IE Circular No. 80-15, Loss of Reactor Coolant Pump Cooling and Natural Circulation Cooldown
SSINS No.: 6835
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D. C. 20555
June 20, 1980
IE Circular No. 80-15
LOSS OF REACTOR COOLANT PUMP COOLING AND NATURAL CIRCULATION COOLDOWN
Description of Circumstances:
This Circular contains information on the St. Lucie Unit 1 response to a
total loss of component cooling water (CCW) flow to reactor coolant pumps.
Pressurizer level and primary system pressure response indicate that voids
were formed in the reactor vessel during the ensuing natural circulation
cooldown. The void was believe to have been steam in the area located
directly under the reactor vessel head.
At time 0226 on June 11, 1980 with St. Lucie Unit 1 at full power, an
electrical short across a solenoid valve terminal board caused one of two
series containment isolation valves in the CCW return from all reactor
coolant pumps (RCP) to fail shut. The terminal short resulted from
environmental effects of a minor steam leak in the immediate vicinity of the
solenoid valve. After unsuccessful attempts to restore CCW flow, the reactor
was tripped manually at time 0233. Within two minutes, all four RCPs were
also manually tripped. A natural circulation cooldown was initiated at
Component cooling water flow to RCPs was restored at 0400. The solenoid
operated air valve whose terminal board had shorted was bypassed with a
temporary air line to reopen the CCW valve (HCV-14-6). Although variations
in seal leakoff flowrates were observed, the seals on the four idle RCPs did
not fail. St. Lucie has Byron Jackson reactor coolant pumps with three
stage mechanical seals plus a vapor seal. Controlled reactor coolant
bleedoff flow is used for seal cooling and lubrication. The pumps do not
have a seal water injection system.
The natural circulation cooldown continued uneventfully until after time
0600. The highest cooldown rate achieved was approximately 65 to 70 F per
hour. Between 0600 and 0630 RCS pressure was reduced from 1140 to 690 psi by
charging water through the pressurizer auxiliary spray line. Pressurizer
level increased rapidly around 0700 while charging via the auxiliary spray
line. Pressurizer level continued variations for approximately five hours
while the cooldown and depressurization continued. When charging was
shifted to the RCS loops, pressurizer level decreased at rates lower than
the rates of increase in level when charging through the auxiliary spray
line. During a two minute interval while charging into the pressurizer
auxiliary spray line at 88 gpm, pressurizer level rose at a rate
approximately ten times greater than the charging flowrate.
IE Circular No. 80-15 June 20, 1980
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The cause of pressurizer level variations appears to be formation of a steam
void in the reactor head area due to a temperature lag with respect to bulk
coolant temperature that developed because of insufficient cooling flow in
that area during natural circulation cooldown. Gas concentration in a 0730
RCS sample was determined by the licensee to be 32 cc/kg hydrogen and 16
cc/kg nitrogen. These concentrations were not high enough to cause a
significant volume of gas to come out of solution. The indicated subcooling
margin (Tsat - Thot) ranged between 220 and 150 F when the reactor head
steam void developed between 0600 and 0700. The minimum required subcooling
of 50 F was not approached during the cooldown until around time 1219 at a
pressurizer pressure of approximately 110 psig. An annotated record of
pressurizer level is enclosed only to illustrate the high amplitudes and
rates of pressurizer level variations. Insufficient information is
contained in the enclosure to enable detailed evaluation of the RCS
Shutdown cooling using low pressure safety injection (LPSI) pump 1B was
established at time 1051. At time 1227, LPSI pump 1A was started taking
suction from the refueling water tank (RWT) and discharging into the low
pressure safety injection header common to both LPSI pumps. The common
recirculation line motor operated isolation valves to the RWT were opened at
1226. LPSI 1A was operating with its recirculation (miniflow) line open.
The LPSI pump 1B recirculation line should have been closed. The
pressurizer was filled to what was believed to be a water solid condition by
charging at 88 gpm and using LPSI pump 1A to inject water and maintain near
shutoff head pressure on the RCS. RCS pressure rose from the minimum 110
psig to 2000 psig (time 1300 reading) during the time LPSI pump 1A was
The cold calibrated pressurizer level instrument indication roseto 64
percent and remained constant while hot calibrated channels rose to 100%
level. Temperature correction data for the cold calibrated level instrument
didn't extend to 360 F, which was the approximate pressurizer temperature
when the pressurizer indicated full. Constant level on the cold calibrated
channel indicated the pressurizer was solid, but continued charging flow at
88 gpm was not causing pressure to rise above 200 psig,as it should have had
the RCS been solid. Letdown had been secured while filling the pressurizer.
The indications of a steam void in the reactor vessel head were no longer
evident after RCS pressure increased although the exact time when the void
disappeared has not been established.
During the time LPSI pump 1A was operating with miniflow recirculation to
the RWT, the absence of rising pressure in response to charging flow as
investigated. RWT level increased 0.3 feet (approximately 4500 gal.) during
this period. Miniflow from LPSI pump 1B operating in the shutdown cooling
mode is believed to have been the discharge path from the RCS to the RWT.
After shutdown cooling system warmup, the LPSI pump 1B miniflow manual
isolation valve had been shut. After RWT level increased, this valve was
found one turn open, which would have allowed miniflow back to the RWT from
LPSI pump 1B. At time 1357, LPSI pump 1A operating in the injection mode
was secured and miniflow was isolated. Continued charging with two pumps at
88 gpm total flowrate caused a slight rise in both pressurizer (to 260 psig)
and cold calibrated pressurizer level. Letdown in excess of charging
IE Circular No. 80-15 June 20, 1980
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established at 1430, and by 1500 a steam bubble was drawn in the pressurizer
with level in the indicating range.
The RCS was degasified over the next day, then depressurized and drained to
inspect and replace all RCP seals. All seals have been removed and visual
inspection showed no signs of degradation of these seals.
This event is significant for several reasons. It is an example of a
natural circulation cooldown during which a steam void formed under
seemingly normal conditions that was large enough to cause large, rapid
variations in pressurizer level. The possibility of a total loss of
component cooling water to reactor coolant pumps due to the single failure
of any one of four CCW containment isolation valves was highlighted.
Inadequate control of LPSI system alignment allowed an unanticipated
discharge of reactor coolant directly to the refueling water tank apparently
through a pump recirculation line.
Recommended Action for Power Reactor Licensee Consideration:
1. Disseminate this information to all licensed operating personnel
working at power reactor facilities. These personnel should become
aware of the possibility of steam void information in the reactor
vessel head during natural circulation cooldown even when a high
subcooling margin exists in the reactor coolant loops.
2. Review and revise natural circulation cooldown and shutdown cooling
procedures to caution operators against the anomalous conditions that
occurred and to include appropriate recovery action if they do occur.
3. Establish a natural circulation cooldown and depressurization rate
envelop that will both preclude steam void formation and assure
adequate core cooling. incorporate this envelop in cooldown
4. Evaluate the design of component cooling water systems to determine
vulnerability to single failures that could cause loss of RCP cooling,
common cause failures of RCP seals and reactor coolant systems leaks
through failed seals at multiple locations.
5. Consideration installation of a reactor vessel head metal temperature
monitoring system, if not already installed. It should aid the
operator in preventing a reactor head to bulk coolant temperature
differential large enough to form a steam void during natural
If you have questions regarding this matter, please contact the Director of
the appropriate NRC Regional Office.
No written response to this Circular is required.
Annotated record of
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