Auxiliary FW header via
the EFW bypass valves. The
feedrate was very slow in
order to minimize thermal
shock to the OTSG and
resultant depressurization
of the RCS. RCS pressure
control was very unstable
at this time.
14:57:09 Bypassed the B-OTSG This was done to regain FW
Rupture Matrix control of the B-OTSG.
Level was still high in
this OTSG (approximately
65% Operating Range).
Therefore, feed was not
necessary at this time. The
Main Steam Isolation valves
were open in preparation
for bypass valve operation
(when necessary).
.
26 February Transient CR-3 - 8 -
Time Event Cause/Comments
14:57:15 Established RC Pump This was done in
Seal Return preparation for a RCP start
(when necessary) and to
minimize pump seal
degradation.
15:00-09 Reestablished Level This verified feedwater was
in A-OTSG being admitted to the OTSG
and made it available for
core cooling via natural
circulation. Feed to this
generator was continued
with the intent of
proceeding to 95% on the
Operating Range.
15:00-09 77F Subcooled "A" Loop This value was based upon
"A" RCS loop parameters
at this time. The "A" loop
was being cooled down at
this time by the A-OTSG
fill and the operator was
attempting to equalize loop
temperatures.
15:15 23F Delta-T/Manned the At this time, loop
Technical Support Center temperatures were nearing
equalization. This delta-T
was calculated from loop
A & B T C's and core exit
thermocouples.
15:17 Declared Class "B" This was done based on the
Emergency fact there was a loss of
coolant through RCV-8 into
the containment and HPI
had been initiated. All
non-essential CR-3
personnel were directed to
evacuate and contact of
off-site agencies began.
Survey team was sent to
Auxiliary Building.
15:19 Opened Emergency FW At this point the A-OTSG
Block to B-OTSG level was increasing and
the decision was made to
commence filling the B-OTSG
simultaneously. The intent
was to go 95% on both
OTSG's without exceeding
RCS cooldown limits (100
F/hr) while maintaining
RCS pressure control.
.
26 February Transient CR-3 - 9 -
Time Event Cause/Comments
15:26 Lo Level Alarm in Sodium This was resultant from the
Hydroxide Tank tank supply valve opening
when the 4 psig RB
isolation and cooling
signal actuated. The
sodium hydroxide was
released to both LPI
trains. Sodium Hydroxide
was admitted to the RCS
via HPI from the BWST.
(Approximately 2 ppm
injected into the RCS.)
15:50 Terminated HPI At this time, all
conditions had been
satisfied (per small break
operating guidelines) to
terminate HPI. RCS
pressure control had been
established using normal
makeup and letdown. HPI
was terminated and
essentially all releases
to the RB were
discontinued.
16:00 Commenced Pressurizer At this time, RCS pressure
Heatup and temperature were well
under control. Natural
circulation was functioning
as-designed (approximately
23F delta-T). RCS
temperature was being
maintained at approximately
450F. RCS pressure was
approximately 2300 psig.
The decision was made at
this point to commence
pressurizer heatup in
preparation to re-establish
a steam space in the
pressurizer.
16:07 Survey Team Report The Emergency Survey Team
reported no radiation
survey results taken
offsite were above
background.
16:08:04 Shutdown Steam Driven The motor driven Emergency
Emergency FW Pump FW pump was running,
therefore, the steam driven
pump was not needed. The
plant remained in this
condition for approximately
2 hours, while heating up
the pressurizer to
saturation temperature for
1800 psig.
.
26 February Transient CR-3 -10-
Time Event Cause/Comments
18:05 Established Steam Space At this point, pressurizer
in Pressurizer temperature was
approximately 620F.
Pressurizer level was
brought back on scale by
increasing letdown. From
this point pressurizer
level was reduced to
normal operating level and
normal pressure was
established via
pressurizer heaters.
18:30 Terminated Class B State and Federal Agencies
Emergency notified.
21:07 Forced Flow Initiated The decision was made to
in RCS re-establish forced flow
cooling in the RCS at this
time. B&W and NRC were
consulted. RCP-1B and 1D
were started. At this
point, RCS parameters were
stabilized and maintained
at RC pressure-2000 psig,
RCS temperature-420F.
Pressurizer level-235
inches. The plant was
considered in a normal
configuration.
.
Page Last Reviewed/Updated Tuesday, March 09, 2021