Information Notice 1986-06, Failure of Lifting Rig Attachment, While Lifting Upper Guide Structure at St. Lucie Unit 1

Failure of Lifting Rig Attachment, While Lifting Upper Guide Structure at St. Lucie Unit 1

ML031220538
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:	02/03/1986
From:	Jordan E NRC/IE
To:
References
IN-86-006, NUDOCS 8601290056
Download: ML031220538 (7)

SSINS No.: 6835 IN 86-06 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 February 3, 1986 IE INFORMATION NOTICE NO. 86-06: FAILURE OF LIFTING RIG ATTACHMENT,WHILE

LIFTING THE UPPER GUIDE STRUCTURE AT

ST. LUCIE UNIT 1

Addressees

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

permit (CP). .

Purpose

This notice is provided to advise licensees of a potentially significant problem

that occurred during the movement of a heavy load over the reactor core. It is

expected that recipients will review this information for applicability to their

facilities and consider actions, if appropriate, to preclude similar problems at

their facilities. However, the suggestions contained in this notice do not

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

is required.

Description of Circumstances

On November 6, 1985, while lifting the upper guide structure from the St.

Lucie Unit 1 reactor vessel, licensee personnel noticed the lifting rig tilt.

The lift was immediately stopped, with the lifting rig canted upward about 6 inches and the guide structure canted downward about 6 inches at one of the

three attachment points. An attempt was promptly made to lower the'load back

to its installed position, but the load cells indicated binding, so the attempt

was terminated after lowering the load a few inches. The 50-ton load was left

suspended about 8 feet above the reactor core.

The licensee declared an Unusual Event. Core alteration containment integrity

was enhanced by resuming full use of the airlocks. Temporary primary manway

covers were installed on bcth hot and cold legs to enhance the nozzle dams.

Survey transits were set up and procedures implemented to monitor the rig and

load for any shifts in their positions.

The licensee and the nuclear steam system supplier, Combustion Engineering, designed and tested a supplementary lifting rig to support the upper guide

structure from the upper portion of the normal rig, using a cable and J-hook

system. On November 9 with the supplementary rig installed, the load was

jacked to a level orientation and moved to its normal parking position in the

refueling pool.

8601290056

IN 86-06 February 3, 1986 Damage caused during the incident included bending the lifting rig and one of

the two guide pins that align the rig with the reactor vessel.

Discussion:

The upper guide structure is shown in Figure 1. It is supported in the reactor

vessel by it§..upper flange. It is aligned by eight alignment keys, four at the

top and fourat the bottom. The structure fits down inside the core support

barrel, just'above the fuel assemblies (see Figure 2). The fuel assembly

alignment plate is the bottom component of the upper guide structure.

The lifting rig is attached to the upper guide structure by three vertically

oriented bolts. These bolts are attached from above the water line by torque

tools that run down the hollow columns of the rig (see Figure 3). Combustion

Engineering's procedure for attaching the rig calls for checking for thread

engagement and torquing each bolt to 50 ft-lbs. The licensee's procedure

omitted the step concerning the check for thread engagement. Subsequent

inspection of the bolt that had pulled loose indicated that part of the last

thread was stripped. It is assumed that this bolt cross-threaded or bound due

to rig to guide structure misalignment during attachment and reached the 50

ft-lb torque requirement with only part of one thread engaged. During the

lift, the few inches of unengaged bolt shaft were pulled through the lifting

rig until the bolt head rested on the rig's surface at the bottom of the

column, resulting in an imperceptible tilt. The resulting lateral load was

initially s6pported by the guide pins. When the rig and guide structure were

lifted about 8 feet, where the guide bushings on the lifting rig reached the

tapered portion of the guide pins, it is surmised that sufficient lateral

motion was permitted to allow the thread of the improperly engaged bolt to slip

free. This caused the observed motion and tilt.

After the guide structure was supported by the supplemental lifting rig and

leveled, it was moved to its normal parking position in the refueling pool.

The short attachment bolts and torque tools were then replaced with full-length

bolts.- The long bolts are designed with heads that rest on surfaces at the top

of the three hollow columns of the lifting rig. This has the advantage of

making anyflmck of full thread engagement more apparent to the personnel

attaching the rig. The guide structure was subsequently returned to its

installed position using the long attachment bolts.

The licensee has not yet decided whether to permanently modify the attachment

bolts. The licensee plans to review all reactor-related lifts for adequacy of

the procedures to ensure proper lift rig attachment, including provisions for

measuring thread engagement.

The potential consequences of dropping heavy loads into the open reactor vessel

were addressed by Unresolved Safety Issue (USI) A-36, "Control of Heavy Loads

Near Spent Fuel." The concern for a UGS drop is that fuel assemblies might be

sufficiently damaged to release the radioactive gases and iodines held within

the fuel-clad gap. Under the reduced containment integrity requirements for

the refueling mode, damage to several fuel assemblies might cause the radiation

dose limits of 10 CFR 100 to be exceeded.

IN 86-06 February 3, 1986 Plant specific calculations were not made for a UGS drop at St. Lucie because

the NRC determined that further calculations were not required after reviewing

initial calculations previously submitted by other reactor facilities in

response to Phase II of USI A-36. Some indication of the consequences of a

UGS drop at St. Lucie can be gained from calculations performed by Combustion

Engineering for a reactor vessel head drop at Waterford 3. The head drop

calculations assumed the reactor vessel head was sufficiently tilted at impact

to directly strike the UGS with the UGS at rest in its normal installed posi- tion. The calculated response velocity of the Waterford UGS was 28 feet per

second, and the resulting vertical stresses imposed on the fuel were not

sufficient to rupture the cladding.

If the St. Lucie UGS had dropped from an 8 foot elevation, its striking velocity

would have been substantially less than the UGS response velocity calculated

for the Waterford head drop. However, the potential for misalignment of the

recesses in the bottom of the UGS (i.e., the fuel assembly alignment plate)

with the fuel assembly upper end fitting posts was not addressed by the Waterford

scenario. If substantial misalignment occurred, the fuel could be subjected to

additional axial loading. Significant misalignment could not occur without

substantial impact damage to the eight keys and keyways, which would also

result in a reduced striking velocity of the UGS as it reached the fuel. On

this basis, significant radioactive gas release is considered to be unlikely, although it has not been shown to be impossible.

No specific action or written response is required by this notice. If you

have any questions regarding this matter, please contact the Regional

Administrator of the appropriate NRC regional office or this office.

ar Jordan, Director

Divisio of Emergency Preparedness

and gineering Response

Office of Inspection and Enforcement

Technical Contacts: S. M. Long, IE

(301) 492-7159 D. E. Sells, NRR

(301) 492-9735 Attachments:

1. Figure 1, Upper Guide Structure Assembly

2. Figure 2, Reactor Internals Assembly

3 Figure 3, Upper Guide Structure Lifting Rig

4. List of Recently Issued IE Information Notices

Attachment 1 IN 86-06 February 3, 1986 EXPANSION

COMPENSATING

RING

CEA

SHROUD

GRID

ASSEMBLY.

CEA SHROUDS

FUEL ASSEMBLY

ALIGNMENT PLATE

Figure 1: Upper Guide Structure Assembly

(St. Lucie Unit 1 FSAR-figure 4.2-10)

UPPER GUIDE

-STRUCTURE SUPPORT

PLATE

Attachment 2 IN 86-06 February 3, 1986 CEA

SHROUD

IN-CORE

-INSTRUMENTATION

OUTLET GUIDE TUBE

CORE

' SUPPORT

BARREL

FUEL

ALIGNMENT

PINS

CORE

-SUPPORT

ASSEMBLY

Am. 3-7/85 Figure 2: Reactor Internals Assembly

(St. Lucie Unit 1 FSAR figure 4.2-7)

At ta chment3 IN 86-06 February 3, 1986 LIFT POINT FOR CRANE HOOK

CLEVIS ASSEMBLY

WORKING PLATFORM

I NSTRUMENT

STALK OPENING

- COLUMN

GUIDE BUSHING FOR

R.V. GUIDE PIN

a PLACES)

- 00

LIFT BOLT

Fi gure 3: Upper Guide Structure Lifting Rig

(St. Lucie Unit 1 FSAR figure 9.1-8)

Attachment 4 IN 86-06 February 3, 1986 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Tnfn,'mnfi n

Notice No. Subject

Date oT

Issue T-Issud tn

Issu---- ud -- t

86-05 Main Steam Safety Valve Test 1/31/86 All PWR facilities

Failures And Ring Setting holding an OL or

Adjustments CP

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