Information Notice No. 93-89: Potential Problems with BWR Level Instrumentation Backfill Modifications

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D.C.  20555

                               November 26, 1993



All holders of operating licenses or construction permits for boiling water
reactors (BWRs).


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.

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 


                                                            IN 93-89
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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

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..

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                                                            Page 3 of 3

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,

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.

                                    /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

1.  Simplified Sketch of Backfill Modification
2.  List of Recently Issued NRC Information Notices

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