Information Notice 1986-05, Main Steam Safety Valve Test Failures and Ring Setting Adjustments

Main Steam Safety Valve Test Failures and Ring Setting Adjustments

ML031220529
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, Barnwell
Issue date:	01/31/1986
From:	Jordan E NRC/IE
To:
References
IN-86-005, NUDOCS 8601290054
Download: ML031220529 (4)

11$ ORIGINAL SSINS No.: 6835 IN 86-05 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 January 31, 1986 IE INFORMATION NOTICE NO. 86-05: MAIN STEAM SAFETY VALVE TEST FAILURES AND

RING SETTING ADJUSTMENTS

Addressees

All pressurized-water-reactor (PWR) facilities holding an operating license

(OL) or a construction permit (CP).

Purpose

This notice is being provided to alert recipients of a potentially significant

problem pertaining to spring-actuated main steam safety valves that may possess

less than the full-rated flow capacity. It is expected that recipients will

review the information for applicability to their facilities and consider

actions, if appropriate, to preclude a similar problem at their facilities.

However, suggestions contained in this information notice do not constitute NRC

requirements; therefore, no specific action or written response is required.

K..i NRC is continuing to obtain and evaluate pertinent information. If specific

actions are determined to be required by NRC, an additional notification will

be made.

Description of Circumstances

In the fall of 1984, Public Service of New Hampshire sent the main steam safety

valves (MSSVs) for its Seabrook plant to Wyle Laboratories for full-flow

testing to determine the proper vent stack size. To determine full flow, Wyle

measured disc travel of the model number 6R10 valves manufactured by the Crosby

Valve and Gage Company. The results of the tests indicated that the valves

could not achieve the required disc travel with the factory-set ring setting

(+155 notches). The disc travel achieved was 50% of the full lift necessary to

develop required steam flow capacity. Adequate lift was not attainable even

with the largest diameter tailpipe.

Additional tests were done in July 1985 to determine the appropriateness of the

ring settings. Specifically, the tests were to determine if the "as-shipped"

ring settings of the valves would allow the required disc travel with minimum

tailpipe backpressure and to determine the effects on valve disc travel for a

range of backpressures between 180 and 390 psig. During these tests, the upper

(guide) ring setting was adjusted from +155 to 0 and then to +25 to achieve the

required disc travel. This is a substantial adjustment. Subsequently, the

8601290054 I

IN 86-05 January 31, 1986 licensee consulted with the valve manufacturer and agreed on ring settings of

+25 for the guide ring and -25 (the original setting) for the lower (nozzle)

ring (see figure 1).

Full flow, full size tests of the sort described in this notice are not normal- ly performed by the licensee or valve vendor for large secondary safety valves, nor are they required by the ASME Code,Section III. Instead the valves are

certified by extrapolations on data from tests of smaller valves.

The MSSVs on most PWRs, while not necessarily the same model or manufacturer as

those at Seabrook, are generally at the upper end of the valve size range.

This raises the concern that full-sized flow demonstration may never have been

performed for many MSSVs and these may have incorrect ring settings. In

addition, similar problems with ring settings have been found when full-size

tests were performed for PWR primary safety valves. Thus, these MSSVs may not

be capable of providing full-relief capacity in accordance with facility design

requirements.

NRC is continuing to obtain and evaluate pertinent information. If specific

actions are determined to be required by NRC, an additional notification will

be made.

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.

LA

Jordan, Director

Oivisi2n of Emergency Preparedness

and Engineering Response

Office of Inspection and Enforcement

Mary S. Wegner

(301) 492-4511 Attachments:

1. Figure 1, Typical Main Steam Safety Valve

2. List of Recently Issued IE Information Notices

Attachment 1 IN 86-05 January 31, 1986

9 FIG. I

TYPICAL MAIN STEAM SAFETY VALVE

I A .-.

Attachment 2 IN 86-05 January 31, 1986 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information Date of

Notice No. Subject Issue Issued to

86-04 Transient Due To Loss Of 1/31/86 All power reactor

Power To Integrated Control facilities holding

System At A Pressurized Water an OL or CP

Reactor Designed By Babcock

& Wilcox

86-03 Potential Deficiencies In 1/14/86 All power reactor

Environmental Qualification facilities holding

Of Limitorque Motor Valve an OL or CP

Operator Wiring

86-02 Failure Of Valve Operator 1/6/86 All power reactor

Motor During Environmental facilities holding

Qualification Testing an OL or CP

86-01 Failure Of Main Feedwater 1/6/86 All power reactor

Check Valve Causes Loss Of facilities holding

Feedwater System Integrity an OL or CP

And Water-Hammer Damage

85-101 Applicability of 10 CFR 21 12/31/85 All power reactor

To Consulting Firms Providing facilities holding

Training an OL or CP

85-100 Rosemount Differential 12/31/85 All power reactor

Pressure Transmitter Zero facilities holding

Point Shift an OL or CP

85-99 Cracking In Boiling-Water- 12/31/85 All BWR facilities

Reactor Mark I And Mark II having a Mark I or

Containments Caused By Failure Mark II containment

Of The Inerting System

85-98 Missing Jumpers From Westing- 12/26/85 All Westinghouse

house Reactor Protection designed PWR

System Cards For The Over- facilities holding

Power Delta Temperature Trip an OL or CP

Function

OL = Operating License

CP = Construction Permit