Information Notice 1997-31, Failures of Reactor Coolant Pump Thermal Barriers and Check Valves in Foreign Plants

Failures of Reactor Coolant Pump Thermal Barriers and Check Valves in Foreign Plants
	ML031050614
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, Three Mile Island, 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
Issue date:	06/03/1997
From:	Slosson M; Office of Nuclear Reactor Regulation
To:	;
References
	IN-97-031, NUDOCS 9705300075
	Download: ML031050614 (8)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 June 3, 1997 NRC INFORMATION NOTICE 97-31: FAILURES OF REACTOR COOLANT PUMP

THERMAL BARRIERS AND CHECK VALVES IN

FOREIGN PLANTS

Addressees

All holders of operating licenses or construction permits for pressurized-water reactor (PWR)

plants.

Purpose

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

addressees to reported problems discovered at foreign pressurized-water reactor (PWR)

plants. Degraded conditions of the reactor cooling pump (RCP) thermal barriers were found

after 10 years of service in the French 900-MWe PWR plants. 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.

DescriDtion of Circumstances

The degraded conditions were described in a paper presented during the July 1996 NRC/American Society of Mechanical Engineers (ASME) Symposium on Pump and Valve

Testing (NUREG/CP-0152, July 1996) and subsequently discussed with the French

authorities. After 10 years of service a sample inspection was conducted to determine the

condition of a 900-MWe French PWR RCP. The inspection was carried out on RCP No. 2 at

the Fessenheim Unit 2 Power Station. The RCP had completed 95,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation.

During the visual examination, a crack was detected on the outside surface of the thermal

barrier housing. Further detailed examinations revealed other cracks inside the housing and

on the underside of the thermal barrier flange. Additionally, the thermal barrier cooling coil

isolation check valves were found to be jammed open.

Subsequently, other inspections revealed that cracks existed in almost all of the 900-MWe

plant RCP thermal barrier housings. Examinations performed on a 1300-MWe power station

did not reveal the cracking phenomenon. About two-thirds of the check valves in the

900-MWe and 1300-MWe plants were found jammed open.

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IN 97-31 June 3, 1997

- Discussion

The RCP thermal barrier, located directly above the pump impeller, prevents the hot reactor

primary coolant (RPC) from reaching the RCP shaft seals. Exposure of the shaft seals

directly to hot versus cooled RPC. would degrade the seals. The thermal barrier is cooled by

a cooling coil, which circulates water supplied from the component cooling water (CCW)

system. The thermal barrier housing shields the cooling coil from the direct thermal effects of

the hot primary coolant. Check valves are located in the supply lines to the thermal barrier

cooling coil for the purpose of isolating the CCW system from the reactor coolant system in

the event of a cooling coil rupture.

The safety concerns were identified and evaluated with regard to the potential consequences

of (1) the formation of loose parts, generated by the thermal barrier housing, which could

damage the pump seals, (2) the jamming of the pump impeller, (3) the rupture of the cooling

coil without the check valve isolation of the CCW system, and (4) the disintegration of the

thermal barrier flange.

On the basis of metallographic examinations, it was concluded that intergranular brittle

rupture of niobium carbides occurred at the grain boundaries in some housings as a result of

the very low ferrite content of the material (UNS S34700 SS). Further, it was concluded that

the circumferential cracking located on the inside cylindrical surface at the interface with the

base of the thermal barrier housing resulted from a fatigue phenomenon. The thermal barrier

flange required further evaluation and testing to understand the root cause of the cracking of

the surface.

The cause of the jammed-open check valves was attributed to a layer of metallic oxide

deposits generated in the CCW system carbon steel piping. The valves were a lift-type

check valve. Lift-type check valve internals, particularly in the smaller valves, typically have

small clearances and passages for coolant flow and are highly susceptible to corrosion

products buildup.

The corrective actions taken for these plants included (1) the installation of alert alarms and

monitoring programs to detect high temperatures in the CCW system coolant at the thermal

barrier outlet, (2) the replacement of the thermal barrier housings with a configuration- modified unit and a material change, (3) a check on the condition of thermal barrier flanges in

3 years, (4) the inservice testing of the lift-type check valves during refueling outages, anct

(5) the systematic replacement of the lift check valves with swing check valves during

refueling outages.

In some recent events, similar failures at domestic plants have been reported with regard to

RCP thermal barrier check valves. PWR components in safety-related code class systems

such as the RCP check valves located in CCW system flowpaths are subject to the

ASME/OM Code inservice testing requirements.

On November 16, 1993, with Sequoyah Nuclear Plant Unit I in cold shutdown and Unit 2 at

100-percent power, the Tennessee Valley Authority reported that both units were outside

their design basis as a result of eight inoperable check valves in the Unit 1 component

KU--1 IN 97-31 June 3, 1997 cooling system piping, located upstream of the RCP thermal barrier heat exchangers. The

Unit I condition was discovered by radiographic inspections of the check valves. The

inspections indicated that seven of the eight check valves were stuck in the open position.

The eighth valve was found to be improperly assembled. Subsequent inspections of

corresponding valves in Unit 2 indicated that seven valves were stuck in the open position.

The cause of the condition was iron oxide corrosion product buildup wedging between the

valve piston and the bonnet. The Unit 2 valves were cleaned, reassembled, and returned to

service. The carbon steel bonnets on the Unit 1 valves were replaced with stainless steel

bonnets.

On October 29, 1991, the Comanche Peak Unit 1 stop check valves (SCVs) in the CCW lines

to the RCP thermal barriers were being tested to satisfy inservice testing requirements.

During the test, five of the eight SCVs failed to close. The valves were subsequently

manually exercised, after which they operated as designed. On November 6, 1991, two of

the failed valves were inspected. A small accumulation of corrosion products between the

plug and the stem of the valves, and a slight scaling along the bore, were found. Larger

accumulations of corrosion products that may have been present were flushed out when the

valves were manually exercised. The root cause of the event was attributed to the

accumulation of corrosion products and less than adequate preventive maintenance.

Corrective action included the development of a preventative maintenance procedure to

manually exercise these valves.

Similar valve fouling conditions were Indicated during inservice testing at several other plants

over the past 5 years.

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 contacts

listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

Ma e M. Slosson, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Francis Grubelich, NRR

301-415-2784 E-mail: fxg@nrc.gov

Eric J. Benner, NRR

301-415-1171 E-mail: ejgl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

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Attachment

IN 97-31 June 3, 1997 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

97-30 Control of Licensed 06/03/97 All material and fuel

Material During Reorgan- cycle licensees

izations, Employee- Management Disagreements, and Financial Crises

97-29 Containment Inspection 05/30/97 All holders of OLs or CPs

Rule for nuclear power reactors

97-28 Elimination of Instrument 05/30/97 All holders of OLs or CPs

Response Time Testing

Under the Requirements of

10 CFR 50.59

97-27 Effect of Incorrect 05/16/97 All holders of OLs or CPs

Strainer Pressure Drop for light-water power

on Available Net Positive reactors, except those

Suction Head licensees who have per- manently ceased operations

and have certified that

fuel has been permanently

removed from the reactor

vessel

97-26 Degradation in Small- 05/19/97 All holders of OLs or CPs

Radius U-Bend Regions for pressurized-water

of Steam Generator Tubes reactors

87-10, Potential for Water 05/15/97 All holders of OLs or CPs

Sup. 1 Hammer During Restart of for boiling-water reactors

Residual Heat Removal

Pumps

97-25 Dynamic Range Uncertain- 05/09/97 All holders of OLs or CPs

ties in the Reactor Vessel for Westinghouse pressurized- Level Instrumentation water reactors

OL = Operating License

CP = Construction Permit

IN 97-31 June 3, 1997 cooling system piping, located upstream of the RCP thermal barrier heat exchangers. The