Information Notice 1993-27, Level Instrumentation Inaccuracies Observed During Normal Plant Depressurization
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C.
20555
April 8, 1993
LEVEL INSTRUMENTATION INACCURACIES OBSERVED
DURING NORMAL PLANT DEPRESSURIZATION
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
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.
Background
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, i993, 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
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IN 93-27 April 8, 1993 WNP-2 was first observed on channel NC" at a pressure of approximately
827 kPa (120 psig].
Channel IBS 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 IC' 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 IC"
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 suctlon
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 IC' reference leg.
Discussion
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
IN 93-27 April 8, 1993 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.
Brian K. Grimes, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical contact:
Amy Cubbage, NRR
(301) 504-2875 Attachment:
List of Recently Issued NRC Information Notices
Attachnent
April 8, 1993
Pge I of I
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LIST OF RECENTLY ISSUED
HRC INFORMATION NOTICES
inroruauon
Inforuti0n
Notice No.
93-26
)
93-25
93-24
93-23
93-22
93-21 Subject
Grease Solidification
Causes Molded Case
Circuit Breaker
Failure to Close
Electrical fenetration
Assembly Degradation
Distribution of
Revision 7 of NUREG-1021,
- Operator Licensing
Examiner Standards'
Veschler Instruments
Model 252 Switchboard
Meters
Tripping of Ilockner- toeller Molded-Case
Circuit Breakers due to
Support Level Failure
Sumary of NRC Staff
Observations Compiled
'during Engineering Audits
or Inspections of Licen- see Erosion/Corrosion
Programs
Thermal Fatigue Cracking
of Feedwater Piping to
Stem Generators
Slab Hopper Bulging
Portable Moisture-Density
Gauge User Responsibilities
during Field Operations
u te OT
Issuance
Issued to
04/07/93
All holders of OLs or CPs
for nuclear power reactors.
04/01/93
All holders of OLs or Cps
for nuclear power reactors.
03/31/93
All holders of operator and
senior operator licenses at
nuclear power reactors.
03/31/93
All holders of OLs or CPs
for nuclear power reactors.
03/26/93
All holders of Ots or CPs
for nuclear power reactors.
03/25/93
All holders of Ots or CPs
for light water nuclear
power reactors.
93-20
)
93-19
93-18
03/24/93
All holdefs of Os or CPs
for PFRs supplied by
Westinghouse or Combustion
Engineering.
03/17/92
All nuclear fuel cycle
licensees.
03/10/93
All U.S. Nuclear Regulatory
Couuission licensees that
possess moisture-density
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IN 93-27 April 8, 1993 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. Grime:
Brian K. Grimes, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical contact:
Amy Cubbage, NRR
(301) 504-2875 Attachment:
List of Recently Issued NRC Information Notices
- See previous concurrence
- OGCB:DORS:NRR
JLBirmingham
04/01/93
- SRXB:DSSA:NRR
ACubbage
03/19/93
- C/OGCB:DORS:NRR
GHMarcus
04/01/93
- C/SRXB:DSSA:NRR
RJones
03/26/93
- TECH:ED
RSanders
03/18/93
- D/DSSA:NRR
AThadani
03/26/93 Document name: 93-27.IN
IN 93-XX
March XX, 1993 errors on 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
(one of) the technical contact(s) listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical contact:
Amy Cubbage, NRR
(301) 504-2875 Attachment:
List of Recently Issued NRC Information Notices
- See previous c
OGCB:DORS:NRR
JLBirmingham
03"'
3 gY.Z-
- SRXB:DSSA:NRR
ACubbage
03/19/93 concurrence
C/OGCB:DORS:NRR
GHMarcusas
°f 1 /93C tW
- C/SRXB:DSSA:NRR
RJones
03/26/93 D/DORS:NRR
BKGrimesrMk
03/ /931'
- D/DSSA:NRR
AThadani
03/26/93
- TECH:ED
RSanders
03/18/93 Document name:
RVLEVEL.IN
\\
I
This information notice requires no specific action or written response.
If
you have any questions regarding the information in this notice, please
contact the technical contact listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contact:
Amy Cubbage, NRR
(301) 504-2875 Attachment: List of Recently Issued NRC Information Notices
Document name:
RVLEVEL.IN
- SEE PREVIOUS CONCURRENCE
OGCB:DORS:NRR
C/OGCB:DORS:NRR
D/DORS:NRR
JLBirmingham
GHMarcus
BKGrimes
03//903/
93
03
/93
- SRXB:DSSA:NRR
B:DSSA:NR &
D/DSSA:NR
ACubbage
RJ nes
Thadani
03/ /93
0 t;/
/93
03 X/931
- TECHED:ADM
RSanders
03/ /93
IN 93-XX
March XX, 1993 This information notice requires no specific action or written response. If
you have any questions regarding the information in this notice, please
contact the technical contact listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contact:
Amy Cubbage, NRR
(301) 504-2875 Attachment: List of Recently Issued NRC Information Notices
Document name:
INFONOT2.RVL
OGCB:DORS:NRR
JLBirmingham
03/ /93 C/OGCB: DORS:NRR
GHMarcus
03/ /93
0/DORS:NRR
BKGrimes
03/ /93 TECHED LADM
JMain
03/8 /93 SRXB:DSSA~:NlRR
ACubbagqAtf-~
03/lcj/93 C/SRXB:DSSA:NRR
RJones
03/ /93 D/DSSA:NRR
AThadani
03/ /93