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 NRC INFORMATION NOTICE 93-27: 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
gradually recovered over a period of two hours. The licensee postulatedand was
this large error in level indication was caused by gas released in the that
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
dissolved noncondensible gas in that reference leg. of
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
the RHR system during shutdown cooling. The design basis for WNP-2 includes in
postulated leak in the RHR system piping outside containment while the plant a
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
1993 plant cooldown, the licensee concluded that, with the observed errors21, level indication, the shutdown cooling suction valves may not have in
automatically isolated the RHR system on low reactor vessel water level
designed. The licensee has implemented compensatory measures for future as;
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
the IC' reference leg. from
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.
large sustained level error is of concern because of the potential for A
complicating long-term operator actions. In addition, the scenario of
postulated leak in the RHR system evaluated by WNP-2 suggests that some a
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
that licensees determine the impact of potential level indication errors part, 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
IN 93-27 April 8, 1993 Pge I of I
dc
LIST OF RECENTLY ISSUED ME0
HRC INFORMATION NOTICES
inroruauon a- Inforuti0n u te OT
Notice No. Subject Issuance Issued to
93-26 Grease Solidification 04/07/93 All holders of OLs or CPs
Causes Molded Case for nuclear power reactors.
Circuit Breaker
Failure to Close
) 93-25 Electrical fenetration
Assembly Degradation
04/01/93 All holders of OLs or Cps
for nuclear power reactors.
93-24 Distribution of 03/31/93 All holders of operator and
Revision 7 of NUREG-1021, senior operator licenses at
- Operator Licensing nuclear power reactors.
Examiner Standards'
93-23 Veschler Instruments 03/31/93 All holders of OLs or CPs
Model 252 Switchboard for nuclear power reactors.
Meters
93-22 Tripping of Ilockner- 03/26/93 All holders of Ots or CPs
toeller Molded-Case for nuclear power reactors.
Circuit Breakers due to
Support Level Failure
93-21 Sumary of NRCStaff 03/25/93 All holders of Ots or CPs
Observations Compiled for light water nuclear
'during Engineering Audits power reactors.
or Inspections of Licen- see Erosion/Corrosion
Programs
2
93-20 Thermal Fatigue Cracking 03/24/93 All holdefs of Os or CPs ° _
of Feedwater Piping to for PFRs supplied by
) Stem Generators Westinghouse or Combustion
Engineering.
co
clo
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0
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93-19 Slab Hopper Bulging 03/17/92 All nuclear fuel cycle
licensees. II
93-18 Portable Moisture-Density 03/10/93 All U.S. Nuclear Regulatory
Gauge User Responsibilities Couuission licensees that
during Field Operations possess moisture-density a.
- )4
2 u
OL - Operating License z aL.
- Construction Permit
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- 3
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 *C/OGCB:DORS:NRR *TECH:ED
JLBirmingham GHMarcus RSanders
04/01/93 04/01/93 03/18/93
- SRXB:DSSA:NRR *C/SRXB:DSSA:NRR *D/DSSA:NRR
ACubbage RJones AThadani
03/19/93 03/26/93 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 cconcurrence
OGCB:DORS:NRR C/OGCB:DORS:NRR D/DORS:NRR *TECH:ED
JLBirmingham GHMarcusas BKGrimesrMk RSanders
03"' gY.Z-
3 °f 1 /93C tW 03/ /931' 03/18/93
- SRXB:DSSA:NRR *C/SRXB:DSSA:NRR *D/DSSA:NRR
ACubbage RJones AThadani
03/19/93 03/26/93 03/26/93 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 *TECHED:ADM
JLBirmingham GHMarcus BKGrimes RSanders
03//903/ 93 03 /93 03/ /93
- SRXB:DSSA:NRR B:DSSA:NR& D/DSSA:NR
ACubbage RJ nes Thadani
03/ /93 0 t;//93 03 X/931
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 C/OGCB: DORS:NRR 0/DORS:NRR TECHED LADM
JLBirmingham GHMarcus BKGrimes JMain
03/ /93 03/ /93 03/ /93 03/8 /93 SRXB:DSSA~:NlRR C/SRXB:DSSA:NRR D/DSSA:NRR
ACubbagqAtf-~ RJones AThadani
03/lcj/93 03/ /93 03/ /93