Information Notice 1993-27, Level Instrumentation Inaccuracies Observed During Normal Plant Depressurization

Level Instrumentation Inaccuracies Observed During Normal Plant Depressurization

ML031080007
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
Issue date:	04/08/1993
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-93-027, NUDOCS 9304020319
Download: ML031080007 (8)

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 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

IN 93-27

April 8, 1993

Pge I of I

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LIST OF RECENTLY ISSUED

HRC INFORMATION NOTICES

inroruauon

Inforuti0n

Notice No.

93-26

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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

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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

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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

gauges.

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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