Information Notice 1985-14, Failure of a Heavy Control Rod (B4C) Drive Assembly to Insert on a Trip Signal

Failure of a Heavy Control Rod (B4C) Drive Assembly to Insert on a Trip Signal

ML031180614
Person / Time
Site:	Beaver Valley, Millstone, Hatch, 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, 05000262, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	02/22/1985
From:	Jordan E NRC/IE
To:
References
IN-85-014, NUDOCS 8502190265
Download: ML031180614 (5)

SSINS No.: 6835 IN 85-14 UNITED STATES.

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 February 22, 1985 IE INFORMATION NOTICE NO. 85-14: FAILURE OF A HEAVY CONTROL ROD (B4C) DRIVE

ASSEMBLY TO INSERT ON A TRIP SIGNAL

Addressees

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

construction permit (CP).

Purpose

This information notice is to alert recipients of a potentially significant

event pertaining to a stuck drive rod assembly of a control rod drive mechan- ism (CRDM). This event occurred while performing hot rod drop tests at a

foreign pressurized water reactor designed by Westinghouse Electric Corporation

(Westinghouse). This event was caused by a breech guide screw that became

disengaged from the external breech of a drive rod assembly and fell on top of

the CRDM latch assembly where it became lodged and prevented driveline motion.

It is expected that recipients will review the information contained in this

notice for applicability to their facilities and consider actions, if appro- priate, to preclude similar problems from occurring at their facilities.

However, suggestions contained in this notice do not constitute Nuclear

Regulatory Commission (NRC) requirements; therefore, no specific action or

written response is required.

Description of Circumstances

On the afternoon of December 5, 1984, Westinghouse notified Duke Power Company

(Duke) of an event that occurred on November 19, 1984, at a foreign reactor

concerning the CRDM design similar to the one installed at McGuire 2 (the

McGuire 1 design is completely different). Based on initial information, Westinghouse considered the event to be an isolated case. However, on the

afternoon of December 6, 1984, Westinghouse notified Duke of unfavorable inspec- tion results on similar CRDMs at several plants and of the impact on the opera- tion of McGuire Unit 2. Based on this new information, Duke orally informed

NRC Region II of the event and began a safety assessment of continued operation

of McGuire 2, the results of which were provided to NRC personnel by conference

telephone call on the afternoon of December 6, 1984, and subsequently documented

in a letter dated December 12, 1984, from H. B. Tucker to J. P. O'Reilly.

By letter dated December 7, 1984, E. P. Rahe, Jr. to R. C. DeYoung, Westinghouse

documented Mr. Rahe's telephone call of December 7, 1985 with Mr. C. E. Rossi

reporting the event associated with CRDM heavy drive rod assemblies under

10 CFR 21, identified as a Potential Substantial Safety Hazard, for one

8502190265

IN 85-14 February 22, 1985 operating plant (McGuire 2) and for six construction plants (Catawba 1 and 2, Seabrook 1 and 2, Watts Bar 1 and 2). (Note: At that time, Catawba 1 had been

issued a low power testing license but had not yet attained initial criticality.)

The drive rod became stuckrat a foreign reactor during downward stepping while

performing hot rod drop tests as part of preoperational testing prior to

achieving initial criticality. Subsequent on-site investigations revealed that

a breech guide screw from a CRDM heavy drive rod assembly had rotated out of

position and fell on top of the CRDM latch assembly where it became lodged and

prevented driveline motion.

The function of the breech-guide screw is to provide alignment and guidance

during coupling and uncoupling of the drive rod from the rod cluster control

assembly during refueling. The breech guide screw is 0.52 inch long and has

a 0.433 inch diameter. If-a breech guide screw were to rotate out of the drive

rod assembly, it would fall- into the annulus between the external breech and

the rod travel housing. Although this annulus is nominally 3/8 inch wide, it

is sufficiently flexible to allow a loose breech guide screw to travel downward

during rod stepping. The loose breech guide screw would then lodge on top of

the CRDM latch assembly, potentially causing misstepping, intermittent stick- ing of the driveline or a totally stuck driveline. Such a loose breech guide

screw would not be able to pass below the above position because the clearance

between the guide tube and-the drive rod assembly is only 0.055 inch.. Thus, if a breech guide screw should become loose, it would not be able to migrate

into the reactor upperhead-regi-on- during-plant operation;-therefore,--it-would

not subsequently be expected to become a loose part in the reactor coolant

system.

The breech guide screw of a heavy drive rod assembly is designed to be held in

position by a locking pin. The locking pin is inserted in a drilled hole that

intersects the mating threads of the breech guide screw and those of a threaded

hole in the external breech of the drive rod assembly. The locking pin, in turn, is welded to the head of the breech guide screw to ensure that it stays in place.

(See Attachment 1, Drive Rod Assembly.) However, in the foreign plant, the

breech guide screw was drilled at an angle such that the locking pin did not

intersect the threads. Thus, the locking pin was ineffectual in that the breech

guide screw was not actually locked in place but was free to unscrew from the

external breech.

Subsequent to the drive rod becoming stuck at a foreign facility, Westinghouse

recommended that a reverse torque test be conducted on the plant's remaining

51 breech guide screws. This reverse torque test consists of applying twice

the installation torque on the breech guide screw but in the reverse direction, with a test failure being either the complete unscrewing of the breech guide

screw or a rotation of 15° or more of the screw. Three breech guide screws at

the affected foreign plant became unscrewed when tested. Afterwards, the breech

guide screws were similarly tested at another foreign reactor using similar

CRDM heavy drive rods to those at the plant that experienced a stuck drive rod.

These tests revealed several (exact number unknown) breech guide screws that

were classified as being either finger tight or loose. In addition, similar

tests were conducted at Catawba 2, Seabrook 1, and Watts Bar 1 and 2 with the

IN 85-14 February 22, 1985 following results: (1) Catawba 2 - of the 57 rod drive assemblies tested, five

breech guide screws were found to be finger tight and were completely unscrewed, and 20 breech guide screws rotated when reverse torque tested; (2) Seabrook 1 -

of the 57 rod drive assemblies tested, 17 breech guide screws became unscrewed

and one rotated but did not become unscrewed when reverse torque tested; and (3)

Watts Bar 1 and 2 - of the 144 rod drive assemblies tested, two breech guide

screws were found to be finger tight and were completely unscrewed and 20 became

unscrewed when reverse torque tested. Finally, the breech guide screws at

Catawba 1 were reverse torque tested, at which time two breech guide screws

became unscrewed and three breech guide screws were found with disfigured heads

that blocked the attaching of the torque wrench to the screws. Following these

tests, Duke replaced 14 CRDMs at Catawba 1.

Westinghouse has advised its affected utility customers of the event and has

provided them with recommendations for operating plants and plants under

construction. Westinghouse recommends that potentially affected operating

plants take the following actions:

1. Increase the frequency of control rod stepping tests from once every

31 days to once every 7 days.

2. If rod stepping anomalies of a mechanical nature occur during these

stepping tests or during any normal rod stepping, the plant should be

shut-down and-the drive rod assemblies inspected.

3. If no rod stepping anomalies of a mechanical nature occur, the drive

rod assemblies should be inspected at the next scheduled outage.

Westinghouse is inspecting the CRDMs of all affected plants under construction

and will repair those assemblies with loose breech guide screws. Westinghouse

has developed a repair procedure for any breech guide screw that fails the

inspection. The procedure consists of drilling another hole in the breech

guide screw at least 90 degrees from the existing hole, inserting a locking

pin in the new hole and welding the locking pin in place.

No specific action or written response is required by this information notice;

however, if you have any question regarding this notice, please contact the

Regional Administrator of the appropriate NRC regional office or the technical

contact listed below.

rd L. Jordan, Director

Division of Emergency Preparedness

and Engineering Response

Office of Inspection and Enforcement

I. Villalva, IE

(301) 492-9007 Attachments:

1. Drive Rod Assembly

2. List of Recently Issued IE Information Notices

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Attachment 2 IN 85-14 February 22, 1985 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information Date of

Notice No. Subject Issue . Issued to

85-13 Consequences of Using I02/21/85 All power- reactor

Soluble Dams - facilities. holding

a CP 0i

85-12 Recent Fuel Handling Events ID2/11/85 All power reactor

facilities holding

a CP

85-11 Licensee Programs For 2/11/85 All power reactor

Inspection Of Electrical facilities holding

Raceway And Cable Installation a CP

85-10 Posttensioned Containment  ?/6/85 All power reactor

Tendon Anchor Head Failure facilities holding

an OL or CP

85-09 Isolation Transfer Switches 1/31/85 All power reactor

And Post-Fire Shutdown facilities holding

Capability an OL or CP

85-08 Industry Experience Onl 1/30/85 All power reactor

Certain Materials Used In 'facilities holding

Safety-Related Equipment' an OL or CP

85-07 Contaminated Radiography 1/29/85 All NRC licensees

Source Shipments authorized to

possess industrial

radiography sources

85-06 Contamination of Breathing I1/23/85 All power reactor

Air Systems facilities holding

an OL or CP

85-05 Pipe Whip Restraints 1/23/85 All power reactor

facilities holding

an OL or CP

85-04 Inadequate'Management Of 1/17/85 All power reactor

Security Response Drills facilities holding

an OL or CP, & fuel

fabrication & pro- cessing facilities

OL = Operating License .

CP = Construction Permit