Information Notice 1993-89, Potential Problems with BWR Level Instrumentation Backfill Modifications: Difference between revisions

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{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY
{{#Wiki_filter:UNITED STATES


COMMISSION
===NUCLEAR REGULATORY COMMISSION===
OFFICE OF NUCLEAR REACTOR REGULATION


===OFFICE OF NUCLEAR REACTOR REGULATION===
WASHINGTON, D.C.
WASHINGTON, D.C. 20555 November 26, 1993 NRC INFORMATION


NOTICE 93-89: POTENTIAL
20555


PROBLEMS WITH BWR LEVEL INSTRUMENTATION
===November 26, 1993===
NRC INFORMATION NOTICE 93-89: POTENTIAL PROBLEMS WITH BWR LEVEL


===BACKFILL MODIFICATIONS===
===INSTRUMENTATION BACKFILL MODIFICATIONS===


==Addressees==
==Addressees==
All holders of operating
All holders of operating licenses or construction permits for boiling water


licenses or construction
reactors (BWRs).
 
permits for boiling water reactors (BWRs).


==Purpose==
==Purpose==
The U.S. Nuclear Regulatory
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
 
Commission (NRC) is issuing this information
 
notice to alert addressees
 
to potential
 
problems that have been identified
 
by licensees
 
involving
 
hardware modification
 
to the reactor vessel water level instrumentation
 
system. 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.Description
 
of Circumstances
 
NRC Bulletin (NRCB) 93-03, "Resolution
 
of Issues Related to Reactor Vessel Water Level Instrumentation
 
in BWRs," issued on May 28, 1993, requested


that licensees
notice to alert addressees to potential problems that have been identified by


implement
licensees involving hardware modification to the reactor vessel water level


hardware modifications
instrumentation system. It is expected that recipients will review the


necessary
information for applicability to their facilities and consider actions, as


to ensure the level instrumentation
appropriate, to avoid similar problems. However, suggestions contained in


system design is of high functional
this information notice are not NRC requirements; therefore, no specific


reliability
action or written response is required.


for long-term operation.
==Description of Circumstances==
NRC Bulletin (NRCB) 93-03, "Resolution of Issues Related to Reactor Vessel


In response to this bulletin, all BWR licensees
Water Level Instrumentation in BWRs," issued on May 28, 1993, requested that


with the exception of Big Rock Point, which does not use cold reference
licensees implement hardware modifications necessary to ensure the level


leg instrumentation, have either implemented
instrumentation system design is of high functional reliability for long-term


modifications
operation. In response to this bulletin, all BWR licensees with the exception


or have committed
of Big Rock Point, which does not use cold reference leg instrumentation, have


to implement
either implemented modifications or have committed to implement modifications.


modifications.
The majority of these licensees have decided to install a reference leg


The majority of these licensees
backfill system to supply a continuous flow of water from the control rod


have decided to install a reference
drive (CRD) hydraulic system through the reference legs to preclude migration


leg backfill system to supply a continuous
of dissolved noncondensible gases down the legs. In August 1993, a potential


flow of water from the control rod drive (CRD) hydraulic
problem was found at the Susquehanna nuclear power plant during the design of


system through the reference
this backfill modification.
 
legs to preclude migration of dissolved
 
noncondensible
 
gases down the legs. In August 1993, a potential problem was found at the Susquehanna
 
nuclear power plant during the design of this backfill modification.


Discussion
Discussion


It was postulated
It was postulated at Susquehanna that a manual isolation valve in one of the
 
at Susquehanna
 
that a manual isolation
 
valve in one of the reference
 
legs (see Figure 1) could be closed by operator error. Closure of this valve would result in pressurization
 
of that reference
 
leg to CRD system pressure and erroneous
 
indication
 
of low reactor water level and high reactor pressure on all instrumentation
 
associated
 
with that reference
 
leg. The transient
 
resulting
 
from pressurization
 
of the most limiting reference
 
leg 9311190454
1 KE-e 93-o 93 Ill
 
K.,_ IN 93-89 November 26, 1993 includes reactor scram and opening of all safety relief valves (SRVs) due to the false high reactor pressure.
 
===The SRVs would remain open and depressurize===
the reactor until the valves are closed by operator action, or actual reactor pressure falls below approximately
 
446 kPa [50 psig] at which time the valves can no longer stay open. Reactor depressurization
 
and loss of inventory through the SRVs, in combination
 
with the false low water level signal on the affected reference
 
leg, would result in closure of the main steam isolation valves, actuation


of high-pressure
reference legs (see Figure 1) could be closed by operator error.


and low-pressure
Closure of


emergency
this valve would result in pressurization of that reference leg to CRD system


core cooling system (ECCS) and containment
pressure and erroneous indication of low reactor water level and high reactor


isolation.
pressure on all instrumentation associated with that reference leg. The


Low-pressure
transient resulting from pressurization of the most limiting reference leg


ECCS injection
9311190454
1 KE-e


would commence after the low-pressure
93-o


permissive
93 Ill


is satisfied.
K.,_
IN 93-89 November 26, 1993 includes reactor scram and opening of all safety relief valves (SRVs) due to


===This permissive===
the false high reactor pressure. The SRVs would remain open and depressurize
would be satisfied


in this scenario, allowing the low pressure ECCS injection valves to open, because only one of the pressure transmitters
the reactor until the valves are closed by operator action, or actual reactor


is affected and the logic would still be satisfied.
pressure falls below approximately 446 kPa [50 psig] at which time the valves


A single failure could defeat this logic, however, preventing
can no longer stay open. Reactor depressurization and loss of inventory


all low-pressure
through the SRVs, in combination with the false low water level signal on the


ECCS Injection.
affected reference leg, would result in closure of the main steam isolation


The low-pressure
valves, actuation of high-pressure and low-pressure emergency core cooling


permissive
system (ECCS) and containment isolation. Low-pressure ECCS injection would


can be bypassed in the control room to open the injection
commence after the low-pressure permissive is satisfied. This permissive


valves for all four low pressure core spray (LPCS) pumps. The Susquehanna
would be satisfied in this scenario, allowing the low pressure ECCS injection


licensee has informed the NRC that it has physically
valves to open, because only one of the pressure transmitters is affected and


disabled the manual isolation valves to prevent misoperation
the logic would still be satisfied. A single failure could defeat this logic, however, preventing all low-pressure ECCS Injection. The low-pressure


of these valves; in addition, the valves are not readily accessible
permissive can be bypassed in the control room to open the injection valves


as they are located 6.1 meters [20 feet] above the floor.This event was recently analyzed for the LaSalle plant by Commonwealth
for all four low pressure core spray (LPCS) pumps. The Susquehanna licensee


Edison.The analysis indicates
has informed the NRC that it has physically disabled the manual isolation


that the low-pressure
valves to prevent misoperation of these valves; in addition, the valves are


permissive
not readily accessible as they are located 6.1 meters [20 feet] above the


for opening the low-pressure ECCS injection
floor.


valve would be defeated for the LaSalle design due to the false high pressure signal, thus preventing
This event was recently analyzed for the LaSalle plant by Commonwealth Edison.


ECCS injection
The analysis indicates that the low-pressure permissive for opening the low- pressure ECCS injection valve would be defeated for the LaSalle design due to


from the affected division.
the false high pressure signal, thus preventing ECCS injection from the


If a single failure is assumed in the relay for the low-pressure permissive
affected division. If a single failure is assumed in the relay for the low- pressure permissive on the other division, no low-pressure ECCS injection


on the other division, no low-pressure
would be available. Because the induced plant transient is potentially so


ECCS injection would be available.
severe, LaSalle has designed its backfill modification to make the injection


Because the induced plant transient
point for the backfill system on the reactor side of the manual isolation


is potentially
valve and excess flow check valve, thereby precluding the potential for


so severe, LaSalle has designed its backfill modification
pressurization of the reference leg through the backfill system.


to make the injection point for the backfill system on the reactor side of the manual isolation valve and excess flow check valve, thereby precluding
Commonwealth Edison took a different design approach for its Dresden and


the potential
Quad Cities plants. The backfill system design for Dresden and Quad Cities


for pressurization
injects into the reference leg on the instrument rack side of the manual


of the reference
isolation valve and excess flow check valve. Additional administrative


leg through the backfill system.Commonwealth
controls were developed to ensure that the isolation valve would not be


Edison took a different
inadvertently closed. The licensee analyzed the inadvertent closure of the


design approach for its Dresden and Quad Cities plants. The backfill system design for Dresden and Quad Cities injects into the reference
manual isolation valve for the Dresden and Quad Cities plants and concluded


leg on the instrument
that, while multiple SRVs would open, the resultant plant transient could be


rack side of the manual isolation
mitigated by appropriate operator actions. Without operator actions, the low- pressure ECCS would be available for event mitigation; however, a single


valve and excess flow check valve. Additional
failure in the instrumentation system could defeat the low-pressure permissive


administrative
for opening the low-pressure ECCS Injection valves and result in no low- pressure ECCS being available for this transient. The licensee also


controls were developed
determined that this design presented an unreviewed safety question because it


to ensure that the isolation
would increase the probability of a previously analyzed accident, and


valve would not be inadvertently
submitted an application to amend its license pursuant to 10 CFR 50.90. The


closed. The licensee analyzed the inadvertent
NRC is currently reviewing the licensee submittal.


closure of the manual isolation
IN 93-89 November 26, 1993 Other minor problems with the backfill system have been encountered when


valve for the Dresden and Quad Cities plants and concluded that, while multiple SRVs would open, the resultant
installing the system and returning the instrumentation to service after


plant transient
installation was complete. At the Perry plant, a problem occurred when the


could be mitigated
licensee was in the process of venting one of the instrument lines following


by appropriate
the installation of the modification. The job plan directed the operation of


operator actions. Without operator actions, the low-pressure ECCS would be available
the wrong valve, and instead of opening the vent valve the technician opened


for event mitigation;
the isolation valve, allowing air into the reference leg.
however, a single failure in the instrumentation


system could defeat the low-pressure
===As a result, the===
instrumentation associated with the high pressure core spray system (HPCS) was


permissive
inoperable until it was re-filled and vented. Similar events have occurred at


for opening the low-pressure
other plants due to procedural inadequacy or lack of attention to detail.


ECCS Injection
===Related Generic Communications===
*
NRC Information Notice 92-54, "Level Instrumentation Inaccuracies Caused


valves and result in no low-pressure ECCS being available
by Rapid Depressurization," July 24, 1992.


for this transient.
*
Generic Letter 92-04, "Resolution of the Issues Related to Reactor


The licensee also determined
Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f),"
August 19, 1992.


that this design presented
*
NRC Information Notice 93-27, 'Level Instrumentation Inaccuracies


an unreviewed
Observed During Normal Plant Depressurization," April 8, 1993.


safety question because it would increase the probability
*
NRC Bulletin 93-03,


of a previously
===Resolution of Issues Related to Reactor Vessel===
Water Level Instrumentation in BWRs,' May 28, 1993.


analyzed accident, and submitted
This information notice requires no specific action or written response. If


an application
you have any questions regarding the information in this notice, please


to amend its license pursuant to 10 CFR 50.90. The NRC is currently
contact the technical contact listed below or the appropriate Office of


reviewing
Nuclear Reactor Regulation (NRR) project manager.


the licensee submittal.
===Brian K. Grimes, Director===
Division of Operating Reactor Support


IN 93-89 November 26, 1993 Other minor problems with the backfill system have been encountered
===Office of Nuclear Reactor Regulation===
Technical contact: Amy Cubbage, NRR


when installing
(301) 504-2875 Attachments:
1. Simplified Sketch of Backfill Modification


the system and returning
2. List of Recently Issued NRC Information Notices


the instrumentation
rfl


to service after installation
tO


was complete.
===CONDENSATE POT===
REACTOR VESSEL


At the Perry plant, a problem occurred when the licensee was in the process of venting one of the instrument
(


lines following the installation
===QA BOUNDARY U===
DRYWELL


of the modification.
===REACTOR BLDG===
7 MANUAL


The job plan directed the operation
L ISOLATION


of the wrong valve, and instead of opening the vent valve the technician
===C EXCESS FLOW===
$ CHECK VALVE


opened the isolation
===REFERENCE LEG===
CRD CHARGING


valve, allowing air into the reference
===WATER HEADER===
INSTRtUm


leg. As a result, the instrumentation
RACK


associated
VARIABLE


with the high pressure core spray system (HPCS) was inoperable
LEG


until it was re-filled
C


and vented. Similar events have occurred at other plants due to procedural
FIGURE 1 - SIMPLIFIED SKETCH OF BACKFILL MODIFICATION


inadequacy
e0 (
cI-
Ij3 ED


or lack of attention
w,


to detail.Related Generic Communications
===A4...achment 2===
IN 93-89


* NRC Information
===November 26, 1993 LIST OF RECENTLY ISSUED===
NRC INFORMATION NOTICES


Notice 92-54, "Level Instrumentation
Information


Inaccuracies
Date of


Caused by Rapid Depressurization," July 24, 1992.* Generic Letter 92-04, "Resolution
Notice No.


of the Issues Related to Reactor Vessel Water Level Instrumentation
Subject


in BWRs Pursuant to 10 CFR 50.54(f)," August 19, 1992.* NRC Information
Issuance


Notice 93-27, 'Level Instrumentation
Issued to


===Inaccuracies===
93-88
Observed During Normal Plant Depressurization," April 8, 1993.* NRC Bulletin 93-03, Resolution
93-87
93-86
93-85
93-84 Status of Motor-Operated


of Issues Related to Reactor Vessel Water Level Instrumentation
Valve Performance Pre- diction Program by the


in BWRs,' May 28, 1993.This information
===Electric Power Research===
Institute


notice requires no specific action or written response.
===Fuse Problems with===
Westinghouse 7300


If you have any questions
===Printed Circuit Cards===
Identification of Iso- topes in the Production


regarding
and Shipment of Byproduct


the information
Material at Non-power


in this notice, please contact the technical
Reactors


contact listed below or the appropriate
Problems with X-Relays


Office of Nuclear Reactor Regulation (NRR) project manager.Brian K. Grimes, Director Division of Operating
in DB- and DHB-Type


===Reactor Support Office of Nuclear Reactor Regulation===
Circuit Breakers Manu- factured by Westinghouse
Technical


contact: Amy Cubbage, NRR (301) 504-2875 Attachments:
Determination of Westing- house Reactor Coolant
1. Simplified


===Sketch of Backfill Modification===
===Pump Seal Failure===
2. List of Recently Issued NRC Information
11/30/93
11/04/93
10/29/93
10/20/93
10/20/93


Notices
===All holders of OLs or CPs===
for nuclear power reactors.


rfl tO CONDENSATE
===All holders of OLs or CPs===
for nuclear power reactors.


POT REACTOR VESSEL (QA BOUNDARY U DRYWELL REACTOR BLDG 7 MANUAL L ISOLATION C EXCESS FLOW$ CHECK VALVE REFERENCE
===All holders of OLs or CPs===
for test and research


LEG CRD CHARGING WATER HEADER INSTRtUm RACK VARIABLE LEG C FIGURE 1 -SIMPLIFIED
reactors.


===SKETCH OF BACKFILL MODIFICATION===
===All holders of OLs or CPs===
e0 (cI-Ij3 ED w, A4...achment
for nuclear power reactors.


2 IN 93-89 November 26, 1993 LIST OF RECENTLY ISSUED NRC INFORMATION
===All holders of OLs or CPs===
for pressurized water


NOTICES Information
reactors (PWRs).


Date of Notice No. Subject Issuance Issued to 93-88 93-87 93-86 93-85 93-84 Status of Motor-Operated
93-83


Valve Performance
===Potential Loss of Spent===
Fuel Pool Cooling


Pre-diction Program by the Electric Power Research Institute Fuse Problems with Westinghouse
===Following A Loss of===
Coolant Accident (LOCA)
10/07/93 All holders


7300 Printed Circuit Cards Identification
for boiling


of Iso-topes in the Production
(BWRs).


and Shipment of Byproduct Material at Non-power Reactors Problems with X-Relays in DB- and DHB-Type Circuit Breakers Manu-factured by Westinghouse
of OLs or CPs


Determination
water reactors


of Westing-house Reactor Coolant Pump Seal Failure 11/30/93 11/04/93 10/29/93 10/20/93 10/20/93 All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for test and research reactors.All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for pressurized
93-82
93-81


water reactors (PWRs).93-83 Potential
===Recent Fuel and Core===
Performance Problems in


Loss of Spent Fuel Pool Cooling Following
===Operating Reactors===
Implementation of


A Loss of Coolant Accident (LOCA)10/07/93 All holders for boiling (BWRs).of OLs or CPs water reactors 93-82 93-81 Recent Fuel and Core Performance
===Engineering Expertise===
on Shift


Problems in Operating
10/12/93
10/12/93


Reactors Implementation
===All holders of OLs or CPs===
for nuclear power reactors


of Engineering
and all NRC-approved fuel


Expertise on Shift 10/12/93 10/12/93 All holders of OLs or CPs for nuclear power reactors and all NRC-approved
suppliers.


fuel suppliers.
===All holders of OLs or CPs===
for nuclear power reactors.


All holders of OLs or CPs for nuclear power reactors.OL -Operating
OL - Operating License


License CP -Construction
CP - Construction Permit


Permit
IN 93-89 a->
November 26, 1993 Other minor problems with the backfill system have been encountered when


IN 93-89 a-> November 26, 1993 Other minor problems with the backfill system have been encountered
installing the system and returning the instrumentation to service after


when installing
installation was complete. At the Perry plant, a problem occurred when the


the system and returning
licensee was in the process of venting one of the instrument lines following


the instrumentation
the installation of the modification. The Job plan directed the operation of


to service after installation
the wrong valve, and instead of opening the vent valve the technician opened


was complete.
the isolation valve, allowing air into the reference leg. As a result, the


At the Perry plant, a problem occurred when the licensee was in the process of venting one of the instrument
instrumentation associated with the high pressure core spray system (HPCS) was


lines following the installation
inoperable until it was re-filled and vented.


of the modification.
===Similar events have occurred at===
other plants due to procedural inadequacy or lack of attention to detail.


The Job plan directed the operation
===Related Generic Communications===
*
NRC Information Notice 92-54, "Level Instrumentation Inaccuracies Caused


of the wrong valve, and instead of opening the vent valve the technician
by Rapid Depressurization," July 24, 1992.


opened the isolation
*
Generic Letter 92-04, "Resolution of the Issues Related to Reactor


valve, allowing air into the reference
Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f),"
August 19, 1992.


leg. As a result, the instrumentation
*
NRC Information Notice 93-27, 'Level Instrumentation Inaccuracies


associated
Observed During Normal Plant Depressurization," April 8, 1993.


with the high pressure core spray system (HPCS) was inoperable
*
NRC Bulletin 93-03, 'Resolution of Issues Related to Reactor Vessel


until it was re-filled
Water Level Instrumentation in BWRs," May 28, 1993.


and vented. Similar events have occurred at other plants due to procedural
This information notice requires no specific action or written response. If


inadequacy
you have any questions regarding the information in this notice, please


or lack of attention
contact the technical contact listed below or the appropriate Office of


to detail.Related Generic Communications
Nuclear Reactor Regulation (NRR) project manager.


* NRC Information
orig /s/'d by BKGrimes


Notice 92-54, "Level Instrumentation
===Brian K. Grimes, Director===
Division of Operating Reactor Support


Inaccuracies
===Office of Nuclear Reactor Regulation===
Technical contact: Amy Cubbage, NRR


Caused by Rapid Depressurization," July 24, 1992.* Generic Letter 92-04, "Resolution
(301) 504-2875 Attachments:
1. Simplified Sketch of Backfill Modification


of the Issues Related to Reactor Vessel Water Level Instrumentation
2.


in BWRs Pursuant to 10 CFR 50.54(f)," August 19, 1992.* NRC Information
===List of Recently Issued NRC Information Notices===
* SEE PREVIOUS CONCURRENCE


Notice 93-27, 'Level Instrumentation
SRXB:DSSA* OGCB:DORS* TECH ED.*
SRXB:DSSA* SRXB:DSSA* D:DSSA* OGCB:DORS*
ACUBBAGE


===Inaccuracies===
PWEN
Observed During Normal Plant Depressurization," April 8, 1993.* NRC Bulletin 93-03, 'Resolution


of Issues Related to Reactor Vessel Water Level Instrumentation
RSANDERS


in BWRs," May 28, 1993.This information
WLYON


notice requires no specific action or written response.
RJONES


If you have any questions
===ATHADANI GMARCUS===
1
11/16/93
11/15/93
11/16/93
11/18/93
11/20/93
11/23/93
11/2.493 DOCUMENT NAME:
93-89.IN


regarding
IN 93-xx


the information
November xx, 1993 This information notice requires no specific action or written response. If


in this notice, please contact the technical
you have any questions regarding the information in this notice, please


contact listed below or the appropriate
contact the technical contact listed below or the appropriate Office of


Office of Nuclear Reactor Regulation (NRR) project manager.orig /s/'d by BKGrimes Brian K. Grimes, Director Division of Operating
Nuclear Reactor Regulation (NRR) project manager.


===Reactor Support Office of Nuclear Reactor Regulation===
===Brian K. Grimes, Director===
Technical
Division of Operating Reactor Support


contact: Amy Cubbage, NRR (301) 504-2875 Attachments:
===Office of Nuclear Reactor Regulation===
1. Simplified


===Sketch of Backfill Modification===
===Technical Contact:===
2. List of Recently Issued NRC Information


Notices* SEE PREVIOUS CONCURRENCE
===Amy Cubbage, NRR===
(301) 504-2875 Attachments:
1. Simplified Sketch of Backfill Modification


SRXB:DSSA*
2. List of Recently Issued NRC Information Notices
OGCB:DORS*
TECH ED.* SRXB:DSSA*
SRXB:DSSA*
D:DSSA* OGCB:DORS*
ACUBBAGE PWEN RSANDERS WLYON RJONES ATHADANI GMARCUS 1 11/16/93 11/15/93 11/16/93 11/18/93 11/20/93 11/23/93 11/2.493 DOCUMENT NAME: 93-89.IN


IN 93-xx November xx, 1993 This information
* SEE PREVIOUS CONCURRENCE


notice requires no specific action or written response.
SRXB:DSSA* OGCB:DORS* TECH ED.*
ACUBBAGE


If you have any questions
PWEN


regarding
RSANDERS


the information
11/15/93
11/16/93
11/15/93 SRXB:DSSA*
WLYON


in this notice, please contact the technical
11/16/93 SRXB:DSSA*
RJONES


contact listed below or the appropriate
11/18/93 D: DSSA*
ATHADANI


Office of Nuclear Reactor Regulation (NRR) project manager.Brian K. Grimes, Director Division of Operating
11/20/93 OGCB:DORS


===Reactor Support Office of Nuclear Reactor Regulation===
GMARCUS At'i
Technical


Contact: Amy Cubbage, NRR (301) 504-2875 Attachments:
11/23/93 D: DORS
1. Simplified


===Sketch of Backfill Modification===
BGRIMES
2. List of Recently Issued NRC Information


Notices* SEE PREVIOUS CONCURRENCE
11/ /93 DOCUMENT NAME:


SRXB:DSSA*
===BWRWTLVL.WEN===
OGCB:DORS*
TECH ED.*ACUBBAGE PWEN RSANDERS 11/15/93 11/16/93 11/15/93 SRXB:DSSA*
WLYON 11/16/93 SRXB:DSSA*
RJONES 11/18/93 D: DSSA*ATHADANI 11/20/93 OGCB:DORS GMARCUS At'i 11/23/93 D: DORS BGRIMES 11/ /93 DOCUMENT NAME: BWRWTLVL.WEN


instrumentation
instrumentation associated with the high pressure core spray system (HPCS) was


associated
inoperable until it was re-filled and vented.


with the high pressure core spray system (HPCS) was inoperable
===Similar events have occurred at===
other plants due to procedural inadequacy or lack of attention to detail.


until it was re-filled
===Related Generic Communications===
*
NRC Information Notice 92-54, "Level Instrumentation Inaccuracies Caused


and vented. Similar events have occurred at other plants due to procedural
by Rapid Depressurization," July 24, 1992.


inadequacy
*
Generic Letter 92-04, "Resolution of the Issues Related to Reactor


or lack of attention
Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f),"
August 19, 1992.


to detail.Related Generic Communications
*
NRC Information Notice 93-27, "Level Instrumentation Inaccuracies


* NRC Information
Observed During Normal Plant Depressurization," April 8, 1993.


Notice 92-54, "Level Instrumentation
*
NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel


Inaccuracies
Water Level Instrumentation in BWRs," May 28, 1993.


Caused by Rapid Depressurization," July 24, 1992.* Generic Letter 92-04, "Resolution
This information notice requires no specific action or written response. If


of the Issues Related to Reactor Vessel Water Level Instrumentation
you have any questions regarding the information in this notice, please


in BWRs Pursuant to 10 CFR 50.54(f)," August 19, 1992.* NRC Information
contact the technical contact listed below or the appropriate Office of


Notice 93-27, "Level Instrumentation
Nuclear Reactor Regulation (NRR) project manager.


===Inaccuracies===
===Brian K. Grimes, Director===
Observed During Normal Plant Depressurization," April 8, 1993.* NRC Bulletin 93-03, "Resolution
Division of Operating Reactor Support


of Issues Related to Reactor Vessel Water Level Instrumentation
===Office of Nuclear Reactor Regulation===


in BWRs," May 28, 1993.This information
===Technical Contact:===


notice requires no specific action or written response.
===Amy Cubbage, NRR===
(301) 504-2875 Attachments:
1.


If you have any questions
===Simplified Sketch of Backfill Modification===
2.


regarding
===List of Recently Issued NRC Information Notices===
EDITED BY:
R. Sanders


the information
DATED:
11/15/93 SRXB:DSSA* OGCB:DORS* SRXB:DSSA* SRXB:DSSA* D:DSSA


in this notice, please contact the technical
OGCB:DORS D:DORS


contact listed below or the appropriate
ACUBBAGE


Office of Nuclear Reactor Regulation (NRR) project manager.Brian K. Grimes, Director Division of Operating
PWEN


===Reactor Support Office of Nuclear Reactor Regulation===
WLYON
Technical


Contact: Amy Cubbage, NRR (301) 504-2875 Attachments:
RJONES
1. Simplified


===Sketch of Backfill Modification===
ATHAqkNI GMARCUS
2. List of Recently Issued NRC Information


Notices EDITED BY: R. Sanders DATED: 11/15/93 SRXB:DSSA*
BGRIMES
OGCB:DORS*
SRXB:DSSA*
SRXB:DSSA*
D:DSSA OGCB:DORS


D:DORS ACUBBAGE PWEN WLYON RJONES ATHAqkNI GMARCUS BGRIMES 11/15/93 11/16/93 11/16/93 11/18/93 11/?V/93 11/ /93 11/ /93* SEE PREVIOUS CONCURRENCE}}
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/93  
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/93
* SEE PREVIOUS CONCURRENCE}}


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Latest revision as of 10:47, 16 January 2025

Potential Problems with BWR Level Instrumentation Backfill Modifications
ML031070176
Person / Time
Site: Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Crane  Entergy icon.png
Issue date: 11/26/1993
From: Grimes B
Office of Nuclear Reactor Regulation
To:
References
IN-93-089, NUDOCS 9311190454
Download: ML031070176 (8)
v • d • e


UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C.

20555

November 26, 1993

NRC INFORMATION NOTICE 93-89: POTENTIAL PROBLEMS WITH BWR LEVEL

INSTRUMENTATION BACKFILL MODIFICATIONS

Addressees

All holders of operating licenses or construction permits for boiling water

reactors (BWRs).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees to potential problems that have been identified by

licensees involving hardware modification to the reactor vessel water level

instrumentation system. 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.

Description of Circumstances

NRC Bulletin (NRCB) 93-03, "Resolution of Issues Related to Reactor Vessel

Water Level Instrumentation in BWRs," issued on May 28, 1993, requested that

licensees implement hardware modifications necessary to ensure the level

instrumentation system design is of high functional reliability for long-term

operation. In response to this bulletin, all BWR licensees with the exception

of Big Rock Point, which does not use cold reference leg instrumentation, have

either implemented modifications or have committed to implement modifications.

The majority of these licensees have decided to install a reference leg

backfill system to supply a continuous flow of water from the control rod

drive (CRD) hydraulic system through the reference legs to preclude migration

of dissolved noncondensible gases down the legs. In August 1993, a potential

problem was found at the Susquehanna nuclear power plant during the design of

this backfill modification.

Discussion

It was postulated at Susquehanna that a manual isolation valve in one of the

reference legs (see Figure 1) could be closed by operator error.

Closure of

this valve would result in pressurization of that reference leg to CRD system

pressure and erroneous indication of low reactor water level and high reactor

pressure on all instrumentation associated with that reference leg. The

transient resulting from pressurization of the most limiting reference leg

9311190454

1 KE-e

93-o

93 Ill

K.,_

IN 93-89 November 26, 1993 includes reactor scram and opening of all safety relief valves (SRVs) due to

the false high reactor pressure. The SRVs would remain open and depressurize

the reactor until the valves are closed by operator action, or actual reactor

pressure falls below approximately 446 kPa [50 psig] at which time the valves

can no longer stay open. Reactor depressurization and loss of inventory

through the SRVs, in combination with the false low water level signal on the

affected reference leg, would result in closure of the main steam isolation

valves, actuation of high-pressure and low-pressure emergency core cooling

system (ECCS) and containment isolation. Low-pressure ECCS injection would

commence after the low-pressure permissive is satisfied. This permissive

would be satisfied in this scenario, allowing the low pressure ECCS injection

valves to open, because only one of the pressure transmitters is affected and

the logic would still be satisfied. A single failure could defeat this logic, however, preventing all low-pressure ECCS Injection. The low-pressure

permissive can be bypassed in the control room to open the injection valves

for all four low pressure core spray (LPCS) pumps. The Susquehanna licensee

has informed the NRC that it has physically disabled the manual isolation

valves to prevent misoperation of these valves; in addition, the valves are

not readily accessible as they are located 6.1 meters [20 feet] above the

floor.

This event was recently analyzed for the LaSalle plant by Commonwealth Edison.

The analysis indicates that the low-pressure permissive for opening the low- pressure ECCS injection valve would be defeated for the LaSalle design due to

the false high pressure signal, thus preventing ECCS injection from the

affected division. If a single failure is assumed in the relay for the low- pressure permissive on the other division, no low-pressure ECCS injection

would be available. Because the induced plant transient is potentially so

severe, LaSalle has designed its backfill modification to make the injection

point for the backfill system on the reactor side of the manual isolation

valve and excess flow check valve, thereby precluding the potential for

pressurization of the reference leg through the backfill system.

Commonwealth Edison took a different design approach for its Dresden and

Quad Cities plants. The backfill system design for Dresden and Quad Cities

injects into the reference leg on the instrument rack side of the manual

isolation valve and excess flow check valve. Additional administrative

controls were developed to ensure that the isolation valve would not be

inadvertently closed. The licensee analyzed the inadvertent closure of the

manual isolation valve for the Dresden and Quad Cities plants and concluded

that, while multiple SRVs would open, the resultant plant transient could be

mitigated by appropriate operator actions. Without operator actions, the low- pressure ECCS would be available for event mitigation; however, a single

failure in the instrumentation system could defeat the low-pressure permissive

for opening the low-pressure ECCS Injection valves and result in no low- pressure ECCS being available for this transient. The licensee also

determined that this design presented an unreviewed safety question because it

would increase the probability of a previously analyzed accident, and

submitted an application to amend its license pursuant to 10 CFR 50.90. The

NRC is currently reviewing the licensee submittal.

IN 93-89 November 26, 1993 Other minor problems with the backfill system have been encountered when

installing the system and returning the instrumentation to service after

installation was complete. At the Perry plant, a problem occurred when the

licensee was in the process of venting one of the instrument lines following

the installation of the modification. The job plan directed the operation of

the wrong valve, and instead of opening the vent valve the technician opened

the isolation valve, allowing air into the reference leg.

As a result, the

instrumentation associated with the high pressure core spray system (HPCS) was

inoperable until it was re-filled and vented. Similar events have occurred at

other plants due to procedural inadequacy or lack of attention to detail.

Related Generic Communications

NRC Information Notice 92-54, "Level Instrumentation Inaccuracies Caused

by Rapid Depressurization," July 24, 1992.

Generic Letter 92-04, "Resolution of the Issues Related to Reactor

Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f),"

August 19, 1992.

NRC Information Notice 93-27, 'Level Instrumentation Inaccuracies

Observed During Normal Plant Depressurization," April 8, 1993.

NRC Bulletin 93-03,

Resolution of Issues Related to Reactor Vessel

Water Level Instrumentation in BWRs,' May 28, 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 Attachments:

1. Simplified Sketch of Backfill Modification

2. List of Recently Issued NRC Information Notices

rfl

tO

CONDENSATE POT

REACTOR VESSEL

(

QA BOUNDARY U

DRYWELL

REACTOR BLDG

7 MANUAL

L ISOLATION

C EXCESS FLOW

$ CHECK VALVE

REFERENCE LEG

CRD CHARGING

WATER HEADER

INSTRtUm

RACK

VARIABLE

LEG

C

FIGURE 1 - SIMPLIFIED SKETCH OF BACKFILL MODIFICATION

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A4...achment 2

IN 93-89

November 26, 1993 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issuance

Issued to

93-88

93-87

93-86

93-85

93-84 Status of Motor-Operated

Valve Performance Pre- diction Program by the

Electric Power Research

Institute

Fuse Problems with

Westinghouse 7300

Printed Circuit Cards

Identification of Iso- topes in the Production

and Shipment of Byproduct

Material at Non-power

Reactors

Problems with X-Relays

in DB- and DHB-Type

Circuit Breakers Manu- factured by Westinghouse

Determination of Westing- house Reactor Coolant

Pump Seal Failure

11/30/93

11/04/93

10/29/93

10/20/93

10/20/93

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for test and research

reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for pressurized water

reactors (PWRs).

93-83

Potential Loss of Spent

Fuel Pool Cooling

Following A Loss of

Coolant Accident (LOCA)

10/07/93 All holders

for boiling

(BWRs).

of OLs or CPs

water reactors

93-82

93-81

Recent Fuel and Core

Performance Problems in

Operating Reactors

Implementation of

Engineering Expertise

on Shift

10/12/93

10/12/93

All holders of OLs or CPs

for nuclear power reactors

and all NRC-approved fuel

suppliers.

All holders of OLs or CPs

for nuclear power reactors.

OL - Operating License

CP - Construction Permit

IN 93-89 a->

November 26, 1993 Other minor problems with the backfill system have been encountered when

installing the system and returning the instrumentation to service after

installation was complete. At the Perry plant, a problem occurred when the

licensee was in the process of venting one of the instrument lines following

the installation of the modification. The Job plan directed the operation of

the wrong valve, and instead of opening the vent valve the technician opened

the isolation valve, allowing air into the reference leg. As a result, the

instrumentation associated with the high pressure core spray system (HPCS) was

inoperable until it was re-filled and vented.

Similar events have occurred at

other plants due to procedural inadequacy or lack of attention to detail.

Related Generic Communications

NRC Information Notice 92-54, "Level Instrumentation Inaccuracies Caused

by Rapid Depressurization," July 24, 1992.

Generic Letter 92-04, "Resolution of the Issues Related to Reactor

Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f),"

August 19, 1992.

NRC Information Notice 93-27, 'Level Instrumentation Inaccuracies

Observed During Normal Plant Depressurization," April 8, 1993.

NRC Bulletin 93-03, 'Resolution of Issues Related to Reactor Vessel

Water Level Instrumentation in BWRs," May 28, 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.

orig /s/'d by BKGrimes

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact: Amy Cubbage, NRR

(301) 504-2875 Attachments:

1. Simplified Sketch of Backfill Modification

2.

List of Recently Issued NRC Information Notices

  • SEE PREVIOUS CONCURRENCE

SRXB:DSSA* OGCB:DORS* TECH ED.*

SRXB:DSSA* SRXB:DSSA* D:DSSA* OGCB:DORS*

ACUBBAGE

PWEN

RSANDERS

WLYON

RJONES

ATHADANI GMARCUS

1

11/16/93

11/15/93

11/16/93

11/18/93

11/20/93

11/23/93

11/2.493 DOCUMENT NAME:

93-89.IN

IN 93-xx

November 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 Attachments:

1. Simplified Sketch of Backfill Modification

2. List of Recently Issued NRC Information Notices

  • SEE PREVIOUS CONCURRENCE

SRXB:DSSA* OGCB:DORS* TECH ED.*

ACUBBAGE

PWEN

RSANDERS

11/15/93

11/16/93

11/15/93 SRXB:DSSA*

WLYON

11/16/93 SRXB:DSSA*

RJONES

11/18/93 D: DSSA*

ATHADANI

11/20/93 OGCB:DORS

GMARCUS At'i

11/23/93 D: DORS

BGRIMES

11/ /93 DOCUMENT NAME:

BWRWTLVL.WEN

instrumentation associated with the high pressure core spray system (HPCS) was

inoperable until it was re-filled and vented.

Similar events have occurred at

other plants due to procedural inadequacy or lack of attention to detail.

Related Generic Communications

NRC Information Notice 92-54, "Level Instrumentation Inaccuracies Caused

by Rapid Depressurization," July 24, 1992.

Generic Letter 92-04, "Resolution of the Issues Related to Reactor

Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f),"

August 19, 1992.

NRC Information Notice 93-27, "Level Instrumentation Inaccuracies

Observed During Normal Plant Depressurization," April 8, 1993.

NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel

Water Level Instrumentation in BWRs," May 28, 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 Attachments:

1.

Simplified Sketch of Backfill Modification

2.

List of Recently Issued NRC Information Notices

EDITED BY:

R. Sanders

DATED:

11/15/93 SRXB:DSSA* OGCB:DORS* SRXB:DSSA* SRXB:DSSA* D:DSSA

OGCB:DORS D:DORS

ACUBBAGE

PWEN

WLYON

RJONES

ATHAqkNI GMARCUS

BGRIMES

11/15/93

11/16/93

11/16/93

11/18/93

11/?V/93 11/

/93

11/

/93

  • SEE PREVIOUS CONCURRENCE


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