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89

Information Notice No. 93­89: Potential Problems with BWR Level

Instrumentation Backfill Modifications

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
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IN 93‐89

November 26, 1993

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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

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..

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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.

/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

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