Information Notice No. 93-27: Level Instrumentation Inaccuracies Observed During Normal Plant Depressurization
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
April 8, 1993
NRC INFORMATION NOTICE 93-27: LEVEL INSTRUMENTATION INACCURACIES OBSERVED
DURING NORMAL PLANT DEPRESSURIZATION
All holders of operating licenses or construction permits for nuclear power
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to inaccuracies in reactor vessel level indication
that occurred during a normal depressurization of the reactor coolant system
at the Washington Nuclear Plant Unit 2 (WNP-2) and to the fact that errors in
level indication may result in a failure to automatically isolate the residual
heat removal (RHR) system under certain conditions. It is expected that
recipients will review the information for applicability to their facilities
and consider actions, as appropriate, to avoid similar problems. However,
suggestions contained in this information notice are not NRC requirements;
therefore, no specific action or written response is required.
As discussed in NRC Information Notice 92-54, "Level Instrumentation
Inaccuracies Caused by Rapid Depressurization," and Generic Letter 92-04,
"Resolution of the Issues Related to Reactor Vessel Water Level
Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)," noncondensible gas may
become dissolved in the reference leg of water level instrumentation and lead
to false indications of high level after a rapid depressurization event.
Reactor vessel level indication signals are important because these signals
are used for actuating automatic safety systems and for guidance to operators
during and after an event. While Information Notice 92-54 dealt with
potential consequences of rapid system depressurization, this information
notice discusses level indication errors that may occur during normal plant
cooldown and depressurization.
Description of Circumstances
On January 21, 1993, during a plant cooldown following a reactor scram at
WNP-2, "notching" of the level indication was observed on at least two of four
channels of the reactor vessel narrow range level instrumentation. "Notching"
is a momentary increase in indicated water level. This increase occurs when a
gas bubble moves through a vertical portion of the reference leg and causes a
temporary decrease in the static head in the reference leg. The notching at
April 8, 1993
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WNP-2 was first observed on channel "C" at a pressure of approximately
827 kPa [120 psig]. Channel "B" experienced notching starting at
approximately 350 kPa [50 psig]. At these pressures, the level error was on
the order of 10 to 18 centimeters [4 to 7 inches] and persisted for
approximately one minute.
Beginning at a pressure of approximately 240 kPa [35 psig], the level
indication from channel "C" became erratic and, as the plant continued to
depressurize, an 81-centimeter [32-inch] level indication error occurred.
This depressurization was coincident with the initiation of the shutdown
cooling system. The 81-centimeter [32-inch] level error was sustained and was
gradually recovered over a period of two hours. The licensee postulated that
this large error in level indication was caused by gas released in the
reference leg displacing approximately 40 percent of the water volume. The
licensee also postulated that the slow recovery of correct level indication
was a result of the time needed for steam to condense in the condensate
chamber and refill the reference leg. The licensee inspected the "C"
reference leg and discovered leakage through reference leg fittings. This
leakage may have been a contributing factor for an increased accumulation of
dissolved noncondensible gas in that reference leg.
The licensee determined that the type of errors observed in level indication
during this event could result in a failure to automatically isolate a leak in
the RHR system during shutdown cooling. The design basis for WNP-2 includes a
postulated leak in the RHR system piping outside containment while the plant
is in the shutdown cooling mode. For this event, the shutdown cooling suction
valves are assumed to automatically isolate on a low reactor vessel water
level signal to mitigate the consequences of the event. For the January 21,
1993 plant cooldown, the licensee concluded that, with the observed errors in
level indication, the shutdown cooling suction valves may not have
automatically isolated the RHR system on low reactor vessel water level as
designed. The licensee has implemented compensatory measures for future plant
cooldowns to ensure that a leak that occurs in the RHR system during shutdown
cooling operation would be isolated promptly. These measures include touring
the associated RHR pump room hourly during shutdown cooling and backfilling
the water level instrument reference legs after entry into mode 3 (hot
shutdown). The licensee is also evaluating measures to minimize leakage from
the "C" reference leg.
The event described above is different than events previously reported because
of the large magnitude and sustained duration (as opposed to momentary
notching) of the level error that occurred during normal plant cooldown. A
large sustained level error is of concern because of the potential for
complicating long-term operator actions. In addition, the scenario of a
postulated leak in the RHR system evaluated by WNP-2 suggests that some safety
systems may not automatically actuate should an event occur while the reactor
is in a reduced pressure condition. Generic Letter 92-04 requested, in part,
that licensees determine the impact of potential level indication errors on.
April 8, 1993
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automatic safety system response during licensing basis transients and
accidents. The information in this notice indicates that sustained level
instrument inaccuracies can occur during a normal reactor depressurization.
Therefore, events occurring during low pressure conditions may also be
complicated by level indication errors.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.
ORIGINAL SIGNED BY
Brian K. Grimes, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical contact: Amy Cubbage, NRR
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