Information Notice 1985-59, Valve Stem Corrosion Failures

Valve Stem Corrosion Failures

ML031180175
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, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	07/17/1985
From:	Jordan E NRC/IE
To:
References
IN-85-059, NUDOCS 8507120320
Download: ML031180175 (4)

SSINS No.: 6835 IN 85-59 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 July 17, 1985 IE INFORMATION NOTICE NO. 85-59: VALVE STEM CORROSION FAILURES

Addressees

All nuclear power reactor facilities holding an operating license (OL) or a

construction permit (CP).

Purpose

This information notice is provided to alert recipients of a potentially

significant problem pertaining to stress corrosion failures of valve stems and

shafts; items that are not routinely examined. It is suggested that recipients

review the information for applicability to their facilities and consider

actions, if appropriate, to preclude a similar problem occurring at their

facilities. However, suggestions contained in this information notice do not

constitute NRC requirements; therefore, no specific action or written response

is required.

Description of Circumstances

There have been four instances where cracks were found in 410 stainless steel

valve stems. These instances involved different licensees and different

manufacturers. Such cracks cannot be observed without the disassembly of the

valves, and the valve operability test programs do not provide a means of early

detection. In three of these instances, the cracks grew until the stem sheared

when the valve was activated. Such failures can prevent the system from

performing its safety function.

Uncontrollable leakage from the stem packing of several Velan globe valves was

reported by Oconee 1 in December 1971. Disassembly and examination revealed

cracks for the entire length of the stems and more than half of the diameter in

depth. In order to prevent cracking, 410 stainless steel needs to be tempered

immediately after hardening, but several batches were not tempered. Ultimate- ly, 2600 stems were replaced in the 1-1/2-, 1- and 1/2-inch valves, using

17-4PH and 300-series stainless steel materials.

A 20 inch Anchor/Darling gate valve stem snapped while being manually opened at

Brunswick 2 on August 4, 1982. There was pitting of the 410 stainless steel in

the gland packing section and the crack had initiated from one of these pits.

The cross-section area of the stem of the suppression pool suction valve had

been reduced by 70% by intergranular stress corrosion cracking (IGSCC). The

material had a higher hardness than specified as a result of improper heat

treatment.

8507120320

IN 85-59 July 17, 1985 Excessive hardness is associated with cracking and corrosion. The manufacturer

replaced five lots of valve stems with properly heat-treated 410 stainless

steel, and there have not been any further problems.

An injection valve in the low-pressure coolant injection (LPCI) system broke in

two places during disassembly at Browns Ferry 3 on February 28, 1984. One

break was below the stem packing area and the other was at the gate connection.

Over 50% of the cross-section of the stem of these 24-inch Walworth valves had

been lost in these areas by IGSCC. The stem had higher hardness than speci- fied. New stems made from 17-4 PH stainless steel were installed.

Linear indications were discovered on three main steam isolation valve (MSIV)

shafts at Farley 1 on February 29, 1984. The indications were from 1 to 13 inches long, contained thick oxides, and were located in the packing gland

area. The MSIVs were Atwood-Morrill 32-inch swing check valves and the shaft

hardness exceeded specifications. 17-4 PH stainless steel also was used as the

replacement material for the shafts.

Although 410 stainless steel is defined as a stainless steel because of its

alloy content, it is really a high chromium, very hardenable steel. Cooling

this material in air from the 1700-to-19001F temperature range results in a

surface hardness of up to HRC 45 and high internal stresses. Tempering the

hard and brittle martensite produces a softer and more ductile composition that

has much less chromium available for intergranular corrosion resistance.

Tempering in the 700-to-10500 F range is not recommended because it results in

low and erratic impact properties and poor resistance to corrosion and stress

corrosion.

The following conclusions were reached:

1. The actual hardness of the 410 stainless steel valve stems and shafts was

higher than specified and higher than documented.

2. The excessive hardness is associated with intergranular stress corrosion

cracking.

3. The cracking occurred in internal areas where there could be concentra- tions of corroding chemicals, such as at the gland packing.

4. The oxides found in the cracks showed that the cracks occurred during

service and grew slowly.

5. The cracks were not detected by the routine valve operability test pro- grams, but were only discovered by actual failures or after disassembly

during refueling outages.

6. Failure of these valves would make the specific safety system inoperable.

- - -

IN 85-59 July 17, 1985 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the Regional

Administrator of the appropriate regional office or this office.

.a or

Divisio Emergency Preparedness

and E jineering Response

Office of Inspection and Enforcement

P. Cortland, IE

(301) 492-4175 Attachment: List of Recently Issued IE Information Notices

Attachment 1 IN 85-59 July 17, 1985 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information Date of

Notice No. Subject Issue Issued to

85-58 Failure Of A General Electric 7/17/85 All power reactor

Type AK-2-25 Reactor Trip facilities designed

Breaker by B&W and CE holding

an OL or CP

85-57 Lost Iridium-192 Source 7/16/85 All power reactor

Resulting In The Death Of facilities holding

Eight Persons In Morocco an OL or CP; fuel

facilities; and

material licensees

85-56 Inadequate Environment 7/15/85 All power reactor

Control For Components And facilities holding

Systems In Extended Storage an OL or CP

Or Layup

85-55 Revised Emergency Exercise 7/15/85 All power reactor

Frequency Rule facilities holding

an OL or CP

85-54 Teletheraphy Unit Malfunction 7/15/85 All NRC licensees

authorized to use

teletheraphy units

85-53 Performance Of NRC-Licensed 7/12/85 All power reactor

Individuals While On Duty facilities holding

an OL or CP

85-52 Errors In Dose Assessment 7/10/85 All power reactor

Computer Codes And Reporting facilities holding

Requirements Under 10 CFR an OL or CP

Part 21 -

(

85-51 Inadvertent Loss Or Improper 7/10/85 All power reactor

Actuation Of Safety-Related facilities holding

Equipment an OL or CP

85-50 Complete Loss Of Main And 7/8/85 All power reactor

Auxiliary Feedwater At A PWR facilities holding

Designed By Babcock & Wilcox an OL or CP

OL = Operating License

CP = Construction Permit